ML19321A749
| ML19321A749 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/26/1980 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| REF-GTECI-A-36, REF-GTECI-SP, TASK-A-36, TASK-OR NUDOCS 8007240141 | |
| Download: ML19321A749 (47) | |
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UNITED STATES E
NUCLEAR REGULATORY COMMISSION o
g e.
WASHINGTON, D. C. 20655 k
June 26, 1980 Docket No. 50-289 Mr. R. C. Arnold Senior Vice President Metropolitan Edison Company 100 Interpace Parkway Parsippany, New Jersey 07054
Dear Mr. Arnold:
In January 1978, the NRC published NUREG-0410 entitled, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants -
Report to Congress". As part of this program, the Task Action Plan for Unresolved Safety Issue Task No. A-36
" Control of Heavy Loads Near j _
Spent Fuel," was issued.
We have completed our review of load handling operations at nuclear power plants. A report describing the results of this review will be issued in the near future as NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants - Resolution of TAP A-36."
This report contains several recommendations to be implemented by all licensees to assure the safe handli.ng of heavy loads.
At the Indian Point Units 2 and 3, Zion Units 1 and 2, and Three Mile Island Unit 1 facilities, we are requesting licensee action to begin to implement these reconnendations at this time on the schedule 4
indicated in this letter.
4 To expedite your compliance with this request, we have enclosed the following:
1.
Guidelines _ for Control of Heavy Loads (Enclosure 1).
3-2.
Staff Position - Interim Actions for Control of Heavy Loads (Enclosure 2).
3.
Request for Additional Information on Control of Heavy Loads (Enclosure 3).
You are. requested to review your controls for the handling of heavy loads to determine the extent to which the guidelines of Enclosure 1 are presently satisfied at your facility, and to identify the required changes and modifications in order to fully satisfy these guidelines.
You are requested to implement. the-interim actions described in Enclosure 2'as soon as possible but no later than 90 days from the date of this
' letter.
800.7240lh
Mr. R. C. Arnold June 26, 1980 You are further requested to submit a report documenting the results of your review and the re-- ' red changes and modifications. This report should include the information identified in Sections 2.1 through 2.4 of Enclosure 3, on how tia guidelines of NUREG-0612 will be satisfied. 'this report should be submitted not later than the following schedule.
Submit the Section 2.1 infonnation within three months from the date of this letter.
Submit the Sections 2.2, 2.3, and 2.4 inf;ormation within six months.
You should commence implementation of required changes and modifications as soon as possible without waiting on staff review, with the objective of completing all procedural and documentation changes, beyond the above interim actions, within two years of submittal of Section 2.4 for the above report.
Please notify your assigned NRC Project Manager if you will not be able to maintain these schedules.
- incerely, M
sen L e
Division of icensing
Enclosures:
As stated cc w/ enclosures:
See next 3 pages
Metropolitan Edison Company Dr. Walter H. J3rdan-881 W. Outsr Drive Oak' Ridge, Tennessee 37830 I
ccw/ enclosure (s).
Mr. Marvin I. Lewis Dr. Linda W. Little 6504 Bradford Terrace 5000 Herinitage Drive P, hila del phia,
Pennsylvania 19149 Raleigh, North Carolina 27612
' Walter W. Cohen, Cons.:mer Advocate Holly S. Eeck
, Departmer
- of Justice Anti-Nuclear Group Representing Strawber., Square,14th Floor York l Harrisburg, Pennsylvania York, Pennsylvania 17404 j
17127 245 W. Philadelphia Street
' Robart L. Knupp, Esq.
Assistant Solicitor John Levin, Esq.
Knupp and Andrews
. Pennsylvania Public Utilities Com.
P.O. Box P Box 3265 l 407 N. Front Street Harrisburg, Pennsylvania 17120 Harrisburg, Pennsylvania 17108 Jordan D. Cunningham, Esq.
John E. Minnich, Chairman Fox, Farr and Cunningham Dauphin Co. Board of Commissioners 2320 North 2nd Street Dauphin County Courthouse Harrisburg, Pennsylvania 17110 Front and Market Sts.
Harrisburg, Pennsylvania.17101 Theodore A. Adler, Esq.
WID0FF REAGER SELK0WITZ & ADLER Atomic Safety and Licensing Appeal Board o
Post Office Box 1547 i
U.S. Nuclear Regulatory Commission Harrisburg Penns.ylvania 17105 Washington, D. C.
20555 Ms. Marjorie M. Aamodt 1
Atomic Safety and Licensing Board Panel R.D. f5 U.S. Nuclear Regulatory Commission Coatesville, Pennsylvania 19320 Washington, D. C.
20555 Ms. Karen Sheldon Docketing and Service Section Sheldon, Harmon & Weiss o
U.S. Nuclear Regulatory Commission 1725 I Street, N.W.
Suite 506 Washington, D. C.
20555 Washington, D. C.
20006 Robert Q. Pollard Earl B. Hoffman 609 Montpelier Street Dauphin County Comissioner Baltimore, Maryland 21?l8 Dauphin County Courthouse Front and Market Streets Harrisb.Jrg, Pennsylvania 17101 Chauncey Kepford Judith H. Johnsrud Ms. Ellen R. Weiss, Esq.
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Environmental Coalition on Nuclear Power Sheldon, Harmon & Weiss 1725 I Street, N.W.
433 Orlando Avenue Suite 506 State College, Pennsylvania 16801 Washington, D. C.
20006 Ms. Frieda Berryhill, Chairman Mr. Steven C. Sholly Coalition for Nuclear Power Plant 304 South tiarket Street
. Postponement Mechanicsburg, Pennsylvania 17055 2610 Grendon Drive Wilmington, Delaware 19838 Mr. Thomas Gerusky Mrs. Rhoda D. Carr Bureau of Radiation Protection 1402 Marene Drive Department of Environmental Resources P.O. Box 2063 Harrisburg, Pennsylvania 17109 Harrisburg, Pennsylvania 17120
Metropolitan Edison Company Karin W, Carter, Esq.
505 Executive House Mr. G. P. Miller-P. O. Box 2357 Mr. R. F. Wilson Harrisburg, Pennsylvania 17120 Mr. J. J. Barton Metropol' tan Edison Company Honorable Mark Cohen 512 D-3 Main Capital Building P. O. Box 480 Harrisburg, Pennsylvania 17,120 Middletown, Pennsylvania 17057 G. F. Trowbridge', Esquire Dauphin County Office Emergency Shaw, Pittman, Potts & Trowbridge Preparedness 1800 M Street, N.W.
Court House, Room 7 Washington, D. C.
20036 Front & Market Streets Harrisburg, Pennsylvania 17101 Mr. E. G. Wallace
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Licensing & nager Department of Environmental Resources t on ntepcePr[e ATTN:
Director, Office of Radiological Parsippany, New Jersey 07054 Post Offi e Box 2063 Harrisburg, Pennsylvania 17105 Pennsylvania Electric Company Mr. R. W. Conrad Director, Technical Assessment Vice President, Generation Division 1001 Broad Street Office of Radiation Programs Johnstown, Pennsylvania 15907 (AW-459)
U. S. Environmental Protection Agency Miss Mary V. Southard, Chairman Crystal Mall #2 Citizens for a Safe Environment Arlington, Virginia 20460 Post Office Box 405 Harrisburg, Pennsylvania 17108 Mr. Robert B. Borsum Babcock & Wilcox Government Publications Section Nuclear Power Generation Division i
State Library of Pennsylvania Suite 420, 7735 Old Georgetown Road Box 1601 (Education Building).
Bethesda, Maryland 20014 Harrisburg, Pennsylvania 17126
- Ivan W. Smith, Esq.
Mr. David D. Maxwell, Chairman Atomic Safety & Licensing Board Panel Board of Supervisors U.S. Nuclear Regulatory Commission Londonderry Township Washington, D. C.
20555 RFD#1 - Geyers Church Road Middletown, Pennsylvania - 17057 Ms. Kathy McCaughin Three Mile Island Alert, Inc.
U. S. Environmental Protection Agency 23. South 21st Street Region III Office Harrisburg, Pennsylvania 17104
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ATTN:
EIS COORDINATOR Curtis Building (Sixth Floor)
Mr. L. W. Harding 6th and Walnut Streets Supervisor of Licensing Philadelphia, Pennsylvania 19106 Metropolitan Edison Company P. O. Box 480 Metropolitan Edison Company Middletown, Pennsylvania 17057 ATTN:
J. G. Herbein, Vice President P. O. Box 542 Reading, Pennsylvania 19603 Ms. Jane Lee R.D. 3; Box 3521 Etters, Pennsylvan,ia 17319 a
Metropolitan Edis"n Company -
Mr. R. J. Toole Manager, THI-1 Metropolitan Edison Company Governor's Office of State Planning P. O. dox 480 and Development Middletown, PA 17057 ATTN: Coordinator, Pennsylvania State Clearinghouse Mr. W. E. Potts P. O. Box 1323 Radiological Controls Manager, TMI-1 Harrisburg, Pennsylvania 17120 Metropolitan Edison Company P. O. Box 480 Middletown, PA -17057 Allen R. Carter, Chainnan Joint Legislative Comittee on Energy Mr. I. R. Finfrock, Jr.
Post Office Box 142 Jersey Central Power & Light Company Suite 513 Madison Avenue at Punch Bowl Road Senate Gressette Building Morristown, New Jersey 07950 Columbia, South Carolina 29202 J. B. Lieberman, Esq.
