ML20117H683
| ML20117H683 | |
| Person / Time | |
|---|---|
| Site: | 05000496, 05000497 |
| Issue date: | 03/03/1978 |
| From: | Switzer D NORTHEAST NUCLEAR ENERGY CO. |
| To: | Regan W Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20114E366 | List: |
| References | |
| FOIA-96-214 NUDOCS 9609090370 | |
| Download: ML20117H683 (10) | |
Text
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PO ECX 270 MArrC:C CCNNECT: CUT 06'31 NoRTHEA$T NUC.. EAR ENERGY COMPANY '\ 2W66& i A NCATHEAST UTILtilES COMPANY
~< a March 3, 1978
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YdkII p/ l Office of Nuclear Reactor Regulation
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ATTN: Mr. William H. Regan, Jr., Chief M ,'j - Environmental Projects Branch 2 U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 ' G:ntlemen: Docket Nos. 50-496 50-497
SUBJECT:
Montague Nuclear Power Station Units 1 and 2 Request for Additional Information As requested in your letter of February 24, 1978, Northeast Nuclear Energy Company will file a supplement to its Environmental Report on or about March 31, 1978. This update will be based upon our best available estimate of the impact of a 1990 in-service date for Unit 1 on the issues identified in your letter. Very truly yours, NORTHEAST NUCLEAR ENERGY COMPAtlY By Northeast Nuclear Energy Company Their Agent [ M D. C. Switzer /\ President J cc: Copy List i I l-f@ 96 790750048 k l 9609090370 960820 PDR FOIA l$,, ,DEKOK96-214_ PDR
i 1 l COPY LIST i 1 FREDERIC J. CCUFAL, ESo., CHAIRMAN l
'ATCMIC SAFETY & LICENSING APPEAL ATCMIC SAFETY & LICENSING 80ARO BOARD PANEL U.S. NUCLEAR REGULATCRY CCtHISSICN '
I U.S. WCLEAR REGU.ATORY CCtHISSICN WASHINGTON, D. C. 20555 WASHINGTCN, O. C. 20555 l CR. RCCERT L. HOLTON ;
- m. RALPH S. DECKER SCmCL OF OCEANCGRAPHY ROUTE 1 OREGCN STATE UNIVERSITY Ecx 19Co CCRVALLIS, CREGCN 97331 CAMBRIDGE, MARYLANDs 21613 i
i j EDWARD G. KETCHEN, JR., ESo. CARoEGIE LIERARY j AVENUE A EDWIN J. REIS. ESo. TURSERS FALLS, MASSACHUSETTS 01376 i OFFICE CF THE EXECUTIVE LEGAL DIRECTCR ! U.S. NUCLEAR REGULATCRY CCMMISSICN ; WASHINGTON, D. C. 20555 JAMES R. TOURTELLOTTE, ESo. JACK 0. CURTISS. esc. CALLAhAN, CURTISS Ato CAREY ASST. CHIEF HEARING COUNSEL 173 MAIN STREET CFFICE OF T:-E EXECUTIVE LEGAL DIRECTCR GREENFIELD. MASSACHUSETTS 01301 U.S. NUCLEAR REGULATORY COMMISSION WASHINGTCN. D. C. 20555 LAURIE BURT, ESo.
- m. FREDERICK J. N E' JOSE R. ALLEN, ESo.
COUNTY CF FRANKLIN PLANNING CEPARTMENT ENVIRONMENTAL PROTECTION DIVISION 625 MAIN STREET 04E ASHSURTCN PLACE - 19TH FLCCR GREENFIELO, MASSACHUSETTS 01301 BOSTON, MASSACHJSETTS 02108 MARK I. BERSCN, esc. KARIN P. SHELOCN, ESo. LEVY, WINER & HCCCS SHELDCN, HARMON & RCISMAN 1025 ISTH STREET, N.W. - STH FLOCR P. O. Box 840 GREENFIELD, MASSACH.1 Sci is 01301 WASHINGTCN, D. C. 2000S E. TUPPER XINDER. ,ESO. ATCMIC SAFETY & LICENSING 80ARO PANEL ENVIROBNENTAL PROTECTICN DIVISICN U.S. NUCLEAR REGULATCRY CCfMISSICN CFFICE CF ATTORNEY GENEHAL WASHINGTCN, D. C. 20555 STATE HCUSE ANNEX, RCCM 208 CONCCRo, New HAMPSHIRE 03301 GREGCR I. MCCREGCR, esc. . CCCKETING & S Rv!CE SECTICN 33 MT. VERFCN STREET OFFICE CF TFE SECRETARY BOSTCN, MASSACHJSETTS 02108 U.S. NUCLEAR REGut ATCRY CCtNISSICN MASHINGTCN CC 2C555 GERALD GARFIELO, EscutRE RICHARD L. MCRNINGSTAR. ESo. PEABCOY, BRCWN, RC% LEY L STCREY DAY, 8ERRY Ato HCWARo CNE 8CSTCN PLACE CNE CCNSTITUTICN PLAZA HARTFCRO, CT C6101 8CSTCN MA 02108 i k
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COPY LIST l i J MORRIS K. MCCLINTOCK, ESo. DOCKET CLERK
- CONSERVATION LAW FOUNDATION ENERGY FACILITIES SITING COUt3CIL ONE ASH 3URTCN PLACE - RM. 1413 l TWEE JOY STREET BOSTON, MASSACHJSETTS O2108 EOSTCN, MASSACHUSETTS O2108 l
' THOMAS 8. LESSER, ESo. OAvro S. PINARo! 39 MAIN STREET , C MSTNUT HILL ROAD NCRTHAMPTCN, MASSACHJSETTS 01060 l hDNTAGUE, MASSACHUSETTS 01351 ! i BARRY S. ZITIER, ESo. l JOm F. X. DAVC8EN OFFICE OF CONSUMER COUNSEL BUILDING & CCNSTRUCTION TRADES COUNCIL STATE OFFICE BUILDING ! AFL-CIO , HARTFORD. CCtNECTICUT 06115 l 11. BEACON STREET l BOSTON, MASSACMJSETTS O2108 ! l -** STEVEN FERREY. ESo, REP. RICHARD P. ROCHE THE STATE HOUSE - ROCM 437 i NATIONAL CCNSUMER LAW CENTER j BOSTCN, MASSACHUSETTS O2133 , 11 BEACON SMEET +
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l BOSTCN, MASSACHUSETTS O2108 i PETER SCOTT RIDER, ESo. ELLYN R. WEISS. ESo. , SHELDCN, HARMON & ROISMAN MASS PIRG 1025 1STH STREET, N.'./. , SUITE 500 233 N. PLEASANT SmEET WASHINGTON, D. C. 20005 l AM ERST MASSACHUSETTS 01002 l l WAYNE S. HENCERSON, ESo. OENNIS J. LACROIX, Esc. I HARRISON A. FITCH, ESo. ENERGY FACILITIES SITING CCUNCIL ONE ASHOURTCN PLACE, RM. 1413 Ne;! ENGLAND LEGAL FcCNDATIcN BOSTON, MASSACHUSETTS O2108 110 TremOnt street BOSTON, MASSACHJSETTS 02108 l I Edward J. Dailey, Esq., Executive Director Massachusetts Energy Facilities Sitino Council One Ashburton Place, ROcm 1413 l Boston, Massachusetts 02108
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METROPOLITAN EDISON COMPANY :_ ::+ :. w v _ .,_ : . - . _ - a j POST OFFICE BOX $42 READING. PENNSYLVANI A 19603 TELEPHONE 215 - 929 3601 3 March 16, 1978 J GQL Oh62 i l - 3 i . '.\ j Director of Nuclea*: Reactor Regulation e..,' .f f ; 3 $ , i Attention: Mr. S. A. Varga, Chief g ., / .< . A ., Light Water Reactors Branch, No. 4 . q N ., . , ( .f U. S. Nuclear Regulatory Commission .
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j Washington, D. C. 20555 . D ,],. .
Dear Sir:
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Three Mile Island Nuclear Station Unic 2 (TMI- ) Docket No. 50-320 Operating License No. DPR-73 Additional Information Concerning Fire Protection The following information is submitted at the verbal request of your Mr. T. Lee on March 3, 1978 to Mr. R. C. Cutler (GPUSC) and supplements the information provided by our letter to you dated February 17, 1978.
- 1. Rupture of fire service piping (Operating License Condition 2.c.(3)l.2).
During the evaluation, the postulated fire service piping break was considered to occur anywhere along the length of the pipe, not just select locations.
- 2. Additional Hose Stations and Diesel Generator Basement Sprinkler Systems (Operating License Conditions 2.c(3)l.1 and G.10 of -
Attachment 2). A. A copy of the Burns and Roe, Inc. Specification 2555-146, revised to incorporate clarifications and corrections as discussed, is forwarded herewith for your information. B. The minimum density coverage requirement of .3 gpm per square foot of diesel generator basement applies to the overall floor square footage as shown on B&R General Arrangement Drawing 2339. Should you have any additional questions, please contact my staff. y truly yours,
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[ / J. . Herbein Vice President-Generation 00 4 e, t sc rn
<~ny4 Enclosure, Burns and Roe, Inc.
*' * [ l Specification 2535-146, Rev.1. I h wa '1{
- g. Enclosure GQL Ch62 4
,) SPECIFICATION 2555-146 SPECIFICATION DETAILS ADDITIONAL FIRE HOSE STATIONS AND WET PIPE SPRINKLER SYSTEMS l l . l l C JERSEY CENTRAL POWER AND LIGHT COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT NO. 2 l l 4 j Burns and Roe, Inc. .
- Engineers and Constructors Rev. 1 29 Park Place l
{]' l Paramus, New Jersey 07652 G 90 '! 0 N E' ' I S )P n
i TECHNICAL SI :CIFICATIONS
- , IJR ADDITIONAL FIRE HOSE STATIONS AND WET PIPE SPRINKLER SYSTEMS TABLE OF CONTENTS ARTICLE PAGE l
1.0 SCOPE 1 2.0 GENERAL 1 2.1 -Work to be Provided 1 2.2 Work by others 2 l 2.3 Codes and Standards 2 l 2.4 Drawings 4 l l 3.0 DETAILED REQUIREMENTS 4 3.1 Design Conditions 4 3.2 Materials of Construction 5 l 3.3 Welding Requirements 7 l 3.4 Hose Stations 7 3.5 Wet Pipe Sprinkler Systems 7 3.6 OS and Y Gate Valves 8 3.7 Hangers for Support of the Piping Systems 9 3.8 Painting and Cleaning 9 _ 3.9 Nameplates 9 4.0 INSTALLATION 10 l 5.0 TESTING 10 6.0 INFORMATION TO BE SU5MITTED 10 Table 1 Hose Stations and Wet Pipe Sprinkler Systems 11 Attachment 1 Seismic Response Curves Attachment 2 Information Drawings - i l
)
l TECHNICAL SPECIFICATIONS FOR .
) ADDITIONAL FIRE HOSE STATIONS AND WET PIPE SPRINKLER SYSTEMS 1.0 SCOPE This Specification covers the furnishing and installing of fire protection piping, valves, sprinkler nozzles, hose stations, wet pipe system alarm check valves, pressure switches and other necessary accessories to provide for the addition of fourteen (14) fire hose stations and two (2) wet pipe sprinkler systems to be added to the existing TMI Unit 42 fire protection systems. The locations of the additional hosa stations are detailed in Table 1 of this Specification and are also shown on the Information Drawings attached to this Specification.
2.0 GENERAL . I 2.1 Work to be Provided Contractor shall furnish and install all piping fittings and valves as required for a complete installation of all Fire Protection Systems specified herein. The work to be provided
~)s under this Specification shall include the following:
. 1 2.1.1 Furnishing and installing of fourteen (14) fire hose l stations, each consisting of a wall reel, 75 feet of hose, hose I angle valve, OS and Y isolation gate valve with monitor switch, and hose nozzle.
2.1.2 Furnishing and installing of two (2) wet pipe sprinkler systems for protection of the basement cf both Diesel Generator __ buildings (El. 230'-6"). Two separate a d independent systems shall be provided, one for each basement area. 2.1.3 Furnishing and installing of all hangers (temporary and permanent) for support of the fire protection piping, and valves. 2.1.4 Engineering and design of the piping and sprinkler lay-out for the wet pipe sprinkler systems. 2.1.5 Engineering and design of the piping layout for the hose stations. 2.1.6 Engineering and' design cf all hangers for support of the fire protection piping and valves. 2.1.7 Submission of Seismic analysis to qualify the hanger sup-port system for the piping. e
- i 2.1.8 All drilling and welding work necessary to tap into the og existing- fire protection piping lines. '
2.1.9 Delivery of all equipment, valves, piping, etc. to jobsite. 2.1.10 Installation of all piping and valves necessary for a complete system installation as specified herein. 2.1.11 All labor, supervision of labor and facilities for packaging, receiving, unloading, checking, storing and handling of.the fire protection equipment. 2.1.12 . Field testing of the equipment per NFPA Code requirements. 2.1.13 Shop testing of fire protection gate va~1ves as specified herein. 2.1.14 Shop painting and cleaning of hangers, piping and valves. 2.1.15 Performing of hydraulic calculations for the piping systems. 2.2 Work by others The following work will be provided by others: 2.2.1 All wiring to the Contractor's supplied electrical equip- , ment where required. 2.2.2 All field finish painting of piping and valves. 2.2.3 All necessary concrete core drilling to permit pipe pene-trations. 2.2.4 Furnishing and installing of pipe sleeves for penetrations. 2.2.5 Hydrostatic testing of piping systems. 2.3 Codes and Standards The codes / standards or regulations listed below are those mentioned or referenced in this Specification. The latest edition of these codes / standards or regulations in effect or promulgated at the time of award shall apply: 2'. 3.1 NFPA Code !13, Sprinkler Systems, Installation 2.3.2 NFPA Code #14, Standpipes and Hose Stations 2.3.3 Underwriters Laboratories (UL) 2.3.4 Factory Mutual Engineering Corporation (FM) 2.3.5 Manufacturer's Standardization of the Valve and Fitting Industry (MSS-SP).
MSS-SP Hydrosta'ic Testing of Steel Valves
)
MSS-SP MSS Standard Marking System for Valves Fittings, Flanges and Unions MSS-SP Pipe Hangers and Supports - Materials and Design MSS-SP Pipe Hangers and Supports - Selection and Application 2.1.6 American National Standards Institute B31.1 Power Piping B16. C.I. Pipe Flanges and Flanged Fittings 2.1.7 American Society of Mechanical Engineers Section IX - Welding and Brazing Qualifications 2.1.8 Steel Structures Painting Council Specifications (SSPCS) SSPC-SP-2-63 Hand Cleaning SSPC-SP-3-63 Power Tool Cleaning 2.1.9 Uniform Building Code, International Conference of Building Officials (']) 2.1.10 American Welding Society (AWS) 1 D1-0-69 Welding in Building Construction ' A2.0 Standard Welding Sy=bols A2.2 Nondestructive Testing Symbols A3.0 Definitions - Welding and Cutting 2.1.11 American Society for Testing and Materials (ASTM)
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A233 Mild Steel Arc - Welding Electrodes A234 Piping Fittings of Wrought Carbon Steel and . Alloy Steel for Moderate and Elevated Temperatures A53 Welded and Seamless Steel Pipe ' Al20 Black and Hot Dipped Zine Coated (Galvanized) Welded and Seamless Steel pipe for ordinary uses. Invocation by title, name and/or number of certain specific codes / standards or regulations, under this paragraph or elsewhere in this Specification, shall in no way diminish Contractor's responsi-bility for ccmpliance with any and all codes / standards or regulations which are generally recognized to be applicable to the work herein specified.
) '
2.4 Drawings , O .) The information drawings (Plant General Arrangement Drawings)
- included with this Specification show the locations of the additional hose stations to be provided by the contractor. Contractor shall prepare detailed piping layout drawings showing the location of all valves, hangers, specialties, floor penetrations, etc. Centracter shall clear all interferences in the field with respect to laying out his piping. The piping drawings generator by the Contractor shall clearly show dimensioned hanger locations and shall show where the new fire piping will tap into the existing fire protecticn system piping lines. In addition, all new fire piping line sizes shall be shown on the drawings and determined by the Contractor based on NFPA code requirements.
Contractor shall also prepare detailed hanger configuration drawings which shall be submitted with the piping seismic analysis. 3.0 DETAILED REQUIREMENTS 3.1 Design Conditions 3.1.1 Piping inside the following buildings shall be created as Seismic Class I: Auxiliary Building Fuel Handling Building Diesel Generator Building } Control Building Control Building Area Reactor Building Seismic analysis shall be performed in accordance with data contained in the respective response spectra curves (See AttacNment 1 to this Specification) which are based upon 0.5% damping. Equivalent static leads shall be developed shrough dynamic analyses of the response cf the piping and equipment to horizontal and vertical accelerations, considered acting simultaneously, using the response spectra curves indicated above and contained in Attach-ment I of this Specificacicn. Note that the value of horizontal acceleration shall be obtained by entering the applicable response spectra curve with the pericd of free vibration (T), to which a " tolerance" of 1 0.02 seconds shall be added. The largest acceleration value for this band of period (T t 0.02 seconds) shall be used for design. The vertical acceleration shall be considered as 2/3 of the horizental acceleration. Ecri cntal and vertical loading shall be applied simultaneously in addition to the normal design loads for idle and/' or operating conditions. 3
Piping and equipment snall be designed using code stress 'g values for normal design loads (AMSI B 31.1) plus the operating basis earthquake load (OBE). t .i addition, equipment must have "no loss of function" based upon normal design loads plus Design Basis
.. Earthquake Load (DBE).
