ML20054M070
| ML20054M070 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/03/1982 |
| From: | Sargent I, Vito D Franklin Research Ctr, Franklin Institute |
| To: | Clemenson F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20054M052 | List: |
| References | |
| CON-NRC-03-79-118, CON-NRC-3-79-118, REF-GTECI-A-36, REF-GTECI-SF, TAC 46646, TASK-A-36, TASK-OR TER-C5257-440, TER-C5257-440-DRFT, NUDOCS 8207090214 | |
| Download: ML20054M070 (35) | |
Text
-
(DRAFT)
TECHNICAL EVALUATION REPORT CONTROL OF HEAVY LOADS (C-10) l CONSUMERS POWER COMPANY' l
BIG ROCK POINT NUCLEAR PLANT NRC DOCKET NO.
50155 FRC PROJECT C5257 NRCTAC NO. 46646 FRC ASS!GNMENT 3 l
l NRC CONTRACT NO. NRC-03-79-118 FRC TASK. 440 Preparedby Author: D. J. Vito Franklin Research Center F C Group Leader: I. H. Sargent P iadelphia PA 103 Prepared for Nuclear Regulatory Commission ~
. Washington, D.C. 20555 Lead NRC Engineer:
F. Clemenson l
l t-l June 3, 1982 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agancy thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such uso, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infrirege privately owned rights.
1
. Franklin Research Center A Division of The Franklin Institute 8207090214 820702 The Bengmm Frenen Parrway, Phda Pa. 19103(215)448-1000 PDR ADOCK 05000155 P
m v
TER-C5257-440 CONTENTS Section Title g
1 INTRODUCTION.
1 a
1.1 Purpose of Review 1
1.2 Generic Background..
1 1.3 Plant-Specific Background 2
2 EVALUATION AND RECOMMENDATIONS 4
2.1 General Guidelines.
4 2.2 Interim Protection Measures.
20 3
CONCLUDING
SUMMARY
25 3.1 General Provisions for Load Handling 25 3.2 Interim Protection Measures.
30 3.3 Sunusary.
30 4
REFERENCES 31 l
l j
4 iii di d Franklin Research Center A Ones.an of The Fransen insatute 4
=
TER-C5257-440 i
.s FORrn*CKD This Technical Evaluation Report was prepared by Franklin Rbsehrch Center under a con. act with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Aegulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted'in accordanceswith criteria established'by, the NRC.
Mr. D. J. Vito and Mr. I. H. Sargent contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.
g..
./
e A Devenson of The FM heeture
TER-C5257-440 1.
INTRODUCTION
, 1.1 PURPOSE OF REVIEW This technical evaluation report. documents the review of gen.eral load handling policy and procedures at the Consumers Power Company's Big Rock Point Nuclear Plant.
This evaluation was performed with the following objectivest o to assess conformance to the general load handling guidelines of NUREG-0612, " Control of Heavy Ioads'at Nuclear Power Plants" [1],
Section 5.1.1
~
~~ '
o to assess conformance to the interim protection measures of NUREG-0612, Section 5.3.
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 [2] to all power reactor licensees, requesting information concerning the control of heavy loads near spent fuel.
The results of Task A-36 were reported in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." The staff's conclusion from this evaluation was that existing measures to control the handling of heavy loads at operating j
plants, although' providing protection from certain potential problems, do not adequately cover the major causes of load handling accidents and should be upgraded.
In order to upgrade measures for the control of heavy loads, the staff j
j 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, Section 5.1.1, is to ensure that all load handling systems at nuclear power plants are designed and operated so that their probability of failure is uniformly small and appropriate for the critical tasks in which
- Ob Franklin Research Center acm.ca.< n. r - m
e TER-C5257-440
. they are employed. The second. portion of the staff 's. objective, achieved through guidelines identified in NUREG-0612, Sections 5.1.2 through 5.1.5,. is
~
to ensure tha t, 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 tne 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:
1.
provide sufficient operator training, handling system design, load _.
handling instructions, and equipment inspection to assure reliable operatio'n of the handling system 2.
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 shutdown equipment 3.
provide mechanical stops or electrical interlocks to prevent movement of heavy loads over irradiated fuel or in. proximity to equipment associated with redundant shutdown paths.
Staff guidelines resulting f:om the foregoing are tabulated in Section 5 of NUREG-0612.
Section 6 of NUREG-0612 recommended that a program be initiated to ensure that these guidelines are implemented at operating plants.
1.3 PLANT-SPECIFIC BACKGROUND On December 22, 1980, the NRC iscued a letter (3) to Consumers Power Company, the Licensee for the Big Rock Point Nuclear Plant, requesting that the Licensee review provisions for handling and control of heavy loads, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determina-tion of conformance to these guidelines.
On June 10, 1981, Consumers Power
, 4 d!! Franklin Research Center A On=eica of The Frannan insonate
TER-C5257-440 provided the initial response [4] to this request.. Additional,information was _
provided by the Licensee on July 1, 1981 [5] and Septerber 23, 1981 [6], March 5, 1982 [7], April 19, 1982 [8], April 26, 1982 [91, April 30, 1982 (10), and May 2, 1982 [11). A site visit on March 23 and 24, 1982 to evaluate the 24-ton fuel transfer cask safety sling provided additional clarification of several NUREG-0612 issues.
e e
e t
G 6
a.-ee.
e +
mea
=
f f"*
l l
i i.
^
A dUbranklin Research Center A Dmason of The Frenamn inesame
~
TER-CS257-440 2
EVALUATION AND RECOMMENDATIONS Evaluation of load handling at the Big Rock Point plant is divided into two categories. These categories deal separately with the general guidelines of NUREG-0612, Section 5.1.1 and the recommended interim seasureh of Section 5.3 or their equivalents from NUREG-0612.
Applicable guide' lines are
~
referenced in each category. Conclusions and recommendations are provided in the summary for each guideline.
2.1 GENERAL GUIDELINES The NRC has established seven general gttidelines which must be met in order to provide the defense-in-depth approach for the handling of heavy loads.
These guidelines consist of the following criteria from Section 5.1.1 of NUREG-0612:
o Guideline 1 - Safe Load Paths o Guideline 2 - Load Handling Procedures o Guideline 3s-Crane Operator Training o Guideline 4 - Special Lifting Devices o Guideline 5 - Lif ting Devices (Not Spacially Designed) o Guideline 6 - Cranes (Inspection, Testing; and Maintenance) o-Guideline 7 - Crane Design.
These seven guidelines should be satisfied for all overhead handling systems and programs for handling heavy loads in the vicinity of the reactor vessel, near spent fuel in the spent fuel pool, or in other areas where a load drop may damage safe shutdown systems. The Licensee's verification of the extent to which these guidelines have been satisfied and an evaluation of this verification are contained in the succeeding paragraphs.
