ML20086L990

Control of Heavy Loads at Nuclear Power Plants,Limerick Generating Station Units 1 & 2 (Phase Ii), Draft Technical Evaluation Rept

ML20086L990
Person / Time
Site:	Limerick
Issue date:	10/30/1983
From:	Friederichs S, Stickley T EG&G, INC.
To:	NRC
Shared Package
ML20086L986	List: ML20086L990
References
CON-FIN-A-6457, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8402150435
Download: ML20086L990 (29)

Text

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1 CONTROL OF HEAVY LOA.DS AT NUCLEAR POWER PLANTS LIMERICK GENERATING STATION UNITS 1 & 2 (PHA5E II)

Docket No. [50/352 and 50/353]

Author S. L. Friederichs Principal Technical Investigator T. H. Stickley Published October 1983 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 fin No. A6457

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ABSTRACT The Nuclear Regulatory Commission (NRC) has requested that all nuclear plants, eitber 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 EGiG's evaluation and recommendations for Limer ick Generating Station Units 1 and 2 for the requirements of Sections 5.1.4, 5.1.5, and 5.1.6 of NUREG-0612 (Phase II). Section 5.1.1 (Phase I) was covered in a separate report [1].

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EXECUTIVE

SUMMARY

Limerick Generating Station Units I and 2 is not totally consistent with the _gqidelines of NUREG-0612. In general, inconsistencies exist in the folicwPng areas:

o Special Lif ting devices o Inter' acing Lift points.

The mai report contains recommendations which will aid in making the above items consistent with tha approp-iate guidelines.

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. 1 CONTENTS 1

ABSTRACT ............................................................. 11 E X E C U T I V, E- St MMA R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii .

1. INTRODUCTION .............. ..................... ............... I 1.1 Furpose of Review ......................................... I 1.2 Generic Background ........................................ I 1.3 Plant-Specific Background ...... ......... ............ ... 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

.............................................. 22 3.1 Guideline Recommendation ................................. 22 3.2 Additional Recommendations ................................ 22 3.3 Summary ................................................... 23

4. REFERENCES .... ................................. ............... 26 TABLCS 2.1 Nonexempt Heavy Load-Handling Systems ........................... 5 3.1 NUREG-0612 Objectives Compliance Matrix ......................... 24 iv

4 CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS LIMERICK GENERATING STATION UNITS 1 AND 2 (PHASE II)

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1. INTRODUCTION 1.1 Purpose of Review 4

This technical evaluation report documents the EG&G Idaho, Inc.,

review of general load-handling policy and procedures at Limerick generating Station Units 1 and 2. This evaluation was performed with the objective of assessing conformance to the general load-handling guicelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" [2], Sections 5.1.4, 5.1.5, and 5.1.6. This constitutes Phase II of a two pbase 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 established by the U.S.

Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to assure the safe handling of heavy loads and to recommend necessary changes to these measures. This activity was initiated by a letter issued by the NRC staff on May 17, 1978 [3], to all power reactor applicants, requesting information concerning the control of heavy loads near spent fuel.

i 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

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heavy loads at operating plants, although providing protection from certain potential problems, do not adequately cover the major causes of load-handling accidents and should be upgraded.

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In order to upgrade measures for the control of heavy loads, the staff developed a series of guidelines designed to achieve a two phase objective using an a:cepted approach or protection philosophy. The first phase 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 secono phase of the staff's objective, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ensure that, for load-handling systems in areas where their failure might result in significant consequences, either (a) features are provided, in addition to those required for all load-handling systems, to ensure that the potential for a load drop is extremely small (e.g., a single-failure proof system) or (b) conservative evaluations of load-handling accidents indicate that the potential consequences of any load drop are acceptably small. Acceptability of accident i

consequences is quantified in NUREG-0612 into four accident analysis evaluation criteria as follows:

4 o " Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than 1/4 of Part 100 limits);

o " Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy load does not result in a configuration of the fuel such that k,77 is larger than 0.95; o " Damage to the reactor vessel or the spent-fuel pool based l on calculations of damage following accidental dropping of a l postulated heavy load is limited so as not to result in-2 I

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water leakage that could uncover the fuel, (makeup water provided to cvercome leakage should be from a borated source of adequate concentration if the water being lost is i

_. borated); and o " Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions."

