ML17277A596

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

ML17277A596
Person / Time
Site:	Columbia
Issue date:	05/24/1983
From:	Dixon B, Stickley T EG&G, INC.
To:	NRC
Shared Package
ML17277A595	List: ML17277A596
References
CON-FIN-A-6457, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8306030088
Download: ML17277A596 (21)

Text

CONTROL OF HEAVY LOADS AT NUCLEAR POlIER PLANTS WASHINGTON NUCLEAR PROJECT NO. 2 (PHASE IIINTERIM)

Docket No., 50-397 Author B. M. Dixon Principal Technical Investigator T. H. Stickley Published May 1983 EGM Idaho, Inc.

Idaho Falls, Idaho 834)5 Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract Nc. DE-AC07-76IDO)570 FIN No. A6457 8306030088 830524

"...A...- <'. PBR.;,

PDR ADQCK 05000397

ABSTRACT The Nuclear Regula.ory Comm'.ssion (NRC) has reouested tha. all nuclear plan.s, either operating or under construction, submit a response of compliancy 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 plan s presently under construction. This report con ains EG&G' evalua ion and recommendations for Washington Nuclear Project No. 2 for the requirements of Sec.ions 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 repor. [1].

EXECUTIVE

SUMMARY

MNP-2 does not totally comply with the guidelimes of NUREG-0612. In oeneral, compliance is insufficient in the following areas:

o Insufficient information has'een provided for review in the areas of lifts over irradiated fuel and lifts by single-failure-proof handling sys.ems.

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

The main report contains recommendations which will aid in bringing the above i .ems into compliance with the appropria.e guidelines.

CONTENTS ArSc I DAB RAG I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

="XECUTIVE

SUMMARY

..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ill 1 ~ INTRODUCTION o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o"'

1.1 Purpose of Review.................................. 1 1.2 Generic Background ......

1. 3 Pl ant-Speci fic Background .................................. 3

2. EVALUATION AND RECOMMENDATIONS 2 .1 Overview . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \ ~ ~ 4 2.2 Heavy Load Overhead Handling Systems .. ~....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 2~3 Guidelines ....................... ~ ~ ~

.3. CONCLUDING

SUMMARY

........................... 13 3.1 Guideline Recommendations ................................. . 13 3.2 Additional Recommendations ................... . . . ... 15 3~3 Summary ................................................... 15 4 bL RcrcRENCES 't e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 16 ~

TABLES 2.1 Nonexempt Heavy Load-Handling Systems ........................... 5

3. 1 NUREG-0612 Objectives Compliance Matrix ......................... 14

CONTROL OF Hc4,VY LOAOS AT NUCLEAR POWER PLANTS WASHINGTON NUCLEAR PROJECT NO. 2

~PLI~CK T

1. INTROOUCTION

1. 1 Purpose of Review This technical evalua ion report documents the EG8G Idaho, Inc.,

review of general load-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 Loads at Nuclear Power Plants" [2], Sections 5.1.4, 5.1.5, and 5. 1.6. This cons:itutes Phase II of a two-phase evaluation. Phase I assesses conformance to Sec.ion 5. 1. 1 of NUREG"0612 and was documented in a separate report

[I].

1.2 Generic Back round Generic Technical Activity Task A-36 was established by the U.S.

Nuclear Regulatory Commission (NRC) staff to systematically examine s.aff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to assure the safe handling of heavy

!oads and to recommend necessary changes to these measures. This ac.ivity was initiated by a le ter issued by the NRC staff on Hay 17, 1978 [3], to all power reactor applicants, requesting information concerning the control of heavy loads near spent fuel.

The results oi Task A"36 were peported in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." The sta;f's conclusion from this evaluation was .hat existing measures to control the handling of heavy loads at operating plants, al hough providing protection from certain potential problems, do not adequately cover the ma'or causes of load-handling accidents and should be upgraded.

