ML20213E536

Control of Heavy Loads at Nuclear Power Plants,Washington Nuclear Project 2 (Phase II-Interim), Technical Evaluation Rept

ML20213E536
Person / Time
Site:	Columbia
Issue date:	05/31/1983
From:	Dixon B EG&G, INC.
To:	NRC
Shared Package
ML17277A735	List: ML20213E536
References
CON-FIN-A-6457, CON-FIN-A-6547, CON-WNP-0632, CON-WNP-632, CON-WNP-63257, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8308150454
Download: ML20213E536 (20)

Text

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. e CONTROL OF HEAVY LCADS AT NUCLEAR FCWER PLANTS WASHINGTON NUCLEAR PROJECT NO. 2 (PHASE II--INTERIM)

Docket No. 50-397 Author B. W. Dixon Principal Technical Investigator T. H. Stickley

. Published May 1983 l%G Idaho, Inc. '

Idaho Falls, Idaho 83415 Drepared for the U.S. Nuclear Regulatory Commission Under COE Contract Nc. DE-AC07-76IC01570 FIN No. A6457 9308150454 030911 #2 pf-SP ADOCK 05000397 f W

', e s ASSTRACT The Nuclear Regulatory Commission (NRC) has recueste: nat all nuclear plants, either 0;erating er under constru:tien, submit a response of c mpliancy with NUREG-0612, " Control cf Heavy Loads at Nvelear Pcwer Plants." EG1G Idaho, Inc., has centractad with the NRC to evaluate the respenses of those plants presently unter construction. This report contains EG&G's evaluation and recemmendations for Washington Nuclear '

Project No. 2 for the retuirements of Sections 5.1.4, 5.1.5, anc 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

'4NF-2 c:es nc; : tally comply with the guicelines of NUREG-0612. In general, c:mplian:e is insuffiefen: in the follcwing areas:

o Insuffi:ient informatien has been provided for review in the areas of lifts over '*r:diated fuel and lifts by single-failure prcof handling systems.

o Lifts over safe shutdewn equipment have not been properly addressed.

The main repor centains ree:mmendati ns which will aid in bringing the above items int: ccmpliance with the a:propriate guidelines.

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6 4 CONTENTS AESTRACT ............................................................. 11 E X E C UT I V E S UMMA RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii .....

1. INTRODUCTICN .................................................... I 1.1 Purpose of Review ......................................... 1 1.2 Generic Eackground ........................................ 1 1.3 P l a n:-S c e ci fi c E ac kg ro und . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

2. EVALUATION AND RECOMMENDATIONS .................................. 4 2.1 Overview .................................................. 4 2.2 Heavy Lead Overhead Handling Systems ...................... 4 2.3 Guidelines ................................................ 4

3. CONCLUDING $UMMARY ...........,.................................. 13 3.1 Guideline Recommendations ................................. 13 3.2 Actitional Recommendations ................................ 15 3.3 S u mm a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4 REFERENCES ....................................................... 16 TAELES 2.1 Nenexempt Heavy Lead-Handling Systems ........................... 5

! 3.1 NUREG-0612 Cbjectives Como11ance Matrix ......................... 14 L

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CONTROL OF HEAVY LCADS AT NUCLEAR :CWER :LANT5 WASHINGTCN NUCLEAR PRCJECT NO. 2 .

(PHASE II) y

1. INTRCOUCTION 1.1 Pur:ose of Review

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

review of general lead-handling policy and procedures at Washington Nuclear Project No. 2 (WNP-2). This evaluation was performed with the objective of assessing conformance to the general load handling guidelines of NUREG-0612, " Control of Heavy Leads at Nuclear Power Plants" [2], Sections 5.1.4, 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

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1.2 Generic Backcround Generic Technical Activity Task A-36 was established by the U.S. '

Nuclear Regulatory Ccmmission (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

! cads 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, reque.st,ing information conherning -the control of heavy leads near spent fuel.

l The results cf Task A-36 were reported in NUREG-0612, " Control of n

Neavy_LoadsatNuclearPowerPlants." The staff's conclusion from this evaluation was that existing measu.es to control the handling of heavy loads at operating plants, although providing protecti.on from certain potential problent, do not adequately cover the major causes t

of Icac-hancling accidents and should be upgraded.

