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e CONTROLOFHEAVYLOADSAT-NUCLEARP6WERPLANTS V0GTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 (PHASEII)
Docket No.
'50-424'
[50-425 Author C. R. Shaber Principal Technical Investigator T. H. Stickley Published January 1984 EG&G Idaho, Inc.
Tdaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6457 e
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ABSTRACT The Nuclear Regulatory Commission (NRC) has requested that all nuclear plants, either operating or under construction, submit a response of consistency with NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." EG&G Idaho, Inc., has contracted with the NRC to evaluate the responses of those plants presently under construction.
This report contains EG&G's evaluation and recommendations for Vogtle Electric Generating Plant Unit 1 and Unit 2 for the requirements of Sections 5.1.2, 5.1.3, 5.1.5, and 5.1.6 of :1UREG-0612 (Phase II).
Section 5.1.1 (Phase I) was covered in a separate report [1].
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EXECUTIVE
SUMMARY
Vogtle Electric Generating Plant Unit I and Unit 2 is not totally consistent with the guidelines of NUREG-0612.
In general, inconsistencies exist in the following areas:
OfinformationonassociatedliftingdevjcesusedwiththeSpent o
Fuel Cask Bridge Crane What NUREG 0612 Article 5.1.2 option will be used and how the o
auxiliary crane and monorail hoist will be made consistent with the article option requirement Response information on the cranes, hoists and associated lifting o
devices in the Containment Building to confirr.: their consistency with NUREG 0612, Article 5.1.3 Other area hoists handling six loads are considered excluded, but o
for them the exemption code indicates the analysis has not been made.
This requires completion.
The main report contains recommendations which will aid in making the above items consistent with the appropriate guidelines.
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l CONTENTS ABSTRACT............................................................
11 EXECUTIVE
SUMMARY
til 1.
INTRODUCTION....................................................
1 1.1 Purpose of Review.........................................
I 1.2 Generic Background.....................<_...................
I 1.3 Pl a n t-S p ec i fi c Ba c kg ro u n d.................................
3 2.
EVALUATION AND RECOMMENDATIONS..................................
4 2.1 Overview..................................................
4 2.2 Heavy Load Overhead Handling Systems......................
4 2.3 Guidelines................................................
4 3.
CONCLUDING
SUMMARY
19 3.1 Guideline Recommendations.................................
19 3.2 S~ummary........
...s 19 4.
REFERENCES......................................................
20 TABLES 2.1 Nonexempt Heavy Load-Handling Systems...........................
5 2.2 FSAR Table 9.1.5-3 Containment Bui Systems..........................lding Overhead Handling 12 i
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REFERENCES
[ Phase I Final Report] Draft, C. R. Shaber and T. H. Stickley, 1.
EG&G Idaho, Control of Heavy Loads at Nuclear Power Plants, Vogtle Electric Generating Plant Units 1 and 2, December 1983.
2.
N'; REG-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 Georgia Power.
Subject:
NRC Request for Addicional Information on Control of Heavy Loads Near Spent Fuel, NRC, 22 December 1980.
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CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS V0GTLE ELECTRIC GENERATING PLANT UNIT 1 AND UNIT 2 (PHASE II) 1.
INTRODUCTION 1.1 Purpose of Review This technical evaluation report documents he EG&G Idaho, Inc.,
review of general load-handling policy and procedures at Vogtle Electric Generating Plant Unit I and Unit 2 (VEGP).
This evaluation was performed with the objective of assessing conformance to the general load-handling guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" [2], Sections 5.1.2, 5.1.3, 5.1.5 and 5.1.6.
This constitutes Phase II of a two phase evaluation.
Phase I assesses conformance to Section 5.1.1 of NUREG-0612 and was documented in a separate report [1].
1.2 Generic Background Generic Technical Activity Task A-36 was estabitshed by the U.S.
Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to assure the safe handling of heavy loads and to recommend necessary changes to these measures.
This activity was initiated by a letter issued by the NRC staff on May 17, 1978 [3], to all power reactor applicants, requesting information concerning the control of heavy loads ec r spent fuel.
The results of Task A-36 were reported in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." The staff's conclusion from this evaluation was that existing measures to control the handling of heavy loads at operating plants,.although providing protection from
'certainpotentialphoblems,donota'equatelycoverthemajorcauses d
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 al1 load-handling system", 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 r.onsequences is quantified in NUREG-0612 into four accident analysis evaluation criteria as follows:
" Releases of radioactive material that may result from o
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;
" Damage to the reactor vessel or the spent-fuel pool based o
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on calculations' of. damage following accidental dropping of a
. postulated heavy load is limited so as not to result in
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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
" Damage to equipment in redundant or dual safe shutdown o
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 staff 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 ciritical areas.
Staff guidelines resulting from the foregoing are tabulated in Section 5 of NUREG-0612.
1.3 plant-Specific Background On December 22, 1980, the NRC issued a letter [4] to Georgia Power Company, the applicant for VEGP requesting that the applicant review provisions for handling and control of heavy loads at VEGP, evaluate these provisions with respect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of conformance to these guidelines.
Georgia Power Company provided responses to this request by use of the V0GTLE FSAR Volume 19 which was transmitted to EG&G Idaho for review.
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EVALUATION AND RECOMMENDATIONS 2.1 Overview 1
The following sections summarize Georgia Power Company's FSAR review of heavy load handling at VEGP. accompanied by EG&G's evaluation, conclusions, and recommendations to the applicant for making the facilities more consistent with the. intent o PNUREG-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 load for the facilities as
~2000 pounds per the NUREG-0612 definition.