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In order to upgrade measures for the con:rol of heavy loads, .ne s.aff developed a series of guidelines designed to achieve a two-phase objec ive using an accepted approach or protec.ion philosophy. The first phase of the objective, achieved through a set of oeneral guidelines identified in NUREG-0612, Article 5.1.1, is to ensure that all load-handling systems at nuclear power plants are desioned and opera ed such that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. The second phase of the staff's objective, achieved hrough 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 additior, to those required for all load-handling sys ems, to ensure that the potential for a load drop is ex remely 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 criter ia 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 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 postulated heavy load does

not result in a configuration of the fuel such that k eff is larger than 0.95; o "Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accidental dropping of postulated heavy load is limited so as not to result in

0 water leakage that could uncover the fule, (makeup wa:er provided to overcome leakage should be ,rom a bora.ed source of adequate concentration if the water be.ng lost is borated); and o "Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a pos.ulated heavy load, will be limited so as not to result in loss of required sa e shutdown func.ions."

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 trainirg, equipment design, and maintenance, coupled with safe load paths and crane interlock devices restric.ing movement over critical areas.

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

1.3 Plant-S ecific Background On December 22, 1980, the NRC issued a letter t'4] to Vashington Public Power Supply Sys em (MPPSS), the applicant for MNP-2 requesting that the applicant review provisions for handling and con.rol of heavy loads at MNP-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. MPPSS provided responses to this request pertinent to Phase II on January 13, February 12, and October 4, 1982 and February 23, 1983 [5,6,7,8].

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2. EYALUATION AND RECOMMENDATiONS

2. 1 Overview The following sections summarize MPPSS's review of heavy load handling at WNP-2 accompanied by EHG's evaluation, conclusions, and recommendations to the applicant for bringing the facilities more completely into compliance with the intent of NUREG-0612.

2.2 Heav Load Overhead Handlin S stems 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 Buiidino 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. t OR (2) "The ef ects 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; s.earn dryers and separators; refueling canal pluas and oates; shielded spent-fuel shipping casks; vessel inspec.ion platform; and any other heavy loads that may be

'rought over or near safe shutdown equipment as well as fuel in the reactor vessel or the spent-fuel pool. Credit may be .aken in this analysis for operation of he 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."

TAOLf. 2.1 NONCXL'HPr IIEAVY I.OAD-IIANOLINO SYSII:HS gfnIIo /~a< /II>mbo r HT-II01-6 I I I:>J, [I))>

Ronclur bui Idlng

~>B Trolloy hoist Sorvlco ~OJIBWA (:f9 I's A- I I ll> 1'i Ilg>NRN I ly RIIR pumps tA!>0)

II09.2 I' electric HT-II01-7 Nonclor bul lding Tro I loy hoist. RCIC pump and A- I Il92. 2 I', o I oclr ic lurb lno HT-li01-0 Roactor building Trolley hoist BIIR pump C A- I II9ll.3 I' ulnctrla H f-IIO1-9 Ronctor building Trolley hoist LPCS pump A-1 ll93 ~ 2 I' electric Hr-N01-10 Reactor building Trol loy hoist IIPCS pump A- I l192.il rt o lnctr lo HT-CRA-6A,(II Stnndby sorvico vator Ovorhond .trnvolllng Stnndby sorviro A-1 pump house crnnu {under hung) va lo r pumps HT-CRA-2 Ronctor building T ra vo I I ing b r I dge Renctor rot'uel lng A-1 125 606 rt crano floor and vossol HT-CRA-1 Turblno building Trnvol ling brldgo Hain turbine and ?00 crane gunora tor HT-II01-18 Roaclor bul ldlng Trol loy hoist Olllhoard main steam A-1 Isolation valve vora'nd plpo tunnel hatch removal

A. Summar of Applicant's Statemen.s The applicant indicated that the Reactor Building Crane is the only crane physically capable of carrying heavy loads over spent fuel in the storage pool or reactor vessel.

