ML20086N079
| ML20086N079 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 01/31/1984 |
| From: | Sekot J EG&G, INC. |
| To: | |
| Shared Package | |
| ML20086N084 | List: |
| References | |
| CON-FIN-A-6457, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8402170216 | |
| Download: ML20086N079 (50) | |
Text
________________ _____ _ __ _ ______
CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS ~
Grand Gulf Nuclear Station Unit 1 (PHASE II)
Docket No. 50/416 Author J. P. Sekot
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Principal Technical Investigator T. H. Stickley Published January 1984 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6457 XA Cop Has Been Sent is PDR Akt>%\\%Mb N
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ABSTRACT The Nuclear Regulatory Commission (NRC) has requested thr.t all nuclear plants, either operating or under construction, submit a response of consistency with NUREG-0612, "Co:strol of Heavy Loads at Nuclear Power Plants." EG&G Idaho, Inc., has contracted with the NRC to evaluate the responses of those plants presently under construction. This report contains EG&G's evaluation and recommendations for Grand Gulf Nuclear Station Unit I for the requirements of Sections 5.1.2, 5.1.3, 5.1.5, and 5.1.6 of NUREG-0612 (Phase II). Section 5.1.1 (Phase I) was covered in a separate report [1].
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EXECUTIVE
SUMMARY
Grand Gulf Nuclear Station Unit 1 is consistent with the guidelines of NUREG-0612.
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CONTENTS ABSTRACT.............................................................
11 EXECUTIVE
SUMMARY
iii 1.
INTRODUCTION....................................................
I 1.1 Purpose of Review.........................................
I 1.2 Generic Background........................................
I 1.3 Pl a n t-S peci fi c Bac kg round.................................
3 1.
EVALUATION AND RECDMMENDATIONS..................................
4 i
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2.1 Overview..................................................
4 l
1 1
2.2 Heavy Load Overhead Handling Systems......................
4 2.3 Guidelines................................................
4 3.
CONCLUDING
SUMMARY
22 3.1 Guideline Recommendations.................................
22 3.2 Addi ti onal Recommendati on s................................
22 3.3 Summary...................................................
22 4.
REFERENCES......................................................
23 j
TABLES 2.1 Nonexempt Heavy Load-Handling Systems...........................
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APPENDIX A-24 e
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CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS Grand Gulf Nuclear Station Unit 1 (PHASE II) 1.
INTRODUCTION 1.1 Purpose of Review i
This technical evaluation report documents the EG&G Idaho, Inc.,
j review of general load-handling policy and procedures at Grand Gulf Nuclear Station Unit 1.
This etaluation was performed with the objective of assessing conformance to the general load-handling j
guidelines of NUREG-0612, " Control cf Heavy Loads at Nuclear Power Plants" [2], Sections 5.1.2, 5.1.3, 5.1.5, and 5.1.6.
This constitutes Pha;e II of a two phase evaluation.
Phase I assesses conformance to Section 5.1.1 of NUREG-0612 and was document?d in a separate report [1].
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1.2 Generic Background l
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Generic Technical Activity Task A-36 was established by the U.S.
Nuclear Regulatory Coramission (NRC) staff to systematic-lly examine staff licensirig criteria and the adequacy of measures in effect at operatir.g nuclear power plants to assure the safe handling of heavy l
loads ano to recommend necessary changes to these measures. This activity was initiated by a letter issued by the NRC staff on May 17, 1978 [3], to all power reactor applicants, 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 bandling of heavy loads at operating plants, although providing protection from certain potential problems, do not adequately cover the major causes of load-handling accidents and should be upgraded.
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In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two phase objective using an accepted approach or protection philosophy. The first portion.of the objective, achieved through a set of general guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load-handling systems at nuclear power plants are designed and operated such that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. The second portion of the staff's objective, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ensure that, for load-handling systems in areas where their failure might result in significant consequences, either (a) 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 (b) conservative evaluations of load-handling accidents indicate that the potential consequences of any load drop are acceptably small. Acceptability of accident consequences is quantified in NUREG-0612 into four accident analysis evaluation criteria as follows:
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" Releases of radioactive material thct 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);
" Damage to fuel and fuel storage racks based on calculations-o 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; o
" 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.
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__._ 7 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 o
" 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."
The approach used to develop the staf* guidelines for minimizing the potential for a load drop was based on defense in depth. This plan includes proper operator training, equipment design, and maintenance coupled with safe load paths and crane interlock devices restricting movement over critical areas.
Staff guidelines resulting from the foregoing are +.abulated in Section 5 of NUREr 0612.
1.3 Plant-Specific Backcround On December 22, 1980, the NRC issued a letter [4] to Mississippi Pcwer and Light Company, the applicant for Grand Gulf Nuclear Station Unit I requesting that the applicant review provisions for handling and control of heavy loads at GGNS Unit 1, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guid,Iines. MP&L Co. provided responses to this request on November 23, 1961 [8], May 4, ISd2 [5], and November 19, 1982 [6]. EG&G performed an evaluation of the information and issued the Phase I report [1].
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2.
EVALUATION AND RECOMMENDATIONS 2.1 Overview The following sections summarize MP&L Co.'s review of heavy load handling at GGNS Unit I accompanied by EG&G's evaluation, conclusions, and recommendations to the applicant for making the facilities more consistent with the intent of NUREG-0612.
2.2 Heavy Load Overhead Handling Systems Table 2.1 presents the applicant's list of overhead handling systems which are subject to the criteria of NUREG-0612.
The applicant has indicated that the weight of a heavy luad for the facilities as 1140 pounds per the NUREG-0612 definition.
2.3 Guidelines Section 5.1.1 of NUREG-0612 includes general guideline for: (1) Safe load paths, (2) procedures, (3) crane ope:rator training and qualification, (4.) special lifting devices, (5) lifting devices that are not specially designed, (6) crane inspection, testing, and maintenance, and (7) crane design. These guidelines were addressed by the applicants and evaluated in a separate report [1].
2.3.1 Spent-Fuel Pool Area [NUREG-0612, Article 5.1.21 (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 provided:
(a) Mechanical stops or electrical interlocks should be provided that prevent movement of the overhead cranes 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 4
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TABLE 2.1.
OVERHEAD HANDLING SYSTEM SUBJECT TO CRITERIA 0F NUREG-0612 GRAND GULF NUCLEAR STATION UNIT 1 Capacity Handling System (tons) location Containment polar crane /
125/35 Containment auxiliary hoist Spent fuel cast crane 150 Auxiliary building New fuel bridge crane 5
Auxiliary Building Monorail for LPCS and 10 Auxiliary Building (el. 139 ft)
RHR "C" hatches f
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9.