Berlock, Israel, Lieberman 26 Broadway New York, NY 10004 Mr. J. J. Colitz Plant Engineering Manager, TMI-1 Metropolitan Edison Company P. O. Box 480 Middletown, Pennsylvania 17057 York College of Pennsylvania Country Club Road York, Pennsylvania 17405 Mr. G, K. Hovey Director, TMI-2 Metropolitan Edison Company P, 0, Box 480 Middletown, PA 17057 Mr. B. Elam Manager, Plant Engineering, Unit 2 Metropolitan Edison Company P. O. Box 480 Middletown, PA 17057 Mr. Jtichard Roberts; The Patriot 812 Market Street Harrisburg, PA.17105 Mr. R. W. Heward Manager, Radiological Control, Unit 2 Metropolitan Edison Company P. O. Box 480 Middletown, Pennsylvania 17057
Enclosure '(1) t 5.
GUIDELINES FOR CONTROL OF HEAVY LOADS Our evaluation of the information provided by licensees indicates that existing measures at operating plants to control the handling of heavy loads cover certain of the potential problem areas, but do not adequately cover the major causes of load handling accidents.
These major causes include operator errors, rigging failures, lack of adequate inspection and inadequate procedures.
The measures in effect vary from plant to plant, with some having detailed procedures while others do not, some have performed analyses of certain postulated load drops, certain plants have single-failure proof cranes, some PWR's have rapid containment isolation on high radiation, and many plants have technical specifi-cations,that prohibit handling of heavy loads or a spent fuel cask over the spent fuel pool.
To provide adequate measures that minimize the occurrence of the principal causes of load handling accidents and to provide an adequate level of defense-in-depth for handling of heavy loads near spent fuel and safe shutdown systems, the measures in effect should be upgraded.
5.1 Recommended Guidelines The following sections describe various alternative approaches which provide acceptable measures for the control of heavy loads.
The objectives of these guidelines are to assure that either (1) the potential for a load drop is extremely small, or (2) for each area addressed, the following evaluation criteria are satisfied:
i I.
Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than 1/4 of Part 100 limits);
'i II.
Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy load does not result in a configuration of the fuel such that k,ff is larger than 0.95; III. Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accidental dropping of a postulated heavy load is limited so as not to result in water leakage that could uncover the fuel, (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated); and IV.. Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions.
After reviewing the historical data available on crane operations, identifying the principal causes of~ load drops, and considering the type and frequency of
. load handling operations at nuclear power plants, the NRC staff has developed an overall philosophy that provides a defense-in-depth approach for controlling the handling of heavy loads.
This philosophy encompasses an intent to prevent
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as well as mitigate the consequences of postulated accidental l'oad drops.
The following summarizes this defense-in-depth approach:
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(1) Provide sufficient operator training, handling system design, load handling instructions, and equipment inspection to assure reliable operation of l
the handling system; and (2) Define safe load travel paths through procedures and operator training so that to the extent practical heavy loads avoid being carried over or near irradiated fuel or safe shutdown equipment; and 1
(3)
Provide mechanical stops or electrical interlocks to prevent novement of heavy loads over irradiated fuel or in proximity to equipment associated 1
with redundant shutdown paths.
Certain alternative measures may be'taken/to compensate for deficiencies in i
(2) and (3) above, such as the inability to prevent a particular heavy load from being brought over spent fuel (e.g., reactor vessel head).
These alterna-tive measures can include:
increasing crane reliability by providing dual load paths for certain components, increased safety factors, and increased inspection as discussed in Section 5.1.6 of this report; restricting crane operations in the spent fuel pool area (PWRs) until fuel has decayed so that off-site releases would be sufficiently low if fuel were damaged; or analyzing the effects of postulated load drops to show that consequences are within acceptable limits.
Even if one of these alternative measures is selected, (1) and (2) above should still be satisfied to provide maximum practical defense-in-depth.
The following sections provide guidelines on how the above defense-in-depth approach may be satisfied for various plant areas.
Fault trees and associated probabilities were developed and used as described in Bases for Guidelines, Section 5.2 of this report, to evaluate the adequacy of these guidelines and to assure a consistent level of protection for the various areas.
5.1.1 General All plants have overhead handling systems that are used to handle heavy loads in the area Of the reactor vessel or spent fuel in the spent fuel pool.
Additionally, loads may be handled in other areas where their accidental drop may damage safe shutdown systems.
Accordingly, all plants should satisfy each of the following for handling heavy loads that could be brought in proximity to or over safe shutdown equipment or irradiated fuel in the spent fuel pool area and in containment (PWRs),- in the reactor building (BWRs), and in other plant areas.
(1)
Safe load paths should be defined for the covement of heavy loads to minimize the potential for heavy leads, if dropped, to impact irradia*.ed fuel in-the reactor vessel and in the pent fuel pool, or to impact safe shutdown equipmenc.
The path shoulo f6ilsw, to the extent practical, structural floor members, beams, etc., such that if the load is dropped, the structure is more likely to withstand the impact.
These load paths should be defined in procedures, shown on' equipment layout drawings, and clearly marked on the floor in the area where the-load is to be handled.
Deviations from defined load paths should require written alternative procedures approved by the plant safety review committee.
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i Procedures should be developed to cover load t,andling operations for (2) 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 These handling of those loads listed in Table 3-1 of this report.
identification of. required equipment; procedures should include:
inspections and acceptance criteria required before movement of load; the j
steps and proper sequence to be followed in handling the load; defining
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the safe load path; and other special precautions.
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Crane operators should be trained, qualified and conduct themselves in i
(3) accordance with Chapter 2-3 of ANSI B30.2-1976, " Overhead and Gantry Cranes."
Special liftina devices should satisfy the guidelines of ANSI N14.6-1978, 4
l (4)
" Standard for Special Lifting Devices for Shipping Containers Weighing This standard 10,000 pounds (4500 kg) or More for Nuclear Materials."
should apply to all special lifting devices which carry heavy loads in For operating plants certain inspections and areas as defined above.
load tests may be accepted in lieu of certain material requirements in In addition, the stress design factor stated in the standard.
Soction 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device i
j This is in based on characteristics of the crane which will be used
- lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor _ on only the weight (static load) of the load and of
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the intervening components of the special handling device.
Lifting devices that are not specially desianed should be installed and (5) used-in accordance with the guidelines of ANSI 830.9-1971, " Slings."
However, in selecting the proper sling, the load used should be the sus l
of the static and maximum dynamic load." The rating identified on the sling should be in terms of the " static load" which produces the maximum j
static and dynamic load. Where this restricts slings to use on on?y certain cranes, the slings should be clearly marked as to the cranes with which they may be used.
The crane should be inspected, tested, and maintained in accordance with (6)
Chapter 2-2 of ANSI B30.2-1976, " Overhead and Gantry Cranes," with the exception that tests and inspections should be performed prior to use where it-is not practical to meet the frequencies of ANSI B30.2 for periodic inspection and test, or where frequency of crane use is less than the specified inspection and test frequency (e.g., the polar crane inside a PWR containment may only be used every 12 to 18 months during refueling operations, and is generally not accessible during power I
ANSI B30.2, however, calls for certain inspections to be operation.
For such cranes having limited usage, the performed daily or monthly.
i' inspections, tests, and maintenance should be performed prior to their
,use.)
u For the purpose of selecting the proper sling, loads imposed by the SSE need a
not be included.in the dynamic loads imposed on the sling or lifting device.
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(7) The crane should be designed to meet the applicable criteria and guide-lines of Chapter 2-1 of ANSI 830.2-1976, " Overhead and Gantry Cranes" and of CMAA-70, " Specifications for Electric Overhead Travelling Cranes." An alternative to a specification in ANSI 830.2 or CMAA-70 may be accepted t
in lieu of specific compliance if the intent of the specification is satisfied.
5.1.2 Spent Fuel Pool Area - PWR I
Many PWR's require that the spent fuel shipping cask be placed in the spent fuel pool for loading. Additionally, other heavy loads may be carried over or near the spent fuel pool using the overhead crane, including plant equipment.
rad-waste shipping casks, the damaged fuel container and replacement fuel storage racks.
Additionally, certain crane failures could cause the crane lower load block to be dropped, and therefore this should also be considered as a heavy load.
The fuel handling crane is used for moving fuel and is generally not used for handling of heavy loads.
To provide assurance that the evaluation criteria of Section 5.1 are met for load handling operations in the spent fuel pool area, in addition to satisfying the general guidelines of Section 5.1.1, one of the following should be satisfied:
i (1) The overhead crane and associated lifting devices used for handling heavy loads in the spent fuel pool area should satisfy the single-failure proof guidelines of Section 5.1.6 of this report.
OR (2) Each of the following is provTded:
(a) Mechanical stops or electrical interlocks should be provided that prevent movement of the overhead crane load block over or within 15 feet horizontal (4.5 meters) of the spent fuel pool.
These mechanical stops or electrical interlocks should not be bypassed when the pool contains " hot" spent fuel, and should not be bypassed without approval from the shift supervisor (or other designated plant management personnel).
The mechanical stops and electrical f.9terlocks should be verified to be in place and operational prior to placing " hot" spent fuel in the pool.
(b) The mechanical stops or electrical interlocks of 5.1.2(2)(a) above should also not be bypassed unless an analysis has demonstrated that damage due to postulated load drops would not result in criticality or cause leakage that could uncover the fuel.
(c) To preclude rolling if dropped, the cask should not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor level of the refueling building or other components and structures along the path of travel.
(d) Mechanical stops or electrical interlocks should be provided to preclude crane travel from areas there a postulated load drop could damage equipment from redundant or alternate safe shutdown. paths.
(e) Analyses should conform to the guidelines of Appendix A.
OR (3). Each of the following are provIded (Note:
This alternative is simlar to (a) above, except it allows movement of a heavy load, such as a cask, into the pool while it contains " hot" spent fuel if the pool is larga enough to maintain wide separation between the load and the " hot" spent fuel.):
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h (a) " Hot" spent fuel should be concentrated in one location in the spent fuel pool that is separated as much as possible from load paths.