2.3 Guidelines The basic guidelines of NUREG 0612 for Phase II evaluations, are quoted and followed with:
A, a summary of the Applicants statements (from the FSAR), B, EG&G's Evaluation, and C, Recommendations.
The criteria includes guidelines 5.1.4 for Boiling Water Reactors only, and guideline 5.1.6 an alternate that may be required.
Vogtle Units 1 and Unit 2 are Pressurized Water Reactors and need to show consistency with guideline 5.1.2, 5.1.3 and 5.1.5 in Phase II.
The alternate guideline 5.1.6 may be required as appropriate.
2.3.1 Spent-Fuel Pool Area [NUREG-0612, Article 5.1.2]
(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.
,(2) "Each of the.following.is provided:
('a) Mechanical stops or electrical interlocks should be provided that prevent movement of the overhead crane 4
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NONEXEMPT HEAVY LOAD-HANDLING SYSTEMS VEGP 1 AND 2 Equipment Designation Load Rating Tons Fuel Butiding Cask lifting device jib A2109R4002 5
Spent fuel cask bridge Main crane hoist 125 Auxiliary hoist Monorail hoist 15 2
Containment Building Polar crane main hoist 2101R4001 475/225 Polar crane auxiliary hoist 50 Radial arm stud tensioner hoist assembly 2
Monorail hoist 2101R4011 2
Wall mounted cantilever jib 2101R4003 3
2101R4004 3
2101R4005 3
2101R4006 3
Wall mounted cable bridge winch 2101R4007 3
2101R4008 3
2101R4009 3
2101R4010 3
Other Areas" Back flushable filters, hatch covers, 3
resin charging tank RHR heat exchanger (2) 15 ESF chilled water chillers 2
Normal chilled water chillers 2
Normal chilled water pumps 2
Designation is based on load to be handled since the FSAR does not a.
identify the crane or hoist, only its capacity.
Exclusion basis remains to be evaluated.
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a load block over or within 15 feet horizontal (4.5 meters) of the spent-fuel pool.
These mechanical stops or electrical interlocks should not be bypassed when the pool contains
" hot" spent fuel, and should not be bypassed without approval from the shift supervisor (or other designated plant management personnel).
The mechanical stops and electrical interlocks should,be verffted to be in place and operational prior to placing " hot" spent fuel in the pool.
(b) The mechanical stops or electrical interlocks of 5.1.2(2)(a) above should also%ot be bypassed unless an analysis has demonstrated that damage due to postulated load drops would not result in criticality or cause leakage that could uncover the fuel.
(c) To preclude rolling if dropped, the cask should not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor level of the refueling building or other components and structures along the path of travel.
(d) Mechanical stops or electrical interlocks should be provided to preclude crane travel from areas where a postulated load drop could damage equipment from redundant or alternate safe shutdown paths.
(e) Analyses should conform to the 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.):
(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 dacayed sufficiently as shown in Table 2.1-1 6
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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 released, 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 c postulated load drop could damage equipment from redundant or alternate safe shutdown paths. Analyses have demonstrated that a postulated load drop in any location not restricted by electrical interlocks or mechanical stops would not cause damage that could result in criticality, cause leakage that could uncover the fuel, or cause loss of safe shutdown equipment.
(d) To preclude rolling, if dropped, the cask should not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor level of the refueling butiding or other components and structures along the path of travel.
(e) Analyses should conform to the guidelines of Appendix A.
E (4)
"Th'e 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."
A.
Summary of Applicant's Statements The overhead load handling systems are designed to minimize the potential for heavy load drops on spent fuel, safe shutdown and decay heat removal equipment.
The Spent Fuel Cask Bridge Crane design general bases are:
Prevention of a load being dropped from a single n
failure event in accord with General,De. sign Criteria (GDC) 2 " '
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Sharing of the Spent Fuel Pool Bridge Crane between o
Unit 1 and Unit 2 does not impair the safety of the plant according to GDL 5 Compliance with GDC 61 regarding safety'of fuel o
handling and storage under normal and postula'ted accident conditions.
Limit switches restrict crane traveling near or over spent fuel pools when the main hoist is handling loads in excess of 15 tons.
The arrangement of switches prohibits bridge and trolley travel with loads >15 tons outside the shaded area illustrated in FSAR drawing 9.1.5-1.
The path of the crane does not pass over either of the spent fuel storage pools.
4-The 125 ton capacity main hoist is single failure proof per NUREG 0554 therefore a spent fuel cask drop is not evaluated. A seismic restraint is mounted on each corner of the bridge to prevent bridge derailing during an earthquake.
Seismic restraints are provided at the trolley end trucks to prevent trolley uplift during an earthquake.
The trolley provides the structural frame support for the crane main hoist, auxiliary hoist and drive mechansism.
Tno 15 ton auxiliary and 2 ton monorail hoist provided on the main crane bridge are used in new fuel handling and maintenance tasks and are not involved with handling of the spent fuel cask.
The auxiliary hoist has two self adjusting de magnetic holding brakes.
The hook is single prong with a safety latch.
The hoist is designed to a capacity of 30 tons but rated,at 15 ton working capacity.
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B.3 EG&G Evaluation i
The data provided which shows that the 125 ton rated Spent Fuel Cask Bridge Crane is single failure proof is consistent with the NUREG 0612 guideline.
However the associated
-lifting devices it uses have not been discussed.' Loads
>l ton, the heavy load classification, and <15 ton apparentlymaybeha'tidledeanywhereinthecoverageareaof the crane.
Since the special limit switch controls establish a restridt$d area only for loads >15 tons, those loadsIfrom 1 ton to,15 tons handled by the main, auxiliary or monorail must be shown consistent with the NUREG 0612 5.1.2 guideline.
This has not been discussed.
