ML20076J271
| ML20076J271 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/31/1983 |
| From: | Jensen S, Stickley T EG&G, INC. |
| To: | NRC |
| Shared Package | |
| ML20076J273 | List: |
| References | |
| CON-FIN-A-6457, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8306240037 | |
| Download: ML20076J271 (21) | |
Text
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4 CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS LASALLE UNITS 1 AND 2 (PHASE II)
Docket No. 373/4 Author S. A. Jensen Principal Technical Investigator T. H. Stickley Published May 1983 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 Fin No. A6457 I3o 6 o 3 72
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CONTENTS ABSTRACT i
EXECUTIVE
SUMMARY
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INTRODUCTION....................................................
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1.1 Purpose of Review.........................................
I 1.2 Generic Background........................................
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1.3 Plant-Specific Background................................
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2.
EVALUATION AND RECOMMENDATIONS..................................
4 2.1 Overview..................................................
4 2.2 Heavy Load Overhead Handling Systems......................
4 2.3 Guidelines................................................
4 3.
CONCLUDING
SUMMARY
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3.1 Guideline Recommendations.................................
16 3.2 Additional Recommendations................................
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4.
REFERENCES.....................................................
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TABLE 2.1 Nonexempt Heavy Load Handling Systems...........................
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- ABSTRACT The Nuclear Regulatory 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. has contracted with the NRC to evaluate the responses of those plants presently under construction.
This report contains EG&G's evaluation and recommendations for La Salle for the requirements of Sections 5.1.4, 5.1.5 and 5.1.6 of NUREG-0612 (Phase II).
Section 5.1.1 (Phase I) was covered in a separate report [1].
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a EXECUTIVE
SUMMARY
La Salle does not totally comply with the guidelines of NUREG-0612.
In genera', compliance is in~ sufficient in the following areas:
o Information on the Reactor Building Crane was not sufficient to determine that the criteria of NUREG 0612 Section 5.1 is met, o
Information on cranes and hoists located over safe shutdown equipment was inadequate for determining full compliance with.
NUREG 0612 criteria.
The main report contains recommendations which will aid in bringing the above items into compliance with the appropriate guidelines.
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TECHNICAL EVALUATION REPORT ES.E LA SALLE NUCLEAR SiATION UNITS 1 AND 2 (PHASE II) 1.
INTRODUCTION 1.1 Purpose of Review This technical evaluation repor *. documents the EG&G Idaho, Inc. review of general load handling policy and procedures at La Salle Station.
This evaluation was performed with the objective of assessing conformance to the general load handling guidelines of NUREG-0612,
" Control of Heavy Loads at Nuclear Power Plants" [2), Sections 5.1.4, 3
5.1.5 and 5.1.6.
This constitutes Phase II of a two phase evaluation.
Phase I assesses confermance to Section 5.1.1 of NUREG-0612 and was documented in a separate report [1].
4 1.2 Generic Background Generic Technical Activity 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 operating 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 [3], to all power reactor applicant, 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 heavy loads at operating plants, although providing protection from i
certain potential problems, do not adequately cover the major causes of load handling accidents and should be upgraded.
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In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two phase objective using an accepted approach or protection philosophy.
The first portion of the objective, achieved through a set of general guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load 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 employed.
The second portion of the staff's objective, 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 might result in significant consequences, either (1) features are provided, in addition to those required for all load handling systems, to ensure that the potential for a load drop is extremely small (e.g., a single-failure proof crane) or (2) conservative evaluations of load handling accidents indicate that the potential consequences of any load drop are acceptably small.
Acceptability of accident consequences is quantified in NUREG-0612 into four accident analysis evaluation criteria.
' The approach used to develop the staff guidelines for minimizing the potential for a load drop was based on defense in depth and is summarized as follows:
I Provide sufficient operator training, handling system o
design, load handling instructions, and equipment inspection to assure reliable operation of the handling system o
Define safe load travel paths through procedures and operator training so that, to the extent practical, heavy loads are not carried over or near irradiated fuel or safe i
shutdown equipment Provide mechanical stops or electrical interlocks to prevent 3
o movement of heavy loads over irradiated fuel or in proximity to equipment associated with redundant shutdown paths.