2.1.1 NUREG-0612, Heavy Load Overhead Handlino Systems a.
Su=marv of Licensee Statements and Conclusions The Licensee stated that the following overhead handling systems at the Big Rock Point plant are subject to the general guidelines of NUREG-0612:
. MO Franklin Research Center
% es w r,.noa w.
~
TER-C5257-440 Reactor crane Reactor auxiliary hoist Reactor depressurization system hoist Cleanup domineralizer hoist SRV hoist Emergency condenser beam Turbine crane.
In addition, portable gantries capable of lifting several tons are located in the reactor building laydown area and the emergency condenser level.
A jib crane is located over the reactor vessel during refueling operations and is rated at 500 lb.
A 500-lb winch is also located on the bridge over the
~
fuel pool and a 2000-lb winch is mounted on the 24-ton fue1 transfer cask.
The Licensee has excluded the following load handling ' systems because equipment required for safe shutdown is not affected by a failure of these systems:
Decontamination room hoist Equipment lock crane Screen hcuse trolley Machine caop trollies CRD hoist and trolley.
b.
Evaluation The Licensee's determination of NUREG-0612 applicability is felt repropriate except for the screen house trr '. ley and equipment lock crane. The intent of NUREG-0612 is to reduce the probability of damage to any system-required for plant shutdown, decay heat removal, or spent fuel handling following a load handling system failure. A review of the Licensee's submittal indicates that a failure of the screen house trolley could affect the circulating and service water systems' ability to provide cooling water for the main condenser, core spray and post-accident heat removal, automatic reactor depressurization, and component cooling. Further, a failure of the equipment lock crane could damage the core spray equipment located in the post-incident room beneath the equipment lock loading dock, affecting post-accident heae removal capability and possibly breaching containment.
-S-
.rTSbs
!!bd!} Franklin Research Center A Dwman of The Frennte insensee
f
- %}
TER-C5257-440 Exclusion of the decontamination room hoist, is epnsistent, with NUREG-0612 on the basis of verification during the recent site visit that the mechanical
. block limiting hoist travel is permanently welded in place.
c.
Conclusions and Recommendations The Big Rock Point plant substantially complies with NUREG-0612 concerning load handling system applicability to the general guidelines included in Section 5.1.1.
In order to fully comply, the ' Licensee should modify the heavy load handling program to include the screen house trolley and equipment lock crane.
2.1.2 Safe Load Paths [ Guideline 1, NUREG-0612, Section 5.1.l(1) ]
" Safe load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to impact irridiated fuel in the reactor vessel and in the spent fuel pool, or to impact safe shutdown equipment. The path should follow, to the extent practical, structural 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."
a.
Summary of Licensee Statements and Conclusions l
The Licensee has indicated that the safe load paths inside the l
containment have been selected to minimize the potential for damage to fuel l
and/or safe shutdown equipment should a heavy load be dropped in these areas.
In addition, the Licensee has stated that restricted areas are designated both in containment and in the turbine building. Lifts within restricted areas are j
procedurally controlled to limit the height of a heavy load lift during power opera tion.
The safe load path for the fuel transfer cask is that path directly between the reactor vessel and the pool over which the operator moves the cask during refueling operation.
This path is clearly marked on diagrams within procedures which the operator is using to perform refueling activities.
l 4 UM Franklin Resensch Center 4cm.onorn. % w
TER-C5257-440-Deviations from the path specified in the refueling procedures occur only when transporting the cask to and from its normal storage location and when activities associated with refueling dictate that the 5-ton auxiliary hook be used to lif t components in other areas of containment.
b.
Evaluation The Licensee's action's concerning the establishment of safe load paths comply with the requirements of Guideline 1 relative to the establishment of reactor building and turbine building exclusion areas and the load path for the 60-ton fuel cask. However, designation of safe load pat'hs for the 24-ten fuel transfer cask covers only the movement between the reactor vessel and the storage pool.
The movements of the 24-ton fuel transfer cask from its stand to the refueling area and during use of the 5-ton auxiliary hook are handled as deviations.
Safe load paths for these additional regular movements of the 24-ton cask should be specified to minimize the movement of heavy loads over the reactor vessel.
In add'. tion, safe load paths should be marked to provide
' visual aids to operators and supervisors during load handling evolutions.
Further, the methodology of handling deviations from safe load paths is not clear.
c.
Conclusions and Recommendations The Big Rock Point plant partially complies with Guideline 1 of e-NUREG-0 612.
In order to fully comply, the Licensee should:
1.
Identify safe load paths for the fuel transfer cask during use of the auxiliary book and movement from its stand.
l l
2.
Provide visual aids to identify safe load paths and restricted areas.
3.
Clarify the procedure for handling deviations from safe load paths.
l 2.1.3 Load Handling Procedures [ Guideline 2, NUREG-0612, Section 5.1.l (2)]
I
" Procedures should be developed to cover load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment. At a minimum, procedures i
1 M 8dj Franklin Research Center A Dme.on of The Franann insonne
_i _ _ _.T.
__.'_,___..Z.~~i TER-CS257-440 should cover handling of those loads listed in Table 3-1 of NUREG-0612. These
prode'aures 'hould include:
identification of required' equipment; inspections s
and acceptance criteria required before movement of load; the steps and proper sequence to be followed in handling the lead; defining the safe path; and other special precautions."
a.
Summarv of Licensee Statements and Conclusions
[
Standard Operating Procedure (SOP) 43, " Control of Heavy Loads,"
generally covers handling operations for heavy loads that.are or could be handled in proximity to irradiated" fuel or safe shutdown equipment inside containment. Additional details were provided in the following procedures:
s SOP Title MFHS-1 Transfer Cask Preparation for Fuel Movement MFHS-2 Puel Handling Cables Inspection / Testing MFHS-4 Test Tripping and Resetting of Fuel Handling Transfer Cask SOP 2 Refueling Operation TR 46 Fuel Bundle Core Loading Procedure l
TR 23 Neutron, Sources Removal, Inspection and Reinstallation 0-RVI-3 Channel and Blade Removal Procedure 0-RVI-4 Reactor-Intervals Reconstitution 0-RVI-5 Drive Operation Correction Procedure 0-RVI-8 Auxiliary Neutron Source Procedure O-FFI-1 Fuel Bundle Removal Procedure b.
Evaluation Load handling procedures involving the fuel transfer cask comply with Section 5.1.l(2) of NUREG-0612. MFHS 1, 2, and 4 delineate the preparation, cable inspection / testing, and the safety sling. testing of the fuel transfer cask.
In addition, the various refueling procedures provide specific 4
-S-Edd Franklin Research Center Acm aen.rr. e u.
1 TER-C5257-440 2
reference to use of the cask during refueling.
However, although MFHS-4 fully describes the procedure for test tripping the safety sling, the periodicity
. for perfor=ing this test is not =entioned in any of the referenced procedures.