An equivalent vibration test using double the design earth-quake load may be substituted for the above analysis. 3.1.2 Seismic design classification, Class II Piping not included above shall be Class II and shall be designed for Zone I loads using methods and stress increase coeffi-cients given in the latest edition of the Uniform Building Code of the International Conference of Building officials. 3.1.3 The wet pipe sprinkler system shall be designed and in-stalled according to the requirements of NFPA Code #13. The place- i ment of all components associated with sprinkler systems shall be in strict accordance with NFPA Standard No. 13. The minimum den-
.- sity coverage for all sprinkler areas shall be .3 gpm per sq. ft.
Minimum orifice size for the sprinkler nozzles shall be 1/2". Temperature ratings of the srinkler fusible links shall be inter- , mediate. 3.1.4 All pipes, valves, fittings, joints, instruments and accessories subjected to fire pump water pressure shall be suit-D able for a 175 psig water working pressure unless otherwise stated. 3.1.5 All valves, spray nozzles, hoses, hose racks, control devices, and other items of equipment shall be U L and/or F.M. approved and so labeled or certified for fire protection service. 3.1.6 Hose Stations shall be designed for Class II services in accordance with the requirements of NFPA Code #14. The piping line sizes to hose stations shall be consistent with the. require-ments stated in NFPA Code 414. Piping line sizes shall be clearly shown on the Contractor's piping drawings and shall require Engineer approval prior to installation. 3.2 Materials of Construction 3.2.1 All piping up to and including 4 inches nominal sizes on the system side of the wet pipe sprinkler valves shall be threaded schedule 40 ASTM Al20 black steel pipe. Fittings shall be standard 150 lb. malleable iron fittings. 1 3.2.2 All 6 inch piping on the system side of the wet pipe sprinxler alarm valves shall be unreaded schedule 40 ASTM A33 black, steel pipe. Fittings shall be standard 150 lb. malleable iron fittings. l1 D
,4 3.2.3 All piping up to and i;cluding 4 inche:; norminal size on the supply side of the wet pipe sprinkler systen valves shall be schedule 40 ASTM Al20 black steel pipe. This pipe shall be welded
" in accordance with paragraph 3.3 of this Specification. Welded fittings shall be standard ASTM A234, WPB forged. steel.
3.2.4 All 6 inch piping on the supply side of the wet pipe sprinkler system valves shall be schedule 40 ASTM A53 black steel pipe. This pipe shall be butt welded in accordance with paragraph 3.3 of this Specification. Welded fittings shall be standard ASTM A234 WPB forged steel. , 1 3.2.5 All piping to hose stations up to and including 4 inches l nominal sizes shall be schedule 40 ASTM A120 black steel pipe. , All 6 inch piping to hose stations shall be schedule 40 ASTM A53 l black steel pipe. Piping 2h inch nominal si cs and above to hose I stations shall be butt welded in accordance with paragraph 3.3 of this Specification. Welded fittings shall be standard ASTM A234 l WPB forged steel. Piping 2 inches nominal sizes and under to hose l
, stations shall be threaded. Fittings shall be standard 150 lb.
I malleable iron fittings. 1 , 3.2.6 Flanged connections shall be provided where required in ) accordance with NFPA 13 requiremencs. 1
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1 3.2.7 Gate Valves (2 1/2" and Larger) - Material & Specification Cast Iron Body l Bronze Mounted ASTM A-126 Class B Type , Bolted Sonnet, O.S. & Y. Rating 175 lb. ANSI Standard Seating Surface, Including Back Seat Bronze Ends Flanged 125 lb. ANSI Standard (316.1) 3'2.8
. Piping Support Hangers - all hangers shall be fabricated from A-36 steel.
3.3 Welding Recuirements - ) The Welding Requirements applied for this Specification shall be the samo as those specified in Specification 2555-72, ! Section 15-H, paragraph 3. 3 entitled "FABRICATICM" . 3.4 Hose Stations Fire hose stations shall consist of a continuous flow wall reel, Wirt & Knox Style FD47 or Engineer approved equal. Hose reels shall be suitable for working pressure up to 300 lbs. and shall be sized to accomodate 75 feet of 1 1/2 inch all service hose. Hose shall be Wirt and Knox Servall non-collapsible hose, 23-0222-2 braid, oil resistant with neoprene cube and cover, or Engineer approved equal. Hose stations shall also include i 1/2 inch Seco type 76U angle valve er engineer approved equal. valve shall incorporate standard tapered iron pipe thread on both ends. Hose no::le shall be Powhatton 02-349 "All Fog" Nozzle, cast brass
- with satin brass finish, for use with 1 1/2 inch hose or Engineer l approved equal.
Each hose station shall be provided with an OS&Y isolation gate valve (minimum size - 2 1/2") with a Grinnel model 4FG40 (or Engineer approved equal) monitor switch. Monitor switch shall be rated 125 volts AC, 10 amps. l 3.5 Wet Pipe Sprinkler Svstems Wet Pipe Sprinkler Systems shall be furnished for the areas specified in Table 1 of this Specification. Each wet pipe sprinkler system shall incorporate an alarm check valve, an OS&Y isolation gate valve (with monitor switch) located upstream of the check valve, pressure switch to indicate system water flow, local water motor gong, pressure gauge for local readout of system pressure, test pipes, drain and ven lines and other necessary accessories which are required by NFPA Code #13. l Minimum size of gate valves and alarm check valves shall be 4". The monitor switch furnished for the gate valve shall be the same as that specified in 3.4 abova for the hose stations. The pressure switen shall be Grinnell model number B-2 or Engineer approved equal. Pressure switch shall be rated 10 amps for 115/ 230 volts AC and shall be provided with one open and one closed circuit. f LO
, 3.6 OS and Y Gate Valves * . The OS and Y gate valves shall be constructed to meet the following requirements in addition to the valve material requirements specified in paragraph 3.2 of this Specification:
Identification plates shall be provided and permanently attached to each valve. Identification plates shall be carbon steel and have black identification figures stamped hereon. Marking on plates shall be in accordance with MSS-SP-25 and shall also state name of system of which they are a component 1 and the tag number of the valve (see Table 1 of this Specifi- l cation). Accessories, if not mounted on the valve, shall be similarly identified with the valve mark number. Rotation arrows for open and close shall be marked on all handwheels. I Valve flange end connections shall be in accordance with ANSI B16.f. 1 Valves shall be designed for repacking at full rated i pressure with valve open. Valves shall be provided with back seat arrangement to prevent leakage-into the gland chamber.
._) The latest revision of UL 262 shall apply to the gate valves to be furnished under this Specification.
3.7 Hangers for Support of the Pipine Systems Contractor shall furnish and install all the necessary hangers, supports, anchors, guides, braces, concrete inserts, supplementary steel, and accessory equipment to independently hold and support the fire protection systems in their proper locat' ions for the designed coverage. Where required to do so because of location, Contractor shall furnish necessary seismic analysis for the proposed hanger configurations and designs and for .the proposed loca: ions of the hangers. Unless otherwise specified, all pipe hangers and support assemblies shall comply with ANSI S31.1, Power and Piping Code, the current issue of the Manufacturer's Standardization Society Standard Practice SP-58 and SP-69 concerning " Pipe Hangers and Supports", and the applicable portions of NFPA Code #13. 8 o
3.8 Painting and Cleaning
]) .1 All unfinished =iscellaneous steel furnished under this Specification for support of piping shall be thoroughly cleaned of dirt, or mill scale in accor-dance with Steel Structures Painting Council Speci-fications SSPC-SP-2-63, " Hand Cleaning" or SSPC-SP-2-63, " Power Tool Cleaning" as required. All of the aforesaid miscellaneous steel shall receive one shop coat of Keeler & Long, Inc., Tri-Polor White Primer 6040 with a dry film thickness as recommended ,
by the paint manufacturer. 1
.2 Outside surface of all pipe shall receive one (1) shop coat of lead-free primer. Shop paint shall be applied to a minimum thickness of 2 mils.
l
.3 Paint shall be cmitted at all areas of field weldine l for a minimum of one (1) inch on each side of the welc preparation. A water-soluble rush-inhibitor shall, instead, be used. All contact surfaces of field bolted connections may be painted.
3.'9 Nameplates All pressure switches, valves and monitor switches shall be labeled with individual nameplates which shall contain the , ]) following information:
- a. Unit tag numbers as designed by the Engineer. Unit tag numbers are specified in Table 1 of this Specifi-cation.
- b. Manufacturer's name.
- c. Manufacturer's serial number.
These nameplates shall be made of suitably inscribed metal or plastic and affixed by screws or attached by stainless steel or copper wire. 4.0 INSTALLATION ' Installation work shall include all recrating, moving from storage, rigging, setting, assembly, alignment, grouting, cleaning, testing and all other necessary work to prepare each system cf equipment and its integral parts for normal service. Contractor shall furnish all labor, materials and equipment necessary to ccmplete the mechanical installation of all fire protection systems and piping specified herein. All field elec-trical work, that is, all external electrical wiring between the Contractor's supplied equipment, will be provided by others.
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The interior of all pipe iTd fittins shall be thoroughly p% cleaned of all foreign matto Iafore being installed and shall be kept clean until the work hc ; been accepted.
.. Every precaution shall be taken to prevent foreign material from entering the pipe while it is being installed. It is essen- :
tial that no foreign matter be permitted,to enter the pipelines at any time. Contractor shall be responsible for laying out and installing his piping systems clear of all service piping, ductwork, equip- ! ment, cable trays, lighting fixtures and structures. It shall l be the Contractor's responsibility to field check his piping lay- 1 out to avoid interferences. Contractor shall be responsible for making all required corrections resulting from faulty layout and , installation work at no additional cost to the Owner. I 1 l 5.0 TESTING !
.1 The wet pipe sprinkler systems shall b2 tested per NFPA Code #13 requirements.
.2 Fire Protection gate valves shall be given a hydrostatic test on the body and seat in accordance with MSS-SP-61.
6.0 INFORMATION TO BE SUBMITTED 5 .1 Detailed piping layout drawings showing the physical location of all valves, hangers and floor penetrations and the sizes of all fire piping lines.
.2 Seismic analysis and hanger designs for the Seismic I piping system.
.3 Specification data sheets (including materials of construction) for all fire equipment to be furnished
_ _. - under this Specification.
.4 Test Data sheets shall be furnished to insure testing compliance with NFPA Ccde #13.
i
.5 Submission of hydraulic calculations substantiating pipe line sizes and orifice no::le sizes for the water 1 systems.
1
s Table 1
/ HOSE STATIONS
- 1. Control Building Mechanical Equipment Room, Elevation 351'-6"; C47-CA (South wall) ; Locate hose station 12 feet from inside of wall. (Tag numbers: Isolation gate valve - FS-V-654, Hose angle valve - FS-V-655, Monitor switch - FS-KS-6700)
- 2. Control Building Mechanical Equipment Room, Elevation 351'-6"; Between C49-C50 & CD, (North wall) ; Locate hose station 10 feet from Panel 709AG (edge of panel)
(Tag numbers : Isolation gate valve - FS-V-656, Hose angle valve - FS-V-657, Monitor switch - FS-KS-6701)
- 3. Control Building Battery & DC Switchgear Room, Elevation l 280'-6", C-49-CA (South wall). (Tag numbers: Isolation gate valve - FS-V-656, Hose angle valve - FS-V-657, Monitor 1 switch - FS-KS-6701) !
- 4. Control Building Battery & DC Switchgear Room, Elevation i 280'-6", C48-CD (North wall;. (Tag numbers: Isolation !
gate valve - FS-V-660, Hose angle valve - FS-V-661, monitor switch - FS-KS-6703)
) 5. Auxiliary Building, Elevation 328'-0", Between AJ-AL, A62; Locate hose station next to 40 B:C fire extinguisher (wall directly north of Unit Substation 2-4 5) . (Tag numbers:
Isolation gate valve - FS-V-662, Hose angle valve - FS-V-663, monitor switch - FS-KS-6704)
- 6. Auxiliary Building, Elevation 305'-0"; Between AE-AG, A60 (East wall) ; Locate hose station seven feet from end o#
wall of Unit Substation Rec = 2-21E. (Tag numbers: I Isolation gate valve - FS-V-664, hose angle valve - FS-V-665, monitor switch - FS-KS-6705)
- 7. Auxiliary Building, Elevation 305'-0"; Between A60-A61, at (North wall) ; Locate hose station 15 feet from end of east wall. (Tag numbers: Isolation gate valve - FS-V-666, hose angle wall - FS-V-667, monitor switch - FS-KS-6706)
- 8. Auxiliary Building, Elevation 280'-6"; Setween AI-AG &
A62-A63; Locate hose station between doors to rooms con-taining WDL-K-2A and WDL-K-2B. Do not locate hose station on block wall. (Tag numbers: Isolation gate valve - FS-V-668, hose angle valve - FS-V-669, monitor switch - FS-KS-6707) . e e
e Table 1 - (Continued)
) HOSE STATIONS
- 9. Control Building Area (West side), Elevation 282'-6",
CAa-C57; Locate hose station on west side of column. (Tag numbers: Isolation gate valve - FS-V-670, hose angle valve - FS-V-671, monitor switch - FS-KS-6708) l
- 10. Diesel Generator Building (West); Between DD and DE, D70, 10 feet south of edge of double door. (Tag numbers:
Isolation gate valve - FS-V-672, hose angle valve - FS-V-673, monitor switch - FS-KS-6709)
- 11. Diesel Generator Building (East); Between DD and DE, D69, 10 feet south of edge of double door. (Tag numbers:
Isolation gate valve - FS-V-674, hose angle valve - FS-V-675, monitor swicch - FS-KS-6710)
- 12. River Water Pump Hose Locate hose station six feet south of double doors of Switengear Room 2-3E. (Tag numbers:
Isolation gate valve - FS-V-676, hose angle valve - FS-V-677, monitor switch - FS-KS-6711)
- 13. River Water Pump House; Locate station near emergency exit fire door, locate station 5 feet west of edge of i
() door. (Tag numbers: Isolation gate valve - FS-V-678, hose angle valve - FS-V-679, monitor switch - FS-KS-6712) l l
- 14. Control Building, Elevation 305'-0". Cable Room C46-47 & l I
CA; Locate hose station four feet west of edge of door. I (Tag numbers: Isolation gate valve - FS-V-680, hose angle { valve - FS-V-681, monitor switch - FS-KS-6713) j l 1 l I
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, ' l l < i WET PIPE SPRINKLER SYSTEM l i i 1. East Diesel Generator Building, Elevation 230'-6", complete , l coverage (one system). (Tag numbers: Isolation gate i valve - FS-V-682, alarm check valve - FS-V-683, monitor l l switch - FS-KS-6714, pressure switch - FS-KS-6715) -l l I i 2. West Diesel Generator Building, Elevation 280'-6", complete l l coverage (one system). (Tag numbers: Isolation gate ' l valve - FS-V-684, alarm check valve - FS-KS-6715, monitor i switch - FS-KS-6716, pressure switch - FS-KS-6717) , l l l ! l ! l l I l i 1 1 l l' I { l
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***.* APR 1 B O Docket No: 50-320 i
l MEMORANDUM FOR: Milton J. Grossman, Hearing Division Director and l Chief Counsel, OELD FROM: D. B. Vassallo, Assistant Director for Light Water Reactors, Division of Project Management, NRR
SUBJECT:
BOARD NOTIFICATION - THREE MILE ISLAND-2 i l 1 recommend that the enclosed document related to the B&W revised small l break Loss-of-Coolant Accident be provided to the Three Mile Island-2 Board. It appears that the interim corrective measure will require procedures for operator action in addition to a power limitation. The plant status 1 7 is such that the proposed limiting power level of 93% has not yet been l l achieved. The staff will take the necessary action to provide for imple- l mentation of the interim corrective action in a timely manner. Long-term solutions to the problem have not been specifically addressed at this time. Q Y D. B. Vassallo, Assistant Director for Light Water Reactors Division of Project Management
Enclosure:
l As stated i cc w/ enclosure: E. Case E. Volgenau R. Boyd ! R. Mattson i H. Denton ( R. DeYoung V. Stello D. Eisenhut T. Engelhardt B. Grimes J. Stolz . K. Kniel j l 0. Parr // j S. Varga /
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7 ort sCE RE Aln90.rE445YLVAY A I ,,,1 ELLr ggt.t g ig . gyg ,yg i April 17, 1978 GQL 071h l Director of fluc J r ni Peactor llegul ation Attn: B. W. Held, Chief Operating Reneters l'rench fio. h, U. S. Nuclear Der,ulatory Cormission
, Washington, D.C. P0555 l'
Director of liuelar.: Itcactor Regulation
, Attn Mr. S . A. Vrti rs , Chief j , Light Water Haite tor n llenneh No, b i , ! U. 8. Iluelear Pr gulatory Corwilesion I,iI , i . , Washington, D.C. 20555 1l% , .