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i Staff guidelines resulting from the foregoing are tabu;ated in Section 5 of NUREG-0612.
1.3 Plant-Soecific Backcrou~nd On December 22, 1980, the NRC issued a letter [4] to Commonwealth Edison, the applicant for la Salle Station requesting that the applicant review provisions for handling and control of heavy loads at La Salle, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guidelines.
Commonwealth Edison provided responses to this request on June 22, 1981 [6], September 22, 1981 [8], and October :9, 1982 [7].
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2.
EVALUATION AND RECOMMENDATIONS 2.1 Overview The following sections summarize Commonwealth Edison's review of heavy load handling at La Salle accompanied by EG&G's evaluation, conclusions, and recommendations to the applicant for bringing the facilities more completely into compliance with the intent of NUREG-0612.
2.2 Heavy Load Overhead Handlino Systems Table 2.1 presents the applicant's list of overhead handling systems which are subject to the criteria of NUREG-0612. The applicant has indicated that the weight of a heavy load for the facilities as 1800 lbs. per the NUREG-3612 definition.
2.3 Gu delines i
2.3.1 Reactor Buildinc [NUREG-0612, Article 5.1.4]
(1) The reactor building crane, and associated lifting devices used for handling the above heavy loads, should satisfy the single-failure proof guidelines of Section 5.1.6 of this report.
OR (2) The effects of heavy load drops in the reactor building should be analyzed to show that the evaluation criteria of Section 5.1 are satisfied. The loads analyzed should
' include:
shield plugs, drywell head, reactor vessel head; steam dryers and separators; refueling canal plugs and i
gates; shielded spent fuel shipping casks; vessel inspection platform; and any other heavy loads that ma be brought over or near safe shutdown equipment as well as fuel in the reactor vessel or the spent fuel pool.
Credit may be taken i
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l TABLE 2.1.
CRANE / HOISTS SYSTEMS CONSIDERED AS POTENTIAL SOURCES FOR DAMAGE OF SAFETY COMPONENTS Eauiement Number Eauipment Name OHC02G Reactor Building Crane OHC22G New Fuel Vault Jib Crane JB-3 1HC29G 12 and 18 Rail Hugger Hoist RH13 IHC70G 22 and 23 Rail Hugger Heist RH34 IHC30G 93 OHC30G 94 OHC31G 95 OHC33G 97 OHC34G 98 OHC38G 127 Trolley Hand Hoist TH04 1HC20G 128 Trolley Wire Rope Hoist TWR01 1HC21G 129 Trolley Wire Rope Hoist TWR02/TWR12 1HC23G 131 Trolley Wire Rope Hoist TWR04/TWR14 IHC24G 132 Trolley Wire Rope Hoist TWR05/TWR15 1HCB8G 168 Trolley Wire Rope Hoist TWR21 i
1HCB9G 169 Trolley Wire Rope Hoist TWR22 1HCB7G 179 and 184 Rail Hugger Hoist RH28 1HC08G 182 Rail Hu;9er Hoist RH05 IHC09G 183 Rail Hugger Hoist RH06 OHC52G 187 RH33 SG1 RH45 SG2 OHC44G DG1 OHC55G DG2 OHC39G PTS 5 OHC04G PTS 6 RH41 and RH42 PTS 10 i
'RH43 and RH44 PTS 11 1HC89G 193 Rail Hugger Hoist RH22 1HC903 194 OHC06GdB2 TWR23 JB6 1HCEIG JB8 Jib Crane No. 8 i
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in this analysis for operation of the Standby Gas Treatment System if facility technical specifications require its operation during periods when the load being analyzed would be handled.
The analysis should also conform to the guidelines of. Appendix A.
A.