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The auxiliary crane and monorail hoist 'are not reported to be single failure proof, therefore any load above 1 ton size that they handle must comply with one of. the options of NUREG 0612 5.1.2 guidelineandtherequired'inalysesmust conform to NUREG 0612 Appendix A.
Since the crane travel' s paths does not go over or into the pool option 5.1.2(3) o
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could not be used.
Strength, depending on safety factors used in design of the auxiliary hoist, may show that it is consistent with Article 5.1.6.
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- EG&GConclusionsandRecommendations Forthoseheavyloads-(>2000 pounds)handledpy~the' single failure proof Spent Fuel Cask Bridge' Crane the associated q
lifting: devices should be identified.
Information on these devices and their loads should be given to verify that they meet NUREG 0612 Section 5.1.6 requirements.
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Identify the option of NUREG 0612 Section 5.1.2 that is to be followed for the Auxiliary Crane and Monorail Hof st; provide information to show how VEGP meets or plans to I\\
comply with the details of the option chosen to be consistent with the guidelin,e.
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.
9.8 (2)
" Rapid containmerc isolation is provided with prompt automatic actuation on high radiation so that postulated releases are within limits of evaluation Criterion I of Section 5.1 taking into account delay times in detection and actuation; and analyses have been performed to show that evaluation criteria.II, III, and Iy of Section 5.1 are satisfied for postulated load drops in this area.
These analyses should conform to the guidelines of Appendix A.
0B (3) "The effects of drops of heavy loads should be analyzed and shown to satisfy tN evaluation criteria of Section 5.1.
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Loads analyzed could include the following:
reactor vessel i -
head; upper v ssel internals; vessel inspection platform; e
cask for damaged fuel;. irradiated sample cask; reactor coolant pump; crane load block; and any other heavy loads brought over or near the reactor vessel or other equipment required for continued decay heat removal and maintaining shutdown.. In this analysis, 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 of Appendix A."
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A.
Summary of Appiteant's Statements A review of overhead heavy load handling systems was performed in accordance with NUREG 0612 and following the guidelines of enclosure 3 to the Nuclear Regulatory Commission generic letter dated December 22, 1980 as amended on February 3,1981.
Included in ghts review were the p
systems in the Containment Building.
A review of plant arrangements was performed to evaluate load drops.
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results of the review for the Containment Butiding is given in FSAR Table 9.1.5-3. A copy of this table is reproduced as Table 2.2 on the following two pages.
The drawings referenced in the table are floor plot plans marked to show the load handling area of the hoists.
B.
EG&G Evaluation The factual ~ details of the options by which consistency with NUREG 0612 Article 5.1.3 may be shown are not addressed, even though the FSAR indicates that a review was made. The FSAR does indicate that the Containment Building crane and lifting devices are not single failure proof, so, option (1) cannot be.used.
The data of the FSAR Table 9.1.5-3 also fails to show consistency with requirements and makes no comm'itments.
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EG&G Conclusions and Recommendations t
There is an insufficient response.
Provide information and facts or a commitment that will show VEGP is consistent with either option (2) or (3) of NUREG 0612 Article 5.1.3 for the l
Containment Building.
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TA8Ld2.2.. TSAR TABLE 9.1.5-3 CONTAINMENT BUILDING OVERHEAD LOAD HANDLING SYSTEMS lleavy Load Equipment Load Load Lifting Re fe rence s '
___Handlina System Deslanator No.
Identification Woleht fIbl Device Drawi rms Polar crane (475/225 1 and 2 - 2101R4001 Integrated head package 420,000 Head lifting rig Figure 9.1.5-5 ton)
(4800 lb)
(sheets 15-18)
Reactor coolant pump 94,400 Reactor coolant ptamp lifting sling Reactor coolant pump 97,600 motor Rerueling machine 36,950 coeponent (maintenance)
Reactor coolant drain 360 tank pump Reactor cavity filtra-tion system ri t ter unit 375
- Pump and motor 250 Upper internals 132,000 Internals Ilft-Ing rig (18,350 lb) e Lowe r inte rna l s 260,000 Interna-Is lirt-Ing rig (18,350 lb)
Regenerative heat 4,200 exchanger Excess letdown heat 1,350 I
exchanger Refueling machine 1 and 2 - 210R6003 fuel assembly 1,600 Refueling machine rigure 9.5.1-5 i
gripper mast (sheets 15 and 16) at node A Radial are stud Reactor stud turnout
)Holst capacity Holst figure 9.1.5-5 tensioner hoist tool assembly is 4000 lb.)
(sheets 15 and 16) a t node B Quick grip stud tensioner 12 e
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2.3.3 Other Areas [NUREG-0612, Article 5.1.51 (1)' "If safe shutdown equipment are beneath or directly adjacent to a potential travel load path of overhead handling' systems, (i.e., a path not restricted by limits of crane travel or by mechanical stops or electrical interlocks) one of the following should be satisfied in addition to satisfying the general guidelines of Section 5.1.1:
(a) The crane and associated lifting devices should conform to the single-failurh proof guidelines of Section 5.1.6 of this report; 0B (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 not i
be taken for intervening floors unless justified by i
analysis.)
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(c) The effects of load drops have been analyzed and the results indicate that damage to safe shutdown equipment would not preclude operation of sufficient equipment to achieve safe shutdown. Analyses should conform to the guidelines of Appendix A, as applicable.
(2) "Where the safe shutdown equipment has a ceiling separating it from an overhead handling system, an alternative to Section 5.1.5(1) above would be to show by analysis that the largest postulated load-handled by the handling system would not penetrate the ceiling or cause spalling that could cause failure of the safe shutdown equipment."
A.
