ML17276B027

Forwards Response to NUREG-0612, Control of Heavy Loads.

ML17276B027
Person / Time
Site:	Columbia
Issue date:	01/13/1982
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
Shared Package
ML17276B029	List: ML17276B027 ML17276B032
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR GO2-82-32, NUDOCS 8202020268
Download: ML17276B027 (29)

Text

REGULATORYjFORMATION DISTRIBUTION SY'M (RIDS)

AO ESSION NBR:8202020268 DOC ~ DATE: 82/0$ /13. NOTARIZED: NO DOCKET CIL:50 397 l<PPSS Nuclear DYNAM)E ProJectF Unit  ? F Washington >Publ ic Powe 05000397 AUTH AUTHOR AFFILIATION BOUCHEYFG AD, l'(ashington .Public Power "Supply System R EC IP ~ NAME REC IP IENT AFFILIATION SCHl)'ENCER P A, L-icensing Branch 2

SUBJECT:

Forvards response to NUREG 0612,,"Contra) of Heavy Loads ~ "

DISTRIBUTION CODE: BOSOS COPIES RECEIVED:LTR . ENCL f SIZE:, J5 TITLE:'ontrol of Heavy l.oads Near Spent Fool (USI,A 26) PRE DL NOTES:2 copies all matl:PP. .05000397 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAPE LTTR ENCL ID CODE/NAPE LTTR ENCL ACTION: A/D LICE NSNG 1 0 LIC BR 02 BC 1 0 LIC BR P2 LA 1 0 NRR CLEMENSON01 AULUCKF'RE 1 1 INTERNAL: ELD '-0 IE 06 3 3 IE/DEP/EPDB 35 1 IE/DEP/EPLB 36 3 3 MPA 1 "0 NRR REQUAFG 1 NRR/DE/CEB 11 1 1 NRR/DE/EQB 13 13 3 NRR/DE/GB 28 1 1 NRR/DE/HGEB 30 2 2 NRR/DE/MEB 18 1 1 NRR/DE/M(TEB 17 1 1 NRR/DE/QAB 21 1 1 NRR/DE/SAB 24 1 NRR/DE/SEB ?5 1 1 NRR/DHFS/HFEBOO 1 1 NRR/DHFS/LQB 32 1 1 NRR/DHFS/OLB 34 1 1 NRR/DHFS/PTRB20 1 1 NRR/DS I/AEB 26 1 1 NRR/DSI/A'SB 27 1 1 NRR/DSI/GPB =

10 1 1 NRR/DSI/CSB 09 1 1 NRR/DS I/ETSB 12 1 1 NRR/DS I/I(CSB 16 1 1 NRR/DSI/PSB 19 1 NRR/DSI/RA 8 22 1 1 NRR/D /RSB 23 1 NRR/DST/LGB 33 1 1 0Q 1 1 EXTERNALS ACRS 16 16 FEMA REP .DIV 39 1 1 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 05 1 1 Ep'7

,TOTAL NUMBER OF COPIES REQUIRED: UTTR 6 *ENCL

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Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372.-5000 January 13, 1982 G02-82-32 SS-L-02-CDT-82-012

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Docket No. 50-397 ~ 198/ e io

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Hr. A. Schwencer, Chief N85@gp g Licensing Branch No, 2 Division of Licensing U.S.'uclear Regulatory Commission Hashington, D.C; "20555

Subject:

.NUCLEAR PROJECT NO. 2 RESPONSE TO NUREG-0612 CONTROL OF HEAVY LOADS Reference': Letter, D.C. Eisenhut to all Licensees, et al, "Control of Heavy Loads," dated December 22, 1980 Enclosed are sixty (60) copies of the HNP-2 r esponse to NUREG-0612, "Control of Heavy Loads" transmitted via the reference letter. The HNP-2 draft SER open item on this subject should be closed upon receipt of this response.

Very truly yours, G.'. Bouchey Deputy Director, Safety and Security CDT/jca Enclosures cc: R Auluck'- NRC HS Chin - BPA R Feil - NRC Site 8202020268 PDR ADQCK 820ii3 A

05000397 PDR I

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CONTROL OF HEAVY LOADS

REFERENCE:

NRC to All Licensees, of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits, December 22, 1980.