Further, although SOP 43 generically addresses heavy load handling, insufficient infor=atio'n has been provided to determine if the iBentified heavy loads (with the exception of the fuel transfer cask) have specific load handling procedures with the detail specified in Section 5.1.l(2) of NUREG-0612, such as 1.
identification of required equipment 2.
inspection and acceptance criteria required before movementi of the load 3.
steps and proper sequence to be followed in handling the load 4.
safe load path S.
special precautions.
c.
Conclusion The Big Rock Point plant partially complies with Guideline 2 of ?
NUREG-0612 in that procedural controls exist for movements of the fuel transfer cask.
In order to fully comply with Guideline 2, the Licensee should verify that procedures containing sufficient detail to adequately define safe load handling are available for the remaining heavy loads which are subject to NUREG-0612.
Also, MFHS-1 should be revised to require the safety sling trip test in MFHS-4 each time the transfer cask is prepared for movement.
2.1.4 Crane Operator Training [ Guideline 3, NUREG-0613, Section 5.1.l(3)1
" Crane operators should be trained, qualified and conduct themselves in accordance with Chapter 2-3'of ANSI B30.2-1976, ' Overhead and Gantry Cranes' [12). "
i a.
Summarv of Licensee Statements and Conclusions The Licensee has stated that the Big Rock Point crane operator training, qualification, and conduct have been reviewed by the Licensee and found in J
4 dij Franklin Research Center 4 % s m re u.
- - ~., - --
a-e------
TER-CS257-440 compliance with the ANSI B30.2 requirements with the exception of visual examinations. Big Rock Point crane operators are presently qualified in
~
visual acuity to ANSI standards comparable to the requirements of their normal job duties, whether they are auxiliary cperators or mechanical repairmen.
While the Licensee intends to upgrade'those standards as necessary to meet crane operator qualification of A..SI B30.2, it is concluded that the present visual testing suffices for the interim.
b.
Evaluation A review of the Licensee's submittal indicates that the Big Rock Point plant complies with Section 5.1.l(3) of NUREG-0612 on the basis of the Licensee's certification of compliance with ANSI B30.2 and a commitment to upgrade visual testing.
c.
Conclusions and Recommendations The Big Rock Point plant complies with Guideline 3 of NUREG-0612.
2.1.5 Special Lifting Devices (Guideline 4, NUREG-0613, Section 5.1.l(411 "Special lifting devices should satisfy the guidelines of ANSI H14.6-1978,
' Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials' [13]. This standard should apply to all special lifting devices which carry heavy loads in areas as defined above. For operating plants certain inspections and load tests may be accepted in lieu,of certain material requirements in the standard.
In' addition, the strass design factor stated in Section 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 based on characteristics of the crane which will be used. This is in 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 the intervening components of the special handling device."
a.
Summarv of Licensee Statements and Conclusions The Licensee has stated that the Big Rock Point plant has only two lifting yokes, which are used for the fuel shipping cask, and which will not
- dLEnklin Research Center A Dwor of The Frannhn insotwee
I 1.
TER-C5257-440 be used in the foreseeable future.
Yokes for the cobalt and T-II casks
~..
accompany their respective casks and are not controlled by the Big Rock Point plant or tested to the requirements of ANSI N14.6-1978.
Procedures will be
~
revised to provide for visual inspection of the yokes prior to their use to meet, in part, the ANSI N14.6 inspection criterion.
b.
Evaluation The review and evaluation of special lif ting devices by the Licensee are-insufficient to satisfy Guideline 4'.
Regardless of the ownership of these
~
devices, ANSI N14.6-1978 should apply to all special lif ting devices that carry heavy loads in proximity to or over safe shutdown equipment or irradiated fuel in the spent fuel pool, in the reactor building, and in other plant areas. The Licensee should evaluate, or require that lif ting device owners evaluate, special lif ting devices used at the Big Rock Point plant, in accordance with NUREG-0612, Guideline 4.
An evaluation and review of ANSI N14.6-1978 has identified several areas which the Licensee should consider Section 3.1:
a.
limitations on the use of the lif ting devices (3.1.1) b.
identification of critical components and definition of critical characteristics (3.1.2) c.
signed stress analyses which demonstrate appropriate margins of safety (3.1.3) d.
indication-of permissible repair procedures (3.1.4) section 3.2:
a.
use of stress design factors of 3 for minimum yield strength and 5 for ultimate strength (3.2.1) b.
similar stress' design factors for load bearing pins, links, and adapters (3.2.4) c.
slings used comply with ANSI B30.9-1971 (3.2.5) d.
subjecting materials to dead weight testing or Charpy impact testing (3. 2. 6) 1 Section 3.3:
a.
consideration of problems related to possible lamellar tearing (3. 3.1) b.
design shall assure even distribution of the load (3.3.4) c.
retainers fitted for load carrying components which may become inadvertently disengaged (3.3.5)
O Nd Franklin Research Center m e m vmmm m
y TER-C5257-440 d.
verification that remote actuating mechanisms securely engage oc disengage (3. 3. 6 )
~
Section 4.1:
a.
verify selection and use of material (4.1.3) b.
compliance with fabrication practice (4.1. 4 )
c.
qualification of welders, procedures, and operators (4. f. 5 )
d.
provisions for a quality assurance program (4.1.6) e.
provisions for identification and certification of equipment (4.1. 7 )
f.
verification that materials or services are produced under appropriate controls and qualifications ( 4.1. 9)
Section 5.1:
implementation of a periodic testing schedule an'd a system to a.
indicate the date of expiration (5.'l.3) b.
provisions for establishing operating procedures (5.1. 4 )
c.
identification of subassemblies which may be exchanged (5.1. 5) d.
suitable markings (5.1.6) e.
maintaining a full record of history (5.1.7) f.
conditions for removal from service (5.1.8)
Section 5.2:
a.
load test to 150% and appropriate inspections prior to initial use (5.2.1) b.
qualification of replacement parts (5.2.2)
Section 5.3:
a.
satisfying annual load test or inspection requirements (5.3.1) i b.
testing following major maintenance (5.3.2)
I c.
testing af ter application of substantial stresses (5.3.4) d.
inspections by operating (5.3.6) and non-operating or maintenance personnel (5.3.7).
c.
Conclusions and Recommendations The Big Rocx Point plant does not comply with Guideline 4 of NUREG-O'612.
i In order to satisfactorily comply, the Licensee should conduct a point-by-point comparison of all special lif ting devices against the criteria of ANSI N14.6-1978, taking into consideration the stress design factor of Guideline 4.
In addition, the Licensee should provide verification that ANSI N14.6-1978 requirements are imposed on all vendors who provide temporary special lif ting devices at the Big Rock Point plant.
)
E' Franklin Research Center t
j
^ em ca er w rr n.en maan.