The approach used to develop the 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 i

movement over critical 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 Philade'phia I

Electric Co., the applicant for Limerick Generating Station Units 1 and 2 requesting that the applicant review provisions for handling and control of heavy loads at Limerick Generating Station

. Units 1 and 2, 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. Philadelphia Electric Co. provided responses to this request on June 18, 1981 [5], April 2, 1982 [11] January 31, 1983 [6],

and June 13, 1983 [12].

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2. EVALUATION AND RECOMMENDATIONS 2.1 Overview ThehhllowingsectionssummarizePhiladelphiaElectricCo.'sreviewof ~

heavy load handling at Limerick Generating Station Units 1 and 2 accompanied by EG&G's evaluation, conclusions, and recommendations to the applicant for making the facilities more consistent with the intent of NUREG-0612.

2.2 Heavy Load Overhead Handling Systems j 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 is 700 lbs per the NUREG-0612 definition.

5 2.3 Guidelines Section 5.1.1 of NUREG 0612 includes general guidelines for (1) safe load paths, (2) procedure (3) crane operator training and qualifications, (4) special lifting devices (5) lifting devices that are not specially designed, (6) crane inspection testing, and -

maintenance, and (7) crane design. These guidelines were addressed by the applicant and evaluated in a separate report [1]. Specific

. requirements for overhead handling of systems operating in the reactor building and other plant areas containing equipment required for reactor shutdown, decay heat removal, or spent fuel pool cooling are indicated below:

2.3.1 Reactor Building {NUREG-0612. Article 5.1.4]

(1) "The reactor building crane, and associated lifting devices used for handling the above heavy -loads, should satisfy the single-failure proof guidelines of Section 5.1.6 of-this-report.

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TABLE 2.1. NONEXEMPT llEAVY LOAD flan 0 LING SYSTEM SUHJfCT TO NUREC 0612 CRl11HIA P

Crane or. Holst Equipment. .

Saf'eLy ReIated Item Safety itoIated Item on hg.m Number Name/Se rvice Drawino . Capacity _ in toad _ Path Next tower Elevallon 1 00-H201 Reactor building overhead crano M122 125/15 Ton Yes Yes i g 2 1A-H501 Diesel generator building cranes M145 15 Ton Yes ' '

IB-H501 N/A 1C-H501 '

10-H901 3 00-11511 Spray pond pump house hoists M388 3 Ton each. Yes Yes 00-H513 4 _a Spray pond RHR and ESW pumps M388 _a Yes Yes ya rd crane 5' 00-Il133 Control room HVAC left beam Mill 3 Ton Yes Yes M126 6 1 A-It203 Recirculation pump motor hoists M119 2f4 Ton Yes Yes

- 1B-6203 7 10-Il216 Core spray pumps hoist M110 5 lon Yes Yes 8 10-11217 Core spray pump hoist MI18 5 Ton Yes Yes 9' 10-Il220 Containnent equipment door hoist M119 6 Ton Yes yes 10 10-H229 - CR0 platform hoist M119 1 Ton Yes N/A 1A-H233 MSRV service hoists M234 2 Ton Yes Yes 1 A-ll234 1A-H235 11 18-H233 MSRV remava l hoist M234 2 Ton Yes Yes 1 B-112318 IB-H235 10-H230.

10-H232 12 10-H237 Containment hydrogen recombiner M120 1 Ton Yes Yes cover hoists 13 00-Il126 - Control room IIVAC equipment M115 2 Ton Yes Yes hoist M130 3 Ton

a. Iloist/ crane to be borrowed f rom other loca tions wh?n needed.

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OR (2) "The effects of heavy load drops in the reactor building should be analyzed to show that the evaluation criteria of Section 5.1 are satisfied. The loads analyzed should include: shield plugs, drywell head, reactor vessel head; .

steam dryers and separators; refueling canal plugs and gates; shielded spent-fuel shipping casks; vessel inspection platfonn; and any other heavy loads that may be brought over or near safe shutdown equipment as well as fuel in the reactor vessel or the spent-fuel pool. Credit may be taken in this analysis for operation of the Standby Gas Treatment System if facility technical specifications require its operation during periods when the load being analyzed would be handled. The analysis should also conform to the guidelines of Appendix A."