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e s In order to upgrade measures for the control of heavy loads, the staff cevelopec a series of guidelines designed to acnieve a two phase .

objective using an accepted aaproacn or protection philosophy. The first phase of the objective, achieved through a set of general guidelines icentified in NUREG-0612, Article 5.1.1, is to ensure snat all load-handling systems at nuclear power plants are designed and operated such that their probability of failure is uniformly small anc appropriate for the critical tasks in which they are employed. The second phase of the staff's objective, achieved through guidelines identified in NUREG-0612, Articles 5.1.2 through 5.1.5, is to ens'ure that, for load-har.dling 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 consequences is quantified in NUREG-0612 into four accident analysis evaluation criteria as follows:

o " Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of postulated heavy load produce doses *

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that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that coses

, 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 postulatec heavy load coes not result in a configuration of the fuel such that k,ff is larger than 0.95; o " Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accicental dropoing of postulated heavy load is limitec so as not to result in 2

water leakage tnat could uncover tne fule, (makeup water provided to overcome leakage shcuid be fr:m a berated scurce .

of adequate concentration if the water being lost is borated); and o "Camage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a pcstulated 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 'or minimi:ing the pctential 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 mcvement over critical areas.

Staff guidelines resulting from the foregoing are tabulated in Section 5 of NUREG-0612.

1.3 Plant-Soecific Backereund On December 22, 1980, the NRC issued a letter [4] to Washington Public Power Supply System (WPPSS), the applicant for WNP-2 requesting that

• the applicant review provisions for handling and :entrol of heavy

• icads at WNP-2, evaluate these provisions with res:ect to the guidelines of NUREG-0612, and provide certain additional information to be used for an independent determination of confermance to these l guidelines. WPPSS provided responses to this request pertinent to Phase II on January 13, Feoruary 12, anc Cctcber 4,1982 and February 23, 1983 [5,6,7,5).

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2. EVALUATICN AND RECOMMENDATIONS 2.1 Overview The following sec-ions summarize WPPSS's review of heavy loac handling at WNP-2 accompantec by EG&G's evaluation, conclusions, and recommendations to the applicant for bringing the facilities more completely into compliance with the intent of NUREG-0612.

2.2 Heavy Lead Overhead Handlino 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 1,200 lbs. per the NUREG-0612 definition.

2.3 Guidelines 2.3.1 Reactor Buildino [NUREG-0612, Article 5.1.41 (1) "The reactor building crane, and associated lifting cevices used for handling the above heavy loads, should satisfy the single-failure proof guidelines of Section 5.1.6 of this report. '

93 (2) "The effects of heavy load drops in the reactor building should be analy:ed 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 platform; and any other heavy loads that may be brought over or near safe shutcown 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 coeratien during periods when the load being analyzed would be handled. The analysis should also conform to the guidelines of Apcendix A."

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O I Allt (. 2. I NOHI XiHPI Ill AVY l OAD-IIAllDI IflG SYSil_HS lini?.

Cf08H8 .liI Humiso r,_ t oca t iote Type Sotvico [IlltA _C pi s g tjapas:l ly .

I Hi-Ilot-6 lteactor bisliding Trolloy holst Hilft pumps ( A&ll) A-l 6 8 t19. 2 f t 1 electric 2 HI-i!01 -7 henctor hullding Trolley hoist itCIC pump and A-1 ',

8 92.2 fl.

8 Oloctric turbino 3 Hi-Il01-fl Itcactor building Trolloy holst tillR pump C A-1 6 8 9fs. 3 f t 4 oluctric Is Hi-Ilot-9 Reactor linilding Trolley holst IlCS pump A-l 1 is9 3. 2 ft electric

's Hi-1101- 10 Itcactor billiding i ro l ley lioi s t IIPCS pump A-1 PH 8:92.84 ft electric 6 HI-CitA-6A,60 Standlay servico water Ove r head .t rave l l i ng Stassilljy servico A-l 85 pump hotaso c ra tio (under hissig ) watnr pumps 7 HT-CRA-2 ficactor buildirig Iravoillrig brlilge Itenctor roruoling A-1 I P's 606 ft c rano floor and vossol 8 Hi-Cit A- 1 lurbino inallilitig T ravolling bridge Halta turbine and Pun crano gono ra to r 9 Hi-ilot-in Reactor building Trolloy hoist Outlanard main steam A-1 a it.ola t f ori va lvo

. Work ared pipo tunnol fiatch removal 9

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o A. Summary of A:clicant's Statements 1

l The a;piicant indicated that the Reacter Builcing Crane is 1 the only crane physically capable Of carrying neavy leads Over spent fuel in the s crage poci er reactor vessel.