TER-CS257-440 2.1.6 Lifting Devices (Guideline 5, NUREG-0613, Section 5.1.l(511
" Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI S30.9-1971, ' Slings' (14].
However, in selecting the proper sling, the load used should be the sum of the static and maximum dynamic load.
The rating identified on the sling should be in terms of the ' static load' which produc,es the maximum static and dynamic load. Where this restricts slings to use on only certain cranes, tne slings should be clearly marked as to the cranes with which they may be used."
a.
Summarv of Licensee Statements and Conclusions The Licensee has stated that slings used in the handling of heavy loads are inspected periodically and prior to use to comply with ANSI B30.9-1971.
Specifically, NX01 slings used to rig the fuel' transfer cask and the reactor vessel head comply with ANSI B30.9-1971, Chapter 9-2 (wire rope).
These slings are stored and used on the reactor deck complying with storage -
and temperature requirements required by ANSI B30.9.
These slings are inspected prior to use in accordance with PEES-1 and thesdaily inspection checklist.
Personnel performing rigging activities are trained with respect to inspection and acceptance criteria and proper rigging practices.
The NX01 sling assembly consists of a sling line rated at 12 tons and a i
l turnbuckle rated at 13.8 tons. The entire sling assembly is certified to 28.6 tons. The sling is not marked with the maximum st.atic load which produces the maximum static and dynamic load because the sling assembly can only be installed on the transfer cask and reactor vessel head.
b.
Evaluation The Licensee has provided limited information to evaluate compliance to l
Guideline 5.
The only slings which can be properly evaluated are those slings used to handle the fuel transfer cask and the reactor vessel head (NX01 slings).
. A Shhranklin Research Center A Drahan of The Frannan armaame l
.. _ i.
___._ i._... l ___.~~ i
_1 TER-C5257-440 The NX01 slings appear to comply with the requirements of ANSI B30.9-1971.
Further, although the issue of dynamic load rating has not been addressed, the slow hook speed (6 ft/ min) of the reactor building crane, the design factor of safety of 5 (based on ANSI B30.9 tables), and the limited use of these lif ting devices is considered adequate to establish that the dynamic loads are insignificant in this application.
In addition, even thoug'h these slings are uniquely designed for a specific application, the Licensee's decision not to mark the slings with the maximum combined. static and dynamic loads does not meet the intent of Guideline 5.
Finally, insufficient information is available to evaluate compliance of 1
the remaining slings used to handle heavy loads subject to Guideline 5 of NUREG-0612 at the Big Rock Point plant.
c.
Conclusions and Recommendations The NX01 slings used to handle the fuel transfer cask and reactor vessel head at the Big Rock Point plant partially comply with Guideline 5 of NUREG-0612. In order to fully comply, the Licensee should perform the following evaluations on the remaining slings used to handle heavy loads and which are subject to NUREG-0612 at Big Rock Point plant:
~
1.
verify that slings are instal 1ed and used in accordance with ANSI B30.9-1971 2.
verify that the load used in selecting and marking the proper sling is based upon the sum of the maximum static and maximum dynamic loads 3.
verify that slings restricted in use to certain cranes are clearly marked to so indicate.
4.
Mark the NX01 slings to indicate the maximum static and dynamic loads or to indicate exclusive use for the 24 spent fuel transfer cask and the reactor vessel head.
2.1.7 Cranes (Inspection, Testing, and Maintenance) [ Guideline 6, NUREG-0612, Section 5.1.l(6) 1 "The crane should be inspected, tested, and maintained in accordance with Cha7ter 2-2 of ANSI B30.2-1976, ' Overhead,and Gantry Cranes,' with the j
. M} Franklin Research Center Jd!h 4c - w m rr.nonw
.m
TER-C5257-440 exception that tests and inspections should be performed prior to use where it is not practical to meet tne 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 Ph*R containment may only be used every 12 to 18 months during refueling operations, and is generally not accessible during power operation.
ANSI B30.2, however, calls for certain inspections to be performed daily or monthly. For such cranes having limited
- usage, the inspecticns, test, and maintenance should be performed prior to their use). "
o a.
Summary of Licensee Statements and Conclusions The Licensee has stated that crane inspection, testing, and maintenance have been conducted in the past in accordance with MIOSHA standards which are comparable to the standards in ANSI B30.2-1976, Chapter 2-2 with the exception of inspection intervals. Consequently, the single-leg gantry crane inspection intervals have been revised to meet the monthly and yearly requirements.
In addition, both the turbine building and loading dock cranes are inspected at quarterly and yearly intervals due to their infrequent use.
Inspections prior-to use are conducted on all cranes, in accordance with the standard.
Crane testing per the requirements of the ANSI standard has not been required as the new requirements apply only to new, reinstalled, altered, extensively repaired, or modified cranes.
The Licensee has stated that Big-Rock Point cranes do not fall within these categories.
Crane maintenance as required by ANSI B30.2 has been included as part of the inspection program.
Although not specifically covered under the ANSI standard, the remaining hoists and lif ting devices used in the handling of heavy loads will be revi ewed and the applicable inspection, testing, and maintenance requirements will be invoked for them as well.
b.
Evaluation Comparison of MIOSHA Standards, Part 18
(" Overhead and Gantry Cranes"),
wi th ANSI B30. 2-1976 indicates that the crane inspection, testing, and l
l l
A '
@.hj Fmnklin Research Center A Oms.cn of The Frwwnninsonne
.. - - -~
y TER-C5257-440 maintenance requirements are comparable. Also, the Licensee's decisions concerning inspection intervals comply with present requirements.
With reference to the rated load test criteria of ANSI B30.2-1976, insufficient information has been provided to evaluate the load test. criteria.
The Licensee should provide crane acceptance test data for review.
The Big Rock Point plant complies with the maintenance criteria of Guideline 6 of NURE-0612 on the basis of the Licensee's certification of compliance with the maintenance criteria of ANSI B30.2-1976.
- ~
c.
Conclusion The Big Rock Point plant substantially complies with Guideline 6 of NUREG-0612.
In order to fully comply, the Licensee should provide initial load test data on all cranes subject to the general guidelines of NUREG-0612.
2.1.8 Crane Design [ Guideline 7, NUREG-0612, Section 5.1.l(7)1 "T.he crane should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976, ' Overhead and Gantry Cranes,' and of CMAA-70, ' Specifications for Electric Overhead Traveling Cranes' (15]. An alternative to a specificatiotr in ANSI B30.2 or CMAA-70 may be accepted in lieu of specific compliance if the intent of the specification is satisfied."
l l
a.
Summarv of' Licensee Statements and Conclusions The Lic'ensee has indicated [16] that the reactor building crane was specified and designed to comply with specification #49 of The Electric I
l Overhead Crane Institute Inc. (EOCI-4 9) [17].