O .4" Oetitlement ' i. 4.' l Three Mile Ielon1 Thielenr Station
'1111-1 Drn 'io, Docket 50-289 l l '114I-2 DPR-73 Docket 50-370 In acconlance ult h yuur orrtl requer.t of April Ib,1970, picane ba neiviced
__ .; ,, , . . that DIN has evalunted the revised small break h0CA nn originally reported on April 1 ), lo70 purnuntit to.10 crn 71. Th e e v<s lu n t. l en r e.a u l t n . ansuming loss ef of fr i t e paver atri the most dnr.nr, lng nint.le f ailure , nre as follows: . i
, J. Annu-Ing no operat e r netton, the tere vill remnin covered nnal
} there i <t no cin.hijng te.mpernture ercursion provltled that thetr.nl power iri limite:1 to lens then:
'l111-1 < 69% of 2535 IGt
'110 -7 '. G3% o f 27 72 Vn't P. Annu
- nr <;pern ter netton in 70 niinit ori t o crons connect thr HPI d i r r h'irr <. n r.1 (ntnuning fnllure of n <11enel) open '.he liri F.' AP-d i r eb p ri.e v 91 v ei: t o n p rc;1 c '. a rr-i n c.t t hrot t.le r e v. t t r g (wht.h pre-vent r !!! }i<u:p runmt ) the core vinil'1 vernnin revereti and there vould be nn clniding temperature excursion for thernni pevers of:
A { -s y ,A v A ,- - , - - 299
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R. W. Ileid, Chiri G. A. Verga, Chlaf .. ?-
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7: M -1 100% of 2335 HWt t ,3 I
. . ' M -e 93% or 2772! Wt
}! ..- e f Nort detailed Nimlysis In4y support a higher power level. '
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- 3. When no cin21e railure is assumed or when operator action la taken
'!N'g jl.b O.";'i'i t Weo r.cci lent conditions do not er.ist (or other LOC /c conditions e ni -
T .! .' ' l 'cIlbt) r ) n:lverse citte.tfon exints or is created t.c.d for::er lIf$EiidE51hk,b,U5.7'.WAncol.3:dsulyr.u,pyA.pvc.11ng...5"' 7g ,
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With respect to i tr:n 2. above the neeenanry operator nction can be cor pletr l under adverse circunntancen within 11 minutce thus allmiing 9 ninuten for cperator re;actl; i. 'd r are in the procenn tf preparing preceilureo to accomptt these operator c.:tiens. We are revievine nt her possible long ter:a relutions for this: situnt Jon nnd vill advine yeu <ir our selecto<1 alternnt.ive.
' In tha int (erim the above men-J tiened operater nction is appropriate and acceptable.
Gincere'y.
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\ .' . . . ' APR 211978 . Docket No: 50-320 hetropolitan Edison Company ATTH: J. G. lierbein Vice President P. O. Box 542 Readino, Pennsylvania 19603 Gentlemen:
SUNECl: STEAM GEi!ERATOR QUESTIONNAIRE - THREE MILE ISLAND UNIT 2 By letter dated December 9,1977 (copy enclosed) we requested other PkR facility licensees to complete and submit a questionnaire on steam generator operating history that was enclosed. The letter stated that the request for information was approved by GA0 under a blanket clearance. Questions have been raised about the appropriateness of this request for information in light of the Federal Reports Act and about the referenced GA0 blanket clearance. These ouestions have been discussed with representatives of GA0 and it was determined that a clarifying letter should be sent to each recipient of our original letter. GA0 has agreed that this request properly .b tits under the GA0 blanket clearance for reports concerning possible generic problems and the applicable GA0 clearance number should have been R0072 rather , than R0071. ! The request f or additional information was prompted by the continuing degra-dation of tubes in all three vendors' steam generators. Such degradation is ! an important safety concern of the NRC because such tubes form part of the primary coolant pressure boundary. Several forms of degradation that have been observed in stean generators in recent months have included the wastage of tubes at Palisades end other facilities, stress corrosion at Ginna and other facilities, vibration cracking and " dinging" of tubes at the Oconee (MW) facilities, antivibration bar fretting at San Onofre, and " denting" of tubes and associated suppert plate "hourglassing" and cracking at Surry, Turkey Point and about 15 ether CE and W facilities. These events have prompted the HRC to issue safety Orders! It is this need for important safety information that has dictaterf this request for additional infonnation. Our original letter to other lir.ensees acknowledged that selected portions of the information being requested aay already be available to the NRC, but not in a convenient format which is readily accessible. We therefore requested that they assist us by returning a single completed copy of the enclosed
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lietropolitan Edison Co npany APR 21 1978 questionnaire. Ve would like to clarify that an acceptaDie response to any item in the questionnaire would be to provide specific reference to any infonnation previously submitted to the NRC, by an original response, or any conbination thereof, whichever and for whatever reasons you elect to use. Our original letter further requested that recipients submit any changes i or additions to their initial submittal to reflect the future operating experience with their steam generators. This would enable us to maintain the information current, which, as we stated, we will periodically publish and send copies to all participants. As we indicated, this would enable i the flRC and others to draw f rom the operating experience of the entire ' nuclear industry on an ongoing basis when making safety and other decisions i concerning steam generators in PLR plants. We are planning to prepare a j submission to GA0 for clearance of a request for reporting infonnation ! regarding changes or additions to your initial submittal under this request. l l Based on the above, we request that you assist us by returning a single completed copy of the enclosed questionnaire to the Director of Nuclear Reactor Regulation, U. S. Iluclear Regulatory Commission, Washington, D. C. 20555, within 60 days of receipt of this letter. Please include any comments , or suggestions for irnproving this information system which you may have. This request for generic information is approved by GA0 under a blanket , clearance Number R0072. This clearance expires December 31, 1980. l Q??4 Sincerely, St n . Varga C Light Water Reacto Branch No. 4 Division of Project Management
Enclosures:
1
- 1. Ltr. dated 12/9/77 to PkR Licensees
- 2. Questionnaire cc w/ enclosures:
See Page 3
l- : l .____ _.._. t I l 11etropolitan Cdison Company APR 211978 i l CCs: George F. Trowbridge, Esq. Shaw, Pittman, Potts & Trowbridge 1800 M Street,tl. W. Washington, D. C. 20036 ilr. I . R. Pinf rock i Jersey Central Power and Light Company . l
!!adison Avenue at Punch Dowl Road .
Florristown, tiew Jersey 07960 Mr. R. Conrad Pennsylvania Clectric Company 1007 Broad Street Jchnstown, Pennsylvania 15907 t l Chauncey R. Kerford, Esq. Chairman York Correnittee for a i Safe Cnviron w nt I 433 Orlando Drive l State College, Pennsylvania 16801 . Mr. Richard W. lle'eard . Project Manager l
- GPU Service Corooration l 260 Cherry 11111_ Road l Parsippany, New Jarsey 07054 l 1
1 fir. T. Gary Broughton i Safety and Licensing Manager GPU Service Corporation 260 Cherry flill Road Parsippany, llew Jersey 07054 I i I i . l l . l l 1 I
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METHOPOLIIAN EDISON COMPANY j - c
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POST OFFICE BOX 542 READING, PENNSYLVANIA 19603 TELEPHONE 2154929 3E0t 21 May 5, 1978 GQL 085h
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Yi ; Director of Nuclear Reactor Regulation O l l l Attn: S. A. Varga, Chief , 7 :n " Light Water Reactors Branch '.;o. h " U. S. Nuclear Regulatory Com=ission ! Washington, D. C. 20555
Dear Sir:
i Three Mile Island Nuclear Station, Unit 2 (TMI-2) l Operating License No. DPR-73 l l Docket No. 50-320 l l Small Break LOCA i l Enclosed please find the results of Babcock and Wilecx's (B&W) mest recent calculations concerning a Small 3reak LOCA at TMI, ( Analysis of Small Breaks l in the Reactor Coolant Pump Discharge Piping for the B&W Lovered Loct 177 FA l Plants, May 1,1978) as well as the analysis presented to the NRC staff by l- B&W at a meeting on April 25, 1978 (Analysis of Small Breaks in the Reactor l Coolant Pump Lischarge Piting for the B&W Lovered Loot 177 FA Plants). Met-Ed !. and GPUSC have reviewed the enclosed analyses and concur with the 3&W finding that full compliance with 10 CFR 50.h6 and Appendix K to 10 CFR 50 is clearly demonstrated for operation at power levels belov 2568 Mv(t) (approximately 925 power for T>E-2 ) . , Recent conversations with B&W have indicated that results of additional ! calculations for power levels up to 2772 Mv(t) vill be available to the NRC ; by approximately June 1, 1973. It is believed that these results vill mere clearly demonstrate complete compliance with 10 CFR 50.h6 and 10 CFR 50 Appen-dix K at power levels up to 2772 Mv('t). Maintenance operaticas at TE-2 are progressing well, and Mode 2 entry (criticality) is expected to be made en May lh, 1978. It is then expected that the pcver level l vill be gradually increased; however, 2568 Mv(t) (925 of ful' power fcr T.c-2) j for which cenpliance with 10 CTR 50.h6 has been demonstrated is not to be achieve 3 prior to June 8,1976. Met-Ed, therefore, propeses to submit (prior to exceeding the 2568 Mv(t) power level) correspondence which, based :n the Sa'i ca'culatiens nov being performed, demonstrates ecmpliance with 10 CFR 50.h6 and 10 CFR 50 Ap- , pendix K for power levels up to 2772 Mv(t) (100", '"C-2 full pcVerl . l i Met-Ed has revised the appropriate !!C-2 procedures (Emergency h ccedre 2202-13, ! Less of Reacter Ccolant/ Reactor Coolant Pressure end Cteratine Freced"~ 0~0h-l.2, ! Makeup and Furification Demineralicatien) as foll:vs: . ! u ', , e h bb (,) s . 'Q.,'! c f i
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1 i . Mr. S. A. Varga - May 5, 1978 1 GC.L 085h
- 1. Energency Procedure 2202-1.3 - revised to detail the operator response (see below).
- 2. Operating Procedure 210h-1.2 - revised to permit operations with one of the makeup pump discharge cross-connect valves open and I the other one closed.
There vill be two (2) operators designated to respond to a small break LOCA, (1) Control Room LOCA Operator, stationed in the control room and trained to reccgnize the symptems and respond to a small break LOCA and (2) Auxiliary Building LOCA Operator, stationed on the primary side of the plant, and i trained to respond to a small break LCCA. The Control Rect LOCA Operator vill, within two (2) minutes after the event, analyze his indications and determine if there is a loss of cffsite power cencurrent with a diesel or l
=akeup pump failure and a small break LOCA. In the event of that occurrence, l by time T = 2 minutes, the Control Room LCCA Operator vill direct the Auxiliary Building LOCA Operator to proceed to the takeup pu=p discharge cross connect valve and open it. The Control Room LOCA Operator vill then proceed to the HPI valves on the affected train. The Auxiliary Building LOCA Operator will take, ,
at a maximum,15 minutes to arrive at the cross connect valve and at time T = l 3.5 minutes, be opening the cross connect valve. Opening the cross connect i valve vill commence within 1.5 minutes (T = 5 0 minutes) and at time T = 10.0 J minutes, the cross connect valve vill be fully open. As described above, when the Control Room LOCA Operator has directed the Auxiliary. / Building LOCA Operator to take the required action, he vill then proceed to the HPI valves on the affected train, arriving in 2.5 minutes. I= mediately upon ar-rival, at the HPI valves, time T = h.5 minutes, the Control Room LOCA Operator vill establish communication on the head set with the Control Room and begin to open the affected train HPI valves and will achieve minimum flov vithin 0.5 min-utes (T = 5 0 minutes). The EPI valves will be opened manually to obtain 125 spm flow per leg concurrent with balancing flov to 125 gpm in the unaffected leg electrically from the Control Foom. This balancing evolution vill take less than 5 0 minutes and vill be ecepleted by time T = 10 minutes. Prior to the balancing evolution, the Control Room CR0 shall verify that the normal makeup valve is closed. These procedure revisions have been fully implemented. Met-Ed review committees have reviewed these precedure revisiens and have determined that (1) there is no increase in the probability of occurrence or the consequences of an accident er calfunction cf equipment important to safety previously evaluated in the safety analysis report in that the procedure revi-sions mitigate the censequences of the accident previcusly analyzed; (2) no pos-sibility for an accident or malfuncticn of a different type that any evaluated in the safety analysis report is created in that the tajor concern, i.e., pump runout, vill not occur under the operator acticn specified above; and (3) the margin of safety as defined in the basis for any technical specificatien is ne reduced in that 10 CFR 50.ho acceptance criteria is not exceeded. In addition, it has been determined that utilization of these procedures uncar any accident cenditien requiring cperation of the epi pumps vill not lead tv
0 :jr. S'. A. Yarga .\ h y 5 , 1 9 7 9 Gql 085L degradation of pu=p performance during any part cf the transient. Performance of these procedures provides assurance that the total EFI flow, whether through two legs or four legs, will not exceed 550 gpt. Further assurance that pump runout will not occur results from B&W's indicating that pu=p runcut vill not occur as long as the back pressure is greater than the pressure equivalent to 1500 ft. of water (approximately 650 psi). Fcr the largest break analyced (0.17 ft ),2 RCS pressure reaches about 650 psia in about 400 seconds, at which time the EPI valves vould already be into the balancing evolution. Conserva-tive calculations based on FSAR and Technical Specificatien data have been per-l formed and indicate that adequate E?SE exists fer at least 7.5 hours while i taking suction from the 3WST. ; Each shift was briefed on the censtraints of the license and the small break LOCA procedure requirements. An Operations Crder is being vritten to require each Operations Depart:ent person to signify understanding of the procedure changes and =anning requirements. Also, each Operations Department person is l to physically locate all equipment required to be operated in accordance with i the procedure changes. The Operations Order vill further require one person on j each shift (who is free to respond to the postulated accident) to be stationed l in the Control Room at all times, and one person on each shift to be stationed i on the primary side of the plant at all times to carry out the required action specified in the procedure changes. A sheet will be attached to the Control Room Log Sheet showing who are the two individuah assigned the responsibilities for carrying out the actions indicated in the procedure changes. Each shift vill be rebriefed at least once per month of the actions required l in the procedures, l i TMI has performed drills to verify that the assumed operator response time is achievable and within the analysis assumptions. l All drills performed to date have shown adequate response (to the point of l l unseating the cross-connect, and HPI discharge valves) in less than 5 minutes. l Met-Ed vill submit a Technical Specification Change Request covering these procedures as soon as possible. ; Met-Ed vill subtit a proposal for a permanent solution by August 5,1973. Should additional analyses be perforted, Met-Ed will make their results available to the NRC, as socn after their completien as possible. Sincerely,
,.!. [ ,
i 1//
' J. G. Eerbein I
/ e', Vice President-Generation
\ b/ , i JGE:RA1:cjg i ec: Harley Silver (:!RC'. l Ene10sure
Babcock &Wilcox %er cemeceuen croun 7735 Old Georgetown Road, Bethesda, Md. 20014 1 Tefephone:(301) 951-0202 l 1
- 1
~
I i April 25, 1978 , 1 l Mr. Robert L. Baer Chief, Reactor Safety Sranch 1 Division of Operating Reactors ; Office of Nuclear Reactor Pagulation j U. S. Nuclear Regulatory Comission Washington, D. C. 20555 i
Dear Mr. Baer:
a l Attached is a report describing the methods used and the i results obtained from B&W's recent studies of small breaks ; in the reactor coolant pump discharge piping for the B&W lowered loop 177 - fuel assembly plants. This report shows l that at power levels up to at least 2558 MWt, full compli-ance with 10 CTR 50.46 is achieved. B6W is currently. extending this study to cover power levels u,7 to 2772 !!Wt and will supplemen the attached information as soon as the ; studies are complete. We request an expeditious review of J this information by the staff. l Very truly yours,
\
l p y W I' _ 2 L-
/
James H. Taylcr f , l'anager, Licensing l 3 i JHT/bh l i i j h ft$r7 iu~ n< ~7 A O W' tt i
d l 1 1 1 i < l ANALYSIS OF SMALL BREAKS IN THE PIACTOR COOLANT PUMP DISCHARGE PIPING l FOR THE B&W LOWERED LOOP 177 FA PLANTS i APRIL 24, 1978 n% a n 1pm ? cf t sv i- n--
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i s
- a. The reactor is operating at 102% of the steady-state power level of 2568 MWt. For breaks greater than 0.1 f t, the analysis utilized a j power level of 102% of 2772 MWe. )
- b. The leak occurs instantaneously, and a discharge coefficient of 1.0 e l
1s used for the entire analysis. Bernoulli's equation was used for ' the subcooled portion of the transient, while Moody's correlation was ... l used in the two-phase portion.
- c. No offsite power is available.
- d. The reactor trips on low pressure at 1900 psia.
- e. The safety rods begin entering the core af ter a 0.5 second delay from the time the reactor trip signal is reached.
- f. The RC pumps trip and coast down coincident with reactor trip.
- g. One complete train of the emergency safeguards system fails to operate, leaving two CFTs and only one HPI and one LPI system available for pumped injection to mitigate the consequences of a cold leg break.
- h. The auxiliary feedwater (FW) system is assumed to be available during the transient. Its main function is to remove heat from the upper half of the steam generator during the initial stages of the transient.
