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 CIL:50 397 l<PPSS Nuclear ProJectF Unit

? F Washington

>Publ ic Powe AUTH DYNAM)E 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

DOCKET 05000397

SUBJECT:

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

DISTRIBUTION CODE:

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J5 TITLE:'ontrol of Heavy l.oads Near Spent Fool (USI,A 26)

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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 Docket No. 50-397 Hr. A. Schwencer, Chief Licensing Branch No, 2

Division of Licensing U.S.'uclear Regulatory Commission Hashington, D.C; "20555

Dear Hr. Schwencer:

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Subject:

Reference':

.NUCLEAR PROJECT NO.

2

RESPONSE

TO NUREG-0612 CONTROL OF HEAVY LOADS

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

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Feil NRC Site 8202020268 820ii3 PDR ADQCK 05000397 A

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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 2.1.2 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.

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 2.1.3 a,b,c 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.

Table 2 lists the heavy loads to be handled by the Reactor Building 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 (d)

Verification that the lifting devices for the RPV/Dry-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

2.1.3 (e) 2.1.3 (g) 2.2.1 2.2.2 2.2.3 2.3.2.b (3)

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.

1I Plant procedure 10.4.1 (enclosure

2) invokes the requirements of ANSI B30.2 1976, Chapter 2-2.

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

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

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.

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.

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.

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 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 MT-H01-7 MT-H01-8 Services RHR Pumps A h E Services RCIC Pumps and Turbine 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 MT-CRA-1 Services Outboard Main Steam Isolation Valves 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 TAG NUMBER MT-HOl-19AhB LOCATION Reactor Building Inside Containment TYPE Trolly Hoist Manual Chain SERVICE/

Main Steam Relief Valves TONS SEISMIC QUALITY.

CMMA CLASS CAPACITY CATEGORY CLASS A-1 2

II II MT-H01-36 MT-HOl-6 Reactor Building Inside Containment Reactor Building 489'-2'rolly Hoist Manual Chain Trolly Hoist Electric Recirc Flow Control Valve (RCC-V-60)

RHR Pumps (AAB)

A-1 A-1 II MT-H01-7 Reactor Building 492'-2n Trolly Hoist Electric RCIC Pump h Turbine A-1

,5 MT-HOl-8 Reactor Building 494'-3" Trolly Hoist Electric RHR Pump C

A-1 MT-H01-9 Reactor Building 493-2" Trolly Hoist Electric LPCS Pump A-l MT-HOl-10 Reactor Building 492'-5" Trolly Hoist Electric-HPCS Pump A-1 20 ii MT-H01-16

. Reactor Building (Inside Containment)

Trolly Hoist Electric Recirc Pump 30 MT-HOl-19C Reactor Building Trolly Hoist (Inside Containment)

Electric Main Steam Relief Valves

TAG NlNBER TABLE 1 (Continued)

TONS SEISMIC QUALITY-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 Reactor Building Travelling Bridge 606'rane Reactor Refueling Floor h Vessel A-1 125 Tons I

MT-CRA-1 Turbine Building Travelling Bridge Crane Main Turbine Generator 200 Tons MT-CRA-9A,9B Reactor Building Gib Cranes Fuel Pool Electric Vessel Service Platform Spare and New Fuel Handling A-1 MT-CRA-11 Reactor Building Jib Crane Reactor Service

- Platform Channel Handling Boom A-1 2008 MT-HOl-18 Reactor Building Trolley Hoist Outboard Main Steam Isolation Valve Work h Pipe Tunnel Hatch Removal A-1 8 Tons MT-HOl-13 Reactor Building Trolley Hoist Onside Containment Inboard Main Steam Isolation Valve Work A-1 8 Tons

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

PPM 10.3.2 ffl Dryer-Separator Storage Pool Plugs Fueling Slot Plugs 120)0008 16)000'SSE

- EQ 40 (slings)

NSSE - EQ 41

-"(slings)

PPM 10.3.2 82 PPM 10.3.2 83 Drywell Head Insulation Head Reactor Pressure Vessel Head "Cattle" Chute 104,000/I NSSE EQ - 15 PPM 10.3.3 84 50,0008 Not Made Yet PPM 10.3.4 85 22)000' IIEST II)

Slings PPM 10.3.5 86 186)700!k NSSE - EQ 15 PPM 10.3.5 86 Vessel Service Platform 12)0008

( IIES TII)

NSSE EQ '- 39 (slings)

PPM 10.3.7 87 RPV Steam Dryer RPV - Moisture Separator In-Vessel Rack Fuel Cask 6008 Slings Not Purchased Not Purchased est.

to weight t'o date Approx. 100 Tons PPM 10.3.8 PPM 10.3.9 (not yet written) 80)0008 NSSE EQ 14 PPM 10.3.6 88 146)50'SSE EQ 14 PPM 10. 3. 6 88

TABLE 3 NAME TYPES Refueling Platform h Service Platform Covers Bib Crane Reactor Building Crane Travelling Bridge Crane EQUIPMENT CAPACITY DESIGNATION 125/15 Tons MT-CRA-2 120Ã

( MT-CRA-9A,98 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 0-2 hour 0-2 hour Power Level (Mwt)

X/Q (Exclusion area boundary),

sec/M3 X/Q LPZ, sec/M3 Peaking Factor No. of Assemblies in Core Pool Water Decontamination Factor Filter Efficiency X:

Elemental Iodine Organic Iodine Cooling Time (hours) 3,000 1.0xlO = 3 1/

1.0xl0--4 1/

1 22/

193(PWR),

760(BWR) 1003/ (for radio-)

active iodines) 95X4/

95X 100 or greater 31 323 l.xlO 3 l.xlO 4 764 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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MASHINGTON PUBLIC POllER SUPPLY SYSTEM GENERAL ARRANGEMENT PLAN EL, 6']2 ~

NUCLEAR PROJECT NO.

2 EL. 606'-

10 I/2M REACTOR BUILOIN