ML20080S350
| ML20080S350 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/31/1984 |
| From: | Stickley T EG&G, INC. |
| To: | NRC |
| Shared Package | |
| ML20080S355 | List: |
| References | |
| CON-FIN-A-6457, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR EGG-HS-6294, EGG-HS-6294-R03, EGG-HS-6294-R3, NUDOCS 8402280725 | |
| Download: ML20080S350 (27) | |
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EGG-HS-6294 Revision 3 9
CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS FORT ST. VRAIN NUCLEAR GENERATING STATION PHASE I Docket No. 50-267 Author T. H. Stickley a
Principal Technical Investigator T. H. Stickley 3
Published May 1983 Revised November 1983 Revised January 1984 EG&G Idaho, Inc.
Prspared for the U.S. Nuclear Regulatory Commi.<sion Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6457 2247W qq$
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ABSTRACT The Nuclear Reculatory Commission (NRC) has requested that all nuclear plants either operating or under construction submit a response of compliancy with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants."
EG&G Idaho, Inc., ha.s contracted with the NRC to evaluate ne responses of those plants presently under construction. This report contains EG&G's evaluation and reco.nmendations for " Fort St. Vrain Nuclear Generating Station (F5VNGS).
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EhdCUTIVE
SUMMARY
Fort St. Vrain Nuclear.. tnerating Station (FSVNGS) is censis ent with.
5.1.1 of NUREG-0512.
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CCNTENTS ABSTRACT.............................................................
11 EXECUTIVE
SUMMARY
111 1.
INTRODUCTION...................................................
I 1.1 Purpose of Review.....................,...................
I 1.2 Ge n e : i c Ba c kg ro und........................................
1 1.3 Diant-Specific Background.................................
3 2.*
EVALUATION AND RECOMMENDATIONS.................................. 4 2.1 Overview.................................................. 4 2.2 Heavy Load Overhead Handling Systems.....................
A 2.3 General Guide 11nes........................................
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-2.4 Interim Protection Measures............................... 15 3.
CONCLUDING
SUMMARY
.........................................,..... 17 3.1 Applicable Load-Handling Systems.......................... 17 3.2 Guideline Recommendations................................. 17 3.3 Interim Protection........................................ 18
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TABLES Table 3.1 NUREG-0612 Compliance Matrix, Fort St. Vrain Nuclear Generating Station........................ 19 e
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CONTROL OF HEAVY LhADS AT NUCLEAR POWER PLANTS
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FORT ST. VRAIN NUCLEAR GENERATING STATION (PHASE I')
3 1.
INTRODUCTION 1.1 Purcose of Review This technical evaluation report documents the EG&G Idaho, Inc.,
review of general load-handling policy and procedures at Fort St. Vrain Nuclehr Generating Station (FSVNGS). This evaluation was performed with the objective of assessing conformance to the general lead-hancling guidelines of NUREG-0612, " Control of Heavy Leads at Nuclear Power Plants" [1], Section 5.1.1.
1.2 Generic Backcround Generic Technical Activdty. Task A-36 was established by the U.S.
Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at opera. ting nuclear power plants to assure the safe handling of heavy loads and to recommend necessary changes to these measures.
This activity was initiated by a letter issued by the NRC staff on.May 17, 1978 [2] to all power reactor licensees, requesting information concerning the control of heavy loads near spent fuel.
The results of Task A-36 were reported in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." The staff's conclusion from this evaluation was that existing measures to control the handling of h,eavy loac: at operating plants, although providing protection frcm certair. potential problems, do not adecuately cover the major causes of load-handling accidents and should 'e upgraded.
o In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a twc chase 1
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4 ebjective* using an accepted toproach or protection philosophy. The first portion of the objective, achieved through a set of general guidelines icentified in NUREG-0612, Article 5.1.1, is to ensure that all lead-handling systems at nuclear power plants are designed and operated such that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employec.