(b) Mechanical stops or electrical interlocks should be provided to prevent. movement of the overhead crane load block over or within 25 feet horizontal (7.5 m) of the " hot" spent fuel.
To the extent practical, loads should be moved over load paths that avoid the spent fuel pool and kept at least 25 feet (7.5 m) from the " hot" spent fuel unless necessary.
When it is necessary to bring loads within 25 feet of the restricted region, these mechanical stops or electrical interlocks should not be bypassed unless the spent fuel has decayed sufficiently as shown in Table 2.1-1 and 2.1-2, or unless the total inventory of gap activity for fuel within the protected area would result in offsite doses less than % of 10 CFR Part 100 if released, and such bypassing should 'equire the approval from the shift supervisor (or other designated plant management individual).
The mechanical stops er electrical interlocks should be verified to be in place and operational prior to placing " hot" spent fuel in the pool.
(c) Mechanical stops or electrical interlocks should be provided to restrict crane travel from areas where a postulated load drop could damage equipment from redundant or alternate safe shutdown paths.
Analyses have demonstrated that a postulated load drop in any location not restricted by electrical interlocks or mechanical stops would not cause damage that could result in criticality, cause leakage that could uncover the fuel, or cause loss of safe shutdown
. equipment.
(d) To preclude rolling, if dropped, the cask should not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor level of the refueling building or other components and structures along the path of travel.
(e) Analyses should conform to the guidelines of Appendix A.
OR (4) The effects of drops of heavy Ioads should be analyzed and shown to satisfy the evaluation criteria of Section 5.1 of this report.
These analys=3 should conform to the guidelines of Appendix A.
5.1.3 Containment Building - PWR PWR containment buildings contain a polar crane that is used for removing and reinstalling shield plugs, the reactor vessel head, upper vessel internals, and on occasion, other heavy equipment such as the reactor coolant pump, the reactor vessel inspection platform, and the cask used for damaged fuel.
Additionally the crane load block may be moved over fuel in the reactor when handling smaller loads or no load at all.
Due to the weight of the load block alone, this should also be considered as a heavy load.
To provide assurance that the criteria of Section 5.1 are met for load handling operations in the containment building, in addition to satisfying the general guidelines of Section 5.1.1, one of the following should be satisfied:
(1) The crane and associated lifting devices used for handling heavy loads in the containment building should satisfy the single-failure proof guidelines of Section 5.1.6 of this report.
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1 (2) Rapid containment isolation is provided with prompt automatic actuation on high radiation so that postulated releases are within limits of evaluation Criterion I of Section 5.1 taking into account delay times in detection and actuation; and analyses have been performed to show that evaluation i
criteria II, III, and IV of Section 5.1 are satisfied for postulated load drops in this area.
These analyses should conform to the guidelines of Appendix A.
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(3) The effects of drops.of heavy Ioads should be analyzed and shown to i
satisfy the evaluation criteria of Section 5.1.
Loads analyzed should include the following:
reactor vessel head; upper vessel internals; vessel inspection platform; cask for damaged fuel; irradiated sample cask; reactor coolant pump; crane load block; and any other heavy loads brought over or near the reactor vessel or other equipment required for continued decay heat removal and maintaining shutdown.
In this analysis, l
credit may be taken for containment isolation if such is provided; however analyses should establish adequate detection and isolation time.
Addi-tionally, the analysis should conform to the guidelines of Appendix A.
5.1.4 Reactor Building - BWR The reactor building in BWRs typically contains the reactor vessel and spent fuel pool, as well as various safety-related equipment.
The reactor building overhead crane may be used in many day-to-day operations such as moving various shielded shipping casks or handling plant equipment related to maintenance or modification activities.
The crane is also used during refueling operations for removal and reinstallation of shield plugs, drywell head, reactor vessel head, steam dryers and separators, and refueling canal plugs and gates.
The crane would also be used subsequent to refueling for handling of the spent fuel shipping cask. This cask may be lifted as high as 100 feet (30 m) above the grade elevation at which the cask is brought into the reactor building.
Additionally the overhead crane's load block may be moved over fuel in the reactor or over the spent fuel pool when handling smaller load or no load at all.
Due to the weight of the load block alone, this should also be considered as a heavy load.
To assure that the evaluation criteria of Section 5.1 are satisfied one of the following should be met in addition to satisfying the general guidelines of Section 5.1.1:
(1) _The reactor building crane, and associated liftin'g devices used for
. handling the above heavy loads, should satisfy the single-failure proof guidelines of Section 5.1.6 of this report.
OR j
(2) The effects of heavy load drops in the reactor building should be analyzed
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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 gates; shielded spent fuel shipping casks; vessel inspection platform; and any other heavy loads that may be brought over or near safe shutdown a
equipment as well as fuel in the reactor vessel or the spent fuel pool.
Credit may be taken in this analysis for operation of the Standby Gas 5-6
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6 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.
5.1.5 Other Areas In other plant areas, loads may be handled which, if dropped in a certain location, may damage safe shutdown equipment.
Although this is not a concern i
at all plants, loads that may damage safe shutdown equipment at some plants include the spent fuel shipping cask, turbine generator parts in the turbine building, and plant equipment such as pumps, motors, valves, heat exchangers, and switchgear.
Some of these loads may be less than the weight of a fuel assembly with its handling tool, but may be sufficient to damage safe shutdown equipment.
(1) If safe shutdown equipment are beneath or directly adjacent to a potential travel load path of overhead handling systems, (i.e., a path not restricted by limits of crane travel or by mechanical stops or electrical interlocks) one of the following should be satisfied in addition to satisfying the general guidelines of Section 5.1.1:
(a) The crane and associated lifting devices should conform to the single-failure proof guidelines of Section 5.1.6 of this report; OR (b)
If the load drop could impair the operation of equipment or cabling associated with redundant or dual safe shutdown paths, mechanical stcps or electrical interlocks should be provided to prevent movement of loads in proximity to these redundant or dual safe shutdown equipment (In this case credit should not be taken for intervening floors unless justified by analysis).
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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 cenform 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.
5.1.6 Single-Failure-Proof Handling Systems i
For certain areas, to meet the guidelines of Sections 5.1.2, 5.1.3, 5.1.4, or 5.1.5, the alternative of upgrading the crane and lifting devices may be l
chosen.
The purpose of the upgrading is to improve the reliability of the handling system through increased factors of safety and through redundancy or l
duality in certain active components.
NUREG-0554, " Single-Failure-Proof l
Cranes for Nuclear Power Plants," provides guidance for design, fabrication, installation,.and testing of new cranes that are of a high reliability design.
For operating plants, Appendix C to this report, " Modification of Existing Cranes," provides guidelines on implementation of NUREG-0554 for operating plants and plants under construction.
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Section 5.1.1 of this report provides certain guidance on slings and special handling devices.
Where the alternative is chosen of upgrading the handling system to be " single-failure proof", then steps beyond the general guidelines of Section 5.1.1 should be taken.
Therefore, the following additional guidelines should be met where the alterna-tive of upgrading handling system reliability is chosen:
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(1) Lifting Devices:
(a) Special lifting devices that are used for heavy loads in the area where the crane is to be upgraded should meet ANSI N14.6 1978,
" 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 N14.6-1978.
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 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 of a drop of the carticular load have been analyzed and shown to satisfy the evaluat1on 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; OR (ii) In selecting the proper sling, the load used should be twice what is called for in meeting Section 5.1.1(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 wit 5 the imple-mentation guidelines of Appendix C of this report.
(3)
Interfacing 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.
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(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.
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ENCLOSURE 2 STAFF POSITION -
INTERIM ACTIONS FOR CONTROL OF HEAVY LOADS e
(1) Safe loa 6 paths should be defined per the guidelines of Section 5.1.l(1) (See Enclosure 1);
(2)
Procedures should be developed and implemented per the guidelines of Section 5.1.1(2) (See Enclosure 1);
(3) Crane operators should be trained, qualified and conduct themselves per the guidelines of Section 5.1.1(3) (See Enclosure 1);
(d) Cranes should be inspected, tested, and maintained in accordance with the guidelines of Section 5.1.l(6) (See Enclosure 1); and (5)
In addition to the above, special attention should be given to procedures, equipment, and personnel for the handling of heavy loads over the core, such as vessel internals or Vessel inspection tools.
This special review should include the following for these loads:
(1) 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 ore clear and concise; (2) visual inspections of load bearing con.ponents of cranes, slings, and special lifting devices to identify flaws or deficiencies that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g., hand signals, conduct of operations, and content of procedures.
h
Enclosure (3)
REQUEST FOR ADDITIONAL INFORMATION ON CONTROL OF HEAVY LOADS 1.
INTRODUCTION Verification by the licensee that the risk associated with load-handling failures at nuclear power plants is extremely low will require a systematic evalua-tion of all load-handling systems at each site.
The following specific information requests hsve been organized to support such a systematic approach, and provide a basis for the staff's review of the licensee's evaluation. Additionally, they have been organized to address separately the two hazards requiring investigation (i.e.,
radiological consequences of damage to fuel and unavailability consequences of damage to certain systems).
The following general information is provided to assist in this evaluation and reduce the need for clarification as to the intent and expect-ed results of this inquiry.
1.
Risk reduction can be demonstrated by either of two approaches:
a.
The possibility of failure is extremely low due to handling-system design features (NUREG 0612, Section 5.1.6).
b.
The consequences of a failure can be shown to be acceptable (NUREG 0612, Section 5.1, Criteria I-IV).
Regardless of the approach selected, the general guidelines of NUREG 0612, Section 5.1.1, should be satisfied to provide maximum practical defense-in-depth.
2.
Evaluations concerning radiological consequences or criticality safety, where used, can rely on either the adoption of generic analyses reported in NUREG 0612, requiring only verfication that these generic assumptions are valid for a specific site, or employ a site-specific analysis.
3.