When the secondary side of the steam generator becomes a source of l heat to the pridary system, the assumption of auxiliary FW maximizes the energy that must be relieved.
- 1. ESFAS signal error band is considered in the analysis to signal the actuation of the HPI system.
- j. The peak linear heat generation rate in the hot pin is the maximum allowed by the Technical Specifications at the 10.5 ft level. .
- k. Operator action is taken to increase the HPI flows to the intact cold legs at 10 minutes following the ECCS initiation signal. This assump-tion is explained more fully below and in section 3.
As most of the breaks evaluated in this spectrum showed core uncovery, temperature calculations were necessary. Once core uncovery occurs a spatial swell distribution an . lysis is necessary to assure that only the core cevered by nincure is included in the swell level. B&W user " c FO A:1 code. The code was utilized under the snee assu=ptions as descr J above with the following additions:
_ . ~ . . . . . _ _ . . _ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . . . _ _ _ _ . . _ J e l was simulated in our present CRAFT code as a step function at 650 seconds 1 i (600 seconds for action, 50 seconds for ECCS signal) . This is illustrated l in Figure 7.
- 2.3. Break Spectrum and Results ~
1 All evaluations reported in this analysis cssume the high pressure injection , l i 1 parformances as described in section 2.2. Breaks of 0.3, 0.2, 0.15, 0.1, ) 2 2 0.07, and 0.04 ft were evaluated. The evaluation of a 0.5 ft break was f a reported in BAW-10103A, Rev 3, and shows complete core covery at all times i 2 j cnd thus no temperature excursion. The 0.5 ft break results are independent i of HPI flow and remain valid. j Figure 2 shows the RCS pressure transient for each break. As shown, each ac-3 2 i cident initiates CFT flow within 2000 seconds except for the 0.04 ft break. Figure 3 shows (CRAFT) mixture height as a function of time for each break of the spectrum. As can be seen, breaks. of approximately 0.3 f t and larger than approximately 0.04 f't uncover part of the core. Various uncovery levels
~
and times are observed but all trends are consistent throughout the spectrum. The 0.04.ft 2break achieves a match up of effective ECCS (the RPI. injected into the intac't cold legs) with the core d'ecay heat and the RCS metal heat at 2500 seconds. Af ter 2500 seconds the mixture level will rise in the core due to excess HPI injection. As the 0.04 ft break has a level of 14 feet at this tice the core never uncovers and no . temperature excursion occurs. For breaks smaller than 0.04, the match up will occur at approximately the same time and the core mixture levels will drop slower; thus, for all smaller breaks tha core will remain covered. Figure 4 shows the time duration of uncovery for four core elevations as a function of break size. These'results are from CRAFT. As can be seen, the naxi=u= degree of uncovery and the maximum tire of uncovery occur for the 0.15 ft break and is the worst case break. This break can thus be identi-2 ficd as the worst case. A siollar uncovery occurs for the 0.1 and 0.07 ft b re C,s . The 0.07, 0.10, and 0.15 ft breaks rere analyzed for temperature response. The results are shown in Figure 5 and are well within the criteria of 10 CFR 50.46. They provide positive assurance that all breaks of the spec-tre are within acceptance criteria.
- 2. If no flow in one train:
-- open pump header cross-connect valves
-- check EPI valve position and open if closed ,-
- 3. Secure flow through normal makeup line li flow is indicated
- 4. Throttle HPI valves as required to balance flow and meet run out limits The above actions initiated at five minutes and completed within 15 minutes subsequent to the ESFAS actuation ensures adequate HPI flow for accident miti-gation. In the analysis, credit is taken for the HPI flow as the EPI injec-tion valves are opened. Figure 7 shows the calculated EPI flow for a typical l
plant as a function of time for a 10 minute valve opening. As shown in Figure 7, the majority of the HPI pump capacity would be delivered with a partial valve opening. For the small break analysis, a linear flow versus valve posi-tion response was simulated by a step function increase,10 minutes af ter ES7AS actuation. I
- 4. Evaluation of Other B&W Supplied Plants
- a. Davis-Besse -- The DB-1, 2 and 3 Plants have been analyzed for a spectrum of small breaks at the RCP discharge in accordance with an approved small break evaluation model. This analysis is reported in BAW-10075A, Rev 1, March 1976. In addition, the Davis-Besse 1, 2, and 3 units have a split high pressure injection and makeup system design. The Davis-Besse EPI pumps, therefore, have considerably higher capacity at the system pres-sures experienced.
- b. 205 and 145 FA - These plants have been analyzed for a spectrum of small breaks at the RCP discharge in accordance with an approved scall break evaluation model. These analyses are reported in BAW-10074A, Rev 1, and BAU-10062A, Rev 1, March 1976. In addition, the 205 and 145 FA HP1 sys-tens contain cross connects between the two HPI trains downstream of the liPI injection valves. These cross connects effectively achieve the same flow split as the operator action assuned in the current 177 FA lowered loop analysis and the flow split is achieved when the HPI pump is started.
_g_
Fig'ure -1. CRAFT 2 Noding Diagram for Small Break lO m _ 1@1
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FIGURE 5. PEAK CLADDING TEMPERATURE q. e '. 2000 - ,.,
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. . i Figure 4 DURATION OF UNC0VERY FOR THREE CORE LEVELS (11.5 FT,11.0 FT,10.5 FT) -
- i (NOTE 10.0 FT DOES NOT UNCOVER) 600 y, ALL AT 2568 M#t t
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i TABLE 1 PEAK CLADDING TEMPERATURE VERSUS BREAK SIZE (All at 2568 MWt) ) l l l Peak Maximum Time of : Break Cladding Local MW Peak ! Size Temperature Reaction Temperature . (Ft2) (oF) (%) (sec.) i 0.04 Core Stays Covered With No Temperature Excursion ! 0.07 1320 .73% 1600 0.1 1440 1.68% 1080 0.13 1551 1.72% 820 0.15 1455 1.67% 740 l l 0.17 1248 0.72% 650 t e a e
. i 1.ocal metal-wstar reaction is shown in Table 1. Tha highest valun is 1.72% ]
2 for the 0.13 ft break. This value is well below the local oxidation limit ] for the large breaks utilized in BAW-10103 for the whole-core metal-water re- i action calculation. Thus, the whole-core metal-water reaction results given ! in section 8 of BAW-10103 is conservative for small breaks. The degree of clad damage is bounded by the large break results which produce higher clad temperatures. Thus, all criteria of 10 CFR 50.46 are met. This analysis is conservative for many reasons as detailed in the writeup and meets all evalu- , ation criteria. This analysis shows that all 177 lowered loop plants meet the criteria of 10 CFR 50.46 if operated at or below 2568 MWe power and in conjunction with the specified operator action. ,
- 3. Operator Action The ECCS analysis used as a basis for this report assumes that the operative HPI train (one train is lost due to a single active failure) provides emer- ,
gency core cooling water to the RC loop containin; the break. It is conser-vatively assumed that the ' break is on the lower portion of the reactor cool-
' ant pump discharge piping resulting in the total loss to the system of 50%
of the available HPI flow. Acceptable mitigation of the accident requires j more than the 50 % of this flow from one HPI pump. If, following the LOCA, it is assumed that one train of HPI does not start it is necessary to take operator action to achieve a flow split wherein no more than 30% of the re-maining pump's flow goes into the cold leg containing the break. The follow- ' j ing is a description of the action required for a typical plant.
- 1. Upon ESFAS signal check for flow through both HPI trains.
- I
- 2. If no flow in one train:
- open pump header cross-connect valves
- check EPI valve position and open '
if closed
- 3. Secure flow through normal makeup line if flow is indicated
- 4. Throttle HPI valves as required to balance flow and ceet run out limits
) . 1 J
- 1. The powar shapa shown in Figure 6 was used but implemented wf th a radial peaking factor of 1.0. This epresents the average channel condition which is appropriate for use in swell level calculations.
- 2. Steam production due to heat from the primary metal, core and lower plenum flashing, was conservatively underpredicted. Although the CRAFT model accurately predicted these effects, full credit was not included in the FOAM simulation as a conservative computational con-venience. This simulation, therefore, underpredicts both the swell level and the steaming rate. Consequently, more core uncovery and lower coolant flow are used in the heat-up evaluation.
The heat-up calculation was performed using the THETA code in the manner described in section 5 of BAW-10104. The following additional assumptions are utilized in the THETA evaluation:
- 1. The power shape of Figure 5 was used with a radial power factor of 1.8.
This -==v4=4 *es steam superheating and sets the peak local power at 10.5 ft at the technical specification LOCA limit.
- 2. Coolant flow and mixture level were taken directly from the FOAM calcu-lations. (
1
- 3. End of life pin pressures were used to conservatively predict the inci-dence of fuel pin rupture.
2.2. High Pressure Injection System Performance . The previous arrangement-of the HPI system allowed for one pump to inject into the reactor coolant system (RCS) at two locations. As one injection point could be in the region of the break, 50% of the one HPI flow could fail to penetrate the reactor vessel. This flow would, therefore, not be avail-able to provide core cooling. The proposed operator action, section 3, will provide four points of penetration of the RCS. Iherefore, only 25% of the HPI flow would be lost. Since the flow from one HPI pump will now be distributed to four injection points and to assure conservatism in allowing for injection line loss dif- ) ferences~, this analysis assumes 30% of the HPI is injected into the broken cold leg. The implemented action starts at 5 minutes after an ECCS signal and is concluded 15 minutes after the signal. The resultant HPI flow can be conservatively represented as a linear ramp from 5 to 15 minutes. This ramp 1
- 1. Introductinn 3 On April 14, 1978, B&W reported that previous small break analyses had not been based on the worst break location. This report indicated that the worst case break had now been determined to be at the reactor coolant pump discharge. A spectrum of small breaks has been examined for the B&W 177-FA lowered loop plants using the small break evaluation model described in .
BAW-10104, Rev 3, "B&W's ECCS Evaluation Model." These results show that it is necessary to use operator action during the early stages of the pos-tulated accident, to effectively mitigate the accident consequances and meet the criteria of 10 CFR 50.46. Operator action is used to achieve suf-ficient and balanced' flow through all four HPI injection lines. This re- . Port shows that operation up to at least 2568 MWt is possible within the criteria of 10 CFR 50.46 and Appendix K.
- 2. Evaluation 2.1. Method of Analysis The analysis method used for this evaluation is that described in Chapter 5 of BAW-10104, Rev 3, "B&W's ECCS Evaluation Model." Specifically, the model, except for break size, break location, and core power, is the same as utilized in Appendix C of BAW-10103A, Rev 3, "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS." The analysis uses the CRAFT 2 code to develop the history of the reactor coolant system hydrodynamics. The CRAFT model uses 19 nodes to simulate the reactor coolant system, two nodes for the secondary system, and one node for the reactor building. A schematic diagram of the model is shown in Figure 1 along with the node descriptions.
Control volumes (nodes) in and around the vessel are all connected by a Pair of flow paths to permit counter-current flow. The break is' assumed to be located at the bottom'of the cold leg piping between the reactor coolant pump discharge and the reactor vessel. The Wilson, Grenda and Patterson average bubble rise model is used for all nodes. Within the , core region, however, a multiplier of 2.38 is applied to the calculated bubble rise velocity. Appendix F of BAW-10104 demonstrates that a multiplier of 2.38 in CRAFT 2 gives a mixture height within +2% of that predicted by FOAM. Thus, no FOAM analysis will be needed if the CRAFT 2 mixture level remains above the core by 2% of the active length. 1 The follcuing assumptions are made for conditions and system responses during the accident:
s / a
/ >;::::,, - -m / .j METROPOLITAN EDISON COMPANY m:< a u ir v . . n.. . = ::: .:a-a POST OFFICE BOX 542 READING, PENNSY LVANI A 19603 TELEPHONE 215 - 929 3601 )
l
- l 4
4 May 11, 1978 _.J l GQL 0915 G l l Director of Nuclear Reactor Regulation _ l Attn: S. A. Varga .' r5.! Light Water Reactors Branch No. h : U. S. Nuclear Regulatory Commission ')
- C Washington, D. C. 20555 l
Dear Sir:
1 Three Mile Island Nuclear Station Unit 2 (TMI-2) l Docket No. 50-320 l Operating License No. DPR-73 l The occurrence, at Crystal River 3, of two separated Burnable Poison Rod ; Assemblies (BPRA's) has raised the concern that a similar incident ; might occur at Three Mile Island, Unit 2. Although such an event is not considered likely, based upon the satisfactory performance of other B&W operating reactors, Metropolitan Edison dee=s it prudent to take certain precautionary measures to provide further assurance that the BPRA's vill remain in place. Of the various options available, we have determined that the best course of action is the installation of positive retention 1 devices, which were recommended by B&W. These retainers have been de- l signed and are undergoing test and evaluation at B&W. ' Currently, it is our intention to install the retainers on all BPRA's following completion of startup and acceptance testing. As discussed belov ve are confident that the plant can be operated for up to 75 full flow days prior to installation of the retainers. TMI-2 NSS operation to date has been with three primary coolant pumps in service. Later portions of the initial startup phase and full power
. operation vill be conducted with all four coolant pumps in service. Crystal River 3 operated for about 300 days with four pump flow before the first indication of BPRA separation occurred. Based on the performance of CR-3 l
and other B&W 177 Mark Bh Fuel Assembly plants and on vear measurements of the fuel assembly BPRA holddown latches at Crystal River 3, Arkansas Nuclear One-1, Oconee-2, it is conservatively estimated that TMI-2 can reliably operate for up to 75 days of the full four pu=p flow.
>l :
t 50
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- Director of Nuclear Reactor Regulation May 11, 1978 GQL 0915 Besults to date of the B&W investigation of the CR-3 event indicate that the separation of BPRA's may be due to a long term wear phenomenon causing separation of the BPRA holddown latch. Coolant flow and the resultant net hydraulic lift compared with the wet weight of a BPRA appears to be a primary factor in the holddown latch wear rate. Latch hardness is also a significant factor. The 68 BPRA's and holddown latches in TMI-2 are of the same design used in all B&W 177 FA reactors.
Analysis of the TMI-2 BPRA hydraulic lift force for four pump flov in-dicates less nominal lift than at Crystal River 3 The holddown latch assembly minimum hardness on TMI-2 fuel assemblies in also equal to or greater than the hardness of the Crystal River 3 holddown latch assemblies which experienced the highest wear. Thus, TMI-2 can be expected to ex-l perience a lower wear rate than Crystal River 3 However, to account for ! other undefined factors which may influence wear rate, a factor of h has been applied to the highest wear rate observed at Crystal River. On this ) basis, an allowable limit of 75 days of TMI-2 four pump operation has been established. l l Wear data from Oconee 2 and ANO-1 for fuel assemblies which operated for l as long as 600 full flow days lend confidence that the use of Crystal River vear data coupled <with a safety factor of h is conservative for TMI-2. Davis Besse with higher calculated lift and comparable minimum holddown assembly hardness has operated without incident for gretter than 150 full ! flow days. Rancho Seco, also with calculated higher lifts but with much higher minimum holddown latch assembly hardness operated for greater than 500 days without incident. Oconee 3 and TMI-1, with calculated lift forces in the same range but slightly lower than THE-2, both operated for greater i than 500 full flow days without incident. i l Operation with three pumps precludes BPRA net lift with a very large margin thus avoiding conditions under which wear can occur. To date, all operation in TMI-2 has been with 3 coolant pumps in service. TMI-2 vill not be operated past 75 days of accu =ulated full flow operation, prior to retainer installation, without further justification. The NRC vill be informed of the results of any investigations which may change the basis for the allowable period of four pump operation. Thus, based on the considerations described above, there is a very low prob- ' ability of a BPRA separating from a fuel assembly in TMI-2 before retainers are installed. These retainers vill insure BPRA locking for the remainder of first cycle operation. In the very unlikely event that a BPRA =ay become separated during plant operation the consequences to the core are within the bounds of the analysis addressed in the FSAR. Depending upcn its locatien within the core a separated BPRA vill have a varying impact upon assembly power peaking. With a significant increase in power peaking the event would be detected by the
. Director of Nuclear Reactor Regulation May 11, 1978 GQL 0915 tilt alarm or power distribution monitoring and appropriate corrective action vould be taken. Lesser power increases vould be within the allovable peaking limits considered in the Technical Specifications.