A detailed comparison was made l
l between the req'uirements of that standard and those of the current standards referenced in Guideline 7 (18]. Additional information was provided concerning the equivalency of actual design features with the requirements of these later standards, in areas associated with structural and mechanical reliability, where specific compliance was not apparent [9].
E gh dd Franklin Research Center 4 > or m r, wen mu.
I
TER-C5257-440 b.
Evaluation The reactor building crane at the Big Rock Point plant was designed and procured prior to the publication of the standards identified in Guideline 7.
Since these standards were not invoked in the original design specification, it is not feasible, in many cases, to determine unequivocally that i specific requirement of these latter standards has been satisfied.
Consequently, design features associated with load handling safety have been reviewed and compared with applicable current requirements employing engineering judgment, where appropriate, to' determine-if 'the intent of the current standard has b'een-
~~ ^
satisfied. The following discussion briefly identifies deviations noted in the course of this review where direct evidence of equivalency could not be established.
o Welding standards - CHAA-70 requires that welding design and procedure conform with the current issue of AWS D14.1.
The Big Rock Point reactor building crane specification invoked AWS standards current at the time of manufacture.
AWS D14.1, copyrighted in 1970, represents the consolidation of various AWS welding design, workmanship, repair, and inspection requirements for specific, applications in load bearing weldments used in the manufacture of industrial and mill cranes. This.
standard, in addition to consolidating welding requirements for
~
specific applications in crane construction, provides some quality improvement by incorporating advances in welding technology 'over that existing in prior AWS standards.
Such improvements are not expected to constitute a significant contribution to* load handling reliability.
o Girder bending stress - CHAA-70 requires that design loading combinations for Class A cranes include an allowance for lateral load, due to acceleration or deceleration, equal to 2.5% of the live load plus the crane bridge (less and trucks and end ties).
The Big Rock Point reactor building crane design included an allowance of 0.834 i
consistent with EOCI 49.
The difference in bending stress between these two loadings is not substantial (approximately 600 psi at 2.54 -
and 200 psi at 0.831).
Further, the principal issue is combined bending stress rather than the contribution from a specific force.
The Licensee has evaluated the combined bending stress for the load combination required to include lateral forces and found it to be. less than 12,000 psi.
This total stress is well within the allowable stress for the crane structural material (16,000 psi).
o Girder shear stresses - CHAA-70 requires that torsional forces be calculated and included in the design loading combination used to determine maximum shear stress.
Appropiate twisting moments were not required in such calculations under EOCI-49.
For cranes, in general, l
- dU Franklin Research Center A Dme.on el The Frwwen insoame
Me.anw er *..
we
-s = -
c-.
j_
_ _ u_. c_ __.
.e==.-%
=
-==-wi=.-.-
-me~===
..==.ee-.--*w.
.-,_+.
%.,m=.e,
.---e---
=. -
.e<
~
~
TER-C5257-440 net twisting moment is not a substantial contribution to overall shear stress (when considered in combination with shecr stresses due to. dead 1 cad, live load, rated load, and i= pact allowance). In the specific case of the Big Rock Point building crane, neither the Licensee's review nor an independent observation of this crane indicated any design feature that would lead to substantial twisting moments due to lateral or overhanging forces. The twisting moment due to starting and stopping of the bridge motor equal to 200% of full motor torque, as specified in CMAA-70, has been estimated at approximately 130 lb-ft.
The additfon of moments of this order to the load combination used to establish shear stress would not be of consequence.
o Girder proportions - 04AA-70 specifies an allowabie web depth to thickness ratio (h/t) of 188 for girders with maximum compression stress equivalent to that allowed in the design of the Big Rock Point reactor building crane.
The girders used in this crane have an h/t ratio of 200 which, while significantly less that the EOCI-49 I
allowable of 240, exceeds the CMAA-70 limit.
The difference between the actual and allowed h/t ratio is small.
The actual compressive stress is less than the allowable stress (as discussed above, the actual combined bending stress is on the order of 12,000 psi).
If the actual compressive stress is approximately this valre, as is -
expected, an h/t ratio of 220 could be used in conformance with (CMAA-70), which supports the conclusion that the intent of this
. requirement has been satisfied.
o Gantry leg structural design - CMAA-70 requires that the design-of leg, end tie, strut, and sill members conform to the requirements of-the current edition of the AISC Manual of Steel Construction at a unit stress proportional to that used for girder design.
BOCI-49, the basis for the Big Rock Point re, actor building crane, was prepared specifically for electric overhead cranes and, consequently, did not identify any structural standards for gantry leg members. While no I
information has been provided by the Licensee concerning design rules employed for the reactor crane gantry, leg, it is judged likely that design rules similar to those contained in the Manual of Steel Construction were used by the manufacturer.
In the specific case of the refueling transfer cask, a significant load due to travel pattern and lif t frequency, the fact that this lif t is approximately 1/3'of design rated load strongly indicates that no reduction in load handling reliability from that provided by strict conformance to CMAA-70 has been incurred.
o Paticue considerations - OdAA-70 requires that fatigue be considered in crane design and provides maximum allowable stress ranges for specific crane members subject to cyclic loading. No similar explicit requirements were provided in EOCI-49.
In the case of the Big Rock Point reactor building crane, the following facts indicate that this crane is protected from fatigue failure in a manner consistent with a crane with a design which explicity included the requirements of CMAA-70:
(1) no significant stress reversals are expected based on
- dJd Frankhn Research Center f
4 mmon w n= r==m >==.
TER-CS257-440 crane performance characteristics; (2) with a total duty cycle of less than 100,000: CHAA-70 previder for maximum stress ranges in the vicinity of allowable compression and tensile stress for only a limited member of types of me=bers and fasteners; (3) the number of lifts at or near design rated load is extremely small; and (4) the lif t making a significant contribution to the overall duty cycle, the transfer cask, is approximately 1/3 of the design load, o Gearing - CHAA-70 provides gearing design criteria based on' AGMA standards.
No similar specific criteria were provided in EOCI-49.
The Licensee has indicated that gearing in the reactor building crane is of steel const'ruction and that good engineering practice was employed in the selection of the herringbone and worm gear designs.
While no specific design information has been provided concerning gear design, it is judged likely that design rules similar to those incorporated in CMAA-70 concerning allowable strength and durability horsepower were employed by the manufacturer.
In t'he specific case of the refueling transfer cask, a significant load due to travel path and lif t frequency, the fact that this lif t is approximately 1/3 of the
' design rated load strongly indicates that no reduction in load handling reliability from that provided by strict conformance to CMAA-70 has been incurred.
o Drum design - CHAA-70 specifies that drum material should be steel or ASTM Grade A48-64 or later, Class 40 cast iron or equivalent, and that the drum be designed to withstand combined crushing and bending loads.
The reactor building crane at Big Rock Point employs a drum fabricated from ASDL A48, Class 35.