Systems required for safe shutdown and continued decay heat removal are site-specific and are not, therefore, identified in this request.
Individual plants should consider systems and components identified in Regulatory Guide 1.29, Position C.1 (except those systems or portions of systems that are required for (a) emergency core cooling, (b) post-accident containment heat removal, or (c) post-accident containment atmosphere cleanup), for evaluation and recognize that the approach taken in this respect is similar to that identified in Regulatory Guide 1.29, Position C.2.
The fact that a load-handling system may be prevented from operating during plant conditions re-quiring the actual or potential use of some of these systems, is re-I i
1
- c gnized in this rs:pset for information.
4 The scope of this systematic review should include all heavy loads carried in areas where the potential for non-compliance with the acceptance criteria (NUREG 0612, Section 5.1) exists. A saamary of typical loads to be considered has been provided in Attachment 6.
It is recog-nized that some cranes will carry additional miscellaneous loads, some of which are not identifiable in detail in advance.
In such cases an evaluation or analysis demon-strating the acceptability of the handling of a range of
-loads should be provided.
S.
At some sites' loads which must be evaluated will include licensed shipping casks provided for the transportation of irradiated fuel, solidified radioactive waste, spent resins, or other byproduct material. Licensing under 10CFR71 is not evidence that lifting devices for these shipping casks meet the criteria specified'in NUREG 0612, Sections 5.1.1(4), 5.1.
1(5), 5.1.6(1), or 5.1.6(3), as appropriate, and thus does not eliminate the need to provide appropriate information concerning these devices. A tabulation (Attachment 7) is provided to indicate multiple-site use of these shipping casks.
The results of the licensee's evaluation, as reported in response to this
~
request, should provide information sufficient for the staff to conduct an in-deper. dent review to determine that the intent of this effore (i.e., the unifbrm reduction of the potential hazard from load-handling-system failures) has been satisfied.
2.
INFORMATION REQUESTED FROM THE LICENSEE 2.1 GENERAL REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS NUREG 0612, Section 5.1.1, identifies several general guidelines related to the design'and operation of overhead load-handling systems in the areas where spent fuel is stored,in the vicinity of the reactor care, and in other areas of
[
the plant where a load drop could result in damage to equipment required for safe shutdown or decay heat removnl.
Information provided in response to this section
, should identify the extent of potentially hazardous load-handling operations at a site, the extent of conformance to appropriate load-handling guidance, and the changes required in order to conform to the guidance.
1.
' Report the results of your review of plant arrangements to identify all overhead handling systems from which a load drop may result in damage to any system required for plant shutdown or decay heat removal (taking no credit for any.
interlocks, technical specifications, operating procedures, or detailed structural analysis).
2.
Justify the exclusion of any overhead handling system from the above category by_ verifying that there is sufficient physical separation from any load-tmpact point and any safety-related component to permit a determination by inspec-tion that no heavy load drop can result in dasage to any system or component required for plant shutdown or core decay heat removal.
3.
With respect to the design and operation of heavy-loss 3andling systems in the containment and the spent-fuel-pool area ind those load-handling. systems identified in 2.1-1, above, provide your evaluation concerning compliance with the guidelines of NUREG 0612, Section 5.1.1.
The following specific information should be included in your reply:
a.
Drawings or sketches sufficient to -clearly identify the location of safe load paths, spent fuel, and safety-related equipment.
b.
A discussion of measures taken to ensure ti.i load-handling operations remain within safe load paths, including procedures, if any, for deviation i
from these pathe.
.c.
A tabulation of heavy loads to be handled by each crane which includes the load identification, load i
weight, its designated lifting device, and verifi-cation that the handling of such load is governed by a written procedure containing, as a minimum, the information identified in NUREG 0612, Section 5.1.1(2).
d.
Verification that lif ting devices identified in 2.1.
3-c, above, s comply with the requirements of ANSI 14 I
6-1978, or ANSI B30.9-1971 as appropriate.
For lifting devices where these standards, as supplemented by NUREG 0612, Section 5.1.1(4) or 5.1.1(5), are not, met, describe any proposed alternatives and demon-strate their equivalency in terms of load-handling reliability.
e.
Verification that ANSI B30.2-1976, Chapter 2-2, has been invoked with respect to crane inspection, testing, and maintenance. Where any exception is taken to this standard, sufficient information should be provided to demonstrate the equivalency of proposed alternatives.
f.
Verification that. crane design complies with the guide-lines of CMAA Specification-.70 and Chapter 2-1 of ANSI B30.2-1976, including the demonstration of equivalency of actual design requirements for instances where spe-
- cific compliance with these standards is not provided.
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4 g.
Exceptions,1f any, taken to ANSI B30.2-1976 with 4
f respect to operator training, qualification, and conduct.
2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE FCOLS 2
4 NUREG 0612, Section'5.1.2, provides guidelines concerning the design and operation of load-handling systems in the vicinity of stored, spent fuel.
Information provided in response to 35's section should demonstract that adt-quate measures have been taken to ensure that in this area, either the likeli-hood of a -load drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria of NUREG 0612. Section 5.1, Criteria I through III.
1.
Identify by name, type, capacity, and equipment designator, any cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carry-ing loads which could, if dropped, land or fall into the spent fuel pool.
2.
Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying heavy loads or are permanently prevented from move-ment of the hook centerline closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load can fall into the fuel-storage pool.
3.
Identify any cranes listed in 2.2-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and-the basis for this evaluation (i.e., complete compliance with NUREG 0612, Section 5.1.6 or partial com-pliance supplemented by suitable alternative or additional design.fectures).
For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) informa-tion specified in Attachment 1.
4.
For cranes identified in 2.2-1, above,-not categorized accord-ing to 2.2-3, demonstrate that the criteria of NUREG 0612,
.Section 5.1, are satisfied. Compliance with Criterion IV i
will be demonstrated in response to Section 2.4 of this request. Wita respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance.
This response should include the following information for each crane:
a.
Which alternatives (e.g., 2, 3, or 4) from those identified in NUREG 0612, Section 5.1.2, have been selected.
o b.
If Alternative 2 or 3 is selected, discuss the crane motion limitation imposed by electrical interlocks or mechanical stops and indicate the circumstances, if any,'under which these protective devices may be bypassed or removed. Discuss any administrative procedures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed technical specification i
(operational and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops.
4
.Where reliance is placed on crane operational c.
8 limitations with respect to the time of the storage of certain quantities of spent fuel at i
specific post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls
.provided to ensure that these assumptions remain valid.
d.
Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed techni-
)
i cal specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
Analyses performed to demonstrate compliance with I
e.
l Criteria I through III should conform to the guide-lines of Attachment 5.
Justify any exception taken to these guidelines, and provide the specific infor-mation requested in Attachment 2, 3, or 4, as appro-priate, for each analysis performed.
2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG 0612,- Section 5.1.3, provides guidelines concerning the design and
. operacion of load-handling systems in the vicinity of the reactor core.
Infor-mation provided in response to this section should be sufficent to demonstrate that adequate measures have been taken to ensure that in this area, either the likelihood of a load drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria 'of NUREG 0612, Section 5.1, Criteria I through III.
1.
Identify by name, type, capacity, and equipment _ designator, i-any cranes physically capable (i.e., taking no credit for ny interlocks or operating procedures) of carrying heavy a
loads over the reactor vessel.
I.
5
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.2. _ Justify the exclusion of any cranes in this area from the above category-by verifying that they are incapable of carrying heavy loads, or are permanently prevented from the movement of any load either directly over the reactor
-vessel or to'such a location where in the event of any load-handling-system failure, the load may land in or on the reactor vessel.
3.
Identify any cranes listed.in 2.3-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for_all loads i~
to be carried.and the basis for this evaluation (i.e., com-plete compliance with NUREG 0612, Section 5.1.6, or partial
-compliance supplemented by suitable alternative or additional design features).
For each crane so evaluated, provide the
- load-handling-system (i.e., crane-load-combination) informa-tion specified in Attachment 1.
4.
For cranes identified in 2.3-1, above, not categorized accord-
.ing to.2.3-3, demonstrate that the evaluation criteria.of l
NUREG 0612^,-Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in your response tc Sec-tion 2.4 of this request. With respect to Criteria I through III,' provide a discussion of your evaluation of crane opera-tion in the containment and your determination of compliance.
This response should include the following information for l
each crane:
4 a.
Wh<re reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate j
the circumstances under which these protective devices cc.n be removed or bypassed and the adminstrative pro-cedures invoked to ensure proper authorization of such action.
Discuss any related or proposed technical specification concerning the bypassing of such interlocks.
b.
Where reliance is placed on other, site-specific con-siderations (e.g., refueling sequencing), provide present or proposed technical specifications and dis-cuss administrative or physical controls provided to ensure the continued validity of such considerations.
Analyses performed to demonstrate compliance with c.
Criteria I through III should conform with the guide-lines.of Attachment 5.
Justify any exception taken to these guidelines, and provide the specific infor-mation requested in Attachment 2, 3, or 4, as appro-priate, for each analysis performed.
2.4 SPECIFIC REQUIREMENTS FOR '0VERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT: REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT
_ REMOVAL. LOR SPENT FUEL POOL COOLING
-NURIG'0612, Section 5.1.5, provides guicelines concerning the design and
. operation of load-handling systems in the vicinity of equipment or components I L L
o required for safe reactor shutdown and decay heat removal.
Information pro-vided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small, or that damage to such equip-ment from load drops will be limited in order not to result in the loss of these safety-related functions.
Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1, and their loads in F
your response to 2.1-3-c.
'l.
Identify any cranes listed in 2.1-1, above, which you have evaluated as having sufficient design features to make the likelihood oi a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete l
compliance with NUREG 0612, Section 5.1.6, or partial com-liance supplemented by suitable alternative or additional design features).