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 interlocks should be verified to be in niace 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 stop: or electrical interlocks should be provided to prec1rde crans travel from areas where a postulated load <
aculd damage equipment from redundant or alte.
afe shutdown paths.
(e) Analyses should confoin.
.he guidelines of Appendix A.
93 (3) "Each of the following are provided (Note: This alternative is similar to (1) 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 large enough to maintain wide separation between the load and the " hot" spent fuel.):
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(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 off-site doses less than 1/4 of 10 CFR Part 100 if rele sed, and such bypassing should require the approval from the shift supervisor (or other designated plant management individual). The mechanical stops or 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.
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(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.
03 (4) "The effects of drops of heavy loads should be analyzed and shown to satisfy the evaluation criteria of Section 5.1 of this report. These analyses should conform to the guidelines of Appendix A."
i A.
Summary of Applicant's Statements At Grand Gulf Unit 1, there are two spent fuel storage pools, one in the auxiliary building at the 208' el. capable of storing up to 158% of a full core and one in containment capable of storing up to 21% of a full core. The new Fuel Bridge Crane with a capacity of 5 tons is capable of carrying loads over the spent fuel pool in the auxiliary building and the Polar Crane which is equipped with main and auxiliary hoists with capacities of 125~and 35 tons, is capable of carrying loads over ~the spect fuel _ storage area and over the Reactor Vessel in containment. -The spent fuel cask crane, located in the auxiliary building is restricted, by-limits of cask crane travel from carrying loads over the spent fuel pool. The cask crane rails do not extend over any portion of the spent' fuel pool; thus, the cask cannot be transported over the spent fuel storage racks. The crane 8
lift system has a dual load path design, with the exception of the main drum (the failure of which is not considered credible), so that'no single component failure will result in a cask drop. There are one-half ton capacity jib cranes in both the auxiliary building and in containment that are apable of carrying loads over the spent fuel pools, however, these cranes are used only to carry loads that do not qualify as " heavy loads." The only other handling system in containment capable of moving loads over the vessel is the refueling platform. No " heavy loads" are handled by this system. MP&L Co. did not feel it necessary to evaluate either of the two cranes identified (New Fuel Bridge Crane and Polar Crane) against the criteria of NUREG-0612, Section 5.1.6.
Both procedural restrictions and technical specifications have been developed to prevent carrying heavy loads over spent fuel in the racks in the two storage pools. Nonetheless, the pool gates (weight approximately 3.5 tons) must be lifted in the pools.
Therefore, Structural Analyses were performed to determine compliance with NUREG-0612 Section 5.1 Criteria I and II as a result of a gate drop onto spent fuel ~in the storage racks. Based on the number of fuel rods that could be damaged as a result of this drop, dose calculations w'ere performed for drops in both the containment and auxiliary buildings using the conservative model and assumptions employed by Fuel Handling accidents described in Section 15.7.4 and 15.7.6 of the FSAR. Those analyses indicate a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dose of 8.5 Rem at the site boundary. This slightly exceeds the NUREG-0612_ limit of 6.25 Rem, however, it is well within the 25 Rem limit of
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10 CFR 100. The calculated inhalation doses were only a small fraction of the NUREG-0612 limit of 75 Ren.. -It was determined that the possibility of a K ff greater than-0.95 could not be precluded in the very unitkely event that a gate drop occurred and impacted spent fuel that contained.~
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f sj
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n.
l substantial U-235, such as could be the case in a core off load situation. This is because the Unit I racks rely on spacing alone to prevent criticality. MP&L Co. has initiated action with the objectives of demonstrating that the NUREG-0612 criteria are met.
Every transfer of a heavy load having the potential of a drop with damage in excess of the criteria will be carried out under administrative controls. Within two years, an evaluation will be conducted to determine if any additional measures beyond administrative controls are required [5]
B.
EG&G Evaluation On the basis of the information submitted, EG&G concludes that MP&L Co. identified all cranes and hoists physically capable of carrying loads over spent fuel in the storage pools located in the auxiliary building and in the containment building and they identified those cranes or hoists that should be excluded from further concern because of their inability to carry heavy loads or by physical limitation of crane travel. MP&L Co. has developed Proposed Technical Specifications (3/4.9.7 Crane Travel-Spent Fuel and Upper Containment Fuel Storage Pools and 3/4.9.13 Jib Crane Loading) that prohibit handling of loads in excess of 1140 pounds over irradiated fuel assemblies in the spent fuel or upper containment fuel storage pool racks. [8]
MP&L Co. discussed results of the structural analysis task that was performed to determine compliance with NUREG-0612 Section 5.1 Criteria I and II as a result of a sate drop onto spent fuel in the storage racks. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dose of 8.5 Rein at the site boundary does exceed the NUREG-0612 limit of 25% of the 10 CFR 100 25 Rem limit, however, it is still only 34% of this limi;. MP&L Co.'s analysis also determined that the possibility of a K,ff
>0.95 could not be precluded in the very unlikely event-that a gate drop occurred and impacted spent fuel that 10 l
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contained substantial U-235. Because of the conservations employed in the calculations, unique conditions that must exist at the time of a load drop, and the administrative controls that are to be employed, it is judged that actions identified by MP&L Co. are consistent with the intent of the NUREG-0612 guideline.
C.
EG&G Conclusions and Recommendations On the basis of the information submitted, EG&G concludes that Grand Gulf Unit 1 is consistent with the intent of NUREG-0612, Guideline 5.1.2 (Spent Fuel Pool Area).
2.3.2 Reactor Building [NUREG-0612. Article 5.1.31 (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.
S (2) " Rapid containment isolation is provided with prompt automatic actu? tion 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 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.
(3) "The effects of drops of heavy 1 cads should be analyzed and shown to satisfy the evalisation 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 i
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shutdown.
In this analysis, credit may be taken for containment isolation if such is provided; however, analyses should establish adequate detection and isolation time.
Additionally, the analysis should conform to the guidelines i
of Appendix A."
A.
Summary of Applicant's Statements The handling systems within containment physically capable of handling loads over the Reactor Vessel are the Containment Polar Crane and the Refueling Platform which is used for refueling operations. No " heavy loads" are handled by the Refueling Platform. MP&L Co. did not feel it i
necessary to evaluate the Polar Crane against the criteria of NUREG-0612 Section 5.1.6.