Summary of Apolicant's Statements I
l Miscellaneous cranes and hoists are provided to service and maintain mechanical equipment. Hoists and cranes have
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adequate capacity to perform lifting of components necessary i
for maintenance.
A review 'of overhead heavy load handling systems was performed in accordance with NUREG 0612 and following the guidelines of enclosure 3 to the NRC generic letter dated December 22, 1980 as amended on February 3, 1981.
The review included the Auxiliary Building, Control Building, Diesel Generator Buildin{, Auxiliary Feedwater Pumphouse and Nuclear Service Cooling Water Pumphouse.
The FSAR Table 9.1.5-2 lists equipment that may be lifted in these areas and includes hoist / crane capacity, load weight, maximum lift height, and the evaluation by code letter justifying an exclusion basis. A reference is given to identify crane load path drawings also.
B.
EG&G Evaluation The referenced drawings by legend symbols indicate the type of lifting unit and the table gives the capacity.
Other identification is not given.
There are 50 loads listed that have been evaluated for the buildings (other areas) including their various floor levels.
The bases for excluding overhead handling systems uses 5 hazard elimination categories that are similar to and generally meet the intent of those specified in Enclosure 3 of the NRC general letter on control of heavy loads.
However there are 6 loads using their Hazard Elimination Category 4.
These cannot be eliminated at present because the category includes a statement that says, "An evaluation will be performed to ensure that the postulated load drop will not affect plant safety."
C.
EG&G Conclusions and Recommendations
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IY the applica~nt's FSAR has included all cranes and hoists for, "Othe'r Areas," all but six of the exclusions are e
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consistent with NUREG 0612 Article 5.1.5.
The six loads identified in the FSAR Table 9.1.5-2 by hazard elimination Category 4 requires the analysis to determine if they can, in fact, be eliminated. Any which fail, by the analysis will required a commitment to show how the NUREG 0612 Article 5.1.5 guideline will be accomplished.
2.3.4 Single-Failure-proof Handling Systems'TNUREG-0612. Article 5.1.61 c.>
(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 evaluat.fon 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 handli.ng 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 satie,fy the evaluation criteria of Section 5.1:
(i) Provide dual or redundant slings or lifting l
devices such that a single component failure or malfunction in the sling will not result in uncontrolled lowering of the load; 1
(ii) In selecting the proper sling, the load used should be twice what 1.s called for in meeting
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Section 5.1.1(5) of this report.
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_,o (2) "New cranes should be designed to meet NUREG-0554,
" Single-Failure-Proof Cranes for Nuclear Power Plants." ' For operating plants or plants under construction, the crane should be upgraded in accordance with the implementation guidelines of Appendix C of this report.
(3) " 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 thp particular load have been evaluated and shown to satisfy the evaluation criteria of Section 5.1:
(a) Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the load; lift points should have a design safety factor with respect to ultimate strength of five (5) times the maximum combined concurrent static and dynamic load after taking the single lift point failure.
QR (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 The Applicant's FSAR does not address an intent to make handling system upgrades using the alternatives of NUREG 0612 Article 5.1.6 B.
EG&G Evaluation i
A suggested need for use of Article 5.1.6 guideline is discussed in 2.3.1 C above for the devices used with the Single Failure Proof Spent Fuel Cask Bridge Crane. All of the other hoists, not excluded, have not had sufficient commitments made to show if the alternative of upgrading system reliability requiring application of Article 5.1.6 will be used.
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EG&G Conclusions and Recommendations There are no recommendations on the basis of presently submitted heavy load handling information.
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3.1 Guideline Recommendations The associated lifting devices used with the Single Failure Proof Main Crane in the Spent Fuel Pool Area need to be shown as single failure proof or to meet NUREG 0617. Article 5.1.6.
u The Auxiliary Crane and Monorail Hoist on the Spent Fuel Cask Bridge Crane require information submittal to show what upgrade will be used to make them consistent with one of the options of NUREG 0612 Article 5.1.2.
The information on the Containment (Reactor) Building cranes and hoists is insufficient to shov consistency with Article 5.1.3.
Provide information or a commitment to show how VEGP will be brought into consistency.
Six loads handled by hoists in other areas require additional study to show if the hoist may be exempt from Article 5.1.5 requirements.
3,2 Summary The VEGP FSAR was used as a basis for review of their compliance with NUREG 0612 Article 5.1.2, 5.1.3 and 5.1.5.
The FSAR provides fine details on some components and does not address others.
It makes statements such as " reviews were performed"... but provides no information on the findings that can be applied to show consistency with the'NUREG guidelines required for this report. As a consequence this Technical Evaluation -Report indicates that addi'tional 'infornration is needed for part are all of each guideline for an acceptable consistent finding.
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NUCLEAR REGULATORY COMMISSION
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j MEMORANDUMJOR:
E. Adensam, Chief. Licensing Branch No. 2, Division of Licensing
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FROM:
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Olan D. Parr, Chief, Auxiliary Systems Branch, Division of Systems Integration
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - AUXILIARY SYSTEMS BRANCH The enclosed request for additional infomation and branch technical positions covers those portions of.the Vogtle FSAR, up to and including FSAR Amendment No. 5, for which the. Auxiliary Systems Branch has pr mary responsibility.
i Attachment I to the enclosed request provides our guidance with respect to.
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the fire protection associated circuits review.
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The enclosure identifies areas for which we need additional infomation or have taken positions. The positions cover internal flooding, internally generated missiles, pipe breaks, spent fuel pool cooling, diesel generator building ventilation 'and water ham.er.