Summary of Applicant Statements The applicant has identified two (2) cranes as being subject to the guidelines of this section.
They are the Reactor Building Crane and the New Fuel Vault Jib Crane.
(Jib
- 3).
The applicant's statements regarding the Reactor Building Crane are summarized as follows:
1.
Potential for load drops are small due to crane design features and testing.
2.
For compliance with the criteria of Section 5.1 of NUREG-0612, reliance is placed on the installation and use of electrical interlocks and mechanical stops, and on site-specific considerations.
The following statements were made with regard to the above mentioned considerations.
" Bridge, hoist, and trolley travel is regulated by limit switches.
Specific limit switches are used to control crane movements when handling the spent fuel shipping cask.
These limit switches prevent movement of the shipping cask over the Reactor Vessel or the Spent Fuel Pool."
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" Procedures or lesson plans have been written for the handling of heavy loads, operator training, and regularly scheduled crane inspections.
In addition, all lead movements are minimized, and follow the safest and shortest route with the load as close to the floor as practical."
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The applicant's statements regarding the New Fuel Vault Jib Crane are summarized as follows:
"New Fuel Vault Jib Crane (Jib #3), with electric Wire Rope hoist TWR08, has a rated capacity of 2000#.. Jib Crane #3 will be used only to transfer new fuel back and forth between the new fuel inspection stand and the new fuel vault.
Therefore, Jib Crane #3 will never be used to lift a load greater than 1800#.
Because Jib Crane #3 has the physical capacity to lift a heavy load, an administrative procedure is being written which will limit the Jib Crane rated capacity to 1800#.
The rated capacity of 1800# will be marked on the Jib Boom and a sig will be placed on the pedestal of the Jib Crane which will refer the operator to the Administrative Procedure which controls the use of the jib."
8.
EG&G Evaluation Insufficient information was provided by the applicant for EG&G to adequately evaluate conformance of the Reactor Building Crane to the requirements of NUREG 0612 Section 5.1.
The applicant states that dependence e meeting the criteria is placed on limit switches anc upon site-specific considerations.
However, no information is provided on the details of these considerations or on how the limit switches can or may be removed or bypassed.
Detailed information on proposed or present technical specifications was not given.
Administrative or physical controls are indicated but no detailed information was given on how they will meet the specific requirements of NUREG 0612 Section 5.1.
The applicant indicates that the new Fuel Va' ult Jib Crane will be made to conform to the criteria of NUREG 0612 Section 5.1 by derating the crane to a load capacity less than that which is considered to be a heavy load.
This complies with the intent of the requirements in EG&G's judgement.
However, the proposed administrative procedure should be available for review by the NRC.
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(c) The effects of load drops have been analyzed and the results indicate that damage to safe shutdown equipment would not preclude operation of sufficient equipment to achieve safe shutdown. Analyses should conform to the guidelines of Appendix A, as applicable.
(2) Where the safe shutdown equipment has a ceiling separating it from an overhead handling system, an alternative to Section 5.1.5(1) above would be to show by analysis that the largest postulated load handled by the handling system would not penetrate the ceiling or cause spalling that could cause failure of the safe shutdown equipment.
A.
Summary of Applicant Statements Crane Systems 12 and 18:
System 12 is used to lift heat recovery coils located in the Unit 1 Turbine Exhaust Filter Room.
System 18 is ; sed to lift Unit 11 heat recovery coils.
The load paths of the two systems are two floors above control panels located in the Control Room.
If a load drop from either system occurs, and damage to Control Room panels result, operation of the systems necessary for safe shutdown can be continued and controlled from the remote shutdown panels located in the Auxiliary Electric Equipment Room (AEER).
Crane System 22 and 23:
System 22 is used to service Unit 1 main steam isolation valves located outside of the primary containment, and System 23 is used for the same purpose in Unit 2.
If a load drop occurs from either crane system and a main steam line is damaged, the release of steam can be stopped by closing the appropriate redundant MSIV located inside the primary containment.