The specification in CMAA-70 of a later, higher tensile strength, material for the drum is similar to the specification of ASTM A-36 in lieu of A-7 structural material and represents recognition of industrial progress in the area of material properties. The use of a similar but lower tensile strength material, of composition based on ASTM standards and with appropriate properties used for design calculations, is judged to result in crane components, or structures, with overall factors of safety and, consequently, load handling reliability equivalent to that produced with higher tensile-strength material.
b.
Conclusions and Recommendations
.The Big Rock Point reactor building crane was designed and fabricated in accordance with appropiate industrial standards existing at the time of manufacture. A detailed comparison of the requirements of those standards and actual design features with the requirements of CMAA-70 indicates that this crane substantially meets the intent of Guideline 7 with respect to load handling reliability. There appear to be three items that require further, A -
Udd Franklin Research Center A Ornsson of The Franssa inesame
__1__'__ ___ ~
-.T_T_~
^
TER-CS257-440 information from the. Licensee in order to fully establish that the reactor building crane, under design rated load, possesses the degree of load handling reliability inherent in compliance with Guideline 7:
- a. Comparison of structural welding standards used with the requirements of AWSD 14.1
- b. Comparison of gear design criteria with the requirements of the AG4A standards invoked by CMAA-70.
- c. Comparison of' gantry leg structural design rules with those required by CMAA-70.
While this information is necessary to establish full compliance with Guideline 7, it should be noted that the basic issue is the approach to allowable stresses and, consequently, its resolution is not judged to be crucial with respect to the reliable handling. of smaller loads, such as the refueling transfer cask, where actual stresses are substantially less than' design values.
2.2 INTERIM PROTECTION MEASURES The NRC has established six interim protection measures to be implemented at operating nuclear power plants to pr, ovide reasonable assurance that no heavy loads will be handled over the spent fuel pool and that measures exist to' reduce the. potential for accidental load drops to impact on fuel in the core or spent fuel pool.
Four of the six interim measures of the report consist of Guideline 1, Safe Load Paths; Guideline 2, Load Handling Procedures; Guiceline 3, Crane Operator Training; and Guideline 6, Cranes (Inspection, Testing, and Maintenance).
The two remaining interim measures cover the following criteria:
1.
Heavy load technical specifications 2.
Special review for heavy loads handled over the core.
Licensee implementation of these interim protection measures is stated and evaluated in the succeeding paragraphs of this section.
l
- O Udd Franklin Research Center A em w w r,e u.
l4 -....
....z....-
- ~ -..
y...
_. w
_.., ~ -. -
n.
^
,, ~...
.z w 2.7
- p. 4- ;
p,.
.,,,s.
=..a,.. _ _.. ;, n...._,.,.
- y -..,
,, y3..
..n. v.
-.~
7, y....
.-. ~, _. ~,
TER-CS257-440 2.2.1 ' Technical Soecifications IInterim Protection Measure 1, NURE -0612, Section 5.3(1)}
"Licenscc for all cperating reacters not having a single-failure-proef eve head crane in the fuel storage pool area shculd be revised to include a specificaticn cocparable to Standard Technical Specification 3.9.7,
' Crane Travel - Spent Fuel Storage pool Building,' for WR's and Standard Technical Specification 3.9.6.2,
' Crane Travel,' fo; B'4R's,'tC prohibit
~
handling cf heavy loads over fuel in the storage pool until implementation of measures which satisfy the guidelines of Section 5.1."
a.
Summarv of Licensee Statements and Conclusions The Licensee has stated that to perform refueling activities the fuel transfer cask must be moved over the reactor vessel and the fuel pool.
b.
Evaluation The Big Rock Point plant has no technical specification prohibiting the handling of heavy loads over spent fuel in the storage pool because of the' unique refueling process.
Refueling procedures require the movement of a 24-ton fuel transfer cask over fuel'in the storage pool in-order to transfer
--fu el.
Further, the reactor building crane is._not qualified as single. failure-.
y..c.. -
proof per NURE -0554.
The 1.icensee has addressed the single-failure-proof
~
requirements of Interim Protection Measure 1 for the movement of the 24-ton cask through the use of a safety sling assembly between the crane and the cask to prevent dropping of the load.
The safety sling assembly is designed to a.
provide a redundant safety feature to mitigate the effects of a cask drop due to a crane failure.
An engineering evaluation indicates that the safety sling asse=bly can be expected to terminate a spent fuel transfer cask drop following a failure of the hoist drive train.
This drop termination is expected to take place af ter a short free fall and to not impose unacceptable loads on the safety cables or crane structures. The Licensee has agreed to provide a comprehensive maintenance and test program suitable to ensure that the safety sling assembly will be operable and able to perform within design limits whenever the l
l g
l
$0 Franklin Research Center A Co a cc of *he Frannan insonAs
TER-C5257-440 transfer cask is being handled. This commitment is supported by written procedures which specify inspecting, rigging, adjusting, and testing requirements for the spent fuel transfer cask and safety slings prior to load
- handling. These procedures have been reviewed and found to be acceptable with the exception of adequate cross refere'ncing of the test t;ipping. guidance in MFHS-4.
The Licensee has agreed to revise MFHS-1 to require the safety cling trip test in MFHS-4 each time the spent fuel transfer cask is prepared for load handling.
Based on the foregoing, it is ' judged that design features provided for
~~ ^
the Big Rock Point 75-ton semi-gantry crane, when handling the spent fuel 1
transfer cask, provides an additional level of protection against facility damage due to a cask drop similar to that which would be provided by a crane conforming to the requirements of NUREG-0554 for mechanical system failures.
In the case of the other load bearing members, such as girders, NUREG-0554 does not require redundancy, but provides an additional level of protection through conservative strutural design requirements imposing, as a designloadingcondition{theforcesinducedbya~safeshutdownearthquake (SSE) with the crane carrying its design rated load. This issue is being evaluated separately by the NRC staff.
c.
Conclusions and Recommendations l
The Big Rock Point plant meets the intent of Interim Protection Measure 1 with the exception of resolution of the issue of seismic qualification.
i
- 2. 2. 2' Administrative Controls [ Interim Protection Measures 2, 3, 4, and 5,_
NUREG-0612, Sections 5.3(2) - 5.3(5)1
" Procedural or administrative measures (including safe load paths, load handling procedures, crane operator training, and crane inspection]...
can be accomplished in a short time period and need not be delayed for completion of evaluations and modifications to satisfy the guidelines of Section 5.1 (of NUREG-0612]. "
f 4 U000 Franklin Research Center 4om.onor w r,m aw.au.
I
_ _. ~.... _ -..
TER-CS257-440 a.
Summary of Licensee Statements and Conclusions Summaries of Licensee statements and conclusions are contained in discussions of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.
b.