In addition, the change in by-pass flow as a result of BPRA removal, is negligible. The consequences of a BPRA separating from the core are bounded by the results of the Ejected Rod Accident analyzed in the FSAR. The reactivity worth of a single BPRA is only 30 to h0% the vorth of a control rod and is less than the maximum e'jected rod worth of 0.65% AK/K used in the FSAR. The consequence of a stuck control rod assembly (CRA) is a normal design 1 consideration for calculating shutdevn margin. All FSAR accidents are analyzed with the reactivity effects of the most reactive control rod , stuck out of the core. The effect of a separated BPRA would be less than a stuck control rod for the same incident. Based upon the above discussion it is concluded that there is a very lov l probability of a BPRA separating frem the core during the linited period { of four pump operation; also, any consequences to the core from such a separation are bounded analyses contained in the TMI-2 FSAR. l l
, neerely l I
l J. G. Herbein Vice President-Generation
- JGH
- RAL:dkf cc: H. Silver
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- ^' & 27H -- ., '=- n, -m l
! METROPOLITAN EDISON COMPANY : - -n - i POST OFFICE BOX 542 READING. PENNSYLVANI A 19603 TELEPHONE 215 - 929 3601 l May 19, 1978 GQL 0970 l l Director of Nuclear Reactor Regulations Attn: Mr. S. A. Varga,_ Chief l Light Water Reactor Branch h l U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Sir:
Three Mile Island Nuclear Station Unit 2 (TMI-2) Docket No. 50-320 Operating License No. DPR-73 Pseudo Ejected Rod Test Deletion i B&W has recc= mended that Met-Ed delete the pseudo ejected rod worth test j performed at h07. power from the TMI-2 Startup and Test Program. B&W's J recommendation is based on the fact that the predicted ejected rod worth j is extremely small, and the prediction has been verified at other plants. We agree with B&W's recommendation. Therefore, we propose to delete the ! l at power ejected rod test in that, performance of the test would generate unnecessary radwaste and consume up to 2h hours of time that could be used more productively. In discussions with members of your staff, several questions were asked concerning deletion of this test. Answers to these questions are attached. l 1 Deletion of the ejected rod worth measurement at power is justified, because: l l (1) the predicted rod worth is substantially less than the safety analysis limits; (2) actual measurement of zero power ejected rod worth was less than i the safety analysis limit; and (3) actual =easurements on similar cores at
- other plants (Davis Besse and Rancho Seco) have confirmed the extremely lov ejected rod worth predictions and the validity of the prediction methods.
l Based en the above, we request your concurrence in deleting the ejected rod verth measurements at h0% power. Sincerely, l
/ J. 3. Herbein
~GH:RAL::as Vice President-Generatien A- achmen, J !';
e 0, -f$$ 4 'kgg b \
Answers to NRC Ouestions Concerning Rod Worth Question 1: Provide a comparison of measured / predicted rod vorth for Groups 1-h, 5, 6, 7, and 8 and ejected rod worth at zero power for TMI-2. Answer: ZERO POWER ROD WORTH Rod / Group Predicted" (%6k/k) Measured (%Ak/k) Group 1-7 10.22 10.0h Group 5 2.22 1.85 (inserted from 79% to 0% vithdravn) Group 6 and 7 3.36 3.165 Group 8 (APSR) 0.k1 0.385 Rod H - 1h 0.87 (Gr. 5 8 h9%) 0.611 (Gr. 5 8 50%)
- Physics test manual values were predicted using 2 zone 20 PDQ, described in BAW-10116A, with all full length rods being inserted from fully withdrawn to fully inserted. Group 8 was inserted from fully withdrawn to its null position.
Question 2: Are the Rancho Seco and TMI-2 cores the same and vere they analyzed identically? Why do the TMI-2/ Rancho Seco predictions differ? Answer: The TMI-2 and Rancho Seco(and Davis Besse) cores are essentially the same. The major difference between the two cores is the presence of the Gadolinia demonstration burnable poison rods in TMI-2. These poison rods, however, have only a minor influence on rod worths. The predictive analyses for the Rancho Seco (and Davis Besse) were performed using 3D - PDQ (1-zone) while TMI-2 analyzed using 3D FLAME (1-zone). The predicted and =easured values for ejected rod worth are given in Table 1 belev.
"'he major difference between the predictions for TMI-2 and those of Davis Besse and Rancho Seco are believed to be a result of the analyses being performed at differing Grcup 6 and 7 withdrawals (75% for Rancho Seco* and 83.3% for TMI-2) rather than the influence of the demonstration poison rods or predictive code utilized. This reasoning is demonstrated in Table 2 as calculated by a pin-by- ;
pin 2D FDQ (see 3AW 10116A). Also, those four assemblies with Gadolinia are l enriched to 1.30 v/o instead of the Batch 1,198 v/o.
*and 75% for Davis 3 esse
. . _ _ . _ _ . - . _ _ _ _ _ . _ - . . . _ . . - . _ - . . ~ . . . _ . _ _ _ _ _ _ _ . _ . _ . . - . _ _ . _ _ . . . _ . . _ . . . . - . _ - . . ~ . _ ______. _.
. TABLE 1: EJECTED ROD WORTH AT POWER 4
Prediction Code Measurement ** (%Ak/k) 4 Rancho Seco 0.035 PDQ 0.0h7 Davis Besse 0.035 PDQ 0.021B TMI-2 0.01 FLAME - i **The acceptance criteria is < 0.65% Ak/k which is more than an order of magnitude greater than predicted or measured. 4 TABLE 2: TMI-2 1 GROUP ROD WORTHS, WITH AITD WITHOUT GADOLINIA POISON RODS Group TMI-2, No Gd. TMI-2, With Gd. 1-7 9 73 9.6h 5 2.08 1.99 (h of 12 CRA's inserted into Gd. bearing ^ assemblies) 6 1.88 1.89 T 1.h8 1.h5 I 1 1 i l 1 4 1 1 _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ ,e ,-.
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l 127M l . G"C n , _ u, METROPOLITAN EDISON COMPANY . :4 - =- ._ . . . POST OFFICE BOX 542 READING, PENNSYLVANI A 19603 TELEPHONE 21E - 929 3601 l l May 25, 1978 1 GQL 0987 - Oh Director of Nuclear Reactor Regulations - u,4 /, IO78 9 l Attn: Mr. S. A. Varga, Chief l l l Operating Reacters Branch No. k U. S. Nuclear Regulatory Cc= mission g
,+
l Washington, D. C. 20555
- l2 gg A9 ,
Dear Sir:
1 Three Mile Island Nuclear Station Unit 2 (TMI-2) Operating License No. DPR-73 Docket No. 50-320 Small Break LOCA l It has come to our attention, through conversations with Mr. Harley i Silver of your staff, that our submittal of May 5,1978, did not ade-l quately address our intentions for power escalation at TMI-2. Please I be assured that Met-Ed vill not escalate power beyond 2568 Mv(t) at TMI-2 until the following events have transpired:
- 1) Babcock and Wilcox (B&W) couplete their analysis of the results of a hall Break LOCA for operation at 2772 Mv(t).
i
- 2) Met-Ed has reviewed and concurred with these analyses, and has l
submitted them to the NRC. i
- 3) The NRC has accepted the validity of the B&W Analysis (as sub-mitted by Met-Ed) for a Small Break LOCA at power levels of 2772 Mw(t).
Sincerely, ' 0 ,I - t
-( _,. O
'fJ. G. Herbein
' 71ca President-Generation
/
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--- *>> 'ay Silver D:RC) '} d '
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ObCOCh&WilCOX JU.V 1 2 W 3 eme ceneracon croup 4 .
] P.O. Box 1260, Lynchburg.Va. 24505 Telephone:(804)'384-s111 June 7, 1973 ,
l Mr. Steven A. Varga, Chief ;
. Light Water Reactors Branch 4 ,
l Division of Project Management ! Office of Nuclear Reacter Regulation !
. U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 ;
i
Dear Mr. Varga:
Earlier today, Messrs. Tokar and Meyer of the NRC asked I several questions of B5N relating to BAN-1496, "BPRA Retainer Design Report", which was forwarded to you by ny letter of June 2, 1978. The cuestions were directed pricarily towards l use of the retainer fer holddown of modified orifice rod ( assemblies (MORA) . This letter is being provided to formalize B4W's responses to these questions so that they may be used in - support of licensing activities associated witn retainer use.
~
The MORA used with primary neutron sources can be visualized by referring to Figure 3-la of BAW-1496. The orifice rod assembly (ORA) spider is the same shape at that shown in Figure 3-la. The MORA is produced fron a sttndard ' ORA by removing the four Y-shaped arms of the spider tnd the cight orifice rods attached to these arns. In additica, the four orifice rods on the inner rod circle created by the remaining straight spider arms (on the fuel assembly diagonals in Figure 3-la) are removed slightly below the spider. A short rod portion remains below the spider arm and the nut above the spider remai'ns. In essence, the MORA is now a spider arrangement with four straight arms and four unmodified orifice rods at the outermost location on these arms. The nuts at the inner location on the straight arms ar'e retained because they " act as locating fixtures for the retainer as described on Page 3-2 of BAN-1496. The weight of the NORA is approximately eight pounds af ter modification as compared to fif teen pounds Prior to modification. ' The minimum holddown' criteria for retainer use is that the margin to component lift must be greater than thirty pounds with the retainer in use. Analyses performed by B6W (taking into account the hydraulic forces acting on the MORA, the NORA weight, and the retainer holddown force) show that the net holddot:n on an NOR/ with a retainer irstalled is approximately thirty-fire pou;1ds in the Davis-Besse 1 reactor. Therefore, this design criteria is met. g E
i ' [ cocka.Wilcox . s Mr. Steven A. Varga Page 2 June 7, 1978 l The fuel assembly growth criteri.a stated on Page B-2 of BAU-1496 is based on a fuel assembly design burnup used as a' basis for the retainer design. Since the maximum burnup seen i in one cycle of operation will be less than the burnup used as a design basis, the fuel assembly growth criteria is met. l 1 l It should be noted that the retainer will be used for only one cycle of operation. , It is postulated that a retainer failure could cause reler.se of a retainer and, possibly, an MORA into the reactor vessel. The neutronic and thermal-hydraulic consequences of i l this event are insignificant. Although interference with control rod motion is very unlikely, this concern has also I been considered. Analyses of stuck .out control rod transients for B6W 177-FA plants have shown that these plants can be safely shut down in this event. Therefore , should interference i with control rod motien occur, the p1, ant could still be safely I shut down. . The major concern associated wit ~h retainer failure is . I plant damage and potential outages for repair. This damage l would be prevented by the Loose Parts Monitoring System (LPMS) i which is provided on all B6W operating plants. The LPMS has the capability to detect a failed retainer in either the reactor vessel or steam generator. The importance of LPMS i indications has been emphasized to plant operating staffs to i preclude possible equipment damage in the unlikely event of retainer failure.
. Even though the retainer is designed for only one cyc1e of operation, B6N uill recommend to utilities using the retainer
.that surveillance inspections be made following retainer use to !
provide additional confirm.ation of acceptable operation. The results of these discussions will be provided to the NRC and definite plans will be provided as they are formulated. We hope that this adequately answers the questions raised in the discussion today. Should any further information be required, please contact Mr. W. R. Gray (Ext. 2553) of my staff. Very tr ly yo .. a J dus,lfsr JmesH. Taylor,
/
l
. Manager, Licensing
~JHT:dsf l
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coc.k.&Wilcox . . . ,- ,- p ,_ ,. . Mr. Steven A. Varga Page 3 June 7, 1978 T:. t .:1. tv t- c- :- ': - ' R. B. Borsum (B6W)
- cc:: - : -- .
1
- . :. - L..B..Engle (NRC) - --
- H. Silver (NRC)
M. Tokar (NRC) . - L. E. Roe (Toledo Edison) . R. W. Heward (General Public Utilities)
~-
'bec: J. S. Tulenko I i'-'. . . .
J. C. Deddens. . W.- R. Gray l
. G..O.' Geissler . l K. O.' Stein i
5 : E. R. Kane . I
..~ '. ' B.'J. Short
- G. A. Meyer .
. )
. M. W. Croft
. R. Berchin -
-L-. R. Pletke .
E. G. Ward ' W.- R. Gibson
;-~ R. E. Kosiba .
J. P. Jones . G. M. Olds - t J. T. Janis .
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4 O . e 1 . . s b o s 6w
1 i l . BAW-1497 Juan 1978 I l l 1 1
- l i
I l i
- I l
l l JUSTIFICATION FOR REMOVAL OF ORIFICE ROD ASSEMBLIES IN l TEREE MILE ISLAND' UNIT 2, CYCLE 1 l l l l l i l l l O i G i i
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1 ( Babcock &Wilcox
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BAW-1497 June 1978 JUSTIFICATION FOR REMOVAL OF ORIFICE ROD ASSEMBLIES IN THREE MILE ISLAND UNIT 2, CYCLE 1 l l l l l l I l t BAECOCK & WICLOX Power Generation Group 2 Nuclear Power Generation Division. P. O. Box 1260 { 2 ' Lynchburg, Virginia 24505 Babcock & Wilcox f I
l 1
- 1. INTRODUCTION _
l This report provides justification for continued operation of the first cycle ) of Three : Mile Island Unit 2 (THI-2) at the rated core power of 2772 MWt follow-ing the removal of orifice rod assemblies (ORAs) from the core. The ORAs l are used to limit bypass flow through fuel assemblies with empty guide tubes. A system flow of 102% of design flow has been used in these analyses which offsets the increased core bypass flow due to renoval of ORAs. An evaluation of thermal-hydraulic performance has been made based on the increase in system flow and removal of ORAs and has been compared to the anal-yses presented in the TMI-2 FSAR I and Fuel Densification Report.2 This evalua-tion shows that the effects of the removal of forty ORAs and the increase in reactor coolant flow rate provide' improved safety margins relative to those reported in the TMI-2 FSAR I and Fuel Densification Report.2 The use of retainers 3 to provide positive holddown of burnable poison rod as-semblies (BPRAs) in the remainder of cycle 1 has also been considered. 1 1-1 Dabcock & \Vilcox
l ) F 1
- 2. THERMAL-HYDRAULIC DESIGN I
The thermal-hydraulic design evaluation supporting continued cycle 1 operation used the methods and models described in reference 2 with the following excep-1 tions:
- 1. An increase in core bypass flow due to ORA removal.
- 2. An increase in system flow.
- 3. The inclusion of retainers to provide positive holddown of BPRAs.
During the initial porties of cycle 1 operation, fuel assemblies which did not contain control rods, 2.?ls, or neutron sources had ORAs installed in the guide tubes to minimize ::re bypass flow. The maximum core bypass flow, with ORAs installed in forty fuel assembly locations, was 6.04% of system flow. Thirty-eight ORAs will be removed for the remainder of cycle 1. Two fuel as-semblies will contain primary neutron sources and modified ORAs. The thermal-hydraulic analysis assumed a total of forty vacant fuel assemblies and resulted in a maximum core bypass flow of 7.6%. As previously noted, a system flow of 102% of design flow was used in the anal-ysis (see Table 2-1) which offsets the affect of the increased bypass flow. This system flow rate is conservatively based on a predicted four-pump flow rate of 105% of design flow as verified during startup testing. Retainers will be installed on all fuel assemblics containing BPRAs and pri-tary neutron sources with modified ORAs. This retainer design is described in reference 3. The additional form loss due to retainer installation has been included in the calculation of core flow distribution. The limiting fuel assembly does not contain a BPRA during cycle 1 operation. ,, Maximum design conditions and significant parameters are shown in Table 2-1 for cycle 1 operation with and without the ORAs. The potential affect of tuel rod bow on DNBR was considered by incorporating suitabl. warcin.- into DNB limited core safety limits and RPS setpoints (pres-sure temperature limit.- ..nd flux / flow setpoint). The maximum rod bow penalty was cniculated from the equction: 2-1 Babcock & Wi!cov
k Table 2-1. Thermal-Hydraulic Design Conditions Densif'n Revised
- TM1-2 FSAR Report Cycle 1 J Design power level, MWt 2772 2772 2772 System pressure, psia 2200 2200 2200 RC flow, spm 369,600 369,600 377,000 Vessel inlet coolant temperature, 100% power, F 557 557 557.2 Reference design radial-local power peaking factor 1.783 1.783 1.,783 Reference design axial flux shape 1.5 cos 1.5 cos 1.5 cos Hot channel factors Enthalpy rise 1.011 1.011 1.011 Heat flux 1.014 1.014 1.014 Flow area 0.98 0.98 0.98 Active fuel length, in. 144.0 141.7 141.7 Average heat flux, 101* power, 185,000(*) 188,000 ID) 188,000(b) -
Btu /h-ft2 CHF correlation W-3 BAW-2 BAW-2 Minimum DNBR, 112% power 1.39 1.62 1.65
") Based on the active fuel length and cold fuel pin diameter.
(b) Based on the densified active fuel length and hot fuel pin diameter. I 1 2-3 Babcock 8.Wilcox
- 3. TRANSIENT ANALYSIS The DNBR related transients presented in reference 2 have been reviewed for
- applicability to operation with the ORAs removed. The four pump coastdown is the loss-of-coolant-flow (LOCF) transient analyzed in the Densification Report. The minimum DNBR during this transient was 1.65 (BAW-2). The initial condi-tions for these transients are at 102% power. Re-analysis at 102% power with ORAs removed shows an increase of 1% in the initial DNBR. The higher initial minimum DNBR makes the results of the transients analyzed for the Densification Report applicable and cc:servative for the revised cycle 1. All loss-of-coolant flow transients, with the exception of the loss of one pump from four pump operation, will result in a reactor trip initiated by the pump monitors. The most limiting LOCF transient for which the pump monitors provide DNBR protection is the four pump coastdown which has been shown to be acceptable. The one pump coastdown from four pump operation is the most limiting flow transient by virtue of its use in determining the flux / flow trip setpoint. The flux / flow trip is based on preventing the minimum DNBR from going below the design value plus the rod bow penalty. Therefore, a one pump coastdown with the resulting flur./ flow reactor trip will result in the most limiting DNBR during normal operation. The TMI-2 FSARI has been reviewed for the most limiting DNBR transients of moderate frequency since the one pump coastdown does not appear directly as an accident. The most liniting FSAR transient is the excessive heat removal accident (feedwater temperature decrease). This transient has been re-analyzed ] [ for revised cycle 1 operation with the same input as used in the FSAR. The l results of the re-analysis are shown on Figure 3-1. The minimum DNBR is 1.58 (BAW-2) versus a 1.43 (W-3) reported in the FSAR. l { i 1 3-1 Babcock 8. VVilcox I
Figure 3-1. Feedwater Temperature Decrease 2.0 - FEEDfATER 1DPERA111RE IIICREASE 1.9 1.8 _ G. E e. Y e 1.7 - u m ' 5 S g .
