ML20090H621
| ML20090H621 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/01/1984 |
| From: | Overbeck G FRANKLIN INSTITUTE |
| To: | NRC |
| Shared Package | |
| ML18026B049 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-07974, TAC-07975, TAC-08438, TAC-7974, TAC-7975, TAC-8438, TER-C5506-483-4, TER-C5506-483-484-48, NUDOCS 8403140281 | |
| Download: ML20090H621 (21) | |
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TECHNICAL EVALUATION REPORT CONTROL OF HEAVY LOADS - PHASE 11 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR POWER STATION UNITS 1, 2, AND 3 NRC DOCKET NO. 50-259. -260. -296 FRC PROJECT C5506 NRC TAC NO. 07974, 07975. 08438 FRC ASSIGNMENT 19
, NRC CONTRACT NO. NRC-03 81 130 FRCTASKS 483, 484, 485 Prepared by Franklin Research Center Author: G. Overbeck 20th and Race Streets Philadelphia, PA 19103 FRC Group Leader: I. H. Sargent Prepared for Lead NRC Engineer: A. Singh Nu font RegulatoryCommission l
Washington, D.C. 20555 '
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March 1. 1984 This report was prepared as an accou.7t of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency 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 use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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l TER-C5506-483/484/485 CONTENTS Section Title a P,ag, l
1 INTRODUCTION . . . . . . . . . . . . . 1 1.1 Purpose . . . . . . ., . . . . . . 1 1.2 Generic Background. . . . . . . . . . . 1 1.3 Plant-Specific Background . ,. . . .' . . . . 2 2 EVALUATION . . . . . . . . . . . . . . 3 2.1 Evaluation Criteria . . . . . . . . . . 3 2.2 Reactor Building Imad Bandling Systems. . . . . . 4 2.3 Load Bandling Systems in Other Areae . . . . . . 10 i
3 CONCLUSION . . . . . . . . . . . . . . 14 3.1 Information Issues. . . . . . . . . . . 14 3.2 Approach Issues . . . . . . . . . . . 15 4 REFERENCES . . . . . . . . . . . . . . 16 l
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TER-C5506-483/484/485 FORENORD This hchnical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear mogulatory Commission (office of maclear Beactor Regulation, Division of operating Reactors) for technical assistance in support of MRC operating reactor licensing actions. The technical evalcation was conducted in accordance with criteria established by the 10tc.
Mr. I. E. Sargent and Mr. G. J. Overbeck contributed to the technical preparation of this report through a subcontract with MEfPTEC Services, Inc.
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- 1. Iwrnococrzou 1.1 PUltPOSE l
Sis technical evaluation report documer.c4 a review of load handling equipment operated in the vicinity of spent fuel and equipment employed for reactor shutdown and fuel element decay heat removal at Browns Ferry nuclear Power Statica Units 1, 2, and 3. Bis review constitutes the second phase of a two-phase review instituted to resolve a genetic issue pertaining to the safe handling of heavy loads at nuclear power plants. ,
s 1.2 GENERIC memeoup .
Generic Technical Activity Task A-36 was established by the nuclear mogulatory commission (nac) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to ensure the safe handling of heavy .3oads kr.d to recosmeend necessary ,
l changes in these measures. This activity was initiated by E letter issued by the unc staff an May 17, 1978 (1] to all power reactor llorasees, req 2 eating information concerning the control of heavy loads near spent fuel.
I he results of Tank A-36 were reported in NUREC-0612 (2] . The staff '
i concluded from this araluation that existing measures to control i:ba handling of heavy loads at operating plants provide protection from certain potantial problems but do not adequately cover the major causes of load handling accidents and should be upgraded.