"The Reactor Building Crane (MT-CRA-2) main hoist meets the requirements for a ' single failure proof crane' as per NUREG-0612, Appendix C.

"The auxiliary hoist will be derated to 7 1/2 tons maximum versus 15 tons design rating for handling heavy loads ever the spent fuel pool or open vessel cavity thus doubling the design safety factor. In addition, travel of the Reactor Building Crane is limited for the main and auxiliary hooks in the area over the spent fuel pool."

B. EG&G Evaluation The single-failure proof status of the Reacter Building Crane (MT-CRA-2) is examined in Section 2.3.3 of this

( report. The entire handling system must be .

single-failure proof, includ,ing slings and lifting points ,

! for this status to be validated.

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The applicant incicated on safe lead path drawing notes that lifts of the shield plugs would be handled by a non-single-failure proof sling system. Therefore, these

! leads fall under the criteria of NUREG-0612 l

Section 5.1.4(2) and should be so addressed.

Currently the applicant has not indicated comoliance to either of NUREG-0612 Sections 5.1.4(1) or 5.1.4(2) for the MT-CRA-2 Auxiliary Hoist. While the increased safety factor f:r this hoist cces provide acditional assurances 6

against a lead dr:p it does not provide single-failure-proof status per NUREG-C612 Ap;endix C nor -

d es it necessarily meet the load crep precability allowable values cutlined in NUREG-C612 Section 5.2.

The applicant should provice more information on the method of travel limitation for the MT-CRA-2 hoists over the Fuel Storage Pool.

C. EG&G Conclusions and Reccamendations WNP-2 is in partial compliance with the requirements of this guideline. The applicant should take the following actions:

(1) Provide an analysis of shield plug lifts per Section (2) of the criteria.

(2) Apply either Section (1) or (2) of the criteria to the Reactor Building Crane Auxiliary Hoist.

(3) Provide information on the limiting method used for the Reactor Building Crane over the Fuel Storage Pool.,

2.3.2 C:her Areas [NUREG-0612, Article 5.1.5]

(1) "If safe shutdown ecuipment 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 steps 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 cenform to ne single-failure proof guidelines cf Section 5.1.6 of this report; 93 7

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(b) If the lead drop could impair the operation of ecuipment or cabling as~sociated with reduncant or dual safe shu:dewn paths, mechanical steps or electrical interlocks shculd be provided to prevent movement of loads in proximity to these redundant or dual safe shu:dewn equipment. (In this case, crecit should not be taken for intervening floors unless justified by analysi s. )

OR (c) The effects of icad drops have been analy:ed and the results indicate that damage to safe shutdown equipment would not preclude operation of suffici'ent 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. Sgmmary of Acolicant's Statements "The following list of cranes and hoists were installed to permit maintenance of a specific piece of equipment. These lifting devices do not meet the requirements of NUREG-0612 and it is not considered economically practical to modify ,

them to meet these requirements. They will be locked out ,

in a safe position and not piaced in use until the equipment they service has been declared inoperable per the Plant Technical Specifications:

MT-HOI-6 Services RHR Pumps A and B MT-HOI-7 Services RCIC Pump and Turbine MT-HOI-8 Services RHR Pump C MT-HOI-9 Services LPCS Pumps MI-HOI-10 Services HPCS Pumps i MT-CRA-6A and 6B Services Standby Service Water Pumps, IA and 18 MT-HCI-la Services Cutboard Main Steam Isolation Valves" i

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3. EG1G Evaluation The applicant should exa-ire tne cranes listed in Sec-ton A above per the criteria of hUREG-0612 Section 5.1.5(1)(c).

A number of these cranes probacly meet these criteria without further modification, although an insufficient amount of information has been provided for EG&G to verify this position. Some cranes may require additional analysis or load handling restrictions due to transport of loads from one train over components in the reduncant train.