"The Reactor Building Crane (NT-CRA-2) main hoist .meets the requirements for a 'single failure proof crane's 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 over the spent fuel pool or open vessel cavity thus doubling the design safety factor. In addition, .ravel of the Reactor Building Crane is limited for the main and auxiliary hooks in the area over the spent fuel pool."

B. EGKG Evaluation The single-failure"proof status of the Reactor Building Crane (MT-CRA-2) is examined in Section 2.3.3 of this report. The entire handling system must be single-failure-proof, including slings and lifting points for this status to be validated.

The applicant indicated on safe load path drawing notes that lifts of the shield plugs would be handled by a non-single-failure-proof sling sys.em. Therefore, these loads fall under the criteria of NUREG-0612 Section 5. 1.4(2) and should be so addressed.

Currently the applicant has not indicated compliance 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 for this hoist does provide addi ional assurances

against a load drop it does not provide single-failure-proof status per NUREG-0612 Appendix C nor does i. necessarily meet the load drop probability allowable values outlined in NUREG-0612 Sec.ion 5.2.

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

C. EG8G Conclusions and Recommendations MNP-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 Reac.or 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 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 s ops or electrical in.erlocks) 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 cf Section 5.1.6 of this report;

(b) If the load drop could impair the operation of equipment or cabling associated with redundant or dual safe shvtdown paths, mechanical stops or electrical interlocks should be provided to prevent movement of loads in proximity to .hese redundant or dual safe shv.down equipment. (In this case, credit shovld not be .aken for intervening floors unless justified by analysis.)

OR (c) The effects of load drops have been analyzed and the results indicate that damace to safe shutdown t would not preclude operation of sufficient equipment to achieve safe shutdown. Analyses s ou conform to the guidelines of Appendix A, as applicable.

(2) "Wh ere thee safe shutdown equipment has a ceiling separating it from an overhead handling system, an alternat ive 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 shu.down equipment."

A. Svmmar of Applicant'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 NUREQ-0612 and it is not considered economically practical to modify them to meet these requirements. They will be locked ovt in a sa e position and not placed in use until the eouipment they service has been declared inoperable per the Plant Technical Specifications:

NT"HOI-6 Services RHR Pumps A and B NT-HOI-7 Services RCIC Pump and Turbine NT-HOI"8 Services RHR Pump C NT-HOI-9 Services LPCS Pumps NI-HOI-10 Services HPCS Pumps NT-CRA-6A and 6B Services Standby Service Water Pumps, 1A and 1B NT"HOI-18 Services Outboard Main Steam Isolation Valves"

B. EG&G Evaluation The applicant should examine the cranes lis ed in Sec.ion A above per the cri.eria of iiUREG-0612 Sec.ion 5.1.5(1)(c).

A number of these cranes probably 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 restric.ions due to transport of loads from one:train over components in the redundant train.

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

C. EG&G Conclusions and Recommenda ions 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 lis.ed in Section A above per Section (1)(c) of the criteria.

2.3.3 Sin le-Failure-Proof Handlina S stems NUREG-0612 Article 5.1.6 (1) "Lifting Devices:

(a) Special lif ina 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 Weiohing 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 N14.6"1978. If only a single lifting device is provided instead of dual devices, the special lifting device should have .wice 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 lif.ing devices that also comply with Section 6 of ANSI N24.6.

(b) Lif.ina devices .hat are not soeciall designed and that are used for hanoling heavy loads in the area where the crane is to be upgraded should meet ANSI B30.9 - 1971, "Slings" as specified in Section 5. l. 1(5) of thi s report, except that .one of the following should also be satisfied unless the effects of a drop of the particular load have been analyzed and shown to satisfy the evaluation criteria of Section 5.1:

(i) Provide 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; OR (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) "Interfacino lift points 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 par icular load have been evaluated and shown to satisfy the evaluation cri:eria 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 fac.or with respect to ultimate strength of five (5) times the maximum combined concurrent s.atic and dynamic load after taking the single lift point failure.

OR (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."