Instead, MP&L Co. performed analysis to demonstrate compliance with NUREG-0612 Section 5.1 Criteria I through III. Analyses were performed to determine the structural consequences of dropping the vessel head on the shroud head assembly during maintenance operation. The consequence of dropping the steam dryer assembly can be extrapolated from the analysis of the shroud head assemb1'y drop, since the steam dryer drop would generate less kinetic energy than the shroud head assembly drop and the impacted structure would be the same in both Results of these analyses show that no damage to cases.
fuel rods or rsles:e of radioactive material is expected.
l Drops of the dry well head onto the. vessel head and the portable radiation shield onto the separator were also evaluated. These drops were bounded by the GE head drop and dryer drop analysis. MP&L Co. concluded that NUREG-0612 criteria I-III are met for all postulated drops into the reactor well. The crane load blocks have not been included in any of the heavy load drop evaluations because the load block is used for handling numerous loads, including the j
Reactor Vessel head,..., and moisture separator. In moving these loads, the ' hook, load block,..., and other load bearing members are subjected to significant stresses 12
approaching the load rating of the crane. By comparison, these components are subjected to a considerably smaller load when only the hook and load block are being moved.
Based on this, it is not considered feasible to postulate a random mechanical failure of the crane load bearing components when moving either the main hoist or auxiliary hoist load block without a load. The only two feasible failure modes for dropping of the main hook and load would be:
1.
A control system or operator error resulting in hoisting of the block to a "two blocking" pcsition with continued hoisting by the motor and subsequent parting of the rope. (This situation can be prevented by operator action prior to "two blocking" or by an upper limit switch to terminate hoisting prior to two blocking); and 2.
Uncontrolled lowering of the load block due to a failure of the holding brake to function (the likelihood of this can be made small by use of redundant brakes).
With the provisions described in Reference 5 (two redundant and diverse upper limic switches, two holding brakes, and loss of power provisions for each hoist), it is concluded that a drop of the load block and hook is of sufficiently low likelihood that it does not require load drop analysis.[5]
Loads only lifted over the vessel when the reactor vessel head or moisture separator is in place were not considered as loads that could potantially drop into the core. These.
are: the dry well head and the steam dryer.
No administrative control are required to enforce this 13 l
situation, because i+ is physically impossible to disassemble or reassemble the reactor such that these loads would be carried over an open vessel.
In addition, the portable radiation shield is installed in the reactor well after the head has been removed, bat before the dryer or separator has been removed.
It is removed from the reactor well after the dryer and sepsrator have been installed.
This sequencing is enforced by written procedures governing the installation snd removal of the radiation shield and will be strictly enforced by individuals in charge of lifts j
by the Polar Crane.
B.
EG&G Evaluation On the basis of the information submitted, EG&G concludes l
that MP&L Co. identified all cranes and hoists physically capable of carrying loads over the Reactor Vessel and identified those cranes or hofsts that should be excluded from further concern because of their inability to carry
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heavy loads. MP&L Co. chose to analyze the effects of heavy load drops in the Reactor Building to demonstrate compliance with NUREG-0612 Section 5.1 Criteria I-III rather than evaluate the cranes or hoists against the criteria of NUREG-0612 Section 5.1.6.
The only identified case where reliance is placed on the installation and use of electrical interlocks or mechanical stops is that of the hoisting and lowering of load blocks.
EG&G concludes that the redundant features prcvided (upper geared limit switch, upper trip bar switch, two holding brakes, each rated at 150% of full motor torque, and loss of power provisions), are adequate for the mitigation of the occurrence of a "two blocking" or " uncontrolled lowering of i
the load" situation.
EG&G also concludes that a drop of the load block and hook is of sufficiently low probability of 14
occurrence because of the extremely high factor of safety for this situation that it does not require load drop analysis.
No mention was made or credit taken by the applicant with regard to reliance placed on operation of a standby gas treatment system. The only identified case where some reliance is placed on site specific considerations (e.g.,
refueling sequence) is the installation and removal of the portable radiation shield in the Reactor Vessel. MP&L stated that "this sequencing is enforced by written procedures... and will be strictly enforced by individuals in charge of lifts by the Polar Crane."[5] MP&L Co., in their response to Phase I, categorized loads handled by the Polar Crane into various safety classes specifying procedural restrictions in handling loads over spent fuel, the reactor vessel or safe shutdown equipment. Both procedural restrictions and technical specification have been developed to prevent carrying heavy loads over spent fuel in the two :,.arage pools.
In addition, MP&L Co. stated "Every transfer of neavy loads having potential of a drop with damage in excess of the NUREG-0612 (5.1) Criteria will bectrriedoutunderadministrativecontrols."[5]
In their response [5], MP&L Co. described the assumptions, methods of analysis, and conclusions of the load drop analysis of Reference [7]. No reference was made by MP&L Co. of any exception taken to the guidelines of NUREG-0612 Appendix A.
MP&L reported conclusions reached with regard to analysis conducted to determine potential radiological releases and criticality effects and identified administrative action toward compliance with NUREG-0612 criteria.
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_ -. = - -
C.
EG&G Conclusions and Recommendations On the basis of the information submitted, EG&G concludes that Grand Gulf Unit 1 is consistent with the intent of NUREG-0612, guideline 5.1.3 Reactor Building.
2.3.3 Other Areas [NUREG-0612, Article 5.1.5]
i (1) "If safe shutdown equipment are beneath or directly adjacent to a potential travel load patn 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 satis'ying 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; t
(b) If the load drop could impair the operation of equipment or cabling associated with redundant or dual safe shutdown paths, mechanical stops 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 ut be taken for intervening floors unless justified by analysis.)
(c) The effects of 104.1 drops have been analyzed and the results indicate that damage to safe shutdown equipment would not preclude operation of sufficient equipment to achieve safe shutdown. Analyses should conform to the-guidelines of Appendix A,-as applicable.
i (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 toad-handled by the handling system would not penetrate the ceiling or cause spalling that could cause failure oi' the safe shutdown equipment "
i A.
Sumrsry of Applicant's Statements l
The handling systems of interest are the Contcinment Polar.
Crane, the New Fuel Bridge Crane, and the LPCS-and RHR 'C' l=
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Equipment and Hatch hoist. MP&L Co. did not feel it necessary to evaluate any of these Handling Systems against the criteria of NUREG-0612, Section 5.1.6.