For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) informa-tion specified in Attachment 1.
2.
For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, a comprehensive hazard evaluation should be provided which includes the following information:
The presentation in a matrix format of all heavy a.
loads and potential impact areas where damage might occur to safety-related equipment. Heavy loads identf fication should include designation and weight or cross-reference to information pro-j i
vided in 2.1-3-c.
Impact areas should be identi-fied by construction zones and elevations or by some other method such that the impact area can be located on the plant general arrangement drawings.
Figure 1 provides a typical matrix.
b.
For each interaction identified, indicate which of the load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, mechanical stops and/or electrical interlocks, or other site-specific considerations.
Elimination on the basis
.of the aforementioned considerations should be supplemented by the following specific information:
i (1)
For load / target combinations eliminated because of separation and redundancy of safety-related equipment, discuss the basis for determining that load drops will not l
affect continued system operation (i.e.,
the ability of the system to perform its safety-related function).
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t (2) Where mechanical stops or electrical inter-locks are to be provided, present details showing the areas where crane travel will be prohibited.
Additionally, provide a discus-sion concerning the procedures that are to be used for authorizing the bypassing of i
interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability af ter cperations which require bypassing have been completed.
(3) Where load / target combinations are eliminated on the basis of other, site-specific consi-derations (e.g., maintenance sequencing), pro-vide present and/or proposed technical speci-fications and discuss administrative procedures or physical constraints invoked to ensure the continued validity of such considerations.
c.
For interactions not eliminated by the analysis of 2.4-2-b, above, identify any handling systems for specific loads which you have evaluated as having sufficient design fea-tures to make the likelihcod of a load drop extremely small and the basis for this evaluation (i.e., ccuplete compliance with NUREG 0612, Section 5.1.6, or partial compliance sup-plemented by suitable alternative or additional design fea-tures).
For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1.
d.
For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate using appropriate analysis that damage would not preclude operation of sufficient equipment to allow the system to perform its safety function following a load drop (NUREG 0612. Section 5.1, Criterion IV?. For each analysis so conducted, the following information should be provided.
(1) An indication of whether or not, for Ebe specific load being investigated, the over-head crane-handling systen is designed and constructed such that the hoisting system will retain its load in the event of' seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).
l (2)
Th: basis for any exceptions taken to the analytical guidelines of Attachment 5.
(3)
The information requested in Attachment 4.
FireRE I Typical Load / Impact Area Matrix CRAME (IDENTIFT THE CRAME BT MAME AMD EQUll'HDrr MtHsER) tsCATluse IMDICATE THE tutt.01MC(5) CORRE5r0MDING TUTilE IMPACT AREA (5) FJLAMPI28. DEACTUR RUILDINC. AUXILIARY SU INFACT AREA (IDENTIFT AREA BT CONSTRUCTION ZOMES)
Esseptes Column Line F-5 Column Line R9-R12 thADS EI.EVATION
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^
5
-R m
M m o R IMINAfi m IX)UlFHEMT CATEconY ELEVATION EQUIPHmT CATF40af (Indicate the vertowe elevatione) toute 1 Mate 2 (Neavy Emed identift-cotton should include Emmaples Elev. 4)$'
deriAnettoe and weight)
Feempl_e Spent Fuel Ceek pt.I 30/24 (100 tone) 0 0
B 9.
9 NOTES TO FIGURE 1 i-
- Note 1:
Indicate by symbols the safety-related equipment.
The licensee should provide a list consistent with the clarification provided in 1,2-3.
a Note 2: Hazard Elimination Categories Crane travel for this area / load combination prohibited a.
by electrical interlocks or mechanical stops.
b.
System redundancy and separation precludes loss of capability of systen to perform its safety-related function following this load drop in this area.
Site-specific considerations eliminate the need to con-c.
sider load / equipment combination.
d.
Likelihood of handling system failure for this load is extremely small (i.e. section 5.1.6 NUREO 0612 satis-fled).
Analysis demonstrates that crane failure and load drop e.
will not damage safety-related equipment.
T 1
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Attachment (1)
SINGLE-FAILURE-PROOF HANDLING SYSTEMS 1.
Provide the name of the manufacturer and the design-rated load (DRL).
If the maximum critical load (MCL), as defined in EUREG 0554, is not the same as the DRL, provide this capacity, 2.
Provide a detailed evaluation of the overhead handling system with respect to the features of design, fabrication, inspection, tssting, and operation as delineated in NUREG 0554 and supplemented by the identified alternatives specified in NUREG 0612, Appendix C.
This evaluation must include a point-by-point comparison for each section of NUREC 0554.
If the alternatives of NUREG 0612, Appendix C, are used for certain applications in lieu of complying with the recommendation of NURIG 0554, this should be explicitly stated.
If an alternative to any of those contained in NUREG 0554 or NUREG 0612, Appendix C, is proposed details must be provided on the proposed alternative to demonstrate its equivalency.
3.
With respect to the' seismic analysis employed to demonstrate that the over-head handling system can retain the load during a seismic event equal to a safe shutdown earthquake, provide a description of the method of analys,is, the assumptions used, and the mathematical model evaluated in the analysis.
The description of assumptions should include the basis for selection of trolley and load position.
4.
Provide an evaluation of the lif ting devices for each single-failure-proof handling system with respect to the guidelines of NUREG 0612, Section 5.1.6.
5.
Provide an evaluation of the interfacing lift points with respect to the guidelines of NUREG 0612, Section 5.1.6.
9
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' 4 l
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i Attachment (2)
ANALYSIS OF RADIOLOGICAL RELEASES I
The following information should be provided for an analysis conducted to demonstrate compliance with Criterion I of NUREG 0612, Section 5.1.
l.
INITIAL CONDITIONS / ASSUMPTIONS I
a.
Identify the time after shutdown, the number of fuel assemblies damaged, and the assumed duration of radio-logical release associated with each accident analyzed.
b.
NUREG 0612, Table 2.1-2, provides the assumptions used to arrive at generic conclusions concerning radiological dose consequences.
To rely on the radiological dose analysis of NUREG 0612, the licensee should verify that these assumptions are conservative with regard to the plant / site evaluated.
If the assumptions are not con-servative for the specific plant, or if a more site-specific analysis is required, the licensee should i
identify plant-specific assumptions used in place of i,
those tabulated.
c.
Identify and provide the basis (e.g., USNRC Regulatory Guide 1.25) for any assumptions employed in site-specific analyses not identified in NUREG 0612, Table 2.1-2.
d.
Dose calculations based on the termination or mitigation of radiological releases should be supported by informa-tion sufficient to demonstrate both that the time delay assumed is conservative and that the system provided to accomplish such termination or mitigation will perform its safety function upon demand (i.e., the system meets the criteria for an Engineered Safety Feature).
Specific information so provided should include the following:
(1)
Details concerning the location of accident 1
j.
sensors, parameters monitored and the values of these parameters at which a safety signal will be initiated, system response time (including valve-operation time), and the total time required to automatically shift from normal operation to isolation or filtra-tion following an accident.
.(2)' A description of the instrumentation and con-l; trols associated with the Engineered Safety l~
Feature which includes information sufficient to-demonstrate that the requirements (Section 4) of IEEE 279-1971, " Criteria for Protection Systems'for Nuclear Power Generating Stations,"
are satisfied.
2-1 i
(3)
A description of any Engineered Safety Feature filter system which includes infor-mation sufficient to demonstrate compliance with the guidelines of USNRC Regulatory Guide 1.52, " Design,3 Testing, and Maintenance Criteria for Engineered Safety Feature Atmos-phere Cleanup System' Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power ?lants.'."
(4) A discussion of any inicial conditions (e.g., manual valves locked shut, containment airlocks or equipment: hatches shut) necessary to ensure that releases will be terminated or mitigated upon Engineered Safety Feature actuation and the measures employed (i.e.
Tech-nical Specification and administrative controls) to ensure that these initial conditions are satisfied and that Engineered Safety Feature systems are operable prior to the load lift.
2.
METHOD OF ANALYSIS Discuss the method of analysis _used to demonstrate that post-accident dose will be well within 10CFR100 limits.
In presenting methodology used in determining the radiological consequences, the following information should be provided.
A description of the mathematical or physical model a.
employed.
b.
An identification and summary of any computer program used in this analysis.
The considerations of uncertainties in calculational c.
methods, equipment performance, instrumentation response characteristics, or other indeterminate effects taken into account in the evaluation of the results.
3.
CONCLUSION Provide an evaluation comparing the results of the analysis to Criterion I of NUREG 0612, Section 5.1.
If the postulated heavy-load-drop accident analyzed bounds other postulated heavy-load drops, a list of' these bounded heavy loads should be provided.
i l
2-2 l
Attachment (3)
CRITICALITY ANALYSIS The following information should be proviqed for analyses conducted to demon-strate compliance with Criterion II of NUkEG 0612, Section 5.1 4
1.
INITIAL CONDITIOUS/ ASSUMPTIONS The conclusions of UUREG 0612, Section Z'.2, are based on a particular model fuel assembly.
If a licensee us'es the results of Section 2.2 rather than performing an independent neutronics analysis, the assump-tions should be verified to be compatible with plant-specific design.
Forany analysis conducted, the following assumptions should be provided as a minimum:
a.
Water /UO2
- 1"'" #*Ei b.
The boron concentration for the refueling water and spent-fuel pool
, s The amount of neutron poison in the fuel c.
d.
Fuel enrichment The reactivity insertion value due to crushing of e.
the core f.
The k,ff value allowed by technical specifications for tne core during refueling 2.
METHOD OF ANALYSIS Provide the method of analysis used to demonstrate that accidental dropping of a heavy load does not result in a configuration of the fuel such that k,gf is larger than 0.95.
The discussion of the method of analysis should include the following information:
Identification of the computer codes employed
{
a.
b.