Our review of the heavy loads h'andling systems and auxiliary,feedwater system reliability are being ~perfomed by our consultants EG&G Idaho and Brookhaven National Laboratory (BNL), respectively. Formal requests for additional information, if required by those labs, will be transmitted under separate cover. By letter dated February 24, 1984 we transmitted to T. M. Novak a draft. technical evaluation report for heavy loads prepared by EG&G and a conference call has been held'between the applicant,~ EG&G and ourselves to discuss the additional infomation required to complete the heavy' loads revi ew.-
The auxiliary. feedwater relf, ability evaluation for Vogtle has not yet been completed by BNL and we do not know if additional information will be required.
Olan D. Parr, Chief
. Auxiliary Systems Branch Division of Systems Integration
Enclosure:
.As Stated
'cc w/ enclosure:
'I R. Mattson J. E. Knight D. Eisenhut--
A. Ungaro J. N. Wilson T. Novak L. Rubenstein M. Miller i
F. Rosa W. LeFave V. Benaroya
Contact:
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l AUXILIARY SYSTEMS BRANCH REQUEST FOR ADDITIONAL INFORMATION V0GTLE ELECTRICAL GENERATING PLANT, UNITS 1 & 2
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DOCKET NOS. 50-424/425
-Provide the results of an analysis to show that site flooding due 410.02 (SRP 3.4.1) 7tD a natural draft cooling tower ba. sin failure or a circulating water system failure in the plant yard will not cause flooding or damage to safety-related equipment. In your analysis consider the possible effec.ts of erosion on underground safety-related piping and tunnels.
410.03 In Section 3.4.1'you state that the nominal finished grade'eJeva-(SRP 3.4.1) tion is 219 feet, 6 inches. To allow us to evaluate the flooding effects from various sources also provide the minimum elevation of entrances to all safety-related structures including the
-ultimate heat sink pump. house, and verify that the 219 f?st, 6 inch grade elevation also applies to the ~pumphouse.
- 410.04(RSP)
In Section 3.4.1.1.2 you state that each area of the plant was (SRP 3.4.1)
, reviewed to detemine the failure of nonseismic Category I tanks, vessels and other process equipment that results in the most adverse flooding conditions. Provide.a discussion of the larger indoor tanksthat were considered in your analysis and show,how it is detemined that no ' safety-related equipment would be affected.
It is our position that a single failure should also be conside' red coincident with the failure of these nanseismic Category I systems.
410. 05(RSP).
It is our' position.that when an internally generated missile source (SRP 3.5.1.1 (inside or outside containment) is a'nonsafety-related system or and 3.5.1.2)-
component. then the single failure criterion should also be met.
To show that your design' meets'this~ position, verify that missiles
' rom nonsafety-related sources will not damage any safety-related (guipment.
410.06 -
Ir. Section 3.5.1 of your FSAR you list gravity-generated missiles
($RP 3.5.1.1 as externally. generated missiles. Verify that gravity-generated and 3.5.1.2) missiles were also considered as internally generated missile sou.ces both inside and outside containment.~ Also verify that nonseismic Category I gravity-generated missile sources are seismically supported,"if they could affect any seismic Category I structures, systems or components.
410.07 In addition to the equipment listed in Table 3.5.1-7 as having (SRP3.5.2) tornado mis'sile protection also verify that tornado missile pro-tection is provided for the nuclear service cooling tower valve Mouse, HVAC intakes and exhausts. Also describe a typical tornado missile barrier for HVAC openings using the control building air intakes as an example.
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410.08 Identify any openings in safety-related structures that are not.
(SRP 3.5.'!)
tornado missile protected and provide justification for not
? having such protection.
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410.09(RSP)...en sheet 2 of FSAR Tat >1e 3.6.1-2 Item B.3, you state that your (SRP 3.6.1, 7 Jiesign confoms to position B.3.b.(3) of BTP ASB 3-1.
You BTPASB3-1) further state that this criterion has also been applied to single-purpose and high-energy systems since the same quality.-
j desta, c:nstruction and inspection standards are used, as for the dual-pt ' nose moderate energy systems. It is our position that you assume a single active failure coincident with all pipe breaks except in the dual-purpose moderate energy systems 'hs.
described in.our branch position, as you have indicated.in'the
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textofyourFSAR(3.6.1'.1.G). Verify that sdch single active 411 ores have been considered ard revise the FSAR accordingly.
j 410.10 Iri Table 3.6.2-2 (Sheet 7) you have provided a 'high energy pipe
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(SRP 3.6-1) break analysis for Room No. R-C83. In this table you refer.to Sheets 88, 89, 90, 91, 92 and 96 of Figure 3.6.1-1 for the high
. energy piping in this room. We have reviewed these sheets and
,they do not appear to coincide with Rm R-C83 which is at the 143 ft.
e 6 in. elevation of the auxiliary building..The piping on the referenced sheets all appear to.be above.that elevation. Also on' Sheet 7 of Table 3.6.2-1. you refer to Table 3F-1 Sheet 16 4
for the' identification of-safety-related' equipment in Room R-C83.
This also appears to be in error and Sheet 14 of Table 3F-1 should i'
be referenced in lieu of Sheet 16. Correct these apparent. dis -
crepancies.and review the contents of Table 3.6.2-2 to ensure no other errors.of this nature exist.- As an example for Rooms R-C88 and C89 Sheets 18 and 20 of Table 3F-1 are referenced in lieu of Sheets 16 and 18 which are the coirect references.
'410.11 In Table 3.6.2-2 (Sheets 7 and 12) you stated that stress analysis (SRP 3.6-1) results confim that no breaks will occur in the high energy lines located in R-C83 and R-C95.
Identify all.high energy. lines in these rooms by system and-line size and provide a basis for not assuming at least one intemediate break location. Also verify that all high energy lines in these rooms are designed to* seismic Category I since for, purpose of equipment protection we assume a break anywhere in nonseismic 1
Category I piping.