The applicant has not addressed the Turbine Building Traveling Bridge Crane MT-CRA-1.

C. EG&G Conclusions and Recommendations WNP-2 is not in compliance with the requirements of this guideline. The applicant should take the following actions:

(1) Address the Turbine Building Bridge Crane MT-CRA-1 per ,

the criteria.

(2) Examine the cranes list.ed in Section A above per ,

Section (1)(c) of the criteria.

2.3.3 Sincle-Failure-Proe'f Handline Systems [NUREG-0612. Article 5.1.6]

, (1) " Lifting Devices:

l (a) Soecial liftino devices that are usec for heavy loacs in the area wnere tne crane is to be upgraded should meet ANSI N14.6-1978, " Standard .:or Special Lifting Devices for Shipping Containers Weighing 10,000 Pouncs (4500 kg) or More For Nuclear Materials," as specified in Secticn 5.1.1(4) of this report except that the l

handling device shoulc also comply with Sect' ion 6 of ANSI N14.6-1978. If only a single lifting device is provided instead of dual devices, the s:ecial lifting l device shoulc have twi:e the cesign safety factor as required to satisfy the guidelines of 1

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c, o Section.5.1.1(c). However, loads that have been evaluated and shown to satisfy the evaluation criteria of Section 5.1 need net-have lifting devices that also ' ,

comply with Section 6 of ANSI N14.c.

(c) Liftine devices that are not scecially desicned and tnat are usec for nanc!ing neavy loacs in :ne area where the crane is to be upgraded should meet ANSI S30.9 - 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; 03 (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 guicelines of Appendix C of :nis recort.

(3) "Interfacino lift ooints such as lifting lugs or cask .

trunions should also meet one of the following for heavy loads handled in the area wh'ere the . crane is to be upgraded' unless the effects of a drop of the particular load have l

been evaluated and shown to satisfy the evaluation criteria of Section 5.1:

(a) Provide recundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the lead; lift points should have a cesign safety factor with respect to ultimate strength of five (5) times the maximum comoined concurrent static and l

dynamic load after taking the single lift point l

I failure.

OR (b) A non-redundant or non-dual lift point system should i have a design safety factor of ten (10) times the maximum comoinec concurrent static and cynamic load."

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A. Su-.ary of Acclicant's Statemer. s -

The applicant indicated that the Reactor Building Crane is a single-failure prcof crane (ree Section 2.5.1A).

Safe load path drawings supplied by the applicant containec the following notes for lifts using the Reactor Building Crane:

"All loads other than shield plugs, lifted with conventional lifting apparatus shall utilize redundant rigging or maintain a safety factor of ten (10). Shield plugs will only be moved when reactor head, RPV space frame and drywell head are in place over the reactor with a lifting apparatus factor of safety of 5 maintained.

" Loads shall be maintained as close to the floor as practical.

The head stcong back and stud tensioner and spreader may be moved as necessary, movement shall be governed by '

appropriate detailed procedure for performance of specific functions."

• B. EGaG Evaluation The applicant has not indicated whether special lifting devices used in conjunction with the Reactor Building Crane meet the requirements of ANSI N14.6 Section 6 as requitec in NUREG-0612 Section 5.1.6 (1)(a).

The applicant also has not indicatec comoliance with Section 5.1.6 (3) of NUREG-0612. ,

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(See Section 2.3.1B for discussion on shield plug lif ts.)

. C. EG&G Conclusions and Recommendatiens

%NP-2 is not in complete compliance with the requirements for single-failure proof handling systems. The applicant should take the following actions:

(1) Provice information pertaining to compliance with ANSI N14.6-1978 Section 6 for all special lifting cevices used in conjunction with the Reactor Building Crane.

(2) Provide information on interfacing lift points for' items lifted by the Reactor Building Crane.

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3. CCNCLUDING SU.vn RY 3.1 Guideline Recommendations WNP-2 is presen-ly not in complete compliance with the recuirements cf NUREG-0612 Section 5.1. Inis conclusion is represented in tabular form as Table 3.1. The following actions should be taken by the applicant: 1 Guidelines Action Section 5.1.4 (a) Provide for review an analysis of shield clug lifts.