The second portion of the staff's object 1.ve, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ensure that, for load-handling systems in areas where their failure mignt result in significant consequences, either (a) features are proviced, in addition to'those required for all load-handling systems, to en,sure that the potential for a lond drnp is extremely small (e.g., a single-failure proof crane) or (b) conservative evaluations of load-handling accidents indicate that the potential consequences of any load drop are acceptably staall. Acceptability af accident consequances is quantified in NUREG-0612 into four accident analysis
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evaluation criteria.
The approach used to develop :he staff guidelines for minimizing the potential for a load drop was based on defense in depth and is summarized as follows:
o Provide sufficient operator training, handling-system design, lead-handling instructions, and equipment inspection to assure reliable operation of the handling system o
Define safe load travel paths through precedures and coerator training so that, to the extent practical, heavy loads are not carried over er near irradiated fuel or safe shutcown equipment
-o Provide mcchanical stops or electrical interlocks to prevent
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movement of heavy leads over irradiated fuel or in proximity to ecuipment associatad with recundant shutcown catns.
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Staffguidelinesresuitingfredtheforegoingaretabulatedin i
Section 5 of NUREG-0612.
1.3 Plan:-Soecific Backorcund On December 22, 1980, the NRC issued a letter [3] to the Public Service Company of Colorado (PSC), the Licensee for FSVNGS, requesting that the Licensee review provisions for handling and control of heavy loads at FSVNGS, evaluate these provisions with respect to the guicelines of NUREG-0612, and provide certain additional information
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to be used for an independent determination of conformance to these guidelines.
On September 16, 1981, PSC provided the initial responsa [4] to this request.
Compliance with Interim Protection Measures was discussed in a letter of September 10, 1982 [4b].
In response to a conference call on September 8, 1982, PSC provided a submittal [4c] dated December 29, 1982.
Some revisions to the December 29,1982, submittal were sent on January 14, 1983 [4d].
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EVALUATION AND RECOMMENCATIONS 2.1 Overvdew The following sections summarize PSC's review of heavy load handling at FSVNGS accompanied sy ECdG's evaluation, conclusions, and recommendations to the Licensee for bringing the facilities more ccmpletely into compliance with the intent of NUREG-0612. The Licensee has not istdicated the weight of a heavy load fo" this-facility (as defined in NUREG-0612, Article 1.2).
The submittal can be interpreted to say that the " Heavy Load" is 165.5 tons.
2.2 Heavy Lead Ove-head Handling Systems This section review's the Licensee's list of overhead handling systems which are subject to the criteria of NUREG-0612 and a review of the
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' ustification for excluding overhead handling systems from the above j
mentioned list.
2.2.1 Scone "Recort the results of your review of plant arrangemants to idantify all overhead handling systems from wnicn a ioac drop may result in damage to any system required for plant shutcown or decay heat remeval'(taking no credit for any interlocks, technical specifications, operating procedures, or detailed structural analysis) and justify the exclusion of any overhead handling system from your list by verifying that there is sufficient physical separation from any load-impact coint and any safety-related component to permit a determination by inspection that no' beavy load drop can result in damage to any system or component recuired for piant snutdown or decay heat removal."
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q A.
,5ummary of Licensee's 5:stements "The results of the revtew of the plan; t-angement nas ioentified two handling systems from vr.': a load drop may result in damage to a system required f:r :lant shutdown.
The two handling systems are the react:r :vilding crane and the turbine building crane."
"The crane in the turbine building can be excluded from the list of potentially hazardous cranes with espec-to load drops since it does not have the requirement nor the capability to carry a heavy load as defined in NUREG-0612.
Loads such as parts from a turbine overhaul that have considerable weight, but not classified as a heavy lead, would not be carried by the turbine building crane when the plant was operating."
B. ~EG&G Evaluation EG&G concurs with the licensee's statements that the reactor building crane is the only crane required to comply wi.th NUREG-0612.
C.
EGLG Conclusions and Recoerendations Based on the information provided, EG&G :: :iuces : hat the Licensee has included all toslicable he's:s and cranes in tneir list of handling systems which mus: ::= ply with the requirements of the general guidelines Of '.L' REG-0512.
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2.3 General Guidelines This section addresses the extent to which the applicable handling '
systems comply with the 3eneral' guidelines of NUREG-0512, Article 5.1.1.