A. Summary of Applicant's Statements The following cranes are capable of carrying loads over spent fuel.

a. Reactor enclosure crane--125/15 T capacity

b. Refueling platform--one fuel handling hoist and I two auxiliary 1000 lb capacity hoists

c. Fuel pool jib cranes--1000 lb capacity

d. Fuel channel handling boom--jib crane '

500 lb. capacity.

The refueling platform crane, the fuel pool jib cranes, and the fuel channel handling boom do not carry loads heavy enough to be classified as heavy loads and therefore the reactor enclosure crane is the only crane capable of carrying heavy loads over. spent fuel in the storage pool or reactor vessel.

The reactor enclosure crane has an overhead bridne crane designed for a rated. load of 125 tons with a.15 ton hoist load. Maximum critical loads (MCL) are 110/6 tons.

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The reactor enclosure crane will lift the following heavy loads:

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-- Weight Load (tons)

1. Reactor well shield plugs 90

( 2. Dry well head 104

3. Reactor vessel head 92

4. Steam dryer 45

5. Steam separator 74

6. Fuel pool stop logs 35

7. Drier separator storage pit canal plugs 45

8. Fuel pool gates 3 (each)

9. Refueling shield 22

10. Spent fuel shipping cask 100

31. Service platform 6

12. RPV head insulation 9

13. Crane load block 5

14. Fuel pool and service platform jib crane 2.2 A detailed evaluation of the crar.e was made with respect to '

the General Requirements of the regulatory Guide 1.104 which preceded NUREG 0612. A point by point comparison of the crane features with the sections of the Regulatory Guide 1.104 is presented in Table 9.1-12 of the Limerick Final Safety Report.

The applicant considers the crane to generally-comply with the requirements of NUREG 0612, except that the auxiliary system does not meet all of the design criteria of either NUREG 0554 or R.G 1.104. The auxiliary System does not meet these requirements in that the means of load attachment is not of a redundant design. The hoist will therefore be downgraded from 15 tons to 6 tons when handling critical 7

loads. This will effectively double the design safety factor and provide additional margin for wear and dynamic load.

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The applicant has performed a seismic analysis of the crane to demonstrate that the overhead handling system can retain the load during a seismic event equal to a safe shutdown I

earthquake. Load bearing members and the main hoist of the crane are designed in accordance with seismic Category I

! criteria so that the crane can structurally withstand the safe shutdown earthquake (SSE) and maintain the fully rated load in a static position during the period following an SSE.

The reactor enclosure crane is prevented from carrying loads over or near the spent fuel pool by zone travel limit j switches on the bridge and trolley. The switches can be l bypassed by key locked switches. Administrative procedures will be developed prior to plant operation to control by passing of the interlocks.

The following loads must be carried over the reactor vessel.

a. Reactor well shi_ eld plug

b. Dry well head

c. RPV head.

d. Steam dryer

e. Steam separator

f. Service platform-

g. Service platform support

h. Refueling shield

i. Jib cra'ne

j. Head stud rack'.

None of the lifting devices for these loads meet the' single-failure free criteria.of NUREG 0612 Section 5.1.6.

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4 They have therefore been evaluated with respect to Criteria I through III of NUREG 0612.

_, The applicant considers the lifts of the following loads to

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meet the criteria of NUREG 0612 although not in full compliance.

a. Spent fuel cask

b. Refueling shield q c. Fuel pool stop logs

d. Fuel pool gates

e. Fuel pool jib cranes, channel handling crane, and head stud rack.

The lifting devices for items a, b, and c will be special 4

lifting devices. The lifting devices for d and e will be conventional slings. .

I The lifting devices for these loads do not fully meet the criteria of NUREG 0612 Section 5.1.6 in that they do not satisfy ANSI N14.6-1978 to use twice the normal design safety factors.

The applicant considers the reactor enclosure crane to have-sufficient design features to make the likelihood of a load drop extremely small. This is based on the crane being in compliance with NUREG 0612 Section 5.1.6.