Crane System 93, 169, and 187:
Systems 93, 169, and 187 are used to service RHR Service Water Pumps and Fuel Pool Emergency Make-Up Pumps.
System 93 is used to lift loads in the vicinity of Unit 1 RHR Service Water Pumps A, B, C, and D and Fuel Pool Emergency Make-Up Pumps A and 3.
RHR Service Water Pumps A and B, and Fuel Pool Emergency Make-Up Pump A are located in a room adjacent to, but separated 9
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from, another room containing RHR Service Water Pumps C and D and Fuel Pool Emergency Make-Up Pump B.
If a lead drop did occur in one of the rooms, damaging any of the pumps, a separate, redundant pump located in the adjacent
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room would be put into service to perform the system function.
Systems 169 and 187 are used to service Unit 11 RHR Service Water Pumps and Fuel Pool Emergency Make-Up Pumps.
System 169 is used to lift loads in the vicinity of RHR Service Water Pumps C and D and Fuel Pool Emergency Make-Up Pump B.
System 187 is used to lift loads in the vicinity of RHR Service Water Pumps A and B and Fuel Pool Emergency Make-Up Pump A.
These pumps are located in a room separate from the pumps serviced by System 169.
If a load drop occurred in either room, a separate, redundant pump, located in the other room, would be put into service to perform the system function.
Crane Systems 94 and 128:
System 94 is used to service the Unit 1 HPCS Pump and System 128 services the HPCS Pump for Unit 11. The lead paths of these cranes are directly above the HPCS Pumps.
If a load drop occurred, and the HPCS Pumps were inoperable, safe shutdown would be achieved by initiating the RCIC System for depressurization, reactor water make-up and initial decay heat removal.
Crane Systems 95, 97, 129, and 131:
Crane Systems 95 and 97 are used to service Unit 1 RHR Pumps B and A, respectively.
Crane Systems 129 and 131 are used to service Unit 11 RHR Pumps B and A respectively.
The load paths of the cranes are directly above the related RHR Pumps.
If a load drop did occur and one of the RHR Pumps was damaged, the separate, redundant RHR Pump would be put into service i
to perform the system function.
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Crane Systems 98 and 132:
Crane System 98 is used to service the Unit 1 LPCS Pump and System 132 is used to service the Unit 11 LPCS Pump.
The crane load paths are located directly above the LPCS Pumps.
If a load drop does occur, and the LPCS Pumps become inoperable, the function; of the LPCS System will be completed by the continued use of the HPCS System.
Crane Systems 127 and 168:
System 127 is used for servicing the Unit 1 HPCS Diesel Cooling Water Pump and Strainer.
System 168 is used for the same purpose of Unit 11.
If a load drop did occur and these components were damaged, resulting in the loss of the HPCS System, the RCIC System would be initiated to perform the required functions for safe shutdown.
Crane Systems 179 and 184:
Crane Systems 179 and 184 are located directly above the RCIC Turbine Driven Pumps for Units 1 and 11.
These cranes will be used only when the RCIC Systems are out-of-service. However, if the RCIC Systems were damaged by a load drop, the HPCS Systems would be used to perform the functions necessary for safe shutdown.
Crane Systems 182 and 183:
Crane Systems 182 and 183 will be used to move the Motor Driven Reactor Feed Pump Motors for Units 1 and 11.
If a load drop did occur, damage may result to the main steam lines below the load paths of the cranes.
If a main steam line did break, the flow of steam could be stopped by closing the appropriate separate and redundant MSIV.
Crane Systems SGI and SG2:
Systems SGI and SG2 will be lifting loads directly above Units 1 and 11 RHR "A" Heat Exchangers.
If a load drop occurs the heat exchangers may l
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be damaged, causing RHR Loop "A" to be inoperable.
If this occurs, RHR Loop "B" will be used to perform the functions necessary for the safe shutdown of the reactor.