To upgrade measures for the control of heavy loads, the staff developed a 1 series of guidelines to implement a two-part objective. The first part of the objective, to be achieved through the implannotation of a set of general ,
- I guidelines espressed in stuleG-0612, section 5.1.1, was to ensure that all land handling systems at seclear power plants have been designed and are operated
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so that their probability of failure is appropriately saml1 for the critical -
tasks in which they are employed. Se results of the reviews associated with this part of the staff's overall objective were provided in a series of technical evaluation reports identified as Phase I reports. Se second part 8'
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~o f the staff's objective, and the subject of this report, was to be achieved through guidelines espressed in NUnaG-0612, sections 5.1.2 through 5.1.5. The purpose of these guidelines was to ensure that, in the case of specific load I handling systems used in areas where their failure might result in significant consequences, either (1) features have been provided, in addition to those required for all load handling systems, to make the potential for a damaging load drop extremely small or (2) conservative evaluations of load handling accidents indicate that the potential consequences of a load drop are acceptably small.
l.3 P!AIFf-8FBCIFIC macurMMUND i
on December 22, 1980, the lulC issued a lettler [3] to the Tennessee Valley Authority (TVA), the Licensee for Browns Perry Units 1, 2, and 3, requesting the review of provisions for handling tm3 control of heavy loads, the i avaluation of these provisions with respect to the guidelines of EUREG-0612, i a,nd the provision of cottain additional information to be used for s.n independent determination of conformance to these guidelines. m roruits of ,
this independent evaluationi with respect to general load handling equipment and procedures (Phase I) were provided on April 5, 1983 [4]. On Septesber 28, l 1982, TVA provided an initial Phase II report [5] concerning conformance with staff guidelines for specific load handling systems operated in areas wbsre a load drop might result in significant consequences. On January 25, 1983 [6]
and on March 28, 1983 [7], TVA supplemented that initial response. The information in maferences 5, 6, and 7 provided the basis for this technical evaluation report.
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- 2. EVALUATION ,
l This section presents an evaluation of critical load handling areas at Browns Ferry tktits 1, 2, and 3. Separate subsections are provided to identify the criteria used in this evaluation and each of the plant areas considered.
For each such area, relevant load handling systems are identified, Licensee-provided information related to the evaluation criteria or proposed citernatives is summarized and evaluated, and a conclusion as to the extent of compliance, including recommended additional action or requirements for additional information, as appropriate, is irovided.
2.1 EVALUATION CRITERIA The objective of this review was to determine if plant arrangements and load handling equipment design were such that either the likelihood of a load handling accident that could tamage spent fuel or equipeent used in reactor shurdown or fuel element decay heat removal is extremely small or that the coraacquences of auch damage, should'it occur, wil.1 be aamptab1w. Guidance '
containe6 in MUREG-0612, Sections 5.1.2, 5.1.3, and 5.1.5 (for pressurized t: ster reactors) and in 5.1.4 and 3.1.5 (fo: boilir.g water reactors) forms the basis for the conclusions reached in thia section and is briefly an==arized as ,
follows.
For a determination that the likelihood of damage is extremely smalls o The design of the load handling system (i.e., crane or hoist and underhook lifting devices) is consistent with, er equivalent to, the MRC staff criteria for single-failure-proof cranes identified in NUREG-0554 [8], or o the plant physical arrangement is such that a crane operated in the '
vicinity of spent fuel or safety-related equipment is prevented j from traeoliaq to a position from which a load drop can be expected to damage such equipment.
yer a determination that the potential consequences of damage following a load drop will be W *1er o In the case of potential damage to spent fuel, calculations have been provided to demonctrate that potential radiological doses at
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l TER-C5506-483/484/485 l the site boundary will not emoeed 25% of the limits specified in *
( 10CFR100 and that the post-accident configuration of the fuel will ,
not result in a k egg larger than 0.95.
o In the case of damage to the reactor vessel or spent fuel pool, it i can be demonstrated that this damage will be limited to the extent ,
that the fuel will not become uncovered. '
- o In-the case of damage to equipment or components employed for reactor shutdown or fuel element decay heat removal, it can be demonstrated that the safety-related function of the affected system will not be lost consequent to a load drop.