A discussion of allowances or conpensation for calculation and physical uncertainties 3.
CONCLUSION Provide an evaluation comparing the results of the analysis to Criterion II of NUREG 0612. Section 5.1.
If the postulated heavy-load-drop accident 3-1
bounds other postulated heavy-load drops, a list of these bounded heavy loads should be provided.
I 9
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3-2
Attachment (4)
ANALYSIS OF PLANT STRUCTURES The following information should be provided for analyses conducted to demon-stratecompliancewithCriteriaIIIandIV/of'NUREG0612,Section5.1.
i 1.
INITIAL CONDITIONS / ASSUMPTIONS Discuss the assumptions used in the analysis, including:
i a.
Weight of heavy load b.
Impact area of load c.
Drop height d.
Drop location e.
Assumptions regarding credit taken in the analysis for the action of impact limiters f.
Thickness of walls or floor slabs impacted g.
Assumptions regarding drag forces caused by the environment h.
Load combinations considered 1.
Material properties of steel and concrete 2.
METHOD OF ANALYSIS Provide the method of analysis used to demonstrate that sufficient load-carrying capability exists within the wall (s) or floor slab (s).
Identify any computer codes employed, and provide a description of their capabilities.
If test data was employed, provide it and describe its applicability.
3.
CONCLUSION Provide an evaluation comparing the results of this analysis with Criteria III and IV of NUREG 0612, Section 5.1.
Whr :e saf e-shutdowp equipment has a-ceiling or wall separating it from an overhead handling' system, provide an evaluation to demonstrate that postulated load drops do not penetrate the ceiling or cause secondary missiles that could prevent a safe-shutdown system from performing its safety function.
Attachment (5)
E APPENDIX A ANALYSES OF POSTULATED LOAD DROPS Certain of the alternatives in Sections 5.1.2 through 5.1.5 of this report call for an analysis of postulated load drops and evaluation of potential consequences to assure that the evaluation: criteria of Section 5.1 are met for such an event.
Section A-1 of this appendix identifies certain considerations that should be included in such evaluation,s. -Sections A-2 and A-3 identify certain additional considerations and assumptions that should be used in analyzing the potential consequences of a drop of the reactor vessel head assembly or the spent fuel shipping cask; other load drops that are analyzed should use similar considerations and assumptions that are appropriate for these other loads.
Section A-4 provides guidance in performing criticality calculations.
1.
GENERAL CONSIDERATIONS Analyses of postulated load drops should as a minimum include the considera-tions listed below.
Other considerations may be appropriate for the particular load drop being analyzed; for example, for a reactor vessel head assembly or a spent fuel cask drop analysis, the additional considerations listed in Sections A-2 or A-3 should be used.
In evaluating the potential for a load drop to result in criticality, the considerations of A-4 should also be followed.
The following should be considered for any load (rop analysis, as appropriate:
(1) That the load is dropped in an orientation that causes the most severe consequences; 1(2) That fuel impacted is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> subcritical (or whatever the minimum that is allowed in facility technical specifications prior to fuel handling);
(3) Thr t the load may be dropped at any location in the crane travel area j
where movement is not restricted by mechanical stops or electrical interlocks;
~(4) That credit may not be taken for spent fuel pool area charcoal filters if hatches, vall, or roof sections are removed during the handling of the heavy load being analyzed,.or whenever the building negative pressure rises _above (-)1/8 inch (-3 m) water' gauge;
~-
(5) Analyses that rely on results of Table 2.1-1 or Figures 2.1-1 or 2.1-2 for potential offsite doses or safe decay times should venify that the assumptions ofLTable 2.1-2 are conservative for the facility under review.
X/Q values should be derived from analysis of on-site meteorological measurements based on 5% worst meteorological conditions.
(6) Analyses should be based on an elastic plastic curve that represents a
.true stress-strain relationship.
e 4
A-1
(7) The analysis should postulate the " maximum damage" that could result, structure, analysis should consider that all energy is absorbed by the i.e., the and/or equipment that is impacted (S) Loads need not be analyzed if their load paths and consequences are scoped by the analysis of some other load.
s (9) To overcome water leakage due to dama'ge from a load drop, credit may be I
taken for borated water makeup of adequate concentration that is required to be available by the technical specifications.
.(10) Credit may not be taken for equipment to operate that may mitigate the effects of the load drop if the. equipment is not required to be operable by the technical specifications when the load could be dropped.
2.
REACTOR VESSEL HEAD DROP ANALYSIS
- i Where a reactor vessel head drop analysis is to be performed to satisfy the PWR Containment or BWR Reactor Building guidelines (Sections 5.1.3 or 5.1.4) of this report, the analysis should consider the following to assure that the evaluation criteria of Section 5.1 are satisfied.
(1) Impact loads should include the weight of the reactor vessel (RV) head assembly (including all appurtances), the crane load block, and other lifting apparatus (i.e., the strongback for a BWR)..
s (2) All potential accident cases during the refueling operations.
Areas of consideration as a minimum should be:
1 (a) Fall of the RV head from it's maximum height while still on the guide studs followed by impact with the RV flange; (b) Fall of.the RV head from its maximum height considering possible objects of impact such as the guide studs, the RV flange, the steam dryer (BWR) or structures beneath the path of travel; and (c) Impact with the fueling cavity wall due to load swing with the subsequent drop of the RV head due to lifting device or wire rope failure.
(3) All cases which are to be considered should be analyze'd in the actual medium present during the postulated accident, e.g., for a PWR prior to reassembly of the reactor, the fueling cavity is drained,after the head engages the guide studs to allow for visual inspection of~the reactor core control drive rods insertion into the head.
During this phase it should be considered that the head will only. fall through air, without any drag forces produced by a water environment.
- These guidelines only consider the dropping of the RV head assembly during refueling and do not apply directly to dropping of the reacter internals such as the steam dryer (BWR), moisture separator (BWR) or the upper core internals (PWR); however, similar assumptions and considerations would apply to analyses of dropping of reactor internals.
A-2
(4)
In those Nuclear Steam Supply Systems where portions of the reactor internals extend above the RV flange, the internals should be analyzed for buckling and resultant adverse effects due to the impact loading of the RV head.
It should be demonstrated that the energy absorption characteristics (causing buckling failure) of these incernals should be such that resultant damage to the core assembly does not cause a condition beyond the acceptance criteria for this analysis.
(5) Reactor vcssel supports should be evaluated for the effects of the transmitted impact loads of the RV head.
In the case of PWRs where the RV is supported at its nozzles, the effects of bending, shear and l
circumferential stresses on the nozzles.should be examined.
For BWRs the effects of these impact loads on the RV ' support skirt should be examined.
i (6) The RV head assembly should be considered rigid and not experience deformation during impact with other components or structures.
3.
SPENT FUEL CASK DROP ANALYSIS Where a cask drop analysis is to be performed to satisfy the guidelines in Sections 5.1.2, 5.1.4, or 5.1.5 of this report, it should consider the following in addition to the general considerations of Section A-1 to assure that the evaluation criteria of Section 5.1 are satisfied:
(1) Applying a single-failure to the lifting assembly, consider that the cask is dropped in an orientation that will result in the mest severe consequences.
(2)
Impact loads should include a fully loaded cask (with water, where applicable) and all equipment requirad for lifting and set down such as baseplates, lifting yokes, wire ropes and crane blocks.
(3) Restricted path travel of the spent fuel cask (defined by electrical interlocks, mechanical stops, and crane travel capability) should be evaluated to determine the locations and probable accident cases along the path where damage could occur to:
(a) the floor and walls of the Spent Fuel Pool (SFP);
(b) racks within the SFP which support the spent fuel; (c) the spent fuel itself; (d) the refueling channel gate; or (e) safety related systems, components and structures:beneath or adjacent to the travel path of the cask.
l (4)
In the analysis consideration may be given to drag forces' caused by the environment of the postulated accident case, e.g., when thh spent fuel cask is postulated.to drop into the SFP, credit may be taken :for drag forces caused by the water in the SFP.
Water level assumed f,or such analyses should be the minimum level allowed by technical specifications.
(5) Credit may be taken for energy absorbing devices integral to the cask if attached during the handling operations in determining the amount of energy imparted to the spent fuel or safety related systems, components l
or structures.
A-3
W (6) For the' purpose of the analysis the cask should be considered rigid (except for devices and appurtences specifically designed for energy absorption and in place) and not to experience deformation during impact.
(7) In the calculating the center of gravity, consideration should be given to modifications made to the cask after purchase, e.g., addition of a 4
perforatedmetalbasketwithintheca[sk.
i 4.
CRITICALITY CONSIDERATIONS 4.1 Spent Fuel Pool Neutronics Analysis InSections5.1.2,"!pentFuelPoolArea-fPR,"and5.1.4,"ReactorBuilding-
.BWR," a number of alternatives are presented for the control of heavy loads in spent fuel pool areas.
Some of these alternatives include neutronics calcula-tions-to demonstrate that crushing the fuel and fuel rack will not result in criticality.. This section is includeo here to give the licensees guidance in performing their neutronics calculation.
i A discussion of the potential for criticality under load drop conditions is discussed in Section 2.2, and summarized in Section 2.2.6.
The results of this section should be used as a guide to determine which neutronics or other l
analyses are required to evaluate the potential for criticality for a specific plant area.
A licensee may choose to use the results of section 2.2, rather than performing an independent neutronics analysis for his giant.
If a licensee uses the results of Section 2.2 rather than performing an independent neutronics analysis, he should verify that the assumptions and model fuel assembly of Section 2.2 are valid for his plant.
.For PWR spent fuel pools, credit may be taken under the accident conditions of a load drop for the baron in the spent fuel pool water to maintain subcriticality.
In this case the required boron concentration should be specified in the facility Technical Specification, and regular monitoring of the baron concentration in the spent fuel pool should also be specified.