EG&G's conclusions and recommendations are provided in summaries for each gu'ideline.
The NRC has established seven general guidelines which must be met in order to provide the defense-in-depth approach for the handling of heavy loads. These guicelines consist of the following criteria from Section 5.1.1 of NUREG-0612:
o' Guideline 1--Safe Lead Paths Guideline 2--Lead Handling Procedures o
Guideline 3--Crane' Operator Training o
Guideline 4--Special Lifting Devices o
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Guidelins 5--Lifting Devices (not specially designed) o Guideline 6--Cranes (Inspection, Testing, and Maintenance) o Guideline 7--Crane Design.
o These seven guidelines should be satisfied for all overhead handling t
j systems and programs in order to handle heavy loads in the vicinity of the reactor vessel, near scent fuel'in the spent-fuel pool, or in other areas where a load drop may camage safe shutdown systems.
The succesding paragraphs adoress the guidelines individually.
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[,,,., '.3.1 Safe I.oad Paths TGuideline 1. NUREG-0512. Article 5.1.1(ih 2
" Safe load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to ' '
impact irradtated fuel in the reactor vessel and in the spent-fuel pool, or to impact safe shutdown equipment. The path
_ should follow, to the extent practical, structural flocr memoers, beams, etc., such that if the load is dropped, the structure is mere likely to withs.tand the impact. These lead paths should be defined in procedures, shown on equipment layout drawings, and clearly marked on the floor in the trea whare the load is to be handled.
Deviations from defined load paths should require written alternative. procedures approved by the plant safety review committee."
A.
Sumary of Licensee's Statements "During plant operation, the area above the prestressed concrete reactor vessel (PCRV) is restricted from travel by the reactor building crane with a load. This is the only restriction on the travel of the reactor building crane. This restriction is in the fom of adninistratively controlled procedures. Since this critical area is restricte:1, PSC feels that there is no need for the nuarking of ofe load paths. "
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"The design of the refueling floor is such that there is no safety-related equipnient in the vicinity of the Reactor Building Crane. The design of the lifting devices, which j
connect the fuel-handling machine to the Reactor Building Crane, is such that there is.a large factor of safety (greater than 6) built into the design. Additionally', the lifting cable has a backup snubber system which, in the unlikely event of a cable break, would become engaged, thus preventing a heavy load drop."
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I "The crane snubber device limits the vertical travel to 14 inches, thereby holding the potential drop distance to a minimum. Because of these reasons, PSC is of the opinion that Guideline 1 has been satisfied and no further action is requi red."
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EG&G Evaluation The information provided in [4]. [4c], and [4d] adequately addresses the concerns of Guideline 1 of NUREG-0612, and is an acceptable deviation from the guideline. Therefore marking of safe load paths is not necessary.
C.
EG&G Conclusions and Recommendations Based. upon the information supplied in [4], [4c], and [4d],
EG&G Idaho considers F5VNGS to be censistent with the intent of Guideline 1 of NUREG 0612.
2.3.2 Lead-Handline procedures (Guideline 2, NUREG-0612. Article 5.1.1(211
" Procedures should be developed to cover load-handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment. At a minimum, procedures should cover haholing of those loads listed in Tacle 3.1-1 of NUREG-0612. These procedures should include:
icentification of required ecuipment; inspections and acceptance criteria required before movement of lead; the steps and proper sequence to be followed in handling the load; defining the safe path; and other special precautions."
A.
Summary of Licensee's 5 atements 1
" Personnel operating the reactor building crane are required by' written approved procedures not to allow any movement of the crane,over the PCRV at any time, except during refueling.
The crane operators are recuired to follow ?SC's Crane Ocerating Procedura Manual in which these procedures of crane operation P.re spelled out.
These crocecures are i
reviewed each refueling with the Fuel Handling People.
Acministrative controls, as defined in the Technical Specifications, Section 7.4.2, would be followed to. deviate from these procecures.
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IAs defined in -be-NUREG-0512, PSC Aas only one netvy leaa tha is handled by the reactor builcing crane.