10

A. Summa. of Aool'icant's Statements The applicant indi ca ed .hat the Reactor Bui 1 ding Crane is a single-failure-proof crane (see Sec:ion 2.3.1A).

Safe load path drawings supplied by the. applicant contained tie 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 li ting apparatus factor of safety of 5 maintained.

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

The head s rong 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. ECEG Evalua.ion The applicant has not indicated whether special liftino devices used in conjunction with the Reactor Building Crane meet he requirements of ANSI N14.6 Section 6 as required i n NUREG-0612 Sec.i on 5. 1. 6 (1) (a) .

The applicant also has not indicated compliance with Section 5.1.6 (3) of NUREG-0612.

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

C. EGKG Conclusions and Recommendations WNP-2 is not in complete compliance with the requiremen s for single-failure-proof handling systems. The applicant .

should take the following actions:

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

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

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3. CDNCLUDING SUi4L4MY

3. 1 Guideline Recommendations WNP-2 is presen ly not in complete compliance with the requirements cf NUREG-0612 Section 5.1. This conclusion is represented in tabular form as Table 3.1. The following ac iions should be taken by the applicant:

Guide 1 ines Action Section 5.1.4 (a) Provide for review an analysis of shield ",lug lifts.

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

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

Sec.ion 5.1.5 (a) Examine the Turbine Building Bridge Crane per the criteria of this section and provide pertinent material for review.

(b) Analyze the effects of load drops from 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 Bui lc'.ng Crane. meet the criteria of ANSI N14.6-1976 Section 6.

13

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TAIII.I: 3.1. ) Plant) NUACG-0612 OIIJLCTIVLS COMPLIANCE HATAIX S ing lu-Fa I luro- Offslto Radio- Damaged Fool Fuel Cover Mater Sal'o Shutdown Pfoof System activo Aoloaso ~CI ): lr<a ) I l.~ Inyoii).os I,ops I<I<<lpm<:ng I <iss

1. ANA pumps ALO hol sL NG

2. ACIC pump hoist

3. AIUl pump hoist II. I.PGS pi<mp hoist NC

5. III'CS pump hoist NC

6. P<imp house ovorhoad crano NG

7. Aoactor bul Idlng brldgo crano

8. Turhlno h<<l ldlng bridge crano

9. MSIV hoist C = Applicant action compl ls vlth NUAFG-0612 Alsk Roductlon ObJoctlvo.

NC = hppl leant acLlon docs noL comply ulLh NUAEG-0612 Risk Reduction Olijoctlvo.

-- = Risk Aoduction ObJocLIvo Is not appl lcablo to this handling systom.

111

(b) Analyze all in erfacing lift points on i.ems lifted by .he Reactor Building Crane per the criteria of .his sec ion and provide per=incr.:

informa ion for review.

3.2 Additional Recommendations This is an interim report. As WNP-2 is a near term operating license plant the applicant is encouraged to provide information on expects" response dates for the items lis.ed in Section 3.1 so as to expedite

.he issuance of the final report. The applicant should arrange for a telephone conference between he applicant, EGM Idaho, and the NRC within 6 weeks of receival of this report.

3. 3 ~Svmmar The applican't 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 per aining .o lifts over safe shutdown equipment 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..

.4. REF=RENCES

1. [Phase I Final Report]

2. NUREG-0612, Control of Heavy Loads a- Nuclear Power Plants, NRC.

3. V. S.ello, Jr. (NRC), Let er to all applican.s.

Subject:

Request for Addi ional Xnformation on Control of Heavy Loads Near Spent Fuel, NRC, 17 May 1978.

4. USNRC, Letter to WPPSS. Subjec-: NRC Request for Control of Heavy Loacs Near Spent Fuel, Additional'nformation on NRC, 22 December 1980.

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

Subject:

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

6. G. D. Bouchey (WPPSS), Let.er 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 Loads, Revision 1; Submittal of, WPPSS, 4 Oc aber 1982

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

Subject:

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