- 2.2 REACTOR BUILDING EANDLING SYSTBIS 2.2.1 Imad sandline systems capable of Carrying a'seavy road (as defined in NUREG-0612) Within the Reactor Buildine i
2.2.1.1 h = mary of Licensee Statements and Conclusions In Reference 5, the Licensee identified the reactor building crane and a i
i 4-ton book-type chain hoist as the load headling systmas within the reactor ,
l building capable of carrying a hoairy load. In Reference 6, the Licensee identified the 3-ton jib crane as being capable of swingir.g into critica.
c19etrical panels. In Anferencs 7, the Licensee revised the response it ,
provided in a June 3,1982 letter [9] to include a 24-ton geaz-type chain hoist and indicated that this hoist irr, used to replace recirculation pump l motors in the drywell.
2.2.1.2 Evaluation and Conclusion The Licensee's statements and conclusions with respect to the load handling systems mentioned above are consistent with the information evaluated in Reference 4 and with the intent of MUREG-0612.
2.2.2 Reactor Buildine Crane
! 2.2.2.1 Summary of Licensee statements and conclusions In Reference 5, the Licensee stated that the reactor-building crane had been evaluated as having sufficient design features to make the likelihood of !
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l TER-C5506-483/484/485 a load drop extremely small. S e Licensee provided an evaluation of the ,
125-ton re' actor building crane with respect to the features of design, fabrication, inspection, testing, and operation as delineated in NUREG-0554 and supplemented by the identified alternatives specified in NUREG-0612 Appendix c. The Licensee also noted that this crane has previously been evaluated with respect to Nnc Regulatory Guide 1.104 and Branch Technical Position APCSB9-1. The Licensee indicated that the evaluation of the reactor building crane design was based upon information'provided in letters dated Pobruary 10, 1981 (10] and June 30, 1976 [11). ,
The Licensee also provided a heavy load / impact area matrix for the .
postulated drop of heavy loads from the reactor building crane within the reactor building.
2.2.2.2 Evaluation The Licensee's evaluatica of the 125-ton reactor building crane with respect to the features of design, fabrication, inspection, testing, and operation as delineated in wptEG-0554 and supplemented by alternatives specified in NUREG-0612, Appendix C was evaluated in a March 31, 1983 report (12]. The evaluation in Reference 12 included additional information provided by the Licensee in letters dated March 11, 1981 (13], May 27, 1981 [14), and December 14, 1982 [15]. In Reference 12, a point-Lrf-point evaluation was 1
completed using Licensee-supplied information. The overall conclusion drawn from that review was that TVA had satisifed the intent of the staff guidance by providing a reactor building crane which provides suitable protection from the effects of operator error and the failure of crane mechanical and electrical components. The evaluation of specific staff requirements indicated that in end issue associated with crane reliability, the Browns Ferry crane satisfied the staff requirements either directly or through features that can be assessed to be technically equivalent, or Tvh had made a .
couaitment to implement modifications which will antisfy the requirements, or TVA had made commitment to implement a test and laspection program to provide appropriate additional confidence in load handling reliability. _+'_:rg :-it to' the evaluation previously d h M , however, the unc issued generic letter q@UU UJ Frenidn Reneerch Center mm . _ _ _ _ _ _
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83-42 [161, which provides an additional evaluation criterion not specifically .
stated in'WOREG-0554. Bis letter notes that it will be the staff's policy to require a desenstration that no single failure in the crane electric power / control system will cause a load drop. B is issue has not been addressed in available Tvh submittals, As indi'cated in section 2.1 and NOREG-0612, section 5.1.4, it is the f
staff's position that if the load handling system is to be considered highly
.. reliable (i.e., it can be established that the likelihood of a load drop is
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extremely small); both the crane, underboed lifting devices, and attachment points should conform to the guidance of NUREG-0612, section 5.1.6. Se i Licensee has not addressed the design onpability of lifting devices or attachments points. As of this review, only the spent fuel shipping cask at i the Browns Ferr'j plant has been reported to be of redundant design, including two lower load block attaching points. Other lifting devices and attachment j schemes, which handle other critical loads (defined as loads which, if dropped, i could result in an offsite dose in emoess of 1/4 of 10CFR100 limits, or damage to fuel or fuel storage racks such that the average k gg la greater than 0.95, or damage to the reactor vessel or spent fuel pool sufficient to cause water leakage to the extent that fuel oculd be uncovered, or damage to the ,
equipment in safety systems sofficient to cause a loss of safe shut.6owa function), have not been addressed by the Licensee.