For these handling systems, a combination of systems and structural evaluations were utilized to determine if Criteria III l
and IV of NUREG-0612 (5.1) are met for all postulated load l
drop scenarios. To assist these evaluations, a set of safety functions were identified corresponding to these l
l criteria. The goal of these evaluations was to demonstrate that the applict.ble safety functions could be accomplished for all load drop scenarios. Tables were prepared to provide a region by region presentation of the loads of interest, evaluations undertaken, results of these evaluations and conclusions based on these results. As l
indicated in the tables, Criteria III and IV have been satisfied for all load drops within each region. The basis for acceptance was, in most cases, that sufficient redundancy and separation of equipment required to accomplish the applicable safety function exists or that equipment was not impacted based on structural analysis.
The basis for acceptance in two specific situations was not equipment separation or redundancy. The first relates to cooling fuel in the in-containment storage pool with the Fuel Pool Cooling and Cleanup System (FPCCU) and the second relatas to the RHR piping used for extended core cooling in Region 8.
Calculations were performed to determine how much time would be available to effect repairs and restore cooling in the event that all cooling to the pool was lost.
Based on the results of this analysis, MF&L Co concluded that there is reasonable assurance that cooling of fuel in the in-containment racks can be accomplished following any load drop scenario in the containment. Load Drop Analyses were perfomed td determine structural damage to the compartment floors above-the steam tunnel. Based on this evaluation, MP&L Co. concluded that there is reasonable assurance that RHR shutdown cooling could be maintained following any postulated heavy load drop in Region 8.
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B.
EG&G Evaluation MP&L Co. did not feel' it necessary to evaluate any of the cranes listed in Table 2.1 Reference [1] agairst the criteria of NUREG-0612 Section 5.1.6 and instead performed a combination of systems and structural evaluations to d+mennrate compliance with NUREG-0612 Section 5.1 Criteria III and IV for all postulated load drop scenarios.
MP&L Co.'s goal was to demonstrate that applicable safety functions could be accomplished for all cases. MP&L Co.
provided a description of the methodology used in their systems and structural evaluation and it is EG&G's understanding that development of a load drop matrix was part of the systems evaluation. MP&L Co. did not present the load drop matrix in their report, however they did present the results of their systems evaluation, and their report indicates that MP&L Co. did perform an in depth tudy to demonstrate compliance with the NUREG-0612 criteria, which is the real objective of this task.
In order to demonstrate that safe shutdown, long term cooling, and fuel pool cooling could be achieved and/or maintained in the event of postulated load drops, MP&L Co. identified:
1.
load impact regions (Figure 1-3 Reference [5])
2.
safety functions required to be accomplished for each region (Table 1 Reference [5])
3.
plant systems required to accomplish the identified safety functions (Figure 5-P Reference [5])
4.
equipment associated with those systems or their support systems that could potentially be lost if a load drop were to occur in the region, and 5.
resultant effects of the_ loss of this equipment on the ability to accomplish the identified safety function.
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MP&L Co. presented the results of their evaluation in Tables 2-15[5]. The results were well presented and provide a region-by-region presentation of the loads of interest, evaluations undertaken, the results of the evaluations, and conclusions based on these results. The above noted figures and tables of Reference [5] are included in this report as Appendix A.
It should be noted that the structural analysis of the postulated Reactor Vessel Head and Dry Well Head drops that were to be performed, have been completed and MP&L Co. reported that "results of the analysis indicate that no damage to the equipment inside the dry well is expected, and therefore, the assumptions made in the nine-month report are valid"[6].
MP&L Co., in their Phase I report and in their Systems Evaluation, addressed load / target combinations and provided justification for the exclusion of those combinations on the basis of separation and redundancy of safety related equipment. No combinations were. eliminated because of mechanical stops or interlocks. MP&L Co. does not anticipate that any of the heavy loads handled by the containment polar crane will be lifted until the Plant has been shut down for some time. EG&G concludes that this assumption is valid.
On the basis of the information provided in the description of the Structural Evaluation Methodology, EG&G concludes that the analysis performed by MP&L Co. of postulated load drops and structural response of the impact area is complete and in accordance with the intent of NUREG-0612. No exceptions to the analytical guide 11ne of NUREG-0612 Appendix A 3
were identified by MP&L Co.
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C.
EG&G Conclusions and Recommendations On the basis of the information submitted, EG&G concludes that Grand Gulf Unit 1 is consistent with the intent of NUREG-0612 Guideline 5.1.5, other areas.
2.3.4 Single-Failure-Proof Handling Systems [NUREG-0612, Article 5.1.6]'
(1) " Lifting Devices:
(a) Special lifting devices that are used for heavy loads in the 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 N14.5-1978.
If ~only a single lifting device is provided instead of dual devices, the special lifting device should have twice the design safety factor as required to satisfy the guidelines of 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 particular load have been analyzed and shown to satisfy the evaluation criteria of Section 5.1:
(i) Provide dualuor 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 i
(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 with the implementation guidelines of Appendix C of-this report.
20
(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 i
l 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.
0R (b) A non-redundant or non-dual lift point system should have a design safety factor of ten (10) times the maximum combined concurrent static and dynamic load."
A.
Summary of Applicant's Statements In response to Sections 5.1.2, 5.1.3, and 5.1.5 of NUREG-0612, MP&L Co. stated "It has not been necessary to evaluate any of these systems a p inst the criteria of NUREG-0612 Sectin 5.1.6."
8.
EG&G Evaluation MP&L Co. exercised their option to demonstrate compliance with Criteria I-IV of NUREG-0612, Section 5.1 rather.than evaluate their handling systems against criteria of NUREG-0612, Section 5.1.6.
C.
EG&G Conclusions and Recommendations On the basis of the information submitted, EG&G concludes that MP&L Co. has adequately demonstrated consistency with the 4
intent of NUREG-0612 Guidelines 5.1.2, 5.1.3, and 5.1.5 by exercisir.g permitted options rather than demonstrate consistency with NUREG-0612 Guidelines 5.1.6, Single-Failure Proof Handling Systems.
l l
21 l
1 3.
CONCLUDING
SUMMARY
3.1 Guideline Recommendations None.
3.2 Additional Recommendations None.
3.3 Summary The report provided by MP&L Co. was well organized and complete in fulfilling requirements of Guidelines 5.1.2, 5.1.3, and 5.1.5 of NUREG-0612. Consistency with these guidelines is now satisfied at i
Grand Gulf Nuclear Station Unit 1.
l
' 1 I
I 22 j
4.