The applicant has conducted load drop analyses of the reactor well shield plugs, dry well head, steam dryer shroud / head separator assembly into the reactor vessel.

i-Refueling floor. heavy load maximum height restrictions are established in order to prevent a sequential failure which could ultimately jeopardize major safety related equipment.

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k The analyses showed that no damage to the reactor fuel or leakage from the reactor vessel would occur.

_ . A load drop analysis of the service platform from its

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maximum height indicated no pressure vessel damage but some fuel damage would occur. However the drop conditions were based on very conservative assumptions and in light of these assumptions the probability of a service platform drop causing fuel damage was judged to be very unlikely. To further reduce the likelihood of a service platform drop causing fuel damage, administrative controls will be imposed to control the manner of installation of the service l platform.

The service platform slings and lift points have a safety factor of 5 and the service platform is lifted by a single failure proof crane.

B. EG&G Evaluation The entire overhead handling system should meet the single failure free guidelines of Section 5.1.6 of NUREG 0612.

Although the reactor enclosure crane can be considered to meet the guidelines, the lifting devices do not. The applicant has proposed as an alternate to compliance with single failure free proof status to show lifting device reliability by load testing the refueling shield, the fuel pool stop log, and the fuel pool jib crane and channel handling boom lifting devices to 150% of design rated load followed by a nondestructive test.

The applicant has also made load drop analyses of the following heavy loads which must be carried over the reactor vessel.

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a. Reactor well shield plugs

b. The dry well head

c. RPV head steam dryer shroud head / separator

_. d. Service platform

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e. Service pictform support

f. Refueling shield

g. Other loads carried over reactor with RPV head on.

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) These analyses were made with respect to Criteria I through III of NUREG 0612 and are discussed on pages 8 through 9, and Table 2 of Reference 12. The analyses indicated that no fuel damage or leakage from the reactor vessel would occur in all cases.

C. EG&G Conclusions and Recommendations The applicant has shown that the reactor crane meets the single failure proof guidelines of Section 5.1.6 of NUREG 0612. The associated lifting devices do not fully meet these guidclines. The applicant has demonstrated by load drop analyses that the Criteria I through.III of Section 5.1 are satisfied. The applicant has evaluated all cranes and hoists with respect to a dropped load affecting the ability to shut down the reactor and continues decay heat removal capability from the reactor or fuel pool. The evaluations indicated that in most cases this was no hazard. In some cases it was necessary to establish load carrying height restrictions or to establish other administrative controls to eliminate damage potential.

Therefore the Criteria IV of Section 5.1 was also satisfied.

EG&G therefore concludes that the intent of Guideline NUREG 0612, Article 5.1.5 ;;as been satisfied.

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2.3.2 Other Areas [NUREG-0612, Article 5.1.5]

(1) "If safe shutdown equipment are beneath or directly adjacent to a potential travel load path of overhead handling systems, (i.e., a path not restricted by limits of crane .

travel or by mechanical stops or electrical interlocks) one of the following should be satisfied in addition to satisfying the general guidelines of Section 5.1.1:

(a) The crane and associated lifting devices should conform to the single-failure proof guidelines of Section 5.1.6 of this report; F

S3 (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 be taken for intervening floors unless justified by analysis.)

S3 (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:

i failure of the safe shutdown equipment."

A. Summary of Applicant's Statements i

The applicant has prepared an index of overhead handling systems Unit I and common which indicates all plant cranes, their service, safety related items in the load _ path or on the next lower' level, and the exclusion criteria. (See 12

Table I--Reference 12.) A load / impact area matrix was also provided for each crane and hoist and the hazard elimination category for each load was indicated. (See Table 2 of

__ Reference 12.) Hazard evaluations were performed for each crane / hoist which could affect the capability to safely shut down the reactor or prevent a fuel pool drainage. These evaluations are indicated in Appendix B of Reference 12.

Major safety,related items located in the load path and/or on the next lower elevation are listed. A description of the effect of a load drop on systems required for safe shutdown or decay heat removals is provided. Safe shutdown capability was also determined to include the ability to establish or maintain a means of decay heat removal from the reactor vessel and the spent fuel pool. Apppendix A of Reference 12 presented a description of the two methods to achieve safe shutdown without offsite power.