The o 11 weigr.: Of the heavy icad is 155 5 tons.
This is ne weient ei the fuel handiing machine plus the weight cf a fuel c '. iaent.
The casignated lifting device is the rea: tor building crane.
The operation of the crane when engaging the fuel handling machine is governed by PSC's Fuel Handling Procedure and by PSC's Crane Operating Pro:edure.
These procedures contain the information required in NUREG-0512, Section 5.1.1(2)."
B.
EG&G Evaluation PSC has. generated procedures to comp s y control the operation of the reactor Building crane. The procedures are reviewed with the Fuel Handling people at each refueling.
Supervisory level personnel, or above, are required to approve procedural changes.
C.-
EG&G Conclusions and Recommendations Based on the infomation supplied in [4], [4c], and [4d],
EG&G concludes that FSVNGS is consistent with the intent of,
Guideline 2.
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i 2.3.2 Crane Ocerator Traininc IGuideline 3,.NL' REG-0612; Article 5.1~1(3)1 "C-ane Operators should be rained, quali~iec, and :endu:t themselves in ac:orcance with Chapter 2-3 of ANSI E30.2-1976,
' Overhead and Gantry Cranes' [5]."
A.
Summary of Licensee's Statements "There are no exceptions taken to ANSI'330.2-1976 with respect to operator training, qualification, anc concuct."
3.
EG&G Evaluation EG&G assumes that the statement in 2.3.3A constitutes a commit-ment to comply with Chapter 2-3 of ANSI B30.2-1976.
C..
EG&G Conclusion and Recommendations 4
FSVNGS is consistent with Guideline 3.
2.3.4 Scecial Liftine De,1ces TGuideline a. NUREG-0612, Article 5.1.1(4)1 "Soecial lifting devices should satisfy the guicelines of ANSI N14.6-1978, 'Stancard for Special Lifting Devices for Shi: ping Containers Weighing 10,000 counds'(4500 kg) or More for Nuclear Materiais' [6). This standard shcald apply to all special lifting devices which carry heavy leads in areas as cefined i
above.
For operating piants, certain inspec-ions and Icad tests
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may be ac:aptec in lieu of certain.aterial recuirements in -ne standard.
In addi-ion, the stress cesign fa :or stated in Section 3.2.1.'1 of ANSI N14.6 should be based on the ::mcined maximum static and dynamic leads that could be imparted on the handline device based en characteristics of the crane which will te used! This is in lieu of :n'e cuideline in Section 3.2.1.* of
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ANSI N14. 6 which bases tne stress design fac :t en only :ne weight (static load) or the load and of the intervening components of the special har.dling cevice."
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Su= mary of Licensee's Statements
" Guideline 4 recuires special lif-ing cevices c de in comoliance with a Modified Version of ANSI N14.6-1978,
' Standard for 5pecial Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials."
" ANSI N14.6-1978 as modified by NUREG-0612, deais with design., maintenance and inspection of special lifting devices used for handling heavy loads at nuclear facilities.
Fort St. Vrain Nuclear Generating Station (FSVNGS) uses such devices only for lifting and positioning of the fuel handling machine.
As specified in ANSI N14.6, these devices are, specifically, a pair of shackles which loop over -he crane lifting hook and are pin connected o a specially designed lifting ' mushroom'.
The designiof the mushroom is such that it afterds positive ma-ing to the upper head of the fuel handling machine.
A light and switch assembly indicates positive connection of the special lif ting devices to the fuel handling machine."
"The fuel handling machine's lifting device was designed and fr.bricated in the late 1960's prior to the issuance of ANSI N14.6-1978.
It is PSC's position that the mushroom and shackle assemblies will meet the ANSI-N14.6-1978 specification for any current operation, maintenance, and
- esting of the lif-ing cevice."
"The shackles and the mushroom-were analyzed and have factors of safety that arceed the requirements of ANSI N14.6-1978. "
" Snubbers limit any load lift to a maximum of 14 inches."
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EG&G EvaTuat' ion The response [4c], [4d] addresses compliance with the design and construction requirements of Guideline 4.