Evaluations, Conclusions, and Recommendations o
The evaluations, conclusions, ard recommendations are contained in discuss'ons of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.
2.2.3 Special Reviews for Heavy Loads over the Core (Interim Protection Measure 6, NUREG-0612, Section 5.3(6)]
"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 foll'owing for these loads:
(1) review of procedures for installation of rigging or lif ting devices and movement of the load to assure that sufficient detail is provided and that instructions are clear and concdse; (2) visual inspections of load bearing components of cranes, slings, and special lif ting 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."
6 l
a.
Summarv of Licensee Statements and Conclusions The Licensee has stated that the interim actions. specified in this-interim protect. ion measure have been completed according to the Licensee's understanding of the measure, except for the specific requirements associated with crane operator visual examination.
b.
Evaluation The Big Rock Point plant complies with Interim Protection Measure 6 of NUREG-0612 on the basis of the Licensee's certification of compliance. However, 4 Ed Franklin Research Center A c==.an er w r, n w.
~ ~
~
j TER-C5257-440 the Licensee should ensure that aopropriate, documentation is, available to document this one-time detailed review of procedures, equipment, and personnel training ir.volved with the handling of heavy loads over the core.
4 e
I
.i l!
~. -
i i
c 9
t i
4 i
~
I l 4
i Udl.fd Franklin Research Center A W of The Frenen Inseture l
r.
TER-C5257-440 3.
CONCLUDING
SUMMARY
" tis summary is provided to consolidate the conclusions and recommenda-tions of Section 2 and to document the overall evaluation of the handling of heavy loacs at Big Rock Point Nuclear' Plant.
It is divi (ed into. two sections, one dealing with general provisions for load handling at nuclear power plants (NUREG-0612, Section 5.1.1) and the other with staf f recommendations for interim protection (NUREG-0612, Section 5.3) pending complete implementation of the NUREG-0612 guidelines.
In each case, recommendations for additional, Licensee action, and additional NRC staff action where appropriate, are provided.
3,1 GENERAL PROVISIONS FOR LOAD HANDLING The NP.C staff has established seven guidelines concerning provisions 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.- Compliance with these guidelines is necessary to ensure that load,
handling system design, administrative controls, and operator training and qualification are such that the possibility of a load drop is very small for the critical functions performed by crepes at nuclear power plants.
These -
j guidelines are partially satisfied at Big Rock Point Nuclear Plant. This conclusion is presented in tabular form as Table 3.1.
Specific recommendations for achieving full compliance with these guidelines are provided as follows:
Guideline Recommendation 1
s.
Identify safe load paths for the spent fuel transfer cask during use of the auxiliary hook and movements from its stand.
b.
Verify that safe load paths are marked to provide visual references for both operators and supervisory personnel.
c.
Clarify the procedure for handling deviations from safe load pa ths.
2 a.
Verify that procedures contain sufficient detail to edequately define safe load handling (i.e., identification of equipment, inspection and acceptance criteria, sequence of steps, safo load l
paths, and precautions).
1 !!C Franklin Research Center o~ on at The Fr.n.a.n in
3 Table 2.1 alg Rock relat/NUREG-0612 Compliance Matris az Neight Interin Int,e r le
- =
or Guideline 1 Guideline 2 Guideline 3 Guideline 4 Guideline 5 Guideline 6 Guideline 7 Measu o 1 Measure 6 b]1
]lj Capacity Safe load Crane Operator Special Lifting Crane - Test Technical Special Weavy toad s (tone)_
Pathe _ Procedures Treintne
_ Devices Silpas and Inspectlan Crane Deston frecjficatione Attention ic 25
- 1. Deactor Crane 75 C
P I
~
r
- a. Shield Flug 75 P
P lq IEC
~
~
C NC C
e g8
- b. Core Spray 8
P P
Line Shield ggn
=3
- c. Core Spray
<1 P
P IEC
~
C
'4 Line s
leC
- d. Vessel Beed 1
P P
C Insulation
- s. Beactor 24 P
P C
C Vessel Head
- f. Puel Trans-24 P
C C
~
P C
1 for Cask DJ m
I
- g. Refueling 0.5 P
P IIC C
Platfore
- h. ISI Inspec-0.5 P
P IIC C
tion Device i
- 1. Pilter Sock 2
P P
NC C
Cask
- j. T-II Cask 7.5 P
P leC
- k. Cobalt Cask 15 9
P
- IOC e
- 1. Stud 3
P
,P 30C C
+
Tensioner f
l C = Licensee action compiles with NUREG-0612 Guideline.
NC = Cicensee action does not compir with WunsG-0612 Guideline.
A = Licensee has proposed revletons/modificatione designed to comply with NUREG-0612 Guideline.
b 1 = Insufflatent information provided by the Licenees.
- = Not applicable.
M P = Licensee action partia!!y comp!!as with the NURSG-0612 Guideline.
M (si 4
I 6
O a
e l
l I
l e
e f5L
, ' "a Table 3 1 (Cont.)
5 k
Wolght Interla Intette
- a. 5' or caldettaa 1 Guideline 2 ouldeline 3 ouldeline a cuideline s cuideline a culdeline 7 nose.te 1 ees.eure s
-4 m Capacity Safe load Csane Operator Special Lifting Crane - Test Technical Special h Neavy inade (tonal PAthe Procedure L Trainine Devlees 5 tines and Inspection Crane Daejm 1:cecif f rat ima Attention 34 (a7
- e. Vessel Head 1
P P
MC C
Q 8tand I
En k
- n. Spent Puel 1-10 P
P IIC
-~
~
Rack O. Puol Ship-2 P
P MC ping Con-toinere
- p. RCP Matchee 16 P
P IIC 9
- q. Came
<1 P
P IIC N
4
- r. Puel Ship-se P
P I
IIC
=
ping Caek
- a. Recircula-4 P
P tion rumpo I
IIC C
- t. Mino. Puel
<0.3-1 P
P IIC C
R&D Toole
- u. Misc.
40.3-1 P
P IIC C
Internale
- v. weste caske
<1-25 P
P IIC C
.and Liners w
- w. Iced Blocks 1
P P
I C
2.
Cleanup Holst 3
C P
- a. Deminere-3 P
P IIC M
t.
Ilser Flug M
q 8
3.
RDS Holst 2
A C
s P
o
- a. sRV
<1 P
P gIC e
S
+
4 Er
'51 L=
im r
I$y b
. 9, 3 fe7 Table 3.1 (Cont.)
nu
{M 5h weight Interla Interle
{n or Guideline 1 Guideline 2 Guideline 3 Guideline 4 culdeltne 5 Guideline 5 Guldettne 7 Measure '
seesente 4 E@
Capacity Sete imod Crane Operator Special L!tting Crane - Test Technicek Special
- p Meevy inade (t onal Pathe Procedures Traintne Devices _
Allege _
and inspection Crane D*alqvi Frecjfications Attentloa s
4, 8 sty Holst 0.3 C
P
- e. SRV (1
P P
IIC f.