-3 1.6 -
1.5 - to cu er O - o x-P 1.4 ' ' '
- e .
s g 0 10 20 30 40 50 60 o . o x
- time, sec
. 1 i I
- 4. CORE LOADING PLAN i l l 1
1 Figure 4-1 shows the revised core loading plan for the remainder of cycle 1. All fuel assemblies are remaining in their original core locations, i.e., no I fuel shuffle will take place. The changes occurring are:
- 1. Retainers will be installed on all BPRAs. ;
- 2. Thirty-eight ORAs will be removed.
- 3. Two ORAs will be modified and installed in the primary neutron source locations (B-12 and ?-4). j l
l l l l \ .. 1 5 1 4-1 Babcock & \Vilcox
l l 1 l l.. ( ! 5. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for the remainder of cycle 1 operation. Changes were the result of the following:
- 1. The pressure-temperature limits have been revised to incorporate the affects of ORA removal, retainer installation, and rod bow penalty.
- 2. System flow of 102% of design flow was used. l i l
- 3. The low pressure se: point has been raised to account for the LOCA small l l
break analysis (ba:iup function only).
- 4. Instrument drift nuchers have been included for calibration drift in 1
accordance with ite: 2.C.(3)f. of the operating license, j
\
l Figures 2.1-1, 2.1-2, 2.2-1, 2.2-2, and 2.1 (Tech Spec numbering) illustrate l the revisions to previous Technical Specification limits. 1 i l 5-1 Babcock a vvitcox
i Figure 2.1-2. Reactor Core Safety Limits THERMAL POWER,5 120 DNBR LIMIT . (.u..iin - O -
- (24.s.iin 110 - KW/FT Lillif ACCEPTA8LE
( 49.2.100) 4 pyyp - - 110 (40.8.100) OPERAfl0N (.33.g,84.6) ^ ^ (24.5,8 4. 6)
. - 80 ACCEPTABLE
(-49.2,73.1) 3 3 4 pyyp (40.s,73.1)
,, ,g .
OPERATION (.33.8,57.4)
^ @. - 80 -
^ (24. 5,57. 4 )
I ACCEPTABLE -- 50 l (-49.2,45.9) 2,3,1 4 PUMP (40.8,45.9) l CPERATICH -- 40
-- 30
. . 20
. . 10 t i f I e f f f f f f 8 50 50 40 -30 20 -10 0 10 20 33 43 50 63 l l
Atlal Power imealance, 5 9 CURVE REACTOR COOLANT FLO2 (GPW) 1 377,000 2 280,400 3 182,800 5-3 Babcock & Wilcox
Figure 2.2-2. Allowable Value for Nuclear Overpower Based on l RCS Flow and Axial Power Imbalance 1 1 ) 5 0F RATED THERMAL POWER , 9" (105.125) 1
- - 100 l 2
l ACCEPTABLE l l4 PUMP l (93.54)
'(93.54) , i I
IOPERATION -- 90 l 80 (78.225) i ACCEPTABLE I -- 70 1 l 3 & 4 PUMP , (66.64) (66.64) I i OPERATION l
- - 60 l l (51.025) 50 -
ACCEPTABLE l
! l 40 l . (39.44)
(39.44) [ PUMP y l OPERATION ,, j l y
- - 40 l 1
20 , l
= 9 a.
=. i E* E 8
NlI 11 10 nSl ,, l i l m , e Et i E!, t Ug! a t 2
-50 -40 -30 -20 -10 0 10 20 30 40 50 Axlai Power imoalance,f, ,
1 I l 5-5 Babcock 8.Wilcox i i l
o-l l . l ' REFERENCES 1 Three Mile Island Nuclear Station, Unit 2 - Final Safety Analysis Report, Docket No. 50-320. 2 Three Mile Island, Unit 2 Fuel Densification Report, BAW-1455, Babcock & Wilcox, Lynchburg, Virginia, July 1977. 3 BPRA Retainer Design Report, BAW-1496, Babcock 6 Wilcox, Lynchburg, Virginia, May, 1978.' 4 NUREG-0432, Three Mile Island Nuclear Station Unit 2 Technical Specification, Appendix A to Lice.se No. DPR-73, February 8, 1978. l l A-1 Babcock & Wilcox
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, METROPOLITAN EDISON COMPANY .,- .: . r . :. c ; . ..
s_ .; . v POST OFFICE 80X 542 READING. PENNSYLVANIA 19603 TELEPHONE 215 - 929 3601 i June 30, 1978 i GOL 1138 I 1 Director of Nuclear Reactor Regulatien Attn: S. A. Varga Light Water Reactors Branch No. '4 U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Sir:
Three Mile Island Nuclear Station Unit 2 (SE-2) Operating License No. DPR-73 1 Docket No. 50-320 ! Steam Generater Questiennaire ! l Attached for your information and use is a completed copy of the Steam Generator Operating History Questiennaire which was enclosed with your letter dated April 21, 1978. This questiennaire has been answered with respect to the 3C-2 (3&W) Once "hrough Steam Generaters. Sincerely, 0
/ /
/
/ /
/ ,
/J.-i.Herbein Vice President-Generatien 1
JOH:JES:cjg
Attachment:
Inclesure 1; Steam Generator @erating History Questionnaire r j' , fa h J ty
- y. t.. . .
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ENCLOSURE 1 . STEAM GENERATOR OPERATING . HISTORY QUESTIONNAIRE r l NOTE: All percentages should be reported to four significant figures. *- I. BASIC PLANT INFORMATION Plant: Three Mile Island - Unit 2 Startup Date: July 1978 (Hot Functional Testing) Utility: Metropolitan Edison oo. Plant Location; Middletown, PA Thermal Power Level: 2772 Mw Nuclear Steam Supply System (NSSS) Supplier: 3&W Number of Loops: 2
) Steam Generator Supplier, Model No. and Type: 3&W OTSG i Number of Tubes Per Generator: 15,531 Tube Size and Materic'.: 0.625" 00, 0.03h" W 3 11, 56' 2 3/8" Lgth, Inconel II. STEA'i GENERATOR OPERATING CONDITIONS .
l Normal Operation i .- rinar y-70CF608CF Inlet l Temperature: FDW STM - 5700F - Flow Rate: 68.cLx;oo #4rAllowable Leakage Rate: 1 GPM Total i i Fri=gry STM-o.12 X 10 5 #/hr l Primary Pressure: 2155 Psiz L" ! Secondary Pressure: 925Psid . [2 l Accidents 1," l
,a 1 Design Base LOCA Max. Delta-P: 925 ?sie ;,.
Main Steam Line Break (itSts) Max. Delta-P: 2200 F3is - l l 111. STEAM GENERATOR SUPPORT PLATE INFCRMATICN [3: 4-1 Material: Car ~cen 5 teel l l Design Type: 3reeched Openin?
- Design Code
- sA_g ;_3 ,
l Dimer.s icas : _13 3/3",:ia. j
. 713o Rate: 53 ". 1:' '3'hr. !
Tube Ecie Dimensicr.s: =0.32 in. =in. rudius nute ( l
-, . . , - ...z.... .--e :- . . . . . ,
i i f' i l IV. STEM GENERATOR BLOWDOWN INFORMATION .
~
frequency of Blowdown: Continuous 8 45',FP; o ! >15% FP }' Normal Blowdown Rate: 5 OFM J w Elowdown Rate w/ Condenser Leakage: N/A , ., p 1 l Chemical Analysis Results @ J. g l l Results Parameter Control Limits . l l l ! h:icn Cond - 1 u=ho < 10 u=ho
, )
j :ia - 0.05 p:= < 1 pp: l
- - 0.05 pp= .51 pp l l
. I 1
l V. WATER CHEMISTRY INF0FdiATION l l Q-l Eecondary Water . i Tj;e of Treatment (AUT) and Effective Full Power 0(ZF3r) EFP Months of Operation: Hydrazine/.t.=cnia l T.r;,ical Chemistry? pH 01<oc l vat= ion.Cond. 58 o - 9 ' 0 no 3 u=:Ta 02 <30=-50 FPS l fci.jnater
*=nuritv Lix.its Fe < 100 ::b Ob = 0 neb 00 1 7ob Hydra:ihb = > <0.1 3!'Teg <5 enb Cati:n Cond. 0 pp=ho.,c=
5 ut pc : . ; = ) . 3- " C...de:. ar Coolir.g Water _ . . m ! irr,ical Chemistry or Impurity Limits: :To 3;ecifi: 'imits - Ij u
.. .ar31izers - Type: :rene s
,]
I;
- ~ . ling Tower (o
- en cycle, closed cycle or none): Closed Cycle (;-
0 m e I
i h ~ 3-VI. TURBINE STOP VALVE TESTING (applicable to Babcock & Wilcox (B&W) S.G. only)
- r-
- 4
- Frecuency of Testing 2
- y j Actual: Not yet performed Manufacturer Recomnendation: '4eekly j l ll y
Power Level At Which Testing is Conducted Actual: Not yet performed g Manuf acturer Recomendation: o to 225 or 50 to loos Testing Procedures (Stroke length, stroke rate, etc.) Actual: Full stroke, approx. 13 see. full cycle Manufacturer Recommendation: full stroke, approx. 13 sec. full cycle VII. STEAMGENERdTORTUSEDEGRADATIONHISTORY (See attached sheet, Page lo) (The following is to be repeated for each scheduled ISI) Inservice Inspection (ISI) Date: NA Number of EFP Days of Operation Since Last Inspection: O (The following is to be repeated for each steam generator) . Steam Generator Number: Percentage of Tubes Inspected At This 151: Percentage of Tubes Inspected At This ISI That Had Been Inspected At The Previous Scheduled ISI: N/A Percentage of Tubes Plugged Prior to This ISI: ' Percentage of Tubes Plugged At This ISI: Percentage of Tubes Plugged That Did Not Exceed Degradation Limits: Percentage of Tubes Plugged As A Result of Exce'edance of Degradation .
,r-Limits: p Sludge Layer Material Chemical Analysis Results: N /A Sludge Lancing (date): 3/A
[ Ave. Height of Sludge Before Lancing: 3/A M . Height of Slud;c After Lancing: 3/A {L (Sriefly
- . e;1acement, Retubing or Other Remedial Action Considered:
Specify Details) 7:ne
) Su ;or: Plate Hourglassing: Ncne Su ;;rt Plate Islanding: 3cne T.::e "etalurgical Exam Results: I
l l h ! W l
]
. i U .
Fretting or Vibration in U-Bend Area (not applicable to B&W S.G.) AS OF (4) ij 3 Percentage of Tubes Plugged Other Preventive Measures ;.?.
. G ,
1 l l
' Wastaae/ Cavitation Erosion AS OF (4)
Hot.Lec: (Repect this information for the cold leg on Combustion Engineering (C.E.) and IIestinghouse (W) 5.G.) Area of Tube Bundle (1) a b c d e l 1
% of Tubes Affected by i Wastage / Cavitation Erosion l
% of Tubes Plugged Due to '
C \ l l Exceedance of Allowable Limit (2) I, of Tubes Plugged That Did not Exceed Degradation Limit Location Above Tube Sr.eet (3) Max. 'r'astage/ Cavitation Erosion - Rate for Any Single Tube (Tube Circum. Ave) (Mills / Month) ]h: Max. nastage/ Cavitation E-rsion '
- f. {
ir. Ar./ Single Unploeged Tube - (Tube Circum. Ave) (Mills) ; s
~
Cra: v:; AS OF (4) C a.:Iti Stress Corrosion Induced in C.E. and W S.G. - F'.c., :nfuced Vibration Caused in B&W S.G. C
1
) '
Crackino (Con't) ' ! Rot Leg: (Kepeat this information for the cold leg on C.E. and W S.G.) . Area of Tube Bundle II) a b c d e ,
% of Tubes Affected L By Cracking
% of Tubes Plugged Due to Cracking
% of Tubes Plugged That Did Hot Exceed Degradation L.irit ,
Location Tube Sheet Above (3) Rate of Leakage From Leaking Cracks (gpm)
) .
Denting (Not applicable to B&W S.G.) AS OF (4) . . Hot teg: (Repeat this information for the cold leg on C.E. and W S.G.) a b c d e , Area of Tube Bundle (1)
% of Tubes Affected by Denting
% of Tubes Plugged Due to '
Exceedance of Alloaable ;- Limit (2) !.J s .q
% of Tuties Plugged That . .' j Lid Hot Exceed Degradation ,
'1 Limit j
- .3 Rate of Leakage Frca Leaking Dents (gpm) y i.
Max. Denting Rate for Any . Single Tuce (Tube s Circut. Ave) (Mills /P.cnth) . _ .
!:n.. Centing in Any Sir.gle Unplugged Tube (Tube Circur.. Ave) (Milis) -
O ! O - Denting (Con't) . Support Max. Denting in Any Single % of Tubes Affected By Plate Tube in Bundle Area Denting in Bundle Levels (Tube Ave) (Mills)-(1) Area , a b c d . e a b c d e 1 2 3 4 5 6 7 8 10 9 0 11. 12 l e Ei' t ;
!;]
U2 ($ 1_! O
~
J -
.y
~
~
. TABLE KEY NOTE: All percentages refer to the perce'nt of the tubes within a given ,
area of the tube bundle. j (1) F Area of the Tube Bundle No. of Tubes Within the Area i
- a. Periphery of Bundle (wi/20 rows for B&W; wi/10 rows for C.E. and W)
- b. Patch Plate (wi/4 rows) I
- c. Missing Tube Lane (B&W only)
(wi/5 rows)
- c. Flow Slot Areas (C.E. and W only)
-) wi/10 rows)
- d. Wedge Regions (C.E. and W only)
(wi/S rows) .
- e. Interior of Bundle (remainder of tubes)
(2) , l Allowable Limit for Wastage / Cavitation Erosion: Allcwable Limit For Denting: 4; (3) .
- 1. Specifies area between the tube sheet and the first support plate .
',1 l 2. Specifies in the following locations: (list the addi:.icnal locations) -
! M ! Wastage / Cavitation Erosion: Cracking: . (; ! Specify the date of the inspection for which results have been tabulated.
- l l
i O
.8- .
1 - g h VIII. SIGNIFICANT STEAM GENERATOR ABNORtiAL ORERATIONAL EVENTS s; t f A DATE
SUMMARY
k (Include event description; unscheduled ISI results, if performed; and subsequent remedial actions) L-23-78 Rapid pressure transient due to improper MS Safety Valve Operation I, l IX. CONDENS$R INFORMATION 1 Condenser Tube Leakage Detectable Detection Material Date Rate (gpm) Limit Method I (?abe s ) l SS-ASTM 0 Cation Cond. l A2Lo Na+ ,
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I l l l X. RADIATION EXF05URE HISTORY WITH RESPECT TO STEAM GENERATORS l 1 Date ExamDosage(Man-Rem) Repair Desage (Man-Rem) Comments
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r 1 1 , - T C . . 5. U a n XI. CiGRADATION HISTORY FOR EACH TYPE OF DEGRADATION EXPERIENCED FOR TEN lj REPRESENTATIVE, UNPLUGGED TUBES FOR WHICH THE.RESULTS OF TWO OR MORE [g ISI'S ARE AVAILABLE N/A E f If the results for ten tubes a're not available, specify this infor- j l mation for all those tubes for which results are available. ( (repeatthefollowinginformationforeachtubeanddegradationtype) l Steam Generator No:
. Tube Identification:
Type of Degradation: (specify denting, wastage, cavitation erosion, caustic stress corrosion cracking, or flow induced vibratian cracking) (repeat the following information chronologically for each ISI for which re:,ults are available) m ISI Date: g . Amount of Degradation: (specify amount and units) : EFP Honths of Operation Since Last ISI for Which Results are Given: 1 1 l 1 1 m I I' l r-
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i d l o - 10 - ! l Item VII. Steam Generator Tube Degradation History: This information ' documents the result of the preservice examination. An ISI l has not yet been conducted. Preservice Inspection Date: December 1977; IFP Days of Operation is not applicable. ; OTSG A OTSG B i Percentage of tubes inspected at PSI: 100% 1005 Percentage of tubes plugged prior to this PSI: unknown unknown Percentage of tubes plugged at this PSI: . 08h5 .155% Percentage of tubes plugged that did not exceed degradation limits: 23.07% 12.5% Percentage of tubes plugged as a result of exceedance of degradation limits : 76.k2% 87.5%
. Tube Metallurgical Exam Results
- None i
"A tube with an eddy current indication of a 55% 0.D. defect was removed.