I ANSI N14.6-1978 is not intended to be applied specifically to the type of device which has been designed to handle the fuel handling machine. However, the general principles outlined in that standard have been applied to this device. :q Further, PSC states that the mushroom and tackle assemblies meet the ANSI N14.6-1978 specification for any current a
operation, maintenance, and testing of the lifting devic'e.
i C.
EG&G Conclusions anc Recemmencations FSVNGS is consistent with the intent of Guideline A.
2.3.5 Liftino Devices (Not Soecially Desiened) (Guideline 5.
NUREG-0612. Article 5.1.1(5)1
" Lifting cevices that are not speciaily designed should be installed and used in accordance with the guidelines of ANSI B30.g-1971, ' Slings' [7].
However, in selecting tne arcoer sling, the loac used should be the sum of the static and maximum dynamic load. The rating identified on the sling sh:uld :e in
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terms of the ' static load' wnich creduces the maximum stati: ar.c dynamic load. Where this restricts slings to use on only certain cranes, the slings should be clearly marxec as to :ne cranes with which't.'ey may be used."
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50mmary of License'e's 5:stements "Accorcing to Section 9-0.1 of ANSI E30.9, this i
soecification applies to. slings."
"The operation.of the Reactor Building Crane for lifts of
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heavy loads is limited to the fuel handling machine. Since this operation does not involve the use of slings, PSC is of the opinion that Guideline 5 is not specifically applicable to heavy load operations at Fort St. Vrain Nuclear Genera' ting Station."
B.
EG&G Evaluation PSC takes the position [4c) that slings are not used to lift
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heavy loads at FSVNGS.
Consecuently, Guideline 5 is not applicable.
C.
EG&G~ Conclusions and Recommendations EG&G Idahe concurs that Guideline 5 is not applicable at FSVNGS.
2.3.5 Cranes (Insoection. Testine and Maintenance) TGuideline 6.
NUREG-0612. Ar -tcie 5.1.I(631 "The crane should be inspected, insted, and maintained in
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accordance with~ Chap er 2-2 of ANSI 530.2-1976, ' Overhead and Gantry Cra.nes,' with th'e exceptien tha-tests and inspections should be performed pri~or to use where it is no: practicti o 0
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mee- -he frecuencies of ANSI 330.2 for periccic ins:e::'.:n anc est, er wnere frequenpy of crane use is, less -han ne s:e:ified inssee:1on anc tes-frequency (e.g., :ne polar crane insice a PWR c:n: tinmen: may only be usec every 12 : 18 men:ns :uring refueling coerations, and is generally not ac:essible curing power coeration.
ANSI 530.2, however, calis for cer ain inspections to be performed dai / or men:nly For su:h cranes having limited usage, the inspections, test, and main:entn=e sh:uld be performed prior : :neir use)."
A.
Summarv of Licensee's Statements
" ANSI B30.2-1976 has been invoked wi-h respect to crane inspection, testing, and maintenance.
These requirements are contained in PSC's Crane Operating Inspection and Maintenance Procedure. All reactor crane cperators at Fort St. Vrain are required to follow these pre:edures.
No exceptions are taken to this standard."
B.
EG&G Evaluation PSC has made an unreserved commitment to cemply with Chapter 2-2 of ANSI B30.2-1976, without the relief afforded by Guideline 6.
C.
EG&G Conclusions and Recommendations EG&G concludes that FSVNG5isconsistentwithGuideline6.
2.3.7 Crane Desien IGuideline 7. NUREG-0612. Article 5.1.1(7'f.
"The crane shouic be designed c mee; the a:plicacie criteria ar.d guiceiines of Chapter 2-1 of ANSI 530.2-1976, 'Overheac and Gantry Cranes,' and of CMAA-70, 'Soecifications for Electric Overhead Traveling Cranes' [8].
An alternative to a s:ecifica:icn in ANSI.530.2 or CMAA-70 may oe a:ce;;ed in iteu :f specific comoliance if the inten; of ne specift:sti:n is
.satisfiey."