As an alternative to providing a single-failure-proof protection through lifting devloe design, W0R30-0612, section 5.1.4 indicates that the effects of l heavy load dropa in the reacter building can be analysed to show that the evaluation criteria of section 5.1 are satisfied (i.e., demonstrate that certain heavy loads -are not critical leads) . So I.ioensee has provided a heavy load / impact matrix for various heavy lands carried by the reactor building orane. Se heavy loads identified in this matrix are the same as those identified la anderence 4 with the following esoeptions. De equipment ,
pool shield plays (5s tons), immend ship box (1.25 tems), equipment pool covers (1.5 esas), motor generator sets (7.5 tems), surge tank plug (2.1 tons), reactor water aleanup (mecc) domineraliser vault plugis and vessel head (6 tons), and the control rod rank (0.5 ton) have not been included in the 6-am w tw _
Tsm-c5506-483/484/485 Licensee's heavy load / impact matrix. Per the heavy loads included in the ,
load / impact: matriz, the Licensee has indicated that the likelihood of the reactor building crane load handling system failure is estremely small for each hasard; however, the Licensee has not indicated that analysis demonstrates that the load bandling system failure (i.e., failure of lifting device) will result in consequences that satisfy the evaluation criteria of Section 5.1. , j 2.2.2'.3 Conclusions and Recommendations
- The Licensee has demonstrated, with the emoeption of specific considera-tion of single failures in the electric power /oontrol system, that the reactor building crane has sufficient design features to make the likelihood of a load drop extremely small. The Licensee has not, however, demonstrated that the l associated lifting devices comply with the guidelines of MUREG-0612, Section ;
> 5.1.6. As a onenequence, the Licensee has not demonstrated that the reactor building crane load handling sys'tes meets the requirements of uuass-0612, Section 5.1.4.
10 demonstrate compliance with NUREG-0612, Section 5.1.4, the Licensee i
should address how the additional guidelines of Section 5.1.6(1) and 5.1.6(3) have been invoked at the Browns Ferry plant. Per those lifting devices and
, interfacing lift points that do not satisfy these additional guidelines, the Licensee should analyse the effects of heavy load drops to show that the evaluation criteria of Section 5.1 are satisfied. In addition, the Licensee should ensure that all heavy loads handled by the reactor building crane are considered or should provide justification for oweinaion of any heavy load previously identified. ,
1 2.2.3 3-Ton Jib Crano 2.2.3.1 Seamary of Licensee Statements and Conclusions ,
In Beforense 9, the 1.tosasse esempted the 3-ton jib crane because the loads carried by this load handling system weigh'less than 1000 lb. During an Ontcher 28, 1982 conference call with TVA, the NRC espressed a concern that
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, TER-C5506-483/484/485 the 3-ton jib crane has a capacity to carry loads greater than this weight .
over Class II equipment within its area of ooverage. pollowing the conference call, the Licensee stated in anference 6 that a design change request would be initiated by .7une 1, 1983 requiring a stop be installed to prevent the 3-ton jin crane from swinging into the critical electrical panels.
l j 2.2.3.2 Evaluation 4 .
l, In Reference 4, the Licensee's commitment to install a stop to prevent movement of a heavy load over or into Class II equipment was considered j' sufficient to esclude the load handling system from further evaluation with
! respect to NuneG-0612, Section 5.1.1. To ensure that the evaluation criteria of NUREG-0612, section 5.1 are satisfied for heavy load drops within the -
reactor building, additional guidelines were provided in Section 5.1.4.