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L 410.12 '
Our review of your pipin'g isometrics and P& ids is hampered by (SRP3.6.1) the fact that we do not have a legend that indicates which system
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identification number corresponds to what system (e.g..1201 i
4 refers to reactor ecolant system). Please provide such a legend 1
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fn order that we may complete our review in the scheduled time frame.
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.. flooding from sources within the room (R-B15) ysis states that
. In Table 3.6.2-2 (Sheet 1), your ficoding anal 410.13(RSP) will affect only -
(SRP 3.6.1
-$s,guipment within the same train / subsystem and, therefore safe BTP ASB 3-1)
..sbutdown will not be compromised. This is not acceptable
..3ntles: the only flooding source is a dual-purpose moderate
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eiergy systen since otherwise an additional single active failure must be assumed. Revise your flooding analysis for this room and for all other rooms where you have made the same assumption.
In FSAR Section 3.F.2 of your hazard analysis you state that when 410.14(RSP)
(SRP 3.6.1 the postulated hazard occurs and results in damage to one of two i
and SRP or more redundant. trains, single failure of components in other.
- 3. 5.1.1 )
trains (and supporting systems) are not' assumed. Again this.
assumptiem is only valid when the hazard is a failure of a dual-purpose m:derate energy piping system or when an internally generated missile source is a safety-related s'eismic Category I system. For all other failures, a coincident single active failure must be assumed. Revise your FSAR and design as necessary to meet the single failure criterion for all oder hazards.
410.15(RSP)
In FSAR Sections 3F.2.2 and 3F.2.4 regarding pipe break and (SRP 3.6.11 flooding assumptions you state your analysis includes the effect-of flooding from the worst-case pipe crack in each roon or general area. 'It i's our position that for flooding analysis i
purposes, the complete failure of ncnseismic Category I moderate energy piping systems should be considered in lieu of cracks in -
detennining the worst case flooding. condition. Revise your analysis and FSAR as necessary to include the worst case flooding' condition for each room or area in the cent of a complete failure of the most limiting nonseismic Catt.r. cy I moderate energy' line.
410.16 Table'3F-1 provides your hazards analysis for the auxiliary building, i.evels B, C and D.
In your pipe break analysis for the (SRP 3.6.1) room identified in this table, you have not made any checks in the column for " moderate-energy cracks within the room do not
. adversely affect saf ety-related equipment in the room.". Identify why this category has not been checked for any of these rooms since it appears that a moderate energy pipe crack evaluation was not perfomed in these rooms.
For the flooding analysis results for each of the rooms identified 410.17 fn Table 3F-1, identify the worst case flooding source, and as (SRP3.6.1) an example of your analysis, provide all the assumptions,made in arriving at the maximum flooding level of one inch for area
. Jt-C88 (Sheet 15 of Table 3F-1). The information provided should
" include how you arrived at the flooding rate, the flooding source
.aht' other possible sources, the level necessary to affect safety-related equipment, and a description of how the flood level is I
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limited to one inch.
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410.18 In FSAR Section 3F.4.2 you state that the blowdown from a main (SRP 3.6.1) feedwater line break results in the maximum flood level. How:
ever, you have not provided sufficient infomation for us to
" detemine if the resulting flooding is acceptable. Provide an
.- svaluation of the resulting flooding, including how the accumulated. '
" iater drains from the areas. ari,d verify that flooding of other
. safety-related areas will not result. Your evaluation should also identify the maximum resulting flood level for each main feedwater piping area and the minimum flood level necessary to
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affect safety-related equipment.
410.19(RSP)
In FSAR Section 3F and in Table 3.6.1-2 you have deviated from (SRP'3.6.1,.
Position B.1.a(1) of BTP ASB 3-1 in that you.have not provided BTP ASB 3-1) for jet im;iingement effects from the nonmechanistic postulated steam and feedwater line breaks. Provide justification for deviating from this, pos'ition or provide the results of an analysis to show that jet impingement will not prevent safe plant shutdown.
410.20 Table 3F-3 is intended to provide the peak values of MSIV/MFIV SRP 3.6.1)
' compartment pre,ssure and temperature. However it only provides J
,the design temperature conditions. Revis.e this table to include the calculated temperatures. Also revise the table'to include the analysis fo'r pipe breaks in areas outside the restraint wall of the auxiliary building. Fpr pipe breaks in areas outside the restraint walls of the control and auxiliary building verify that
' double-ended. ruptures of piping were considered in the pressure and temperature. analyses and identify any safety-related equip,
ment in the areas.
41 0.21'
-Verify that'the safety-related Train A and Train B electrical
- (SRP 3.'6.1) conduits identified on Sheet 14 of Table 3F-1 are not necessary for safe shutdown.
410.22 In various rooms identified in Table 3F-1 a general statement is i
.(SRP3.6.1) made regarding " flooding from sources within this room will not impair the safe shutdown capability, of the safety-related equipment."
Provide a basis for this assumption for each of the rooms identified in Table 3F-1 that has this statement.
410.23 Your pipe break analysis on Sheet 45 of Table 3F-1 indicates there (SRP 3.6.1) are no high energy lines in the centrifugal charging pump room.
train A (Room R-C115), Please correct this obvious error.,
410.24 With respect to your AFW pump rooms pipe break analysis, verify (SRP 3.6.1) --that a pipe break or cract in the comon area of the pump house cannot result in loss of more than one AFW train. Also revise Table 3F-2 to include the calculated temperatures following a steam line break and the calculated flood levels following an AFW discharge line break.