5.
Turbine Crane 25 C
P.
I
~
- s. Esciter 7
P P
IIC
~
~
8
- b. Co.inge 10 F
P' HC I
- c. Condeneste 5
P P
IIC Pump Co.-nent.
MC
- d. Feed Pump 7
P P
Componente
- e. Turbine le P
P NC ustches i
e 6.
Deergency 1
P P
IIC P
Condenser Bene 8
a vi N
ui M
8 3
6 I
i 1
I O
l I
- i e
a l
b e
y I i
-.... + _. - -...
.-4-.
e
~.
TER-C5257-440~
i Guideline Recommendation b.
Revise MFHS-1 to require the safety sling trip test in MFHS-4 each time the transfer cask is prepared for move': tent.
3 (Tne Big Rock Point plant complies with this guideline.)
4 a.
Conduct a point-by-point review of all special lifting deyices against the criteria of ANSI N14.6-1978 b.
Verify that ANSI N14.6-1978 requirements are' imposed on subcontractors providing special lifting devices.
5 a.
Verify that slings are installed and used in accordance with ANSI B30.9-1971.
b.
Verify that the load used in selecting and marking the slings is based upon the sum of the maximum static and maximum dynamic
- loads, c.
Verify that slings restricted in use to certain cranes are clearly marked to indicate the crane (s) with which they may be used.
d.
Mark the NX01 slings' to indicate the maximum static and dynamic.
loads or to indicate exclusive use for the 24-ton spent fuel transfer cask and RV bead.
6 Provide initial load test data on all cranes subject to the general guidelines of NUREG-0612.
Provide information concerning the crane design criteria for the 7
a.
turbine crane.
b.
Provide additional crane design information for the 75-ton semi-g'antry in the reactor building comparing welding standards used to AWS D14.1, gear design to to AGMA standards, and gantry leg structural design to CMAA-70.
In addition, the Licensee should review load handling systems at Big Rock Point for NUREG-0612 applicability and modify the heavy load handling program to include the screen house trolley and equipment lock crane.
4 Udd Franklin Research Center A Dma on of The Frnruen insmute
TER-CS257-440 i
l 3.2 INTERIM PROTECTION The NRC staff has established (NUREG-0612, Section 5.3) certain measures tha t should be initiated to provide reasonable assurance that handling of heavy loads will be performed in a safe manner until implementation of the general guidelines of NUREG-0612, Section 5.1 is complete.
Specified. measures include the implementation of a technical specificatica to prohibit the handling of heavy loads o'ver fuel in the storage pool; compliance with Guidelines 1, 2, 3, and 6 of NUREG-0612, Section 5.1.1; a review of load handling procedures and operator training; and a visual inspection program, including component repair or replacement as necessary of dranes, slings, and special lif ting devices to eliminate deficiencies' that could lead to component failurel This evaluation of information provided by the Licensee indicates that the following actions are necessary to ensure that the staff's measures for interim protection at Big Rock Point Nuclear Plant are met:
Interim Measure Recommenda tion 1
Full compliance is contingent on resolution of seismic qualification.
2, 3 Implement the recommendations of Guidelines 1 and 2 identified in Section 3.1.
4 (The Big Rock Point plant complies with this interim protection measure.)
5 Implement the recommendations of Guideline 6 identified in Section 3.1.
l 6
(The Big Rock Point plant complies with this interim protection' measure.)
3.3
SUMMARY
l NRC's general guidelines and interim protection measures of NUREG-0612 1
I have been partially complied with at Big Rock Point Nuclear Plant. Several programs have been installed which comply with NRC staff guidelines, inclu' ding operator training and interim inspections.
In order for the Licensee to fully comply with NUREG-0612, Licensee action is required on the remaining general guidelines and interim actions, nklin Research Center A Dms on of The F.ansen ansonee
TER-CS257-440 e...
4.
REFERENCES 1.
" Control of Heavy Icads at Nuclear Power Plants" NRC, July 1980 NUREG-0612 2.
V. Stello, Jr. (NRC)
Letter to all licensees.
Subject:
Request for Additional Information on Control of Heavy Loads Near Spent Puel NRC, 17 May 1978 3.
NRC Letter to Consumers Power Company (CPC)
Subject:
Request for Review of Heavy Load Handling at. Big Rock Point 22 Deceter 1980 7
4.
G. C. Withrow (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Control of Heavy Loads - Big Rock Point 10 June 1981
~5.
D. P. Hof fman (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Control of Heavy Loads - Big Rock Point 1 July 1981 6.
T. C. Bordine (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Control of Heavy Loads - Big Rock Point 23 September 1981 7.
M. J. McMahon-(Whiting Corp.)
Letter to J. R. Schaub (CPC)
Subject:
Specification Comparison, Report 5 March 1982 8.
P.
P. Steptoe (CPC)
Letter to R. Bachman (NRC)
Subject:
Response to Interrogatories 10 and 11 (Spent Fuel Modification) 19 April 1982 9.
Joseph Gallo (CPC)
Letter to R. Bachman (NRC)
Subject:
Response to Interrogatory 12 (Spent Fuel Modification) 26 April 1982 l
. idif Franklin Research Center A Cwasen of The Frarwan ensatute
~
TER-C5257-440 t...
10.
P. Steptoe (CPC)
Le t ter to P Ba rt.n
{ND.C)
Subject:
Response to Draft Control of Heavy Loads TER for Big Rock Point 30 April 1982 11.
J. Gallo Letter to R. Bachman (NRC)
Subject:
Response to Interrogatory 12 (Spent Fuel Modification) 2 May 1982
~
12.
" Overhead and Gantry Cranes" American National Standards Institute, New York ANSI B30.2-1976 13.
" Standard for Lif ting Devices for Shipping Containers hTeighing 10,000 Pounds (4500 kg) or More for Nuclear Materials" American National Standards Institute, New York ANSI N14.6-1978 14.
" Slings "
American National Standards Institute, New York ANSI B30.9-1971 "SpecificationsforElectricdverheadTravelingCranes" 15.
Crane Manufacturer's Association of America, Pittsburgh, PA, 1975 CMAA-70 16.
CPC Report on the Analysis and Evaluation of the Consequences of Postulated Fuel Cask Drop Accidents at Big Rock Point t
1 July 1974 17.
" Specifications on Electric Overhead Cranes for Standard Service" EOCI 49 L
v.
18.
G. C. Withrow (CPC)
Letter to D. M. Crutchfield (NRC)
Subject:
Study of 24-ton Spent Fuel Transfer Cask Redundant Support i
System (MPR Associates)
- 5. December 1980 1
. 200!! Franklin Research Center f
A Ohemen of The Frannha nnaaewie e
i