Reliance on plant physical arrangement to prevent a load handling system in i the vicinity of safe shutdown equipment from traveling to a position from l which a load drop can be espected tc. damage such equipment is a means of f ensuring that the likelihood of damage is estremely erall. The Licensee's
! commitment, to install a permanent stop to prevent the 3-ton jib crane from
. swinging into Class II equipment is consistent with the intent of NOREG-0512, 1
1 section 5.1.4.
2.2.3.3 Conclusion and neocamendations The Licensee has demonstrated that the 3-tan jib crane load handling system meets the guidelines of NORBG-0612, Section 5.1.4.
2.2.4 24-Ten Gear-Tree Chain Epist 2.2.4.1 summary of Licensee statements sad conclusions In anforeuse 9, the Licensee desorlhed this load handling device as a
" twenty-four-ton onpacity, low headroom, spur geared type chain hoist w/ geared trolley and shain con E iner.' ghe Licensee indicated that this hoist la used to replace roeirculation pump motore '(20 tems each) in the drywell.
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2.2.4.2 Evaluation .
In Reference 5, the Licensee described the load handling system and indicated how compliance with the guidelines of 3033G-0612, section 5.1.1 was to be accomplished. S e purpose of Reference 9 was to inform the NEC that TVA had identified additional load handling systems which had not been included in its initial response concerning NunsG-0412, Section 5.1.1. Bowever, the Licensee's response in Reference 9 did not address how the additional guidelines of NUREG-0612, section 5.1.4 were to be satisfied, h -
! 2.2.4.3 Conclusion and W tions i
j The Licensee has not demonstrated that the 24-ton gear-type chain hoist load handling system meets the guidelines of ugueG-0612, Section 5.1.4.' To demonstrate compliance with these guidelines, the Licensee should revise its i
response provided in Reference 5 to include the 24-ton gear-type chain hoist load handling system.
I 4-Ton Book Tvoe Botet 2.2.5 2.2.5.1 Summary of Licensee Statements and Cocclusions In Reference 5, the Licensee has stated that the 4-ton book-type chain hoist does not have sufficient design features to make the likelihood of a load drop entremely small. Because the 4-ton book-type chain hoist has the physical oapability of oarrying loads over safe shutdown equipment, the I
Licensee provided a hoewy land / impact matrix for the various heavy loads hanM ed and described the consequences. WDr each postulated load drop and impect area, the Licensee stated that system redumesney and separation ensure that the safety-related fumation of the affected system is not precluded following the postulated load drops. The Licensee further esplained that the load drops are postulated in separate pump rooms omstaining core spray pumps &
and C or' pumps 3 and D. Doomune of physical separation of the two loops, a
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load drop in one zooniwould not adversely affect the other redundant system,
. provided that hand control valve scT-75-1 is closed during repair on pumps A and C, and valve scv-7F 29 is olosed during repair on pumps 3 and D.
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2.2.5.2 Evaluation l Analysis of the effects of heavy load drops to demonstrate that the evaluation criteria of MUmsG-0612, Section 5.1 are satisfied is consistent with the guidelines of Section 5.1.4 for load handling systems within the I
! reactor building. He 4-ton book-type chain boist is used over hatches to remove various equipment from lowne floors to the 545-ft elevation floor. He ,
loads analysed by the Licensee are consistent with those previously reviewed l cnd identified in Reference 4. H e results of the Licensee's load drop analysis indicate that the physical separation of the two core spray loops precludes a load drop in one room from adveesely affecting the other redundant loop as long as the suppression pool suction valve is closed following the i load drop. moview of the affected equipment following a load drop indicates l that the load drop will not initiate an event that will require the use of the l core spray system (i.e., damage to a core spray loop will not initiate a loss-
! cf-ooolant accident). Secefore, sufficient time should be available for the
- operator to assess which core spray loop has been damaged and to shut the cuppression pool suction isolation valve in the affected Icop. !
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2.2.5.3 Conclusior.s and Recommendatione Se Licensee has demonstrated that criterion IV of NORsG-0612, Section 5.1 is satisfied for the 4-ton book-type chain hoist and that the load j handling system neets the guidelines of NUERG-0412.