Crane Systems DG1 and DG2:
Crane Systems DG1 and DG2 will be used to service the Turbine Driven Reactor Feed Pumps for Units 1 and 2.
If a load drop did occur, damage to the HPCS Diesel Control Panels, located two floors below may result.
If, as a result of a load drep, the HPCS System is inoperable the RCIC System will be used to perform the functions required for the safe shutdown of the reactor.
Crane Systems 193 and 194:
Crane Systems 193 and 194 are used to lift flowmeters installed in the cain steam lines for Units 1 and 2.
If a load drop did occur from one of these systems, a main steam line could be damaged.
If a break in a main steam line occurs, the release of steam could be stopped by closing the appropriate MSIV.
Cranes JB2, JB6 and JB8: Jib Cranes 2, 6, and 8 are used to lift various equipment through the equipment hatches located above the equipment access way.
The items are being lifted above the Unit 1 RCIC Pump, and damage to the pump may result if a load drop occurs.
If the RCIC System is damaged, the Unit 1 HPCS System will be used to perform the functions necessary for the safe shutdown of the reactor.
Patented Track Crane Systems PTSS and PTS 6:
Crane Systems PTSS and PTS 6 are used to lift Turbine Building Exhaust Fans for Units I and 2.
The load paths of these cranes are two floors above control panels located in the Control Room.
If a load drop from either system occurs, and damage to Control Room Panels result, operation of the systems necessary for safe shutdown can be continued and controlled from the remote shutdown panels located in the Auxiliary Electric Equipment Room.
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Patented Track Crane Systems PTS 10 and PTS 11:
Crane Systems PTS 10 and PTS 11 are used to service the Motor Driven Reactor Feed Pumps for Units 1 and 2.
The load paths of these systems'will permit loads to be carried above the HPCS Diesel Control Panels.
If, as a result of a load drop, the HPCS System is inoperable, the RCIC System will be used to perform the functions required for the safe shutdown of the reactor.
B.
EG&G Evaluation All the above mentioned handling systems are eliminated from further analysis by the applicant on the basis of separation and redundancy of safety-related equipment.
However, the information provided was insufficient in many cases.
For instance a load drop by crane System 22 and 23 could result in a steam release but no indication as to the safety related consequences of such a release were included in the applicant's statements. Other crane systems have a similar problem.
In other cases crane systems operate near or over several safety-related systems but no information is provided to show that a load orop damaging one system will not damage other systems.
Statements about separate rooms are made but the possibility of a load drop damaging more than one piece of equipment in the same room is not precluded. Another problem that is not addressed is the possibility of safety related equipment being damaged by a load drop related to the servicing of redundant equipment.
Administrative or physical controls to prevent such occurrences are not addressed in the applicant's response.
No indication is made that an analysis conforming to the guidelines of Appendix A of NUREG 0612 has been made concerning the effects of load drops.
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C.
EG&G Conclusions and Recommendations EG&G concludes that the applicant has not provided enough information to adequately evaluated conformance to the requirement of this section.
We recommend that the applicant provide a more thorough evaluation of the effects of load drops on the involved safety-related systems.
The information should be detailed enough to determine whether the criteria of NUREG 0612 Section 5.1 is met.
2.3.3 Single-Failure-Proof Handling Systems [NUREG-0612, Article 5.1.6]
(1) Lifting Devices:
(a) Special lifting devices that are used for heavy loads in the a'ea where the crane is to be upgraded should meet ANSI N14.61978, " Standard For Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More For Nuclear Materials," as specified in Section 5.1.1(4) of this report except that the handling device should also comply with Section 6 of ANSI N14.5-1978.
If only a single lifting device is provided instead of dual devices, the special lifting device should have twice the design safety factor as required to satisfy the guidelines of
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Section 5.1.1(4). However, loads that have been evaluated and shown to satisfy the evaluation criteria of Section 5.1 need not have lifting devices that also comply with Section 6 of ANSI N14.6.