Evaltation of Appendix B cranes and hoists were made using the Limerick Fire Protection Evaluation Report and the separation drawings to determine whether there was any safety related equipment in the load path or next lower level. Except for the refueling floor where there were lifts of_ very heavy loads, assumptions were made that sequential failures could not occur on more than one floor below the load path. These assumptions were based on the presence of large quantities of reinforcing bar in the floor preventing large concrete spalling fragments, which could damage equipment on the next lower floor. Detailed studies i

were made of the equipment separation drawings to determine that there was sufficient separation of safe shutdown equipment so that only one set of safe shutdown equipment was affected.

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4 In most cases it was shown that based on separation and redundancy of safety related systems or other plant considerations, no real hazard exists.

In some cases, it was found advisable to establish load carrying. height restrictions or other administrative controls to eliminate concern about potential damage to safety related systems. This was the case with loads carried by the reactor enclosure crane over the refueling floor. The refueling floor heavy load height restrictions are shown on Table 4 of Reference 12.

B. EG&G Evaluation

) The applicant appears to have conducted thorough analyses of potential load drops which could impair safe shutdown or decay heat removal capability. In most cases it was possible to show that based on redundancy and separation of safety related systems or other plant considerations no serious hazard exists.

A situation which was not addressed in the report was the case where movement of a load was determined to be so infrequent that no crane or hoist or other lifting device was provided. This situation will be addressed as it arises cnd treated with guidelines based on NUREG 0612 to reduce the probability of a load drop.

2.3.2.C. EG&G Conclusions and Recommendations Based on the information presented the applicant has made'an in-depth load drop analysis of all heavy loads carried over safe shutdown equipment and showed that no loads would penetrate the ceiling or that no spalling would occur of such magnitude to incapacitate both shutdown units of a

! redundant safe shutdown system.

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2.3.3 Single-Failure-Proof Handling Systems [NUREG-0612, Article 5.1.6]

(1) "Lif ting 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 i ANSI N14.6-1978. If only a single lifting device is  !

provided instead of dual devices, the special lifting )

device should have twice the design safety factor as required to satisfy the guidelines of Section 5.1.1(4).

However, loads that have been evaluated and shown to satisfy the evaluation criteria of Section 5.1 need not have lifting devices that also comply with Section 6 of '

ANSI N14.6.

(b) Lifting devices that are not specially dpsigned and l that are used for handling heavy loads in the area where the crane is to be upgraded should meet ANSI B30.9-1971, " Slings" as specified in Section 5.1.1(5) of this report, except that one of the following should also be satisfied unless the effects of a drop of the particular load have been analyzed and shown to satisfy the evaluation criteria of Section 5.1:

(1) Provide dual or redundant slings or lifting devices such that a single component failure or malfunction in the sling will not result in uncontrolled lowering of the load; QR (ii) In selecting the proper sling, the load used should be twice what is called for in meeting Section 5.1.1(5) of this report.

(2) "New cranes should be designed to meet NUREG-0554,

" Single-Failure-Proof Cranes for Nuclear Power' Plants." For operating plants or plants under construction, the crane should be upgraded in accordance with the implementation guidelines of Appendix C of this report.

(3) " Interfacing lift points such as lifting lugs or cask trun-ions should also meet one of the following for heavy loads handled in the area where the crane is to be upgraded unless 15

the effects of a drop of the particular load have been evaluated and shown to satisfy the evaluation criteria of Section 5.1:

(a) Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering ..

of the load; lift points should have a design safety factor with respect to ultimate strength of five (5) times the maximum combined concurrent static and dynamic load af ter taking the single lift point failure.

OB (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 has listed in Table 3 of Reference 7, the special lifting devices to be used at Limerick 1, the load weight of each, the design safety factor, and whether there was compliance with ANSI N14.6-1978.

It was indicated that none of the lifting devices meet the requirements of ANSI N14.6-1978 because they do not use twice the normal design safety factors.

The applicant states that the Lifting Devices (nct specially designed) " Heavy load slings" that will be used at Limerick will meet the requirements of ANSI B30.9-1971. In addition a 25% dynamic load factor will be applied to these slings.