The following are the Supply System's WNP-2 plant's responses to the request for information made in enclosure 3 to the referenced letter 2.1.1 Table I (attached) provides a listing of all overhead handling equipment with the potential to damage systems required for plant shutdown or decay heat removal.

2.1.2 The following cranes listed in Table I can be excluded for the following reasons:

MT-H01-36 This hoist is inside containment and dedicated to handling reactor recirculation flow control valve internals. It can only be used when the reactor is shutdown and the containment is open. It does not pass over the RHR return suction lines.

MT-HOl-13 This hoist is inside containment and is used for working on the 4 in board main steam containment isolation valves. It will only be used when the reactor is shutdown and containment is opened for mainten-ance work. These monorails and the hoist lie below both RHR supply lines and thus could not cause failure of the shutdown cooling mode of that system.

MT-HOl-19AAB These manual hoists are located inside containment and their mono-rail systems do NOT pass over the RHR lines or electrical cables required for shutdown cooling.

2.1.3 Table 2 lists the heavy loads to be handled by the Reactor Building a,b,c Crane (MT-CRA-2).

Sketches 1. through 8 are enclosed herein and are a part of the procedures listed in Table 2.

The control for assuring that these loads are moved in the prescribed path is procedural. However, the Reactor Building Crane is provided with limit switches to prevent travel over the spent fuel pool as shown on Drawing M570, (enclosure 1).

2.1.3 Verification that the lifting devices for the RPV/Dry-(d) well Head (NSSE EQ15), Dryer/Separator (NSSE - EQ - 14) and the Vessel Service Platform (NSSE EQ - 39) meet the requirements of ANSI N14.6 1978 is currently being analysed.

t 1

The slings procured for the Vessel Cavity and Oryer/Separator Pool Plugs (NSSE EQ 40) were purchased to a 39.6 ton safe working load. The four slings with their spread angle conform to ANSI 830.0 1971 requirements.

of 1I 2.1.3 Plant procedure 10.4.1 (enclosure 2) invokes the requirements (e) ANSI B30.2 1976, Chapter 2-2.

The reactor building crane meets or exceeds the design of CMAA specification 70 (enclosure 4).

2.1.3 There are no exceptions taken to ANSI 830.2 - 1976 with respect to (g) operator training, qualification and conduct.

2.2.1 Table 3 lists the cranes and hoists that are physically capable of carrying loads over spent fuel in the storage pool or in the reactor vessel.

2.2.2 The Refueling Platform/Service Platform Jib Crane (MT-CRA-9A,9B) are provided with load limiting devices that limit the load to 1200 pounds.

The Channel Handling Boom (MT-CRA-ll) is only designed for 200 pound loads. On this basis, these three cranes should be excluded from the criteria of 2.2.1.

2.2.3 The Reactor Building Crane (MT-CRA-2) meets the requirements for a "Single failure proof crane" as per NUREG 0612, Appendix C. See Enclosure 3 - " Letter "Single Failure Proof Reactor Building'Crane"

- B. A; Holmberg to 3. W. Hedges dated November 2, 1981, and the Whiting Bid proposal (enclosure 4).

In addition, travel of the Reactor Building Crane is limited for the main and auxiliary hooks as shown on enclosure l.

2.3.2.b The following list of cranes and hoists were installed to permit (3) maintenance of a specific piece of equipment. These lifting devices do not meet the requirements of NUREG OG12 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 placed in use until the equipment they service has been declared inoperable per the Plant Technical Specifications.

MT-HOl-'6 Services RHR Pumps A h E MT-H01-7 Services RCIC Pumps and Turbine MT-H01-8 Services RHR Pump C

MT-H01-9 Services LPCS Pumps MT-HOl-10 Services HPCS Pumps MT-CRA-6A h 6B Services Standby Service Water Pumps, lA h 1B MT-H01-18 Services Outboard Main Steam Isolation Valves MT-CRA-1 The Turbine Building Crane has the potential for damaging the high pressure turbine governor valves with their associated fast closure reactor shutdown switches. Procedural control consisting of a warn-ing NOTICE posted in the crane cab will restrict travel to areas outside the turbines when the turbines are in use. In addition, the RPS turbine control valve fast closure logic is "de-energize to trip" such that failure of the shutdown switches would produce a master scram.