(b) Lifting devices that are not specially designed and that are used for handling heavy loads in the area where the crane is to be upgraded should meet ANSI B30.9 - 1971, " Slings" as specified in Section 5.1.1(5) of this report, except that one of the following should also be satisfied unless the effects
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of a drcp of the particular load have been analyzed and l
shown to satisfy the evaluation criteria of Section 5.1:
(i) Provide dual or redundant slings or lifting devices such that a single component failure or malfunction in the sling will not result in uncontrolled lowering of the load; i
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9B (ii) In 3 electing the proper sling, the load used should be twice what is called for in meeting Section 5.1.}(5) of this report.
(2) New cranes should be designed to meet NUREG-0554,
" Single-Failure-Proof Cranes for Nuclear Power Plants." For operating plants or plants under construction, the crane should be upgraded in accordance with the implementation guidelines of Appendix C of this report.
(3)
Interfacino lift points such as lifting lugs or cask trunions should also meet one of the following for heavy loads handled in the area where the crane is to be upgraded unless the effects of a drop of the particular load have been evaluated and shown to satisfy the evaluation criteria of Section 5.1:
(a) Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the load; lift points should have a design safety factor with respect to ultimate strength of five (5) times the maximum combined concurrent static and dynamic load after taking the single lift point failure.
93 (b) A non-redundant or non-dual lift point system should have a design safety factor of ten (10) times the maximum combined concurrent static and dynamic load.
A.
Summary of Applicant Statements The applicant has stated that they have no single-failure proof handling systems.
B.
EG&G Evaluation No evaluation is necessary.
C.
ES&G Conclusions and Recommendations No conclusions or recommendations are necessary.
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3.
CONCLUDING
SUMMARY
3.1 Guideline Recommendations The NRC staff has established guidelines for judging the safety implications for handling heavy loads in the area of the reactor vessel, near stored spent fuel, or in other areas where an accidental load drop could damage safe shutdown systems.
These guidelines are established to insure that potential for load drops are extremely small or that potential consequences of load drops are acceptably small.
3.2 Additional Recommendations Crane Category Recommendations 1.
Reactor Building Cranes a.
The applicant should NUREG-0612 Article 5.1.4 provide additional information a.
Reactor Building Crane showing that chances of a serious load drop are small or that there will be minimal consequences.
b.
New Fuel Vault Jio Crane b.
This crane is judged by EG&G to satisfy NUREG-0612 criteria based on applicant information.
2.
Cranes Over Safe Shutdown The applicant should provide Equipment additional information showing NUREG-0612 Article 5.1.5 that the guidelines of NUREG-0612 are met.
3.
Single-Failure-Proof The applicant has not Handling Systems designated any systems as NUREG.0612 Artic1s 5.1.6 single-failure proof.
No recommendations.
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4.
REFERENCES 1.
Centrol of Heavy Loads at Nuclear Power Plants LaSalle Nuclear Station Ur.its 1 and 2 (Phase I), April 1983.
2.
NUREG-0612 Centrol of Heavy Loads at Nuclear Power Plants NRC 3.
V. Stello, Jr. (NRC)
Letter to all applicants.
Subject:
Request for Additional Information on Control of Heavy Loads Near Spent Fuel NRC, 17 May 1978 4.
USNRC Letter to Commonwealth Edison.
Subject:
NRC Request for Additional Information on Control of Heavy Loads Near Spent Fuel NRC, 22 December 1980 5.
J. S. Abel (CE), Letter to D. G. Eisenhut (NRC).
Subject:
Control of Heavy Loads, May 15, 1981.
6.
E. D. Swartz (CE), Letter to D. G. Eisenhut (NRC).
Subject:
Response
to NUREG-0612, June 22,1981.
7.
E. D. Swartz (CE), Letter to D. G. Eisenhut (NRC).
Subject:
Supplemental Response to NUREG-0612, October 19, 1982.
8.
E. D. Swartz (CE), Letter to D. G. Eisenhut (NRC).
Subject:
Response
to NUREG-0612, September 22, 1981.
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