REFERENCES 1.
Control of Heavy Loads at Nuclear Power Plants Grand Gulf Nuclear Station Units 1 and 2 (Final--Phase I), Docket No.s 50/415 and 50/417, N. Maringas and T. H. Stickley, March 1983.
2.
NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, NRC.
3.
V. Stello, Jr. (NRC), Letter to all applicants.
Subject:
Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 17 May 1978.
4.
USNRC, Letter to MP&L Co.
Subject:
NRC Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 22 December 1980.
5.
L. F. Dale (MP&L Co.) letter to D. G. Eisenhut (USNRC).
Subject:
Response to Requests for Information in Sections 2.2, 2.3, and 2.4 of Enclosure 3 to NRC December 22, 1980 letter, May 4, 1982.
6.
L. F. Dale (MP&L Co.) letter to H. R. Denton (USNRC).
Subject:
Amended Response to EG&G Comments on Six Month Heavy Load Report, Nine Month Report Load Drop Analysis, November 19, 1982.
7.
NEDC-23566, Structural Analysis of Reactor Vessel and Internals.for i
Vessel Head Drop, Shroud Head Assembly Drop, and Steam Dryer Assembly Drop Conditions.
I 8.
L. F. Dale (MP&L Co.) letter to D. G. Eisenhut (USNRC).
Subject:
Response to Request for Information in Section 2.1 of to NRC December 22, 1980 letter, November 23, 1981.
23
APPENDIX A CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANT GRAND GULF NUCLEAR STATION UNIT 1 Tables and figures included in:
L. F. Dale (MP&L Co.) letter to D. G. Eisenhut (USNRC)
Subject:
Response to requests for information in Section 2.2, 2.3, and 2.4 of December 22, 1980 NRC letter enclosure 3, May 4, 1982.
Figure 1 Plan View--Auxiliary Bldg / Containment, elev. 208' 10" Figure 2 Plan View--Auxiliary Bldg / Containment, elev. 114' 6",
1195 0",
120' 10" Figure 3 Plan Viev--Auxiliary Bldg / Containment, elev. 135' 4", 139' 0",
147' 7" Figure 4 Heavy Load Drop into Region 8 Figure 5 Safety Function No. 1 Spent Fuel Cooling Figure 6 Safety Function No. 2 Extended Core Cooling--Head in Place--Systems Selected for Purpose of Evaluation Figure 7 Safety Function No. 3 Extended Core Coolir.g--Head Removed--System Selected for Purposes of Evaluation Figure 8 Safety Function No. 4 Shutdown Pcwer/Cooldown Systems Selected for Purposes of Evaluation Table 1 Safety Functions--Applicable Load Impact Regions Table 2 Region 1 System Evaluation Table 3 Region 2 System Evaluation Table 4 Region 3 System Evaluation Table 5 Region 4 System Evaluation l
Table 6 Region 5 System Evaluation Table 7 Region 5 System Evaluation Table 8 Region 7 System Evaluation Table 9 Region 8 System Evaluation 4
24 1
1 m
Table 10 Region 9 System Evaluation Table 11 Region 10 System Evaluation Table 12 Region 11 System Evaluation Table 13 Region 12 System Evaluation Table 14 Region 13 System Evaluation Table 15 Region 14 System Evaluation Table 16 Summary of Controlling Structural Behavior Resulting from Postulated Heavy Load Drops i
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COOLir4G DIVISION i i
2 DIVl510H 2 KMPEIMNT PAllis ALTERNATIVL ,.l ALTERNATIVE PATHS PATH $
e RHR "A" RHR '11" Si fuTOOWN LPCS St IUTDOWN LPCI"C"
?,
COOLING MODE COOLil4 MODE o
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RELIEF VALVt-ADS RELIEF VALVE ADS OPERATION OPEllATION E
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- yst.::as shown in the (i.;ure we <md flua aced to evoluole wi=ther sie safely funcilon con te accomp!Isl.ed following f
a poslulated load drop, i.e., all,3g e;s.1 ovoilable to accornpihh Ilie sadely function are not necessurily simwn in the fi<Jwe.
l EXTEt1DED COttE COOLit 4G DIV15 TON I DIVISION 2 PATH 5 4
ALTERf 8.'. live ALTERNATIVE Pali s-PATHS
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COOLlHG MODE COOLING MODE L
I FIGUltE 7 SAFETY FUNCTION NO. 3 EXTEtOED CORE COOLING-HEAD REMOVED SYSTEMS SELECTED FOR PURPOSES OF EVALUATION
- a The systems duiweiin liia f..
e are only finise useil to evalueen whether lie safety function can be accomplished following a postulated lumi drop, l.v.,, il sysseous available lo accornplid Ihm safely funcilon are not necessarily simwn in the figure.
SIAITDOWN i
Ato CORE COOi.itC FROM POwtR l
1 REQUIRED FUNCTIONS T
l M' " I NE DEPRE55URI.
INITIAL CORE EXTENDED
. SCRAM s
i p,,.j.
ZATION COOLNG/ MAKEUP TERS C
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PATHS PATHS l
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FIGURE 8 SAFETY FUNCTION NO. 4 SHUTDOWN IV)WER/ COOT _DOWN SYSTEMS SELECTED FOR PURPOSES OF EVALUATION *
- Te= sy e sh In n. es,. er..s.i, it u d i..voknis. wheit=, n..ae.sy funcil a can t.cco wid=d son iae o pontclated load drap, l ael systeqi..:,. ikdal. to accampildi um solely temcelan er. rwt racessarIly slown in th. figte..
8 3
TA3 LEI RELATED APPLICABl.2 LOAD NRC NO.
SAFETY FUNCTIONS IMPACT REGIONS CRITED.!A
~
I Accomplish Spent Fuel Cooling All Regions ill in the Spent Fuel Pool and the Upper Containment Pool 2
Maintain Extended Core Cooling -
All Regions IV Reactor Vessel Head Bolted in Place 3
Maintain Extended Core Cooling -
All Regions IV Recetor Vessel Head Removed 4
Accomplish Reactor Shutdown, Regions 1-4 IV Depressurization and Core Cocling S
Limit Radiological Dose at Regions I,12 & 13 i
Site Boundary from impact of Spent Fuel to 1/4 of 10 CFh Part 100 i
6 Limit Keff to Less than 0.95 Regions I,12 & 13 II Crushing of Fuel i
l i
A-9
TAIILE 2 FINat AITLICAILE LOADS ARE CONCLU540N titC l
SAFETY HAf0 LING OF ANALYSES I1E5ULTS BASIS FOR
RESPONSE
REG Q
_ FUNCTIONS
_ SYSTEM NTEREST PEftFOftMED ACCEPTAILE
, ACCEPTANCE _ ITEM NO.