By using this load factor which is conservative for the.

fastest hook speed to be encountered (27 ft/ min for the-reactor feed pump area Bridge Crane) it will _not be necessary to associate the slings with specific cranes.

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The applicant considers the lifts of the following loads to meet the criteria of NUREG 0612 Section 5.1.6. '

! _. . 1. Spent fuel cask

2. Refueling shield .

3. Fuel pool stop logs

4. Fuel pool gates

5. Fuel pool jib crane and channel handling boom.

Loads 1, 2, and 3 will be carried by special lifting devices.

Loads 4 and 5 will be carried by conventional slings.

The special lifting devices for loads 2 and 3 above do not  ;

meet the requirements of NUREG 0612 Section 5.1.6 in that they do not satisfy the requirements of ANSI N14.6-1978 regarding the use of twice the normal safety factors for critical loads.

The applicant states that "the Spent Fuel Cask--Limerick is under construction. Since licensing under 10 CFR 71 is not evidence that the spent fuel shipping cask lifting device and lift points meet the requirement of NUREG 0612, the '

NUREG 0612 requirements will be a hses for selection of shipping cask (s) for Limerick."

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There are four lifting points on the refueling shie1d. They provide a minimum static factor of safety of 4.8 with respect to the material ultimate strength. This does not satisfy the NUREG 0612 Section 5.1.6-safety factor requirement.

There are two lifting lugs on each fuel pool stop log. They provide a minimurfl factor of Rfety of 7.25 with respect to material ultimate strength plus a margin of 25% for dynamic loads. This does not satisfy Section 5.1.6 safety factor-requirements.

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Conventional slings will be used to carry the Fuel Pool Gates and the Fuel Pool Jib Crane and Handling Boom.

_. The slings will meet the requirements of NUREG 0612 Section 5.1.6 (1) (ANSI B30.9-1971). See Reference 6.

There are two lift points on each Fuel Pool Gate which I

provide a minimum static factor of safety of 9.3 with j respect to material ultimate strength. This does not satisfy NUREG 0612 Section 5.1.6 requirements.

There is one lift point on the Fuel Pool Jib Crane and the Channel Handling Boom. The lift points have a minimum  !

design safety factor of 5 with respect to material ultimate strength.

The applicant takes exception to the requirements that the l lifting devices and the interfacing lift points meet the safety factor requirements of NUREG 0612 Section 5.1.6 and does not believe that an increase in safety factor will l produce a proportionate improvement in lifting device reliability. As an alternative to fuel compliance with this requirement, the applicant proposes to perform load tests at 150% of rated capacity followed by nondistructive examination of the load bearing welds to demonstrate design adequacy.

2.3.3.B. EG&G Evaluation There are a number of heavy loads carried by the Reactor Enclosure Crane. The Reactor Enclosure Crane is considered by EG&G to meet single failure proof guidelines.

Load drop analyses tests were conducted on the following load handling systems using special lifting devices and lifted by the Reactor Enclosure Crane. .

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1. Dryer and separator slings

2. RPV head strongbacks

3. Service platform rack sling i

_. 4. Reactor well shield plugs and dryer separator canal plugs with strongback

5. Dryer / separator carial . plugs with strongback.

These analyses with other evaluations show that Criteria I through IV of Section 5.1 have been satisfied, and therefore the lifting devices which are single lifting devices do not have to be designed to have twice the design safety factors as required to satisfy the guide lines of Section 5.1.1(4).

Reactor enclosure Crane load handling system on which load drop analyses were not conducted are:

1. Spent fuel cask

2. Refueling shield 1-

3. Fuel pool stap logs

4. Fuel pool gates

5. Fuel pool jib crane channel handling boom.

These handling systems therefore are subject to the requirements of Section 6 of. ANSI 14.6-1978 or acceptable alternative to meet the requirements of Article 5.1.6.

The above handling systems do not fully meet the requirements of single failure proof status in that some of the lifting devices are not designed to meet Section 5.1.6(a) or 5.1.6(b).