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410.25 For all areas of the plant where watertight doors are relied on (SRP 2.6.1 and for flood mitigation verify that these doors will be indicated
- 3. 4.1 )
" and alanned in the control room, and that the plant technical
.. specifications will include surveillance requirements for these 7 doors with appropriate limiting conditions for operation.
410.26 In Section 3F.1 you state that an analysis for the effects of (SRP3.6.1 a circulating water system failure have been provided. However, and10.4.5) this analysis has apparently been omitted. Provide this analysis and the following information:
The maximum flowrate through' a completely failed expansion a.
joint.
.b, The potential for and the means provided to detect' a failure in the circulating water transport system barrier such as the expansion joints.
Include the design and operating pressures of the various portions of the transport. system barrier and their relation to the pressures which could exist during malfunctions and failures in the' system (rapid valve
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closure).
The time. required to stop the circulating system water flow c.
(time zero beira the instant of' failure) including al1 inherent delays s,uch as operator reaction time, drop out times of the control circuitry and coastdown time.
d.
For the. worst case postulated failure give 'the rate of rise of water in the associated spaces and total; height of the.
water when the circulating water system flow has been stopped or overflows to site grade.
e.
For each flooded space provide a discussion, with the aid of drawings, of the protective barrier provided for all essential systems that could become affected as a result of flooding.
Include a discussion of the consideration given to passageways; pipe chases and/or cableways joining the flooded space to the spaces containing safety-related equipment.
410.27 Verify that the flood levels from a main feedwater line break in (SRP 3.6.1) the turbine building is less than that resulting from a break in the circulating water system. Otherwise provide;an analysis to -
show that resulting flood levels will not affect safety-related-equipment via interconnections between the turbine building and
~ safety-related structures.
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410.28 Provide the results of a flooding analysis for a postulated moderate L
(SRP 3.6.1) energy leakage crack in the CST and RWST. suction lines for the.
. various plant areas that may be affected.
410.29 In Amendment 3 to the FSAR you deleted the statement that the spent (SRP9.1.2)
.iffel pool liner was seismic Category I.
If the fuel pool liner 4
is not seismic Category I provide the information identified in SRP Section 9.1.2, Item III.3.b regarding failure of the fuel pool liner.
410.30(RSP)
In FSAR Section 9.1.3.1 you state that the design decay heat load (SRP 9.1.3) for the spent fuel' pool cooling system was calculated following the guidance of ANS 5.1.
In FSAR Section 9.1.3.7 you state, that standard. Westinghouse methods were used'for decay heat -load cal-culations.
It is our position that either ANS 5.1.1978 or 'BTP ASB 9-2 be used to calculate decay heat loads. ' Clarify what- '
methodology was used to calculate the design basis heat load for the spent fuel pool cooling system.
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410.31 The standby nuclear service coolingwater (NSCW) pump for each train (SRP 9.2.1) starts automatically on low pressure in the discharge manifold during accident conditions. Verify that the loss of one of the two operating pump.s will result in a low enough discharge pressure to start the. standby pump,.and also 'specify.if the standby pump will automatically start on loss of discharge manifold pressure during normal operati.ng conditions.
J 410.32 Air operated valve CV-94461and 9447'are the seismic Category I (SRP. 9.2.1) boundaries between the NSCW system and the nonseismic Category I blowdown line. Describe what signals close CV-9446 and 9447 to prevent drainage from the NSCW system causing a loss of system
. function or flooding problems.
If manual i. solation is relied on' describe the method of detecting the leakage and ve'rffy that ade-quate time is available for' operator action.
410.33 In FSAR Amendment No. 4, you revised Figure'9.2.1-1, Sheet 5, (SRP 9.2.1) to include.a two-inch intertie from the Train B NSCW discharge header to Train A (Figure 9.2.1-1, Sheet 1). Presumably, the interconnection goes to the train A discharge header.
- However, i
the interconnection is not shown on Sheet 1 of Figure 9.2.1-1.
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Revise Sheet I to be consistent with Sheet 5.
Also provide a.
discussion of the purpose of this intertie including any safety-related function the interti,e may have.
4)0.34 -
,For the component cooling l water (CCW): system identify the minimum (SRP 9.2.2 flow requirements and maximum allowable CCW temperature at the and9.2.1)
~. inlet of each component served by the CCW system. Provide the same information for equipment cooled by the NSCW system.
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4 410.35 Verify that flooding analyses have been performed for a failure (SRP9.2.3) o.f the nonseismic Category I demineralized water makeup system
.'; where the piping runs through safety-related structures such as the auxiliary building", control building, and tunnels containing
- safety-related equipment.
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410.36 Provide the component design data, including the minimum net posi-i (SRP 9.2.5) tive suction head (NPSH) requirements for the NSCW pumps and NSCW I
transfer pumps, in order that'we may detemine that the minimum system flow requirements and NPSH requirements are met.
410.37 In order to pemit an evaluation of the ultimate heat sink and j
(SRP9.2.5) other heat removal systems, provide an analysis of the thirty-day l
period following a design basis accident listing'the total heat i
raiected, the sensible heat rejected, the station auxiliary heat rejectedand the decay heat released from the reactor.
In submitting the results of the analysis requested, include the following infomation in both tabular and graphical fam:
1.
The total integrated decay heat; I
i
'2.
The-heat rejection rate and integrated heat rejected by the station auxiliary systems, including all operating pumps, i
ventilation equipment, diesels and other sources; 3.
The heat rejection rate and integrated heat rejected due to sensible heat removed from the containment and the primary system;
-4.
The total integrated heat due to the above; 5.
The maximum allowable inlet water temperature taking into account' the rate at which the heat energy must be removed, cooling water flow rate, and the capabilities of the respective heat exchangers.
6.
The available NPSH to the NSCW pumps and tiransfer pumps' at"the minimum ultimate. heat sink. water level versus the r,equired NPSH.