REMARKS
- I AS. Dose
- New Fuel 4 Fuel Pool A. Structrual A. Demoge to NUREG-0612 2.2.4 As described in Spent Fuel Bridge Crone Cole Analysis of rods in es guidelines the response Pool
- 4. Criticality Drop Oneo many as 10 con be met to item 2.2.4, (see Table 1) e'FPCCU Rocks (Gole) assemblies acceptable i
SeaFipre1 Filter predicIed solutions are Denim B. Dose Assess-ovoilable.164 flosch rnent B. Colculated possible solu-Covers Whole Body lions will be C. Criticality Duse of 8.5 investigated
. e' Miscel-Evoluotion tem is slight-ond an accept-laneous ly peoter obte soio lon loads as lluvi HUREG.
S7:2..Ied.
much as 0612 timit of 5 sons 025 rem, but well within 10 CFR, Pori 100, limit of
- 1 25 eem.
'o C. A Kegt poeter Ihan 0.95 can-not be pre.
ciuded uruler worst cose
, condillons.
Al. Fuel Pool -
D. Seructural D. Liner pene.
NUftEG-0612 2.4.2 Ceoling Analysis of -
trollon-minor guidelines Drop to Pool leakage only.
we met.
Flow (Gos )
Thereiere, no e
significant
.~
loss of pool inventcry predictei.
4
~
~..
TAIILE 2 (continue <g FINAL APPLICADLE LOADS AHE CONCLitSION NRC SAFE lY HAFOLING OF ANALYSES HESULTS BASIS FOR IIESPONSE REGCN FUNC110NS SYSTEM PITEREST PEftFORMED ACCEPTAILE ACCEPTAIEE 11EM NO._
REMARKS
- 1
- 2, #3, ed #4.
New Fuel e FuelPool E. Systems E. Minor leakage NUHEC-0612 2.4.2 Shutdown md Drldge Crane Gate Evoluosion inom pool waadd # sidelines inveen eliecI ore met.
Core Coating en Safety Func.
eFPCCU Filler tions 2,3, el 4 Domin tiecause (1) there Holcit is no equipment related to these Covers functions Miow e Miscellmeous Ilie puoi that could be 5-ton load domoged and (2) minor weepage woo:d not tesull in flooding that
.: 8.,
could allect
- -4 brooder oreos.
+
e e
4 I
e 4
i TAI 15 3 1
FNAL ARE CONCLUSION t41C APPLICADLE LOADS
. ANALYSES RESULTS 9 ASIS F0ft f1ESPONSE SAFETY HAtiX.ING OF
. (1EGION FLNCTIONS SYSTEM INTElWST PERFOflMfD ACCEPTAILE ACCEPTANCE ITEM NO.
I1EMARKS
- 2
- l. Fuel Pool New Fuel e Fuel Pool Structural Yes, spent fuel NUREG-tMl2 2.4.2 Brid p Crone Gale onelysis of cooling and guidelines Aux. Bldg.
Coolins i
20drel.
5-ton load core cooling are met: no i
floor eFPCCU drop. Anolyzed con be accom-eqdunent
- 2, #3, and #4.
Demin for local md plished.
Impacted at See Figure l Shutdown and Hutch overall floora below.
Cor.s Cooling Covers
- response, o Miscel-loneous 5-son load Y
to TAILE 4 FNAL ARE CONCLUSION ti1C APPLICA0LE LOADS REStA.TS BASIS FOR
RESPONSE
SAFEIY HAteX.ING OF ANALYSES REClON FutCTIONS
_ SYSTEM HTEllEST PEIFOflMED ACCEPTAILE ACCEPTANCE ITEM NO.
IlEMAllKS
- 3
- 1. Fuel Puol How Fuel e New Fuel Systems Yes. spent HUREG EMl2 2.4.2 Aun. Didg.
Conling Bridge Crorm Cantoirwrs Evoluutions fuel cooling guidelines orul core cool-are ont Miscellwwous Hotch
- 2, #3, and #4.
e Miscellaneous ing con la Shutdown wwl loods os accomplisted See Figure 1 Core Cooling nwch as 5 sans s
i TAflLE 5 FNAL APPLICAllt.E LOADS ARE CONCLUSION fitC SAFETY llAtiX.ING OF ANALYSES ITESULTS BASIS F(Mt iiESPONSE FIEGION FttacTIOts '
- SYSTEM, HTEf1EST ITitroitMrD ACCEPTAlll E ACCEPTAtCE ITEM NO.
IIEMATIKS
- 4
- 1. Spent Fuel LPCS &
e Hatch Systems Yes, spen 1 FAJREG-0612 2.4.2 Aum. Ilidg.
Pool Cooling flift "C" Covers Evoluotions fuel cooling geldelines LPCS &
Espipment and core cool-are met ill f1 "C" anilluech e Miscel-ing can he ilusches
- 2,#3,md#4. Manoroil loneous accomplisted Shutdown ed leads us 97 el.
Core Cooling heavy as listel.
10 tons 139 el See Figpro 2 4
?
.M TAllLE 6 FWM.
APPLICABLE LOAOS ARE COW 8.*JSION tutC SAFETY llANDLING-OF ANALYSES RESULTS llAal5 t'OR
RESPONSE
HEClON FUNC100NS SYSTEM N100 TEST pef tFOilMf D.
ACCEPTAlv.E AcrFPTANCE ITEM NO.
REMARKS
- 5-
- 1. Spent Fuel Conseinment Miscel.
A. Systems A. Yes, Spent The intent of 2.4.2 A. Even though Cantoinmeni Cooling Polor Crane lanuus Evoluollen fuel couting NUllEG-0612 TPCCu could
. Esptpment E<pipenent con he oc-guidelines are potentially be Hatch Area Being Removed couplahed.
- met, lostsoin-or installed containnwnt See Figpre 1 of ConIain-fuel seorage ment 208'el.
pool, eines e is i
ecasonalate
- 2 and #3.