2.3 EOhD EAMDLING SYSTMS IN OTEIR ARBAS 2.3.1 Load Eandline Systems Capable of Carryine Imada in Plant Areas outside of the Reactor Buildine 2
2.3.1.1 Summary of Licensee Statements and Conclusions In Reference 5, the Licensee has identified the self-propelled truck crane as the only .i.and handling system outside the reactor building capable of i carrying a heavy hand.
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the Licensee's statement and conclusion with respect to the load handling system mentioned above are consistent wit.h the information evaluated in neforence 4 and with the intent of NUERG-0612.
4 2.3.2 Self-Propelled Truck Crane 2.3.2.1 Summary of Licensee statements and Conclusions In Reference 5, the Licensee stated that the self-propelled truck crane does not have sufficient design features to make the likelihood of a load d:op estremely small. Because the self-propelled truck crane has the physical i
capability of carrying loads over safe shutdown e'quipment, the Licensee provided a heavy load / impact area matriz. This matrix identified the various J
loads handled and described the consequences when these loads are dropped in ;
specific areas. For each load postulated to drop (ezoopt the residual heat removal service water (EmmsW] pump) and impact a safety-related system, the Licensee stated that system r:^-- ':xy and separation ensure that the '
j safety-related function of the affected system is not pescluded. For the ,
3 postulated drop of the Mass pump, the Licensee is relying on site-specific considerations to eliminate the need to evaluate load drop consequences. ;
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- 2.3.2.2 Evaluation i
i Analysis of the effects of heavy load drops to demonstrate satisfaction i cf the evaluation criteria of EURsG-0612, section 5.1 is consistent with the i guidelines of section 5.1.5. At the Broens Ferry plant, the two general impact areas within the operating range of the self-propelled truck crane j which contain equipment provided for safe shutdown or decay heat removal were ;
considered. In each case, the loads analysed were consistent with the heavy l loads identified and reviewed in anference 4.
The hJanamae's analysis of lead handling accidente in the vicinity of the diesel generator indicInted that the potential consequences of such an accident would satisfy the relevant acceptance criterion. This analysis was based on the postulated drop of the reactor building embaust fans onto the roof of f qg 11 uimiMonten Reneerch Ceneer ANP
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TREN:5506-483/484/485 either the tatits 1 and 2 or the tanit 3 diesel generator buildings. The air intake, filter, and exhaust for each diesel and the diesel generator air conditioning chilled water panels for the the Unit 3 diesel were identified as the safety-related equipment that might be damaged by such an accident.
Because each diesel generator has an independent air intake, filter, and exhaust system, damage to any one of these systems would not impair the l
operation of the other diesel generators. similar damage to one of the Unit 3 diesel generator air conditioning chilled water' panels would not preclude the proper operation of other panels. In each case, the phy,sical separation between independent or redundant, systems has been found to be sufficient to prevent damage to more than one system. The Browns Ferry Updated FSAR [17]
4 indicates that the operation of only three diesel generators in each building i
i is required. Specifically, three out of the four diesel generators at Units 1
- and 2 in parallel with three out of the four diesel generators at Unit 3 are 1
l sufficient to supply all required loeds for the safe shutdown and cooldown of all three units in the event of loss of offsite power and a design basis I
accident in any one unit. .
! In the case of pot.ential lond handling accidents in the vicinity of the i
l NERSW system, the Licensee's response is not sufficient to support a determination that staff acceptance criteria have been satisfied. TVA's response regarding postulated drops of a aaaa====e circulating water (CCW) i pump, a CCW pump motor, or a fire pump does indicate that the physical separation of the MERSW pumps is sufficient to preclude a common mode failure and the loss of system safety function. Botention of such capability in the case of the postulated drop of an EER5W pump, however, appears to rely on unidentified site-specific considerations.
l At the Browns Ferry plant, the RERSW system is a fairly complex system providing substantial flexibility in operating modes. It consists of four pairs of MRSW pumpe assigned to the EER systems for the three units, plus .