'Use' the methods set forth in either our.BTP ASB 9-2 or ANS 5.1, 1978 to establish the inptrt due to fission product. decay and heavy t
element decay. Assume an initial service water temperature based on the most adverse conditions for nomal operation.
410.38 5 top'of'eachcoolingdower"fanc'ellthere'isadebriscatcher
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(SRP9.2.5),.. designed to prevent trash from entering the fan cells. Describe the details of these debris catchers and verify that they will not become gravitational missiles as a result of an earthquake f
or high winds.
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1 410.39(RSP)
Verify that sufficient condensate storage tank (CST) capacity (SRP 9.2.6
. exists to cool the reactor coolant system to the RHR cut in and 5.4.7) temperature assuming the most limiting single active failure.
In determining the time required to perform such a cooldown
^iEnlysafety-gradeequipmentshouldbeassumedavailablein
- accordance with BTP RSB 5-1.
410.40 With.regards to the heat tracing provided for the. safety-related (SRP 9.2.6 and portions of the piping systems for the condensate makeup system, 9.2.7) and the reactor water makeup system, describe the means of detecting heat tracing system failure 'and whether indication and/or alarms are'provided in the control room. Provide the same information for. storage tank heaters.
410.41 Identify the minimum gravity flow makeup rate.from the reactor (SRP 9.2.7 and water makeup tank to the spent fuel pool and verify that.it is 9.1.3) sufficient to makeup for the maximum possible evaporative losses from the pool.
410.42 In FSAR Section 9.2.8 you state that auxiliary component cooling
~
(SRP 9.2.2 and. water (ACCW) cooling is available irrespective of which NSCW train 9.2.1)
.is in service. From the description in FSAR Section 9.2.1 it was not clear whether one or both trains of NSCW would normally be operating. Please provide a description of the normal mode of operation of. these two systems.
410.43 In Section 9.3.1.4 of your FSAR you state that the compressed air (SRP 9.3.1) system conforms to the standardsof ISA-57.3. Since FSAR Section 9.3.1.4 is only related to testing and inspection, verify that the instrument air
- portion of the compressed air system conforms to the guidelines'of ISA-57.3 (ANSI MC 11.1-1976) regarding air quality standards as identified in III.2 of SRP Section 9.3.1.
410.44 In FSAR Table 3.2.2-1, Item 19 under the instrument and service (SRP9.3.1) air system heading identifies' safety.related piping.and valves (other than containment isolation) associated with the air system. However FSAR Section 9.3.1 indicates that no safet related piping or valves (except for containment isolation)y-exists in the air syst. ems. Clarify this apparent discrepancy.
If there are some safety-related piping and valves associated with safety-related accumulators identify their function and l
provide a typical drawing of the, accumulator system.. Also j
t discuss the testing capability and frequency for such accumulator i
systems.
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410.45 Provide a drawing showing the drain system and sumps for the AFW (SRP 9.3.3) pumphouse and CST and describe the means of preventing flooding
- (due to sump overflow or backflow through drain system) of the AFW pump rooms due to. drainage from the CST area.
410.46 7 Jescribe the means of preventing backflow through the drain systems --.
(SRP 9.3.3) of the control and auxiliary buildings for areas where train A and train B rooms / areas drain to a common header and no check valves u-or closed isolation valves are installed.
410.47 In Amendment 3 to the FSAR you revised Section 9.3.3.3 whereas (SRP 9.3.3) originally there were watertight doors 'for a'll ESF equipment rooms and with the revision watertight doors would only be used' for ESF rooms if a flooding analysis showed they were necessary.. Indicate which ESF rooms will not have watertight doors. and provide the results of your flooding analysis that shows the doors are not necessary. The analysis should show that the doors are not necessary for flooding into or out of the room. Also revise FSAR Tables 9.3.3-3 and 9.3.3-4 to reflect the fact that all ESF equipment
, rooms are not watertight and revise the FSAR layout drawings as
,necessary.
410.48 FSAR Table 9.4.1-2 indicate's that'the control build'ing. ventilation (SRP 9.4.1) system is designed to. maintain a 1/2-inch water gage (WG)' pressure inside the control room. FSAR. Sections 9.4.1 and 6.4 indicate that the control room normal HVAC' system maintains a positive 1/8-inch and 1/4-inch WG pressure respectively in the control. room. Clarify these apparent discrepancies. Also,' verify that positive pressure is maintained by the cmergency control room HVAC system.
In' FSAR Amendment 3, you revised FSAR Section 6'4.'3 to eliminate the 410.49 (SRP 9.4.1) automatic control room isolation signal as a result of a safety injection signal. However, FSAR Section 6.4.2 and FSAR Figure 7.3.6-1 (control room isolation logic) indicate that the safety l
injection automatic isolation capability still exists. Revise the l.
FSAR to show' the actual design and if the safety injection isolation
' signal !)as been deleted, provide your basis for the design change.
410.50 FSAR Figure 9.4.1-2 (sheets 1 through 3) show that the control room (SRP 9 4.1),
air intake smoke detectors have some automatic isolation capability.
l However, the FSAR text indicates that the smoke detectors perform l
no automatic function (except alarm). Clarify this apparent dis-
,crepancy and describe.the details of any automatic, functions,the l?
' smoke detectors may perform.
l' 41 0.51
- Neither FSAR Section 9.4.1 or 9.4.5 provides a description of which (SRPs 9.4.1 essential HVAC system provides cooling to ti.e cable spreading rooms and 9.4.5) during emergency or accident conditions. Provide a description of l:
how the cable spreading rooms and surrounding a mas are ventilated j
during emergencies or accidents.
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