Likewise, if the neutronics analysis postulates a bounding distribution of non-spent fuel within the spent fuel pool, then the Technical Specifications must be modified to require that the actual distribution of fuel is no more deleterious than that assumed in the analysis.
In postulating a limiting distribution of non-spent fuel, the licensee may either assume an irfinite array or a finite array.
The largest finite array of non-spent fuel a licens,ee should have to consider would be that of an off-load core.
In this neutronics analysis the licensee must demonstrate that the fuel remains subcritical in the optimum crushed configuration.
It is adequqte to assume I
that the optimum configuration is with the rack crushed to uniformly reduce the separation between assemblies and the spacing between fuel pins uniformly reduced to maximize k All boral and structural material may be assumed to remaininitsoriginaTbo.nfiguration relative to the fuel, and not forced out of the fuel-array.
The neutronics analysis for the spent fuel pool should consider the case where it has become necessary to off-load an entire core into the spent fuel pool and a heavy load is dropped on fuel in the pool.
A-4
As noted in Section 5.1.4 it is not necessary to analyze the effects of crushing on k for BWR spent fuel pools that use baron plate cans and do not rely on spacTktomaintainsubcriticality.
4.2 Reactor Core Neutronics Analyses 4.2.1 NeutronicsAnalysesforaBWRCore[
For a BWR core, the potential for a load drop to drive control rods out of the core should be analyzed using the appropriate considerations of Sections A-1 and A-2.
If this analysis shows that post.ulated load drops could drive control rods out of the core, the number of rods tha,t could be affected should be determined, and a neutronics analysis performed to determine the potential for criticality to result.
If in the analysis it is assumed that all rods are in the core just prior to the load drop, then the facility technical specifications should require that all rods are in when handling a heavy load over the core.
- 4. 7. 2 Neutronics Analyses for a PWR Core In Table 2.2-2, we see that crushing the model PWR core in 2000 ppm boron refueling water increases k by about 0.02.
Since only one model fuel geometrywasconsideredherIffother fuel geometries could have a slightly higher reactivity insertion due to crushing.
A value of 0.05 may be used as a bounding worst case reactivity insertion value due to crushing of a PWR core.
In performing a neutronics evaluation of a postulated load drop on a PWR core, a licensee may use this estimated reactivity insertion limft in lieu of performing a plant specific calculation.
If a licensee can demonstrate that for his fuel a value less than 0.05 is bounding, then he may use this lower value instead.
The current Technical' Specifications require that during refueling k,N perform should be maintained at 0.95 cr less.
This is based on an uncrushed care.
a neutronics analysis to demonstrate that crushing the core will not drive it critical at least two alternatives for demonstrating this are acceptable.
(1) The licensee can perform a neutronics analysis on his core uniformly crushed in the x y direction to maximize k this option he must demonstrate that the mix $;.
If the licensee chooses aum k,f 7 is no greater than 0.95, with all uncertainties taken into account.
OR (2) Using his core refueling neutroliics analysis (uncrushed), the licensee can demonstrate that k for the uncrushed core is no greater than 0.90.
Then, using the estima N 0.05 maximum reactivity insertion due to crushing, the maximum achievable k,ff is still less than 0.95.
S.
ACCEPTANCE CRITERIA In performing the above analyses, the acceptance criteria #or resultant damage should be that it does not cause a condition that may exceed evaluation criteria I-IV stated in Section 5.1 of this report.
A-5
Attachment (6)
TABLE 3.1-1 SURVEY OF HEAVY LOADS Over (0) or Only Proximity (P) to Approx.
Frequency Area Loads Handlied fuel Weight y Handled 1.
PWR - Refueling 1.
Spent Fuel Shipping Cask (P)15-110 Tons Building (13-100,000 kg) 2.
Pool Divider Gates (some plants)
(P) 2 Tons 2-4 x's (per (1800 kg) refueling) 3.
Fuel Transfer Canal Door (P) 2 Tons 2-4 x's (per (1800 kg) refueling) 4.
Missile Shields (P) 4-20 Tons 2 x's (per (4-19,000 kg) refueling) b 5.
Irradiated Specimen Shipping (P) 3.5-12 Tons Once per year to Cask
.(3-11,000 kg) '-'monce per 10 years 6.
Plant Equipment (some plants)
(0) 2-4 Tons As required for (e.g., pumps, motors, valves, (1800-3600 kg) modification or heat exchangers, etc.)
replacement 7.
Spent resin, filter, or other (P) 5.
fons
~ 5 x's per year radioactive material shipping (4500-33,000 kg) casks
/
- 8. - New fuel shipping containers (P) 3-4. Tons E
with fuel (usually 4 assemblies)
(2700-3300 kg) 9.
Failed Fuel Container (0) 1 Ton Less than once (900 kg) per refueling 6-
TABLE 3.1-1 (Continued)
Over (0) or Only Proximity (P) to Approx.
Frequency Area Loads Handled Fuel Weight 3j Handled
- 1. (cont.)
10.
Fuel transfer carriage (0) or (P) 1.5 Tons Only for main-(1300 kg) tenance or repair
(~ once per 10 years) 1 11.
Crane Load Block (0) 4-10 Tons (4-9,000 kg) 2.
PWR - Containment 1.
Reactor Vessel Head (0) 55-165 Tons 2 x's (per Building (50-150,000 kg) refueling) 2.
Upper Internals (0) 25-65 Tons 2 x's (per (23-33,000 kg) refueling) 3.
In-Service Inspection Tool (0) 4.5 Tons Used at least once (4,000~kg)""
"s~very three years 4.
Reactor Coolant Pump (P) 30-40 Tons 4-10 x's over (27-36,000 kg) life of plant 5.
Missile Shields (P) 10-20 Tons 2 x's (per (9-18,000 kg) refueling) 6.
Cra,ne Load Block
[(0) 4-10 Tons 1
(4-9,000 kg) 3.
BWR-- Reactor 1.
Missile or Shield Plugs (6-12)
(P)15-125 Tons 2 x's (per Building (13-112,000 kg) refueling) 2.
Drywell Head (P) 45-85 Tons 2 x's (per (40-77,000 kg) refueling)
~
TABLE 3.1-1 (Continued)
Over (0) or Only Proximity (P) to Approx.
Frequency Weight 77 Area Loads llandled Fuel Handled
- 3. (cont.)
3.
Reactor Vessel llead (0) (Over 45-96 Tons 2 x's (per reactor)
(40-86,000 kg) refueling) 4.
Steam Dryers 5 (0) (Over 20-40 Tons 2 x's (per reactor)
(18-36,000 kg) refueling) 5.
Moisture SeparatorsE (0) (Over 20-75 Tons 2 x's (per reactor)
(18-68,000 kg) refueling) 6.
Spent Fuel Pool Gates (0) (Over 2-6 Tons 2 x's (par spent fuel (1800-5,000 kg) refueling) pool) 7.
Dryer / Separator Storage Pit (P) 75 Tons 2 x's (per Shield Plugs (some plants)
(68,000 k' )
~ ^ refueling) g 8.
Refueling Slot Plugs (0) (Over 2-6 Tons 2 x's (per spent fuel (1800-5400 kg) refueling) pool)
S 9.
Spent Fuel Shipping Cask (0) (Over 15-110 Tons spent fuel (14-99,000 kg)
[ pool) 10.
Ves'sel Service Platform (0) 1-5 Tons 5-10 x's (per (900-4500 kg) refueling) 11.
Waste and Debris Shipping (0) (Over 8-30 Tons 1-3 x's (per Casks reactor and/
(7-27,000 kg) year) or spent fuel pool) 9
TABLE 3.1-1 (Continued)
Over (0) or Only Proximity (P) to Approx.
frequency Area Loads Handled Fuel Weight yf Handled
- 3. (cont.)
12.
Vessel Head Insulation (P) 4-6 Tons 2 x's (per (4-5,000 kg) refueling) 13.
Replacement Fuel Storage (0) (Over 8 Tons On installation Racks for Spent Fuel spent fuel)
(7,000 kg) 14.
Crane Load Block (0) 4-10 Tons 1
(4-9,000 kg) 15.
Plant Equipment (0) (Over 1 Ton safety equip.)
(900 kg) 2/, y w
4.
Other Plant 1.
Spent fuel Shipping Casks (0) (Over 15-110 Tons On Areas (some plants) safety equip-(14-99,000 kg)
~ ~ "'-
~
ment) s-2.
Turbine or other equipment (0) (Over) 2-150 Tons As required in turbine building (some safety equip-(2-135,000 kg) for equ went plants) ment) overhau; and
~
replacement 3.
Other plant equipment (pumps, (0) (0ver 1-30 Tons As required for motors, valves, heat exchangers,
' safety equip-(1-27,000 kg) equipment overhaul j
etc.:)
ment) and replacement 4
906
~.
TABLE 3.1-1 FOOTNOTES 1
Listed weight for loads does not include weight of load block except where listed separately.
The load block may add 4-10 tons (4,000 -
9,000 kg) to the weight of the dropped load.
Because of this, the load block should be considered a heavy load even if it is not carrying a load, or is being used with a lighter load.
2/
These are presently not being used at most plants.
However, once offsite waste repositories are established, c' asks will be used frequently for shipping spent fuel offsite.
For a typical 1,000 MWe pressurized water reactor, spent fuel casks must be shipped offsite from 7 to 65 times per year depending on the size cask used.
This is based on casks currently licensed for use in the United States.
3/
A typical 1,000 MWe power plant would usually require 16 or 17 new fuel containers (four fuel assemblies each) per year.
d!
These are presently'not being used at most plants.
However, once offsite waste repositories are established, casks will be used frequently for shipping spent fuel offsite.
For a typical 1,000 MWe boiling water reactor, spent fuel casks must be shipped offsite from'12 to 125 times i
per year depending on the size cask used.