(b) Examine the Reactor Building Crane Auxiliary Holst per the criteria of this section anc provide pertinent material for review.

(c) Provide information on limiting devices used with the Reactor Building Crane.

Section 5.1.5 (a) Examine the Turbine Building Bridge Crane per the criteria of this section and provice ~

pertinent mate' rial for review. ~

(b) Analyze the effects of lead drops frem cranes listed in Section 2.3.2A of this report per the criteria of this section and provide pertinent information for review.

Section 5.1.6 (a) Indicate whether all special lifting devi:es used in conjunction with the Reactor Builcing Crane meet the criteria cf ANSI N14.6-1g73 Section 6.

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F b I Alsi l~ 3.1 l Plant l--NtHil G-06l2 OllJI CT 4VI.S COH1't.l ANCE MAIRIX

• S i seg l o- Fa i lle ro- Offsite Radio- Ilamaged Issol

.llaft!!! ! sial. Sys ts;m. , fro!r_Sygten ac t lyo._Ito [ea sg ruol Cover W.itor Saru Sleistdown Criticallty_ j pyijntgry 1,oss, l emlpment I nss -

1. 1 0 :14 pumps Aall ho i s t -- -- -- --

NC

, P. ItCIC pesop lusist -- -- -- --

NC

3. lulli ptemp hoist -- -- -- --

NG is , i PCS permp hoi s t -- -- -- --

NC

• p . lit *CS pump hoi st -- -- -- --

HC

6. Pump house overhead crasio -- -- -- --

HC

1. etcactor hullding bridge crano i I I I --

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8. turhino building bridge crano I -- -- --

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9. Itsly holdt -- -- -- --

NC C = Applicant action cosipils witte NURIG-0612 Risk Reduction Objectivo.

NC = Applicant act iore stocs not comply wi tte NUlttG-0612 itisk ItedescL lon ohjoct ivo.

-- = ltisk Roduction Objective is not. appilcable to this handling system.

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(b) Analy:e all interfacing lift points en 1 ems lifted by tne Reactor Building Crane per tne .

criteria of -his section and provice per-inent information for review.

3.2 Additional Recommendatiens This is an interim report. As WNP-2 is a near term operating license plant the applicant is encouraged to provide information on exoectad response dates for the items listed in Section 3.1 so as to expedite the issuance of the final report. The applicant should arrange for a telephone conference between the applicant, EG&G Idaho, and the NRC within 6 weeks of receiyal of this report.

3.3 Summary The applicant is currently considered to be in partial compliance with each of the guidelines covered in this report.

More information is required to complete the review of compliance with criteria pertaining to lifts over irradiated fuel and single-failure proof handling systems.

The applicant indicated that for economic reasons the guideline pertaining to lifts over safe shutdown eouipment will not be met.

However, EG&G feels that full compliance can be achieved for many of these cranes through the use of proper procedures with minimal economic impact. The applicant has been requested to reexamine these cranes.

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4. REFERENCES

1. [ Phase I Final Report) -

2. NUREG-0612, Control of Heavy Loacs a- Nucisar Power Plants, NRC.

3. V. Stello, Jr. (NRC), Letter to all a:clicants.

Subject:

Request for Additional Information on Control of Heavy Loads Near Spent Fuel, NRC, 17 May 1978.

4. USNRC, Letter to WPPSS.

Subject:

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

5. G. D. Bouchey (WPPSS), Letter to NRC.

Subject:

Nuclear Project No. 2 Response to NUREG-0612 Control of Heavy Loads, WPPSS, 3 January 1982

6. G. D. Bouchey (WPPSS), Letter to NRC.

Subject:

Nuclear Project No. 2 WNP-2 Response to NUREG-0612, Control of Heavy Loads, WPPSS, 12 February 1982

7. G. D. Bouchey (WPPSS), Letter to NRC.

Subject:

Nuclear Project No. 2 Response to NUREG-0612, Control of Heavy Loacs, Revision 1; Submittal of, WPPSS, 4 October 1982

8. G. D. Bouchey (WPPSS), Letter to NRC.

Subject:

Nuclear Project No. 2 Control of Heavy Loads, Revision 2, 23 February 1983

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