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four additional pumps assigned to the emergency coolimp water system (EICW) .
Each of the pump pair's' foods one independent m header which, in turn, feeds one um heat emahanger in seek unit. Two of the individual (EECW-assigned) pumps feed one BBCW boeder. ghe two remaining pumps feed the alternate EECW WWW - - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____
TER-C5506-483/484/485 header (s) . Two RER5W pumps and two RER heat eschangers are required per unit ,
j to effectively remove afterheat under emergency oceditions. Because of pump I
capacity limitations, one pump is necessary to supply enough coolant to serve' one heat enchanger, and no single train's heat enchanger set can be operated
- in all three units simultaneously. This flexibility is further constrained, as identified in Reference 17, by limitations associated with allowable loadings of diesel generators and electrical boards such that certain multiple l failures may disable the system. It is reasonabie to believe that certain site-specific considerations, combined with the fact that damage to the RERSW
. system will not initiate an event that will require the automatic initiation of the RERSW system, can be relied upon to provide the degree of protection l inherent in verbatim ocepliance with NUREG-0612 however, such consideration cannot be determined independently from the information provided.
2.3.2.3 Conclusion The Licensee has demonstrated that criterion IV of EDRBG-0612, Section i 5.1 is satisfied for each load handled by the self-propelled truck crane except for the postulated drop of a Mtasw pump. 10 demonstrate that this :
criterion is satisfied for all loads handled by the self-propelled truck crane, the Licensee should describe the site-specific considerations that j eliminate the need to consider the postulated drop of a Rumsw pump onto an operating RKasw pump.
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- 3. COBCLUSION ,
His sumanary is provided to consolidate the results of crane 1 specific evaluations presented in section 2. It is not meant as a substitute for the specific conclusions reached in the various subsections of section 2. It is l
provided to allow the reader to focus on the key topics which should be addressed in seeking to resolve issues where the degree of load handling reliability provided by cranes at the Browns Ferry plant was not found to meet the objectives of upBEG-0612. S is section addresses issues for which the information provided is felt to be inadoglate to support a definitive conclusion and issues wherein the information provided has been evaluated as proposing an, approach inconsisto.,t with the the guidance of NUREG-0612.
3.1 INFORMATION ISSUES S e information provided by the Licensee has been assessed as insufficient to support an independent conclusion that load handling reliability is consistent with the evaluation criteria of Section 2.1 in the following areas:
meactor Building Crane Lifting Devices (Section 2.2.2)
The Licensee should identify how the guidelines of MUREG-0612, Sections 5.1.6 (1) and 5.1.6(3) have been satisfied for lifting devices and associated lift interfaces for heavy loads handled by the reactor building crane for which the analysis of the drop of such loads has not demonstrated that the criteria of NOREG-0612, Section 5.1 have been satisfied. mere such analyses have been employed to demonstrate that additional design features are unnecessary, the results of these analyses should be provided.
manctor mailding Crane Electrical Power / Control system (Section 2.2.2)
D e Licensee should verify that a single failure in the reactor building crane electrical power / control system will not result in a load drop as discussed in maference 16.
Reactor Emilding 24-Ton Gear-Type Chain Boist (Section 2.2.4) s' l Se Licensee should identify how the guidance of MonaG-0612, Section 5.1.4 is satistied for operation of the reactor building 24-ton gear-type chain hoist.
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i ranhNn RegerCh C4n%K '
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l l I TER-C5506-483/484/485 l
1 Self-Propelled Truck Crane (Section 2.3.2) *
. j
- Se Licensee should identify the site-specific considerations which lead
~
- to the conclusion that lifts of an REkW pump in the vicinity of the IERSW system cannot, in the event of a load drop, cause a loss of RERSW system safety functions.
3.2 APPROhCE ISSUES This review has revealed no issues wherein the approach or position taken by the Licensee, based on information provided thus far, is inconsistent with the staff's objectives as expressed in the evaluation criteria of Section 2.1.