The lifting devices for the Spent Fuel Cask and the cask lift points are believed to be designed to meet and' conform to NUREG 0612 Section 5.1.6 although the statement regarding.

this is somewhat vague. See page 5 Reference 12.

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The lifting devices for the Refueling Shield and the Fuel Pool Stop logs are special lifting devices and have been fabricated. They do not however meet the requirements of

_. NUREG 0612 to use twice the normal safety factors in that their safety factors are 3 (static load) and 4.5 (static 4

load + 25%) respectively, based on yield strength of material instead of 6 (static load + dynamic load).

Conventional slings will be used to carry the Fuel Pool Gates, the Fuel Pool Jib Crane and Handling Boom. These i

slings will be selected in accordance with the requirements of NUREG 0612 Section 5.1.6(1).

The interfacing lift points for the following loads which have been fabricated do not meet the requirements of NUREG 0612 Section 5.1.6.

Refueling Shield i

Fuel Pool Stop Logs Fuel Pool Gates l

Fuel Pool Jib Cranes Channel Handling Boom These loads have interfacing lift points safety factors which vary from 4.8 for the refueling shield to 9.3 for the Fuel Pool Gates based on static loads and ultimate strength of material. This guideline Article 5.1.6 requires safety factors of 10 for the lift points.

1 2.3.3.C. EG1G Conclusions and Recommendations The. applicant has met the intent for compliance for all lifting devices except for the refueling shield and the Fuel-Pool Stop logs. These devices have safety factors which do not meet NUREG 0612 Article 5.1.6. It is recommended that 20 l

the applicant perform a review of these lifting devices to ANSI N14.6-1978 with special attention to Sections 3, 4, and 5.

The applicant should also confirm the intended design of the Spent Fuel Cask lifting device with respect to ANSI N14.6-1978 and the cask interfacing lift points with respect to Article 5.1.6.

i i

21

3. CONCLUDING

SUMMARY

3.1 Guideline Recommendations The ap'plicant has not fully satisfied the intent of all three guidelines of this renort. The following are the conclusions regarding these guidelines Guideline Action Article 5.1.4 Intent of Guideline Satisfied Article 5.1.5 Intent of Guideline Satisfied Article 5.1.6 The applicant has not fully met the intent of this guideline. Additional information should be provided to

1. Confirm the design of the spent fuel cask lifting devices with respect to ANSI N14.C-1978 and the interfacing lift points to Article 5.1.6.

2. The applicant should perform a review of the refueling shield and the fuel pool stop log lifting devices to ANSI N14.6-1978 with special attention to Sections 3, 4, ar.d 5.

l

! 3.2 Additional Recommendations The Applicant should provide additional detail regarding proposed nondestructive examinations of load bearing welds on lifting devices and interfacing lift points to be examined.

22

t 3.3 Summary i

/

The applicant is consistent with the intent of Guidelines.

l

-2.3.1 Reactor Building (NUREG 0612, Article 5.1.4)

I 2.3.2 Other Areas (NUREG 0612, Article 5.1.5)

The applicant is not fully consistent with the intent of Guidelines:

I 2.3.3 Single Failure Handling Systems (NUREG 0612, Article 5.1.6)

Additional information should be provided regarding comparison of design of the refueling shield and fuel pool stop log lifting devices with respect to N14-6-1978.

Interfacing lift points of critical loads should be designed to meet the requirements of NUREG 0612 Article 5.1.6.

l The Limerick Generating Station Units 1 and 2 NUREG 0612 Objective Compliance Matrix is shown in Table 3.1.  ;

23

, t i s

ns wo

._ ol d

tt un he C C C C C C C C C C C C C Sm p

ei fu aq SE s I e s

o

,6 8 e

_. rl e

vy or A A A A A A A A A A A '.