This is based on casks currently licensed for use in the United States.
5#
Due to certain dimensional restrictions, for most BWR's it would not be possible to drop the dryers or moisture separators onto fuel in the reactor core.
I 3-6 j
D O-Attachment (7) 1 of 1 SHIELDED SHIPPING CASKS CERTIFICATED FOR NUCLEAR POWER PLANTS I - Fuel (New and Spent) l GROSS LOT IN CERT.
'MODEL PRIMARY LICENSEE I
LBS. (APPROX.)
SECONDARY LICENSEE 4986 RA-1, 2, 3, J General Electric Co.
TVA 5450 RCC, 1, 2, 3 Westinghouse Electric VEP, DLC 5805 Vandenburgh C' hem-Nuclear Systems, 70,000 APC, CPL, DLP, DPC,*
Inc.
FPL, FPC, JCP, NPP, VEP 5901 NFS Model 100 Nuclear Fuel Services 126,200 CPC, PGE 5938 HNPF 48,000 PEC 6078 927Al Combustion Engineer-6200 APL
_927Cl ing, Co.
7000 6206 B
Babcock & Wilcox Co.
~
6273 48 (Series) 4500 VEP 6373 PB-1 Chem-Nuclear Systems, 67,050 APC, BEC, CPL, DPC Inc.
FPL, FPC, GPC, JCP, MYA, MIC, NNE NSP, PNY, TVA, VEP 6400 Super Tiger Westinghouse Electric 45,000 APL, CPC, DLP, DLC, Co.
MEC, NPP, SMU, VEP 6698 NFS-4 Nuclear Fuel Services, 50,000 BGE, BEC, CWE, DLP, Inc.
DPC, FPL, FPC, JCP, MYA, RGE, SCE WMP, 9001 IF 300 General Electic Co.
140,000 CPL, CWE 9010 NLI-1/2 NL Industries, Inc.
47,500
'BEC, FPL, VYC 9044 GE-1600 General Electric Co.
23,000 APC, BGE, BEC, CPL, CPC, DPC, FPL, FPC, GPC, IEL, JCP, MEC, NNE, NSP, VEP, VYC
- See attached list of abbreviations.
l l'of 3
-SHIELDED SHIPPING CASKS CERTIFICATED FOR NUCLEAR POWER PLANTS t
II - I!aste
?
GROSS LOT IN CERT.
MODEL PRIMARY LICENSEE I-LBS. (APPROX.)
SECONDAP,Y LICENSEE
- t 5026 BC-48-220 Chem-Nuclear Systems,
' 71,000 APC, BEC, CPL, CWE, Inc.
CYA, DPC, DLC, FPL, FPC, JCP, NPP, VEP, WPS 6058 B3-1 Nuclear Engineering Co.
30,000 APL, CPC, DLP, IEL, MEC, NPP, NSP, PGE, SMU, TEC, VEP i
6144 6144 Nuclear Engineering Co.
42,000 APC, APL, CPL, CEC, CPC, DLP, DPC, FPL, FPC, GPC, IEL, JCP, MEC, NPP, NSP, PGE, PNY, RGE, SMU, VEP 6244 6244 Chem-Nuclear Systems, 46,000 APC, CPL, CWE, DPC,
~
Inc.
FPL, FPC, GPC, JCP, MEC, NMP, NSP, PEC, YEP, WMP 6272 Poly Panther Nuclear Engineering Co.
6100 APL, CPC, DLP, MEC E
NPP, SMU, VEP 6568 LL-60-150 Tennessee Valley Auth.
73,000 6574 RN 200 Hittman Nuclear and 47,000 APL, BGE, CWE, CEC, Development Corp.
DLP, DLC, IME, JCP, 4
MYA, MZC, NPP, PEC,
~-
PNY, VYC, YAC 6601 LL-50-100 Chem-Nuclear Systems, 70,000 l
Inc.
. APC, BEC, CPL, CYA, CEC, CPC, DLP, DPC, 7
FPL, FPC, JCP, NPP, NNE, PEG, RGE, TVA, VEP 6679:
1/2 Super.
Nuclear' Engineering Co.
45,000 APL, CPC, DLP, MEC, 4
Tiger NPP, SMU,' VEP'
. 6722.
BS-33-180:
Tennessee Valley Auth.
51,000
- See attached list-of abbreviations.
V n
e 2 of 3 SHIELDED SHIPPING CASKS CERTIFICATED FOR NUCLEAR POWER PLANTS II - Waste GROSS LOT IN
, CERT.
MODEL PRIMARY LICENSEE (APPROX.)
SECONDARY LICENSEE LES.
6744 Poly Tiger Nuclear Engineering Co. 35,000 APL, BEC, CPC, DLP, MEC, NPP, SMU, VEP i
6771 SN-1 Nuclear Engineering Co.
60,000 APL, CPC, DLP, NPP, SMU, VEP
-9074' AP-100 28,000 DLC
'N-100 Ser. 2 Hittman Nuclear and 98,000 APL, BGE, CEC, CWE, 9079 H
Development Corp.
DLP, IME, JCP, MYA, MEC, NPP, PEC 9080 HN-600 Hittman Nuclear and 42,000 BGE, CWE, CEC, DLP, Development Corp.
IME, IEL, JCP, MYA, N
MEC, NPP, PEC, YAC 9086 HN-100 Ser. 1 aittman Nuclear and 46,000 APL, BGE, CWE, DLP, Developeent Corp.
IME, JCP, NYA, MEC, NPP, NNE, PEC, RGE, VYC l
1
-9089 HN-100S Hittman Nuclear and 36,500 BCE, CWE, CEC, IME, Development Corp.
JCP, MYA, NPP, PEC 9092 HN-300 Hittman Nuclear and 43,000 MYA Development Corp.
I 9093 HN-400 Hittman Nuclear and 43,000 MYA Development Corp.
9094 CNSI-14-195-H.
Chem-Nuclear Systems, 56,500 APC, APL, BEC, CPL, Inc.
'DPC, FPL, FPC, GPC, JCP, MEC, NMP, NNE, NSP, OPP, PGE, PEC,
'I PGC, PNY, PEG, TVA,
~
VEP.
9096 CNSI-21-300 Chem-Nuclear Systems,-
57,450 APC, APL, CPL, CEC, Inc.
DPC, FPL, FFC, GPC, JCP, MEC, NMP, NNE, PNY, PEG, VEP 4
- See attached list of abbreviations.
i
. -,. _ ~,
o o
o 3 of 3 SHIELDED SHIPPING CASKS CERTIFICATED, FOR NUCLEAR POWER PLANTS II - Waste GROSS LOT IN CERT.
MODEL PRIMARY LICENSEE LBS. (APPROX.)
SECONDARY LICENSEE
- 9105 RAD-Waste CR.I Chem-Nuclear Systems,
' 58,400 APC, CPL, DPC, FPL, Inc.
FPC, GPC, JCP, MEC, NMP, VEP 9108 AL-33-90 Chem-Nuclear Systems, 41,300 APC, CPL, CkT, CEC, Inc.
DPC, FPL, FPC, JCP, NPP, NMP, NNE, PCC, VEP, kTP 9111 CN6-80A Chem-Nuclear Systems, 51,500 APC, CPL, CVE, CEC, Inc.
MEC, NNE, PGC, SMU, TEP 9113 7-100 Chem-Nuclear Systems, 7000 2*C, BEC, CPL, CWE, Inc.
CYA, DPC, FPL, FPC, GPC, JCP, MEC, NMP, ENE, NSP, VEP 9123 18-450 Chem-Nuclear Systecs, 61,000 BEC Inc.
1 f.
- See attached list of abbreviations.
J
^D 0
1 of 1 SHIELDED SHIPPING CASKS CERTIFICATED COR NUCLEAR POWER PLANTS III - Byoroducts GROSS LOT IN CERT.
MODEL PRIMARY LICENSEE LBS. (APPROX.)
SECONDARY LICENSEE 5971 GE-200
,,10,000 PEC 5980 GE-600 18,500 NNE. NSP 6275 LL-28-4 Chem-Nuclear Systems, 30,000 APC, CPL, DPC, FPL, Inc.
FPC, NPP, VEP 9001 CNS-1600 Chem-Nuclear Systems, 26,000 APC, BGE, CPL, DPC, Inc.
's 1
- See attsched list of abbreviations.
l 1
s
-o LICENSEE ABBREVIATIONS APC Alabama Power Company APL Arkansas Power and Light Company BEC Boston Edison Company BGE Baltimore Gas and Electric Company CEC Consolidated Edis'on Company CPC Consumers Power Company CPL Carolina Power an'd Light Company GTE Commonwealth Edison Company CYA Connecticut Yankee comic Power Company LLC Duquesne Light Company DLP Dairyland Power Cooperative DPC Duke Power Company FPC Florida Power Corporation FPL Florida Power and Light Company GPC Georgia Power Company IEL Iowa Electric Light and Power Company IME Indiana and Michigan Electric Company JCP Jersey Central Power and Light Company MEC Metropolitan Edison Company MYA Maine Yankee Atomic Power Company NMP Niagara Mohawk Power Corporation NNE Northeast Nuclear Energy Company NPP Nebraska Public Power Corporation NSF Northern States Power Company OPP Omaha Public Power District PEC PhiLdelphia Electric Company PEG Public Service Electric and Gas Cem;pany PGC Portland General Electric Company PNY Power Authority of the State of New York RGE Rochester Gas and Electric Corporation SMU Sacramento Municipal Utilities Corporation TEC Toledo Edison Company TVA Tennessee Valley Authority VEP Virginia Electric and Power Company VYC Vermont Yankee Nuclear Power Corporation YAC Yankee Atomic Electric Corpany WHT Wisconsin-Michigan Power Company WPS Wisconsin Public Service Corporation
._