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Sununary of' Licensee's Statements "The crane was originally specified and designed in 1967, to EOCI Spec. #61 In 1972, the crane was reanalyzed and upgraded to confom to CMAA 70 specifications."
B.
EG&G Evaluation The December 14, 1981, submittal (Phase II). Attachment 1, states that the crane was certified to be in compliance with CMAA-70 (1970) By the Whiting Corporation, the manufacturer.
Compliance with Chapter 2-1 of ANSI B30.2-1976 is addressed in[4c]and[4d].
EG&G accepts the statement of compliance with CMAA-70 and ANSI B30.2 'i976.
C.
EG&G Conclusions and Recemmendations FSVNGS is consistant with the intent of Guideline 7.
-2.4 Interim Protection Measures The NRC staff hds established (NUREG-0612, Article 5.3) that six measures should be initiated to provide reasonable assurance that handling of heavy loads will be performed in a safe manner until final implementation of the general guidelines of NUREG-0612, Article 5.1 is complete.
Four of these six interia measures censist of general
' Guideline 1, Safe Load Paths; Guideline 2, Load-Handling Procedures; Guideline 3, Crane Operator Training; and Guideline 6, Cranes (Inspection,-Testing, and Maintenance).
The two remaining interim measures cover the following criteria:
o Heavy load technical specifications o
Special review for heavy lea:s hanclec over :ne cere..
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. Licensee icplertntation and, evaluation of these interim protecticn l
measures are contained in the succeeding paragraphs of this section.
2.4.1 Interim Drotection Measure 1--Technical Soeci'ications
.- " Licenses for all operating reactors not having a single-failure proof everhead crane in the fuel ster:ge :::1 area should be revised to incluce a specification ccmparable t:
Standard Technical Specification S.9.7, ' Crane Travel--Spent Fuel Storage Pool Building,' for PWRs and Standard Technical Specification 3.9.6.2, ' Crane Travel,' for BWRs, to pr:nibit handling of heavy 1cacs over fuel in -he storage cool.until i=clementation of measures whien satisfy tne guidel'ines of Section 3.1."
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Summary of Licensee's Statements Not addressed by the LicenseeT',..,
B.
EG&G Eualuation Since Interim Protection Measure 1 is written to address fuel storage pool;, it is not directly applicable to FSVNG5.
C.
EG&G Conclusions and Recommendations EG&G concludes that Interim Pr~otection Measure 1 is not f
applicable to FSVNGS.
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2.4.2 Interim protection Measures 2. 3. A. and 5--Administrative Controls
" Procedural or administrative. measures [ including safe load paths, load-handling precedures, crtne operator training, and crane inspection]... can be accomplished in a short time period' and need not be delayed for completion of ' evaluations and modifications to satisfy the guideljnes of Section 5.1 of l
[NUREG-0612]."
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Summarv of Licensen's Statements Summaries of Licensee's statements are contained in discussions of the respective general guidelines in Sections 2.3.1, 2.3.2, 2.3.3, and 2.3.6, respectively.
PC3 also. addressed these measures in reference [4o].
3.
EG&G Evaluations, Conclusiens, and Recommendations FSVNGS is consistent with Interim Protection Measures 2, 3, 4, and 5.
'2.4.3 Interim Protection Measure 6--Soecial Review fer u,,yv te,es Over the Core "Special attention should be given to procedures, ecuipment, and personnel for the handling of heavy loads over the cert, suen as vessel internals for vessel inspection tools. This special review should include the following for these loads:
(a) review of procedures for installation of rigging or lifting devices and movement of the load to assure that sufficient detail is. provided and that instructions are clear and concise; (b) visual inspections of load bearing components of cranes, slings, and special lifting devices to identify flaws or deficiencies tnat could lead to failure of the component; (c) appropriate repair and replacement of defective components; and (d) verify that the crane operators have been properly trained and are familiar with i
specific procedures used in nandling these loads, e.g., hand signals, conduct of operations, and content of procedures."
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Summary of Licensee's Statements In Reference [4b], PSC avows compliance with this interim measure.
B.
EG&G Evaluation From a study of all submittals, EG&G Idaho is able to acquire sufficier.t.information to reach a conclusion.