MT-H01-19C Used for removing and reinstalling main steam relief valves (maximum weight 4,000/$ ), crosses over the 14" RHR loop B return to RPV at Azimuth 170o. The RHR line is approximately 18 feet under the valve.. passage and is protected by steel grating (1 1/2" deep 3/16 bars spaced 1 1/8" apart) supported on a 4'ect'angle of 8" and 14" deep I beams. The RHR line is a 7'adius bend at this location making a direct blow almost impossible even if the grating were .

penetrated. This coupled with the existance of an alternate shut-down cooling system (RHR Loop A) which does not pass under the re-lief valve monorail provide a ample assurance that shutdown cooling capability will not be compromised by a potential drop of a heavy load.

MT-Hol-16 This hoist is inside containment and is used to remove and reinstall the reactor recirculation motor and pump internals. The RRC-P-1A components cross over the single 20" RHR shutdown cooling suction line (elevation 509') from azimath 145o to 195o. This 20" line lies below the 512'rating and support structure. Because of its physical size the 30 ton pump motor cannot be hoisted more than 6" above the 512'rating level. If the pump motor were to drop, the structural *steel framework at elevation 512 would preclude damage to the 20" RHR suction line. As a backup, the RHR system has an alter-nate shutdown cooling path should this shutdown cooling line become inoperable.

TABLE 1 OVERHEAD HANDLING SYSTEM WITH POTENTIAL FOR DAMAGE TO ANY SYSTEM REQUIRED FOR PLANT SHUTDOWN OR HEAT REMOVAL TONS SEISMIC QUALITY.

TAG NUMBER LOCATION TYPE SERVICE/ CMMA CLASS CAPACITY CATEGORY CLASS MT-HOl-19AhB Reactor Building Trolly Hoist Main Steam Relief A-1 2 II II Inside Containment Manual Chain Valves MT-H01-36 Reactor Building Hoist Recirc Flow Control A-1 Inside Containment Manual Chain Valve (RCC-V-60)

MT-HOl-6 489'-2'rolly Reactor Building Trolly Hoist RHR Pumps A-1 II Electric (AAB)

MT-H01-7 Reactor Building Trolly Hoist RCIC Pump h A-1 ,5 492'-2n Electric Turbine MT-HOl-8 Reactor Building Trolly Hoist RHR Pump A-1 494'-3" Electric C MT-H01-9 Reactor Building Trolly Hoist LPCS Pump A-l 493-2" Electric MT-HOl-10 Reactor Building 492'-5" Trolly Hoist Electric-HPCS Pump A-1 20 ii MT-H01-16 . Reactor Building Trolly Hoist Recirc Pump 30 (Inside Containment) Electric MT-HOl-19C Reactor Building Trolly Hoist Main Steam Relief (Inside Containment) . Electric Valves

TABLE 1 (Continued)

TONS SEISMIC QUALITY-TAG NlNBER LOCATION TYPE 'ERVICE CMMA CLASS CAPACITY CATEGORY CLASS MT-CRA-6A,6B Standby Service Overhead Travelling Standby Service Water A-1 8 -. I I Water Pump House Crane (under hung) Pumps MT-CRA-2 606'rane Reactor Building Travelling Bridge Reactor Refueling Floor h Vessel A-1 125 Tons I MT-CRA-1 Turbine Building Travelling Bridge Main Turbine & 200 Tons Crane Generator MT-CRA-9A,9B Reactor Building Gib Cranes Spare and New A-1 Fuel Pool Electric Fuel Handling Vessel Service Platform MT-CRA-11 Reactor Building Jib Crane Reactor Service A-1 2008