Metallurgical examination revealed a scab or lap presumably from tubing manufacture with a depth of about 50%. l 4 3 'T l
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MEMORANDUM FOR: . B. Vassallo, Assistant Director for Light Water Reactors, DPM ? ( FROM: H. Silver, Proj ect Manager, Light Water Reactors Branch No. 4, DPM , , ,
\.4 THRU: S. A. Varga, Chief, Light Nater Reactors
Branch No. 4, DPM
SUBJECT:
ITEMS FOR THI-2 HEARING BOARD The following items have arisen lately on TMI-2 which may be of interest to the hearing board. (The appeal board is still convened.)
- 1. Purge Valve Operability In response to concerns raised by the Containment Systems Branch, the Mechanical Engineering Branch has identified certain requests for additional in-formation regarding operability of the containment purge valves. These requests deal with confirmatory information to more completely document the ability of the containment purge valves to close if they are
%G in use at the time of a LOCA. Present technical specifications restrict the time these valves may be open with the reactor critical to 90 hour: per vear.
Responses to our requests are not expected to raise any issues which represent significant safet' probiens.
- 2. Burnable Poison Rod /0rifice Rod Assemblies At another operating B6N reactor, it was found tLat two burnable poison rod assemblies (BPRA) nad been ejected from the core and pieces of several components had been carried into the steam generator inlet plenums.
The reactor was safely shutdown and no dama;e was done which represented a significant safety issue. 36W concluded that this problem was due te wear in the BPRA ball-lock coupling caused by hydraul;c 1;ftin; of the BPRA during operation with all fou: reactor coolant pumps. B6W proposed installation : f a 3 Pid retainer which would provide positive holdd:wn against
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-D. B. Vassallo JUL ! ! 37e l l
1 all lift forces, and Met Ed has stated their intention l to install this device on all BPRA's. B5. ha s submi t t ed ! B6W-1496, BPRA Retainer Design Repo rt , for our approval, 1 i but our review is not yet complete. l 1 The same ball-lock device is also employed on orifice l rod assemblics (ORA) in B6W reactors. During the inspection of these devices at another B5% reactor, l wear similar to that on the BPRA's was observed. B6N l has concluded that for some plants , including TNI-2, the ORA's should be removed. This will require revi8ed thermal-hydraulic analyses for the core, but based on such analysis already completed for other reactors, these are expected to be acceptable. These analyses for TMI-2 have recently been submitted and will be reviewed and approved prior to plant star:up. TMI-2 has been shut down for several weeks for correction ! of operating prob 1 cms and is not schedulec to 8 tart up prior to early August 1978. Met Ed at its own ri8k has already completed the installation of the BPRA retainers and removal of the ORA's. As noted above, startup will not be permitted prior to approval of all supporting. documentation.
- 3. Auxiliary Trans former W
The licensee has informed us that recent studies have shown that for the normal operating range of the grid, operation with a singic auxiliary transformer will not provide adequate voltage icvels to suppor operation of balance-of-plant auxiliaries and engineered safety features. Met Ed proposed both short tern and long term corrective action which they believe to be in accordance with General Design Criterion 17 and the TMI-2 FSAR. See Met Ed let:cr of May 30, 197S, and LER 78-35/IT attached for additional information. 4 We have not yet completed our review of this information. As noted above, TMI-2 is expected to be shut down until August, 197S, by which ime our r e '; ; c w should i be complete. .
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Ha rley S i.lve r',' i P r'oj ecTMana ge: Light Water Reac to rs Branch ':c . - i Division of Proj ect Management j cc: See next page 1 3 E
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a l METROPOLITAN EDISON COMPANY -- 1 POST OFFICE Sox 542 AE AOING, PENNSYl V ANI A 19603 % p HON (lit$ .g,..;dOf ., .ve
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. Director, lluelear Reactor Ser21stica . .O Atts: :'.r. 5. A. Yarga, Chief l Light Water Reacters 3 ranch Ic. .
i U.S. Nuclear Rega.latory Cc-d ssion j l Was hington , D . . 20555 j
Dear Sir:
, "'hree !ile :sland Nue'. ear 3:atien, '!ni; 2 ( 2C-2' Cperating License :fc. ;FR 73 i
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Decket No. 50-320 < l 1 l In respense te questions raised by Mr. Harley Silver cf ycur staff, J enclosed please find inf =ation concerning the pctentisi pr **** *s-sociated vith the less cf an auxiliary transfer =er a; T:C-2. 44%
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the StI-2 Auxiliary Transfer ers. The results of a Surns and Roe voltage study indicated that with :nly :ne auxiliary transfor=er in operation and with the plant at f'.:7.1 1:ad, the bus val e +.a.,a. veuld b.e redu-a.". ', o . %.. . a. .:c *. e .. . ". . . a*. , ' . '. ". . e =. ". a. .. . . '. ' ~.s'a". e. , _e- m .a. safety loads vould not be picked up. -'his :cu'.d resu'. in -he blevine :r control fuses on the safety-related cenpenents. Met-Id found this situation to be repcrtable in acecrdance with ?ection 6 . 9 . .. 3 ..g o .c . g .. a. -v. . .t -2 %. . ... ". ..' .11 S.,e e . .d .'.'.. - . .' ~.ts . '...=.~a..'.-a., . - . .'.N v. 0, , 1978, Met-Td subnitted ~.icensee Ivent .:.eper (~ E' M -35/1 :: the 2,cn-
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/ METHOPOLITAN EDISON COMPANY . -
l . POST OFFICE BOX 542 REACING. PENNSYLVANIA 19603 TELEPHONE !5 - 9:9 0601 a 3- G01 0090
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?t . 3. H. Orier, Director -
Office of Inspectica and Inforce=ent ~'S' Region I U. S. Nuclear Regulatory Oct=issien 631 Psrk krenue - King of Prussia, Pennsylvania 19hC6
Dear Sir:
Three !-{ile Island :Tuclear Statica "Jnit 2 (DC-2) - Operating License ic. PR 3 Inclcsed please find icensee Ivent Repc-t 73-35/10 vhich is submitted in accorda .ce vith Secticn 6.9.1.8.h of cur Technical 5pecifiesticas. Sincerely, Signe: J. 3. Hi. :3:n J. G. Eerbein Vice ? esident-Generstien JGH : E,*d,: cJ 3 I.clesure: *Z3 73-35/1T 1 l cc: Marley Silver (:iRC)
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,@l (DLE Ass PAier c A TWE ALL AECul A!O INFOAMarios:
I o l t l l ? l A l '.' l :4 lI l 2 l@l0 l0 l -l O l O ! 0 l ]l 0 l - l 010;>!@l ; l ' I 11. f 1[ 4 [ l lk a s .iC Nsercaos ia a uCaNu Nuveta ;4 ..c a s4 %,e .a u a r 34 CON'T ioltl [n"c: , 610 ! 5 l0 IO !0 l' I2 10 a ClO !~ h !1 !~ li
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s sa ei coex a r a. vs t a 42 4vaNr cars 4 s ar :4 5 ao EVENT DE$CAIPr 0N ANO 8ACGA8LE CONSECUENCES h 10121 l'4hile in Mode 5 the TMI-2 a chiteet__ enrineer f3uns i Reel netified -he *.icensee - t iolal l single auxilia y transfer =er : eration vithin the ner:a1 cperstine rsnre of the :-i lotal (vill not provide sufficien; velta:e levels for ceersti:n O. .,.e s - .. l o I s i Iduring periods of peak unit euxiliary demand. 3eesuse the uni- has act been cpers:
! 3 !s t l at 2 ;c er level ree_uiring etxinus unit auxiiiries . here was ne ;- ential f:r a .r l o I 7 I l adverse effect on the hesl h snd safety of the public.
O{dj l 7 9 1 Sv57tM CAUSE CALSE C '.* * .agg COCE CCCE $cSCC08 C098CNEN? 00E $Ls:0:1 sus;;04 3l9l l3 !$ l l 3l 1 1l lCl3!EI 2l$l2l l2l f;g2l 8 3 to it t; 13 is i9 SECutNTIAL OC:,;m a E NCE aceca; a t'. isi:N fvtNTwfan atsC A? NO. C CE WPE s0. O "ts "c e qC Nous"a i l 7131 l-l l 0131 Sl IM 10 ! : ! IT l l- I 13l 25 22 23 ;4 ;6 ;r :s :s ;o ai .; aC?iCN surv a t tastCT / 5>uf 0CV.N .. AT ACMVf N T N 8 8 C-8 88:V f ~ 2 '.t 8 ::v8C N P f a K gre ACT6 C's ON PUNT vtT*C0 a'CL A 5 .- swsus TE3 8Cau sue 568*.!= ua%. sac-_ wgg l xl@l XI@
;J Ja I :l@
JS 126 21 @ s,1 0 ! 0 1 0 4o! 11 IJI@ 4. INI@ 4: 12l@ 42 1:!?l 44 C AUSE OESCAiPTION ANO COA AECTIVE ACT:CNS 77 i l2i ldingle transf:rmer c;erstien vel age studies rendu::ed by the ar:hitect enrineer
!>iil lecerletin: the vcitare cetici:sti:n studies , shewed -hat vithi:- :ne .:- Li :;ers:1-i i i t range of the grid (232 to 23sr/' s sinrle au.:cilisrr ir r.s fe .er 22n't ;reviie ide~
i 131 l voltage levels to supper c;erstien of the uni: Inrineering Safe r ?titure s and b C. i i j j ef plan: 2a iliaries. .'00. inued' , I 4 ) s ACit,' fv lC ME TwCO 08 .- sta tus %80W48 Ofa(*5'a*W5 35COstav Or$0:v t a v
- 13 0- *
- 0 s '"!2"N 7
- i i sa l? I 3 l@ ic 10 10 1 ' I2
- 9Ii3 NA 44
! Icl@l ::ctif t:a-i n hv AI 45 46 1
ACfivir< 00r').?t ;T aELtA510 08 St. fast avouN* 3aC*.i" '-
. Ca*r0N:: zi.i.3i d i l6l l 'I A ! l Si 4 9 33 ti 44 45
*te'.ON'.E. fuscsuats Nyvata tver O t 3;mir ?:CN 09 i i I 7 I l 01010 !@l z l@l ::A a 9 i, 12 13 81 a 5Cr.'.f L **ubmit i d'
su.*Cin O E SC R.* fic's
! !A! 0 010 40 ! IA 8 1 it it L C*,5 0 8 ,t1 #) A *.ta
! 'O 8 &C? Li f g./ vQ rvet Otsc sis ics l 1 l l 2 lk,1 'I .. . . . - . . . . --
h ~l f
o , t i i Cause Descrintion and Corrective Actier.s In order to assure adequate voltage level, one of following ccrree ,ive actions vill be taken: A) the unit vill not be operated above a pcVer level coc:patible to safe single transfor=er cperation; B) selective balance of plant load shedding vill be installed; C) or the aute=atic bus transfers to . the other auxilia:/ trans fo=er for designated bus vill be disabled. 1 i . l l l i l l l _ _ . , . , i
6 lARRATI'/E TO ACC0" PATI 1ER ?S-M/1T n C'n May 3, 1978, the results of the single auxiliarf transfer =er voltage study were received frc= the Architect Engineer (Surns & Ree). D e voltage values calculated for the k80 V Motor Centr 1 Centers, for a nor=21 vcitage of the 220 r/ grid vere below the required h07 7 AC necessary to assure safe Operatica of , the magnetic centro 11ers and prevent control pcver fase bloving. These voltage l- values are based on having the =aximu= unit auxiliaries in sertice during the summer months, with all circulating vater pu=ps in serrice.
.This item vas deter =ined to be a violatien of Technical Specification 6.9.1.8.h, in that the Safety Evaluation Report states that each unit auxiliary transfor=er is sized to carry the unit full lead auxiliaries and the energi:ed safety features auxiliaries. Since the IST bus povered frc= the 1 Operable transfor=er is desisne:
to fast transfer to the re=aining transfor=er, the potential te disable both IS? l trains due to control pcVer fuse bloving exists. Follow up studies have verified that by autc=atien117 shedding selected Salance t ' of Plant loads upen auxiliarf transfor:er failure, adequate tcitage levels are available at the h80 V Motor Centrol Centers thrcushcut the scr-e operating range of the 230 r/ grid. l l i I 1 1 i l l l I l
.- . . ,, ., - - - - . - - - . - - - ~ - - - - - - ~ - - - -
I . . t , c0 % j UNITEo STATES e 'i 't NUCLEAR REGULATORY COMMISSION 1, { j WASHINGTON, D. C. 20555
% 8 July 14,1978
%, .' .u y Docket No.: 50-320 1
l Metropolitan Edison Company l ATTN: Mr. J. G. Herbein Vice President ! P. O. Box 542 Reading, Pennsylvania 19603 l l Gentlemen: RE: THREE fi!LE ISLAND UNIT NO. 2 - We have recently sent the enclosed letter to Babcock & Wilcox Company i regarding the pote-tic 'cr excessive control rod guide tube wear I at facilities using the B&W design. This infonnation is required to assess the significa .ce of control rod guide tube wear at your facility. We requested the required information from the vendor, in order to provide us with an overall assessment of this problem and to minimize the amount of utility, vendor and HRC staff work required. However, , . ,g should this approach to obtain the necessary information not be j successful, a direct response to the enclosed questions, from each ' l'censee utilizing the B&W design, will De required. We request your cooperation with B&W in generating and analyzing the control rod guide tube wear data necessary to address this concern. - Sipcerely, , me i . 'a 1 i , As ta Obector
/ for Light Water Reactor Division of Project Managecegt
Enclosure:
Letter to B&U dated , June 13, 1978 l N' s g, s
$$&7 20' 6 /36 (
. . _ . - _ _ _ . _ _ . _ _ . . _ . _ _ _ _ _ . _ _ _ = _ . . .. _ _ _ _ _ .
. , i s l 1
1 Metropol.i tan Edison Company ! ces: I George F. Troworidge, Esq. Shaw, Pittman, Potts & Trowbridge
. 1800 M Street, H. W. -
.< Washington, D. C. 20036 {
i Mr. I. R. Finfrock Jersey Central Power and Light Company Madison Avenue at Punen Bowl Road l bberistown, New Jersey 07960 l Mr. R. Conrad Pennsylvania Electric Company 1007 Broad Street Johnstown, Pennsylvania 15907 . Chauncey R. Kepford, Esq. Chairman ' York Cocmittee for a Safe Environment 433 Orlando Drive State College, Pennsylvania 16801 Mr. Richard W. Heward l Project Manager ; GPU Service Corporation ,4 260 Cherry Hill Road Parsippany, New Jersey 07054 ! Mr. T. Gary Broughton l Safety and Licensing Manager GPU Service Corporation 260 Cherry Hill Fbad l Parsippany, New Jersey 07054 l l 1 1 i i l i l
. p UNITED STATES f 7 ,
*[
$ NUCLEAR REGULATORY COMMisslON CASHING TON, D. C. 20555 k ...+ j/
- June 13,1978 Babcock & Wilcox Company ATTN: Mr. James H. Taylor Manager, Licensing Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 2'4505 Gentlemen:
Significant wear has been found in control rod guide tubes at the Combustion Engineering (CE) NSSS facilities. The guide tube wear has been primarily located at the axial location where the control rod is " parked" in the fully withdrawn position during nomal oper-ation. CE postulates that the wear is caused by a flow induced , guide tube.of the Inconel control rod against the softer Zircaloy vibration Corrective actions, including increased operability surveillance, step insertion of control rods and extensive sleeving of both new and irradiated guide tubes, have been taken at all affected CE facilities. We realize that your NSSS design is different from the CE system, ' however, we believe that a similar wear problem could exist at facili-ties using your NSSS design. You are requested to provide the enclosed yW, additional infomation for all B&W facilities with operating licenses within 60 days of the date of this letter. Sincerely, Brian K. Grimes, Assistant Director for Engineering & Projects Division of Operating Reactors
Enclosure:
Request for Additional Infomaton
.t l
?
fh f d g)
t REQUEST FOR AD0fTf0NAL fNFORMAT10N INTEGRITY OF CONTROL ROD GUIDE TUBE (CRGT) BABCOCK & WILCOX FACILITIES Answers to the following questions should be supported with data and drawings to the extent possible.
- 1. Describe the details of any routine surveillance of fuel assemblies performed at your facilities using your NSSS design.
- 2. Have examinations of the fuel assembly guide tubes to detect wear been completed at any facility using your NSSS design? If so, provide.the following information:
a, The method of examination (i.e. destructive testing, eddy-cur-rent testing, periscope, borescope, mechanical gage, TV, etc.)
- b. The areas of CRGT examined.
- c. Qualification of the examination procedure.
G l
- d. The number of CRGT sampl'ed at each facility and the applicable '
l operational parameters including: the core location; EFDH; tire 1 in service; related control rod parameters; fluence; etc.
- e. Results of observations or measurements.
- 3. Were any CRGT destructively tested (e.g., by mechanical or metallographic means) and what observations or measurements were made?
'4 .
4, What correlations were suggested between operating parameters and CRGT condition? l 1 l i l ~ . . - - ~ . . . , . . . . _ , , , . = - , - . - , . - - - - - , . . - , .
I . 2 5. If specific examinations for CRGT wear have not been completed at any facility, either provide other evidence for the absence of wear or answer the folicwing:
- a. Are examinations planned? If so, provide details as requested
! in 2 a-d.