C.
EG&G Cenc1'usion EG&G Idaho concludes that FSVNGS is censistant with Interim l
Protection Measure 6.
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CONGLUDING
SUMMARY
3.1 A::lica:1e Lead-Hancline Systems Based on the infor=ation supplied. EG&G concludes that the list of cranes and hoists previoed by tne Licensee as being subject to.ne
~ provisions of WUREG-0612 is adequate (see Section 2.2.1).
3.2 Guideline Recommendations Consistency with the seven NRC guidelines for heavy load handling.
(Section 2.3) is satisfied at F3VNG5. This conclusion is represented in tabular form as Table 3.1.
Guideline Recemmendation 1.
(Section 2.3.1) a.
FSVNGS is consistent with Guideline 1.
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(Section 2.3.2) a.
FSVNGS is consistent with Guioeline 2.
3.
(5ection 2.3.3) a.
FSVNGS is consistent with Guideline 3.
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(Section 2.3.4) a.
F5VNGS is consistent with Guideline 4.
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(Section 2.3.5) a.
Guideline 5 is not applicable to FSVNGS.
6.
(Section 2.3.6')
a.
F5VNG5 is consistent with Guicelino 6.
7.
(5ection 2.3.7) a.
F5v5G5 is consistenc wi-n Guideline 7.
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3.3 Interim Protection EG&G's evaluation of information provided by the Licensee indicates that the following actions are necessary to ensure that the six NRC staff measures for interim protection at FSVNGS are met:
Interim Measure Recommendation Interim Measure Not applicable to FSVNGS.
No. 1 Interim Measures FSVNGS complies with Interim Protection Measures 2, 2, 3, 4, 5, and 6 3, 4, 5, and 6.
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NUREG 0612, Control of Heavy Loads, PS2561.
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Unit 1, dated December 29, 1982.
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Public Service Comoany of Colorado, Letter to Robert A. Clark.
I Suoject:
NUREG 0612, Control of Heavy Loacs, Pi3012.
Fort St. Vrain,
,-Unit 1, dstec January la, 1983.
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ANSI N14.6-1978, " Standard for Lifting Cevices for Shipoing Centainers Weighing.10,000 Peu.-ds (4500 kg) or more for Nu: lear Materials."
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u.s. NuctaA2 CEGULATORY COMutSWON EGG-HS-6294 SISLl GRAPHIC DATA SHEET i TITLE AND SUSTITLE
- 2. ILesse eseel Control of Heavy Loads at Nuclear Power Plants Fort St. Vrain Nuclear Generating Station, Phase I
- s. RECIPIENT *$ ACCESSICW NO.
Docket No. 50-267 7 AUTHOR tS3 S. DATE REPORT COMPLETED MONTM j vgam T. H. Stickley November 1963 9 8ER80RMING ORGANsZATION NAME AND MAILING ACCRESS teacsvee I,a CsetA CATE REPORT ISSUED MONTw lvtA*
January 1984 EG&G Idaho, Inc.
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Idaho Falls, ID 83415 S.(Lasee areai
- 12. SPONSORING CRGANIZATION NAME AND MAILING AOCRESS tiaesvec los Csees Division of Systems Integration Office of Nuclear Reactor Reculatlen gi.,in go.
U.S. Nuclear Regulatory Comission Washington, DC 20555 A6457
- 13. TYPE OF REPORT pe mico cove neo traceus,ve seest
- 15. SUPPLEMENTARY NOTE 5
- 14. Ites,e areas
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- 16. ASSTR ACT 200 werer er desas The Nuclear Regulatory Comission (NRC) has requested that all nuclear plants either operating or under construction submit a response of compliancy with
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WJREG-0612, " Control of Heavy Loads at Nuclear Power Plants." EG&G Idaho, Inc.
has contracted with the NRC to evaluate the responses of those plants presently ut; der construction. This report contains EG&G's evaluation and recomendations for Fort St. Vrain Nuclear Generating Station (FSVNGS).
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- 17. KEY erQROS ANC DOCUMENT ANALY$l$
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