A. Systems A. Yes, fit fl HUREC-061I assurawe shot Core Cooling Evoluotion seusdown g>idelines spent fuel cool-a' cooling are nwt-Ing con be con be fil Nt shutdown provided as accomplished. cooling con be deseribed in 4.
maintained.
the response lo NitC lleen 2.4.2.
i' TADLE 7 FINAL 4
APPLICAllLE LOADS AfiE CONCLUSION titC SAFETY HAtellHG OF ANALYSES RESUt.TS DASl5 FOR
RESPONSE
REGION Ft#4CTIONS SYSTEM INTEREST P0f tFOfthf:0 ACCEPTAlll E ACCEPTANCE ITEM NO.
REMARKS
- 6
- 1. Spent Fuel Containment Drywell A. Systems A. Yes, spent The Intent of 2.4.2 A. See REMARKS Drywell Croling Polar Crone Head Evoluollon fuel and flilR HullEG-0612 for Table 6 sluidown e,pddelines Head Starage
- 2.Entended cooling con are mes-B. Assumes that Area Core Cool-B. Structural be accom-spent fuel ed structor al ing, HV Analyses plished RIlit shutdown analysis will l
i See ripre i Head in cooling con iw verify no demoge l
Place moinseined.
to equipment in B. Structural sto drywell, i
onelysis of o dop of the drywell head oneo ils storage loco-tion ut 20lyel will be per-p forened to verifv its
--ae ossunoption that dume.ge
~ to egsipment l
In the drywell within this t
j region would not occur.
s i
TADLE 8 FINAL l
' APPLICABLE LOADS ARE CONCLUSION NRC SAFETY ltAtot_ING OF ANALYSES RESULTS DASIS FOtt
RESPONSE
REClON FUNC'llONS SYSTEM INif.I EST P0ftFOttMED ACCEPTAlliE ACCEPTANCE _ liEM NO.
REMARKS
- 7
- 1. Spent Fuel Centainment.oRWCU A. Systems A. Yes. spent The intent of 2.4.2 A. See REMAltKS Groting Cooling Polar Crnne HX's
.Evoiuusen
- fuel and ftt IR NUltEG-0612 for Table 4 Ar2s-PE shutdown guidelinea guadrone of
- 2 ed #3.
o Miscellmeous cooling are nwt -
centeinment Eatended Core Equipment con be spent fuel Cooling '
occomplished and Rt IR siwt-See Flpre i cooling con be mainteined.
E TAllLES flNAL AM'llCAtu.E LOADS ARE COtaCLUSIOff MtC i
SAFETY HAtOLING OF ANALYSES RtSULTS 11A515 FOR
RESPONSE
REClot4 FUNCTIONS SYS1F M NTEREST Pf~l4FOllMFD ACCFf *T Alli E ACGPIANCE ITEM NO.
REMARKS
- 8
- 1. Spent Fuel Canlainmen) e Hotch A. Systems A.l Yes, Spent The intent of 2.4.?
A.I. See REMARKS I
RwCU Cooling Polar Crane Cover:
Evoluut!on Fuel Cooling HultEG-06I2 for Table 6 can be ac-g,idelines Hool Eudiengers Areo
- 2 and #3.
e Heat complished are met -
A.2. See Response Emlended Core Excluingers 8. Structural Spent fuel to llem 2.4.2 See Flyne i Cooling Analyses A.2 Yes. AtIR cooling can ior basis for suction and be maintained.
acceptobla result, injection (FW) RI Et shutdown lines are in cooling is not this region likely 10 be below 20tr el. lost.
If piping inleyler is lost, Hiil shut.
down cmling could be lost.
Itowever,it was demonstrated based on 7
structural analysts and proteclian of forded by inter-l vening structures and components tlust loss of einese lines was inconcelwible.
D. Structural l
i analyses I
per formed in supporI con-clushuiin A.2 above. Only scald >ing pos.
silde.
e
TAILE 10 a
FINAL AITLICAIM.E LOADS AflE CONCLUSION titC 5AFETY HAtiX.ING OF ANALYSES flESULTS tlA515 Fult ilE58'ONSE f4EGION FUNCTIO _ts SYSTEM _
NIEftEST PEilFOllMED ACCEL *TAHLE ACCEL'TAICE 11EM NO.
ftEMAftKS
- 9
- l.SpentFuel Conlaiewant e RV timod A. Systems A. Yes, spent Tim intent of 2.4.2 A. SeeIIEMAftKS Groting Cooling 8%Iar Crane insulation Evoluosion fuel and I'dit HUltEG-0612 for Table 6.
IsoSE shutdown psidelines eye.frant
- 2.Entended cooling can ere me --
Core Cool.
be accom-spent fuel See Fleste I ing,IW plisted.
md its ut sinst.
Need in down cool!.g Place -
can be maintained.
3a
.L Ch I
+
9 9
e
TAllLE 11 FINAL-APPL.lCADLE LOADS
. ARE CONCLUSION IkitC SAFETY HAtOI lHG OF ANALYSES RESULTS BASIS FOlt
RESPONSE
REGION ft#8CTION5 SYSTEM NIEREST PEftFOttMFD ACCI.PTAlW.E ACCE"TANCE ITEM NO.
REMARKS Tim intent of
- 10
- 1.Spene Fuel Containment e RVHead A. Systems A. Yes, spent NUREG-0612 2.4.2 A. See REM.'.MS RV Hood Cooling Polar Crane Evoluollan fuelarulfitR guidelines for Tal,le 6.
Steroce o PortaWe sissidown are mel-Aree
- 1 Extended Radiation cooling Spent fuel Care Cooling-Shield can be
<sul Rt H shul-SeeFiprei Hood Removed accomplished down cooling e RV Head can be mainteined insulation B. Structural B. Structural B. Assumes that eRWCU Analyses analysis of strucitval analy-Filler a RPV head ses will verify Dominerol-drop onle its no damage to laer storage loco-equigwnent in Hotches lion et 20r the drywell, el will be periarmed to p
verify the ossumption N
that damage lo equipment l
In the dry-well wiildn
('
~ his region t
woubt not occur.
i TAIM E 12 FINAL 1
I AiPLICADLE LOADS ARE COtELUSION NRC SAFETY HAtOLNG OF ANALYSES RESLA.TS BASIS FOR
RESPONSE
ftEC80H FUNCTIONS SYSTEM NTEftEST PEftFOllMEI)
ACCIPTAIME ACCFI'TANCE ITEM NO.