Co C N N N N N N N N N N N N t

l n ec u%

Fn i

X I

R l T ey A ut M Fi l

E da A A A A A A A A A A A A C

N ec gi C N N N N N N N N N N N N A at I

mi L

P ar DC M

O

_ C S oe E is V da I

T ae Rl C e F eR A A A A A A A A A A A J t C N N N N N N N N N N N B e -

O iSv i 2 rt 1 rc 6 Oa

- 0 C

E -

. R e U rm N ue t t 2 is ay D rS C A A A A A A A A A A

_ N - N N N N N N A

A N N N N N N ef l o 1 go

. nr S iP T S I

N

. U

_ t t t N o s s n n O n s 4 i

s e I

T m

e da nr C p o io mep t s

t t g 2 m A t g r ac A m

u h h i l l s

i s o rr T s y

n o p V p p p u o o o de S ie te m Rd H mm q H h h yv G

S dn an u Hr n u u E ho N g l a ra ir er ps Ra m y o os p p m e - l c t it t r c a t n uc nc o I

e ds d ns rm ts y y nt o i v nr T

A i

b d gg ni oo op ai a a es f v o ee R

l r a n pm a l o r r mi t r m mn E

d n oe ph l e uh p p no a e e ni N

li ye yp ru ob c S S ih l s r ibs a th ed rr a p amt

_. E H cr sl as a tt io e e tr V V tos C ae ru po rWpS nf oi ct r r no D ei R .R ncl K

ev iu eo o o oo R S S oeo Ro Db Sh SE Cl Rm C C Cd C M -M Crh C

I R

E t M tn 1 1111 13 3 33 6 7 0 9 345 34502 7

. I ese1 0 0000 01 3 00 1 1 2 2 333 33333 3 L nl m. 2 5555 5S 1 22 2 2 2 2 222 22222 2 a o po H HHHH HH H HH H H H H HHH HHH1H 8 H

rhin

. 1

. C rqu( 0 0

ABCD 00 1III 00 a_ 0 0

A8 0 0 11 1 0

1 1

0 AAA 1

1I1 B8B00 0 I1I11 1 oE 1

E L m B e I 2 3 4 5 6 7 8 9 0 1 2 A t 1 1 1 T i

TABLE 3.1. (continued)

Cra ne or Holst Equipment Single-Failure- Orr-Site Radio- Damaged fuel ruel cover item (No.1 Handlina System Pruor System Sare Shutdown activo Release C r i t i ca l i t y Inventory lo s e, {3tli pment Lom 13 00-H126 Control Room HVAC NA NA NA NA ' 0 C Equipment hoist ' '

a = Iloist/ crane to be borrowed from other locations when needed.

C = Applicant action complies with Nt? REC-0612 Risk Reduction Objectives NC = Applicant action does not comply with NUREC-0612 Risk Reduction Objective.

NA = Ris4 Reduction Objective is not applicable to this handling system.

25

4 REFERENCES

1. EG&G-HS-6293, Control of Heavy Loads at Nuclear Power Plants, Limerick Generating Station, Units 1 and 2 (Final Phase I), Docket No. 50-352 and 50-353.

2. NURES-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 Philadelphia Electric Company.

Subject:

NRC Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 22 December 1980.

5. John S. Kemper, Philadelphia Electric Company. Letter to D. G.

Eisenhut (NRC) Response to staff position: Interim actions for Control of Heavy Loads dated June 18, 1981.

6. J. S. Kemper, Philadelphia Electric Company. Letter to D. G. Eisenhut (NRC),

Subject:

Forwards additional info re Heavy Loads Analysis in response to 830106 Telcon w/util, EG&G, and NRC dated 1/31/83.

7. ANSI B30.2-1976 " Overhead and Gantry Granes."

8. ANSI N14.6-1978 " Standard for Lif ting Devices for Shipping Containers Weighing 10,000 lbs (4500 kg) or more for Nuclear Materials.

9. ANSI B30.9-1971 " Slings."

10. CMAA-70 " Specifications for Electric Overhead Traveling Cranes.

11. J. S. Kemper, Philadelphia Electric Company. Letter to D. G. Eisenhut (NRC),

Subject:

Control of Heavy Loads at Nuclear Power plants" Final Report " Overhead Handling System Review Dated 4/2/82.

12. J. S. Kemper, Philadelphia Electric Company. Letter to D. G. Eisenhut (NRC),

Subject:

Limerick Generating Station Overhead Handling System Review Final Report dated 6/13/83.

13. Limerick Generating Station, Final Safety Analysis Report (FSAR)

Section 15.7.4 " Fuel Handling Accident."

26

. _ .