- b. Have out-of-pile wear tests been completed? If so, provide details including qualification of the test procedure and answers to 2 a-d. Address vibration, fatique, flow visualization, etc.
- 6. Document any other observations of wear or degradation found in the examination of your fuel assemblies (i.e. , grid wear, post wear,etc.). Provide the results of your assessment of the consequence of these observations. Describe any design changes
*Wi effected to either mitigate the consequences of this wear or eliminate the wear.
- 7.
If CRGT wear has been found at facilities using your NSSS design:
- a. What have been the attributive causes?
- b. Have correlations been made to characterize the phenomena with respect to operating procedures and plant specific core parameters?
.4 l
)
i
- c. Are specific locations within the core or particular CRGT within an assembly more susceptible? :
8. If CRGT wear has been observed at any facility using your NSSS design: a. Describe your efforts to reassess the mechanical integrity > of the core with worn CRGT to demonstrate that cool-ability and scramability exist for the normai, seismic ! and anticipated operational occurrence loading conditions. Describe the worst condition analyzed.
- b. Discuss your structural design bases. Indicate if provisions have been made to accommodate wear in the design. What amount of wear or related degradation would be cause for rejection for reload?
Provide the allowable stresses used in the structural analysis. Discuss the effects of temperature strain rate, D notch severity, irradiation and hydrogen content on mechanical properties used to establis'h the allowable stresses. c. Provide the results of your structural analysis sumarizing the - CRGT loads and the primary and secondary stress intensities for normal, fuel handling, and accident loading conditions. d. Discuss the effects of CRGT wear on the thermal-hydraulic performance of the reactor under nomal and accident conditions. 4 i l l
- 9. \
Discuss any control rod scram testing that has been completed to demonstrate scramability in worn CRGT. Address the effects of worn CRGT on scramability for the worst expected guide tube geomet ry. Include the strain-deflection limits for control rod ' f0nctionability. 10. If examinations for CRGT wear have not been or will not be made at representative facilities using your NSSS design, provide justification for continued operation of these facilities. 11. B&W has redesigned the guide tube lock nuts of later design fuel assemblies by making a change from zircaloy to stainless steel to mitigate the effects of observed wear in the upper region of the guide tube. Indicate which design is employed at each B&W dcrigned facility and indicate any observations that have been made to det ect i' . ,g,; wear in this area and/or verify the adequacy of the redasign. l
; l i
l 1 l e
s c 127M gi -- - ~x.,,u METROPOLITAN EDISON COMPANY i u s- .- -- me -,
'OST OFFICE box N 't %NG. PENNSYLVANIA 19603 TELEPHONE 215 - 92t}3601 July 2L , ' 9-2 Gqt 1260 Diree:cr of 'uclear Reacter Regulation Attn: S. A. Yarga, Chief Light Water Reactors Branch No. L U. S. Nuclear Regulatory Cc= mission Washin6 ten, 3. C. 20555 l
Dear Sir:
Three Mile Island Nuclear Statien Unit 2 (TMI-2) Operating License No. DPR-73 l Docket No. 50-320 Small Break LCCA Inclosed please find the results of Babcock and Wilcox's (B&W) most r* cent 1 calculations concerning'a small break LOCA at the reactor coolant pump dis-charge piping for the S&W lower loop 177 FA plants dated July 18, 1978. : Met-Ed and GPUSC have reviewed the enclesed analysis and concur with BhW's ! finding that full compliance with 10 CFR 50.h6 and Appendix K to 10 CFR 50 is clearly demonstrated for operation at pcVer levels of 2TT2 Mv(t). 2.13 enclosure and the appropriate 'BC-2 procedures (Emergency Precedure 2202-13 Loss of Reactor Coolant / Reactor Coolant ?ressure and Operating Procedure 210h-1.2 Makeup and Purificatien Demineralization) described in Met-FA's letters of May 5, 1978 and May 11, 1978 ccepletely satisP.r the cenditions of Provision (1) Section IV cf the Crder of Mcdification of License dated May 26, 1978. Further: ore, the encicsed B&W analysis pro-vides the necessary . justification for operatien at a pcVer level of 2772
- te(t) which has been restricted 2568 ee(t} by Prevision (2) of Section !?.
Therefore, Met-Id requests that the Order of Mcdifica:icn be smended by deleting ?revisiens (1) and (2) of Section 27 cf the Order of Mcdification of License dated May 26, 1978. Sincerely,
@ bd J. ^. Herbein
' lice President-Generatien 06 ke
.$. .a w 1r.& - s '
e ni . er ;y :.w _ . . 1 ' ~~' ~ ' 30 { l e, , 5'S21944534f Ihh
i Babcach&Viilccx e.,,, a.n.,,,,,, c,c,, P.o. Box 1250. Lynchburg. Va. 245C 5 j Telecnone: (804) 384 5111 1 July 18, 1978 l Mr. S. A. Varga, Chief Light Water Reactors Branch #4 Division of Project Management Office of Nuclear Reactor Regulation l l U.S. Nuclear Regulatory Commission l Washington, D.C. 20555
Dear Mr. Varga:
Attached is additional- ECCS small break analyses for B6N's 177 Fuel Assembly Lowered-Loop NSS. These analyses are in accordance with the small break model as approved in BAW-10104A, Rev. 3, " BEN's ECCS Evaluation Model," except for two of the proposed modifications in my letter to you of May 26, 1978. These analyses differ from -those in my letter to you of June 19, 1978, in that the proposed Zaloudek Corre-lation modification was not utili:ed and two additional breaks were analyzed. These analyses, therefore, are intended to
. replace those of June 19, 1973.
A power level of 2772 MNt is assumed in these analyses. 1 Credit is assumed for operator action as described in my letter l to you of May 1, 1978. Break si:es of .04, 055, .07, .085,
. f
)
.10 and .15 ft 2
are examined. These attached analyses, along
.with the break analyses in BAN-10103A, Rev. 3, "ECCS Analysis of BGW's 177-FA Lowered-Loop NSS," constitute a complete spectrum of small break analyses which we believe to be wholly in conformance with 10 CFR 50.46 and 10 CFR 30, Appendix K.
Your expeditious review of this submittal is requested. If you have any questions, please contact me or Henry Bailey (Ext. 2673) of my staff. l j Very ,truly fours, j
. / -
, /
*/. -
J ar"e s !! . Taylor knager. Licensing JtT:dsi Attachment i cc: R. B. Borsum (35W) I
. *a .. *. .-- a on. . . g -t j gy p, f I La,f I WA Q
- 1. Introduction _ !
Analysis of a spectrum of small breaks at the pump discharge has been perfor ed i for B&W's 177-FA lowered loop plants. The small break evaluation model described in BAW-10104, Rev 3, "B&W's ECCS Evaluation Model," along with two of the pro-posed modifications described in the report of May 26, 1978 (J.H. Taylor to S.A. j Varga) was utilized for this study. Operator action is used to achieve suffi- l cient and balanced flow through'all four high pressure injection (HPI) lines. The operator action is described in detail in the report of May 1,1978 (J.H. Taylor,to R.1;. Baer). The analysis contained herein, coupled with the analyses of BAW-10103A, Rev 3, "ECCS Analysis of BSW's 177-FA Lowered Loop NSS," provide an appropriate spec-l trum of breaks for the evaluation of a small leak transient. The results of l the analyses show that the piants can be operated 'up to a power level crf 2772 l MWt within the criteria of 10 CFR 50.46 and Appendix K of 10 CFR 50. i h l 2. Method of Analysis The analysis method used for this' evaluation is that described in Chapter 5 of BAW-10104, Rev 3, "B&W's ECCS Evaluation Model," along with' two of the codi-ficacions described in the report of May 26, 1978 (J.H. Taylor to S.A. Varga). The two modifications utilized were the two node inner vessel simulation and the phase distributional'cultipliers for bubble rise in all control volumes
~ within the reactor vessel. The CRAFT 2 code is used to develop the history of
~
the reactor coolant system hydrodynamics. The CRAFT model uses 20 nodes to i simulate the reactor coolant system, 2 nodes for the secondary system, and one i node for the reactor building. A schematic diagram of the model is shown in Figura 1 siong with the node descriptions. Control volumes (nodes) in and
~
around the vessel are all connected by a pair of flow paths to permit counter-current flow. The breaks analyzed in this report are assumed to be located at the bottom of the cold leg piping between the reacter coolant pump discharge and the reactor vessel. The Wilson, Grenda, and Facterson average bubble rice model is used for all nodes. Within the reactor vessel, however, multipliers I of 2.38 and 2.0 are applied to the calculated bubble rise velocity in the core l nede and the remaining vessel nodes, respectively. The justification for the i use of 2.38 multiplier value in core node is given in Appendix ? of BAU-10104 The report of May 26, 1978 (J.H. Taylor to S.A. Varga) justifies the use of a uitiplier of 2.0 in the dcwucomer, lower rienum, and the upper plar.um re"icn;. I _1_ l l l
The follais.ing assumptions are made for conditions and syste= responses during the accide nt:
- a. The r.. actor is operating at 1027. of the steady-state power level of 2772 LHt .
- b. The lonk occurs instantaneously, and a discharge coefficient of 1.0 is used for the entire analysis. Bernoulli's equation was used for the sub-coold portion of the transient, while Moody's correlation was used in the tva-phase portion. .. -
- c. No of 'it te power is available. 3
- d. The r*0 actor trips on low pressure at 1900 psia.
- e. The s-7fety rods begin enter ng the core af ter a 0.5 second delay from the 1
time the reactor trip signal is reached. 1
- f. The M pumps trip and coast down coincident with . reactor trip. l l
- g. One e c'mplete train of the emergency safeguards system fails to operate, l leavi r.y. two CFTs and only one HPI and one LPI system available for pumped injee : lon to mitigate the consequencbs of a cold leg break.
h.' The 3m(11ary feedwater (FW) system is assumed to be available during the tran s. . c n t . Its main function is to remove heat from the upper half of the steam generator during the initial stages of.the transient. When the sec-endar . tide of the steam' generator becomes a source of heat to the pricary syste'm. the assumption of auxiliary W maximizes the energy that must be relie "ed. _ l 1
- 1. ESFAS signal error band is considered in the analysis to signal the actua- r l
tion ,.f the HPI system.
- j. The , e. ik linear heat generation rate in the hot pin is the taximum allowed j l
by tN- technical specifications at the 10.5 ft level.
- k. Opera :ar action is taken to increase the HPI flows to the intact cold legs at 1; minutes following the ECCS initiation signal. This action is ex-plain ..I more fully in the May 1,1978, report (J.H. Taylor to R.L. Sacr)
Since the .7 AFT calculations showed partial core uncovery for some of the breaks, se . ! fically the 0.055 , 0.07 , and 0.085-ft 2 breaks. a FOA.'! ana1, sis was perf N! to determine the inner vessel ti::ture height. The FCA2!
-2 i p ,m -- - - . - g- 'y
- ,aws -y y er-p +
e-yo.,-,i---4 2
1 void fraction in the lower regions of the core and, similarly to the dis-cussion in item b above, will result in a conservative mixture height. The heat-up calculation was performed using the THETA code in the manner de-scribed in section 5 of BAW-10104. The follouing additional assu=ptions are utilized in the THETA evaluation:
. a. The power shape of Figure 2 was used with a radial power factor of 1.67.
This maximi es steam superheating and sets the peak local power at 10.5 ft at the technical specif1 cation LO'CA limit.
~
- b.
- Coolant floo and mixture level were taken directly from the FOAM calcula- )
tions. As discussed above, the methods utilized in the FOAM calculations result in conseriative values for'these parameters. ,
- c. ,End of life pin pressures were us,ed to conservatively predict the inci- ,
I dence of fuel pin rupture. I
~
- 3. Break Spectrum anii Results j
)
Topical report BAU-10103A, Rev 3, presents the analysis of a CFT line break, j the.0.5 ft2 break at the RC pump discharge and the spectrum of breaks at the
.RC pump suction. As s'hown in that report, the results of those analyses are wh'olly in compliance with the criteria of 10 CFR 50.46 and Appendix K of 10 CFR 50. Those anal'/ses are still valid and conservative in light of the in~
pact of the model modifications. The report of May 26, 1978 (J.H. Taylor to S. A. Va'rga), describes the impact of the modifications. In the present analysis, breaks of 0.04, 0.055, 0.07, 0.085, 0.10, and 0.15 ft2 at the RC puep discharge are evaluated. Figure 3 shows the RC pressure response for each break. As shown, each accident initiates CFT flow within 2000 seconds except the 0.04 ft 2 break. Figure 4 shows (CRAFT) mixture height as a function of time for each break of 1 the spectrum. As can be seen from the figure, minor core uncovery was calcu-a
}
laced for the 0.055 , 0.07 , and 0.085-ft 2 breaks. For the 0.04 , 0.1 , and 0.15-ft2 breaks no core uncovery was calculated and, thus, no tenperature ex-cursions occur. l l The 0.04 ft! break achieves a match up of effective ECCS (the HP: injected l I into the intact cold icgs) uith the core decay heat and the RCS etal heat at Af:;r 30C0 seconds. The core has a mixture height of 13.5 feet at this time. I 1
calculation included all sources of steam production within the vessel, i.e., steam production due to decay heat, flashing, and primary metal heat. To ex-pedite the FOAM analysis, the distribution of the steam sources was chosen to minimi c the complexity of the input calculations and, as described in the sub-sequent paragraphs, results in an underprediction of the swell level. By un-4 derestimating the core mixture height, the core steaming rate will also be un-deresticated; thereby resulting in an overestimation of the steam superheating and the peak cladding" temperature.. The axial power shape shown in Figure 2 was used in the FOAM calculation and was implemented with a radial peaking factor of 1.0. Thus, the resultant mix-i .ture. height is representative of the average channel conditions and is con-servative relative to that for the hot channel. To utilize the power shape in FOAM,.the shape was divided into 26 axial nodes. Steam production due to primary metal' heating and flashing within the inner
~
I vessel was assuced to have a distribution similar.to that for. decay heat. As such, the complexity in th'e input generation for FOAM was reduced to finding an " equivalent decay power" which would generate the same amount of steam as that which is produced from all sources. Use of this steam production shape rdsults in conservative core mixture heights for the following reasons:
- a. When the core is uncovered, some of the steam production due to primary metal heating and flashing would not be used in calculating the mixture i level. Thus, the mixture height would be underestidated.
1 . b. By using this shape, the void fraction at the core inlet is zero. In ac- l tuality, due to steam production in the lower plenum and the subsequent bubble rise into the core, a void fraction vill exist at the core inlet. I Furthermore, this initial core void fraction results in additional bubbles rising throughout the core mixture and increases the entire core void fraction. Since the assumed shape underestimates the core void fraction, the mixture height is underestimated.
- c. Since the axial power distribution is skewed towards the top cf the cere, (see Figure 2) the majority of the steam production will be calculated to occur towards the outlet of the core. Realistically, the total steam
- production due to primary metal heat and flashing would be skaued towards the bottom of the core. The distribution analyzed will underestimate tha 3000 seconds the mixeure level will rise in the core due to excess HPI. For
' breaks smaller than 0.04 ft2, the match up will occur at approximately the j same time and the core mixture levels will drop slower; thus, for all smaller breaks the core will remain covered and the HPI alone can mitigate the tran-sient.
1n performing the analysis, the historical small break spectrum (0.04 , 0.07 , 0.1 , and 0.15-ft breaks) 2 was performed first. As shown by Figure 4, only the 0.07-ft2 break resulted in core uncovery. To further assure that the i 1 worst case had been obtained, the 0.055- and the 0.085-ft 2breaks were analyzed. These cases resulted in some core uncovery but less than that for the 0.07-ft2 break. All three cases were analyzed for temperature response by utiliaing the THETA code; Figure 5 shows the cladding temperature responses. The peak cladding temperature for the worst case break, the 0.07-f t2break, was only_ j 1092F which is well below the 2200F criteria of 10 CFR 50.46. Thus, the analy- ' sis demonstrates that.B&W's 177-FA lowered loop plants can be operated at power levels up to 2772 .Vit and satisfy the ECCS acceptance criteria. , l S 9 l
Figuro 1. I CRAFT 2 Noding Diagrac for Small Breahs I@l m i @ 2-n
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- Feth Sep.e..n.e Ces..f mee. Ip.sp
( . 31 .. _
!o 1:c>. Iden tifica tion Path No. Identification 4
; Downcomer 1,2 Core 1 Lower Plenens 3,4,18,19. Hot Leg Piping 3 Core 5,20
' . 14 Hot Leg Piping Hot Leg, Upper 1,15 6.21 SG Tubes SC & Upper Head 7.22 1
,,16 SC Lover Head '
Steam Cenerator Tubes 8 Core Sypass 4
- 17
, Secondary, SC 9,13.24 l
?, A8 SC Lower Head Cold Leg Piping l 10.14,25 Pumps
,11.19 Cold Leg Piping 11,12.15,16,26,27 Cold Leg Piping 0,A2,20 Cold Leg Piping 17,31 3 Upper Devncoe.ir Ocvnconer 23 LPI i 1 Pressurizer 28,29 Upper Downce=er 1
2 Centainnent 30 3 Upper Plenu Pressuri:er ' 32 Vent Valve 33,34 35,36 Leak & Return Pach HPI 37 Contain:ent Sprays 4
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