REMARKS Tlie inten; of All
- 1. Spent Fuel Centelnment e PorteNe A. Systems A. Yes, spent NUltEG-0612
.2.4.2 A. See REMARKS f
.. Grating Cooling Polar Crane Radiation Evoluotion fuel and Rift guidelines for Table 6.
l Aree in SW Shield
. shutdown are met -
quadrant of
- 2 and #1 cooling can Spent fuel j
eentoinment Extended Core e RWCU Filter be accom-pool and RHR l
20e el,,
Cec 44ng Domineralizer
- pidshed, shuldown cool-Hetch Covers Ing can le See Flere i maintained.
h l
[
Tant E 13 FINAL APPLICABLE LOADS AftE CONCLU510N NRC SAFETY llAtOI.1HG OF ANALYSES IlESUL15 UA515 FiW1 flESPONSE REGION FUNCTIONS SYSTEM t'lTEREST IYltFOltMED ACCFPIAIME
,ACCFPTANCE ITEM NO.
REMARl(S
- 12
- 5.Dese Centainment e Deam A. Structrual A. Domoge to NUREG-0612 2.2.4 As described in Dryer &
Polar Crane Uryer Analysis of rods in as guidelines the ressense Fuel Storoge
- 6.Crleicallty Drop Onto mmy as 10 can be meI to item 2.2.4, Peel (see Table 1) e Portable Itacks (Gale) assemblies acceptable fladiation predicted solutions are See Flpare i Sideld B. Dose Assess-avalloide. The ment B. Calculated possible solu-e Pool Whole ikxty tiens will be Gales C. Celticelley Dose of 8.5 investigated Evoluotion tem is slight-md an accept-e Miscel-ly penser able solution laneous than NUllEG-Implemented.
Equipment 0612 timie of 6.25 rem, but well within 10 CFR, Pari 100, limit of I.-4
~
25 rem.
03 C. A #<,fg greater tien 0.95 can-not be pre-l cluded under warsi case casullt.ons.
- 1. Spent fuel D. Structural D. Liner pene-HullEG-0612 2.4.2 Cooling enalyses of eruelon and yddelines are drop to pool,
scabbing pos-nwt-gent fuel floor (Dryer) sible, ndoor pool water (Appesulix C) leakage;no Inventery and signaticant cooliswJ con be loss of maintained.
pool inventory.
Hacks are in lower end of pool - would not be uncovered.
E. System E. Yes, spent The Intent of Evoluotion fuel cooling t4JREG-0612 (Appendix B) can be psidelines are accomplished. mel - Spent fuel i
cooling can be maintained l
Tant.E 14 FIN M.
AIPLICAfLE LOADS ARE cot 4CLUS10N titC SAFETY
- HAtotlte, OF ANALYSES RESULTS BASIS FOR
RESPONSE
REClON FUNCTIONS SYS1EM _
NIEflEST PE f tFOllMFD ACCEPTAlit.E ACCEPTAtCE ITEM NO.
ftEMARKS
- 13
- 5. Dose Cas.tainmes.:
e RV tiend A. Structural A. Yes no tANIEG-0K!2 2.3.4 Reactor Iller Crars Anolyses fuel domoge guidelinc. we
+
Well 86.Criticolity e Drywell (GE Analyses, predicted met-no fuel (SeeTatde1)
Head see FSAR, domoneor See Fipre i Table 9.l-7 crushing e Dryer HEDC-23566 predicted.
and CE lesler, o Separator Smith to D,de dated 2/5/82).
e Pereable 1Ef1A performed Radiation evalualian to Shield assure shot GE analyses bounded oiler postulated drops.
Y
- 2 and #3.
A. Structural A. Ye s, vessel tilREG-06l2 2.4.2 G
Estended Care Analyses integrity are met-Coo!ine (CE Analyses, maintained vessel see above)
Integrity g
mainlained.
9 e
1 4
?
(!
9
s TM*E 15 FINAL.
APPt.lCAf1LE LOADS AltE COf K.'t U510N MIC 5AFETY HAfOt.ING OF ANALYSES RESULTS tlASIS FOf t
RESPONSE
REGION ruNCTIONS SYSTEM NTEllEST ITilfOilMFD AC'It'TAllLE ACCEL
- tat 4CE ITEM NO.
REMARKS
- 14
- 1.5 pent Fuel Cantoinmens e 5two ml A. Struetural A. Yes.F.iner pene. HUREG-0612 2.4.2 Separotor Cooline Polar Crus.e Head /
Anotyses tronica a us e,idelines
$sceoge Separosor scolddes w s or e me -
Pool J2 and #1 possit,1q Synt fuel Extended Core e RWCU HX enieuw leukage aul fuiIt SeeFlpareI Cooling Hotch into drywell slutdawn Covera we 4J not of.
cooling can lecs Spent I,e enainseined fuel or Idit simidown crol.
log. No j
electelcol equip-meni in drywell sequ! sed 6o operole to accongdish Sof =Ir Fmesions Y
1,2or3.
l e
e e
f L
l
TABLE.16 SU!HARY OF CONTROLLING STRUCTURAL' BEHAVIOR RESULTING FROM POSTULATED HEAVY LOAD DROPS CONTROLLING MODE OF RESPONSE APPROXIMATE WEIGHT HANDLING OVERALL LOAD TON SYSTEM STRUCTURAL LOCAL 1.
Reactor Pressure 117 Polar Crane X
Vessel Head (RPV) 2.
Steam Dryer 40 Polar Crane X
3.
Shroud Head / Steam 68 Polar Crane X
Separator 4.
Drywell Head 61.5 Polar Crane X
X 5.
Portable Refueling 12 Polar Crane X
X Shield 6.
RPV Head Insula-10.5 Polar Crane N/A tion w/ Support Structure 7.
Reactor Well/ Steam 3.5 Polar Crane X
X Dryer Storage Ares
~
3 ate 8.
Upper Containment 3.5 Polar Crane X
X Fuel Pool / Transfer Pool 9.
Load Block 5.6(M) Polar Crane N/A 1.0(AUX) i
- 10. RWCU Regenerative 15 Polar Crane X
X HX Hatches (2)
- 11. RWCU Non-15-17 Polar Crane X
X Regenerative HX Hatches (3)
- 12. RWCU Filter Demin-20 Polar Crane X
X eralizerHatches(3)
- 13. New Fuel Shipping i.5 New Fuel X
Containers Bridge Crane
- 14. Fuel Pool and Clean 3
New Fuel X
Up Filter Demineral-Bridge Crane ization Hatch (2)
- 15. Spent Fuel Pool Gate 3.5 New Fuel X
X Bridge Crane
'A-21
.