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1 1 N Munida Amamarch Carter
! Test-C5506-433/484/M5
- 4. REFERENCES
! 1. V. Stallo Jr. (NRC)
- Iatter to all Operating Reactor Licensees subjects moquest for Additional Information on Control of Esavy Loads i Bear Spent Puel Ney 17, 1978
- 2. " Control of Beavy Ioads at IRaclear Power Plants
- i M C, July 1980 . I MUREG-0612 l 3. ERC .
Letter to Tennessee Valley Authority l
Subject:
Esquest for Baview of Beavy Imed Bandling at Browns Perry 4
Ikaclear Plant twits 1, 2, and 3
- i December 22, 19'O
- 4. PRC Technical Evaluation Report subject: Control of Beavy Imads Bandling at Browns Ferry Muclear Plant Units 1, 2, and 3 TER-C5506-337/338/339 April 5, 1983 ..
- 5. L. M. Mills (TVA)
Letter to D. V. Vassallo (MRC) subject: Information Regarding Sections 2.2 and 2.3 of NUREG-0612 September 28, 1982
- 6. D. S. Eanner (TVA)
Letter to D. V. Vassallo (MK) subject: It= 4--Item Description of TVA's ansponse Presented to k ,
i Clemenson (MRC) in Conference call on October 28, 1982 I January 25, 1983
Subject:
Taahlon of Two Imad Bandling systems that were Determined to qualify for Inclusion in Aeoordance with 1R3tEG-0612 i i mrah 28, 1983 '
- 8. WC 3D130-0554, ' single-Failure-Proof Cranes at Wenlaar Power Plants
- I May 1979 ,,,
1" .,m- - , . -. , _ ,. ~ - . ,,,e.-..,., _ . . . . . _4,m., -_ __~_r,.._,,m,#,, - - - . - , --, ..w_..- - , , ..,,
l TER-C5506-483/484/485 l
- 9. D. 8. Eammer (TVA) .
Letter to D. V. vassallo (MRC) subjects Information Bogarding MUREG-0612, " Control of Beavy Imada at 3kaclear Power Plants" as Ihaquested by D. G. Eisenhut letter of December 22, 1980 June 3, 1982
- 10. L. M. Mills (TVA)
Letter'to T. A. Ippolito (NBC)
Subjects Response to Request for Information Regarding Browns Ferry Reactor Building Crane '
February 10, 1981
- 11. E. G. Parris (TVA)
- Intter to A. Schwencer (MRC)
Subject:
Response to Branch Stdanical Position APCSB 9-1 June 30, 1976
- 12. FRC Draft Technical Evaluation Report Subject Control of Beavy Imada at Browns Ferry Nuclear Power Station .
TER-C5506-336 March 31, 1983
- 13. L. M. Mills (TVA)'
Letter to T. A. Ippolito (WRC)
Subject:
Revision to Imtter of Pebruary 10, 1981 March 11, 1981
- 14. L. M. Mills (TVA)
Letter to T. A. Ippolito (NBC)
Subject:
Revised Response to Several Questions Addressed in Letter of February 10, 1981 May 27, 1981
- 15. L. M. Mills (T7A) i Intter to D. v. vassallo (MRC)
Subject:
It= Vf-Item Description of TVA's Response Presented to Mr.
Clemenson (IWIC) in Conference Call on October 27, 1982 Dea ==har 14, 1982
- 16. D. G. Eisenhut (NRC)
Intter to all solders of Operating Licensees, Application for Operating Licensees and Bolders of Constructor Premit for Power Reactor (Generic .
Int'.or 53-42)
Subjects Clarifloation to Generic Letter 81-07 Bogarding Response to
- 3055G-4412 De h r 19, 1983 l
l g !
JUU Franh8n Reseemh Cr8mr
t
- a
- 6 I
- TER-C5506-483/484/485
- 17. L. N. Mills (TVAs -
Letter to E. R. Denton (MRC) subjects Updated Final safety Analysis amport hagust 5,1983 l
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