- Platform Channel Handling Boom MT-HOl-18 Reactor Building Trolley Hoist Outboard Main Steam A-1 8 Tons Isolation Valve Work h Pipe Tunnel Hatch Removal MT-HOl-13 Reactor Building Trolley Hoist Inboard Main Steam A-1 8 Tons Onside Containment Isolation Valve Work

TABLE 2 HEAVY LOADS - REACTOR BUILDING CRANE Maximum Travel Path Load to be Handled Wei ht Liftin Device Procedure Sketch Vessel Cavity Shield Plugs 215,0008 NSSE - EQ - 40 PPM 10.3.2 ffl (slings)

Dryer-Separator Storage 120)0008 - EQ 40 PPM 10.3.2 82 Pool Plugs (slings) 16)000'SSE -

Fueling Slot Plugs NSSE EQ 41 PPM 10.3.2 83

-"(slings)

Drywell Head 104,000/I NSSE EQ - 15 PPM 10.3.3 84 Insulation Head 50,0008 Not Made Yet PPM 10.3.4 85 Reactor Pressure Vessel 186)700!k NSSE - EQ 15 PPM 10.3.5 86 Head "Cattle" Chute 22)000' Slings PPM 10.3.5 86 IIEST II

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Vessel Service Platform 12)0008 NSSE EQ '- 39 PPM 10.3.7 87

( IIES T II

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RPV Steam Dryer 80)0008 NSSE EQ 14 PPM 10.3.6 88 RPV - Moisture Separator 146) 50'SSE EQ 14 PPM 10. 3. 6 88 In-Vessel Rack 6008 Slings PPM 10.3.8 Fuel Cask Not Purchased Not Purchased PPM 10.3.9 est. to weight t'o date (not yet written)

Approx. 100 Tons

TABLE 3 EQUIPMENT NAME TYPES CAPACITY DESIGNATION Reactor Building Crane Travelling Bridge Crane 125/15 Tons MT-CRA-2 Refueling Platform Bib Crane 120Ã ( MT-CRA-9A,98 h Service Platform Covers Channel Handling Boom Jib Crane 200// MT-CRA-ll

ATTACHMENT 2 ANALYSIS OF RADIOLOGICAL RELEASES Section 15.7.4 of the WNP-2 Final Safety Analysis Report discusses "Fuel Handling Accident". The modified Table 2.1-2 shown below compares WNP-2 assumptions with those provided in Attachment 7.

, TABLE 2.1-2 HEAVY LOAD DROP ACCIDENT ASSLNPTIONS Reactor Type PWR and BWR WNP-2 Values Power Level (Mwt) 3,000 31 323 0-2 hour X/Q (Exclusion area boundary), sec/M3 1.0xlO = 3 1/ l.xlO 3 0-2 hour X/Q LPZ, sec/M3 1.0xl0--4 1/ l.xlO 4 Peaking Factor 1 22/

No. of Assemblies in Core 193(PWR), 760(BWR) 764 Pool Water Decontamination Factor 1003/ (for radio-)

active iodines)

Filter Efficiency X:

Elemental Iodine 95X4/

Organic Iodine 95X Cooling Time (hours) 100 or greater 24 Hours Based on 5X worst meteorological conditions.

Value is 1.2 for, greater than one damaged fuel assembly. For a single assembly the values are 1.65 and 1.5 for PWRs and BWRs, respectively.

3/ See Reg. Guide 1.25 4/ See Reg. Guide 1.52 NOTE: With the exception .of power level WNP-2 is more conservative than the assumptions shown in Table 2.1-2.

ATTACHMENT 3 CRITICALITY ANALYSIS Section 9.1.2.3 of the WP-2 Final Safety Analysis Report discusses critical-ity control as it is associated with spent fuel storage. Technical Specifica-tion Sections 3/4.9.6 "Refueling Plantform Operability" and 3/4.9.7 "Crane Travel, Spent Fuel Storage Pool" provides limiting conditions for operation to limit loads moved over the spent fuel pool to stay within the bounds of the FSAR Criticality Analysis.

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NUCLEAR PROJECT NO. 2 EL. 606'- 10 I/2M REACTOR BUILOIN