ML19344A866

Submits Info Re Overhead Handling Sys Used to Lift Heavy Objects in Vicinity of Spent Fuel.There Are No Heavy Loads Required to Be Moved Over Equipment Required for Safe Shutdown of Plant While Plant in Operation.W/O Encl

ML19344A866
Person / Time
Site:	FitzPatrick
Issue date:	08/03/1978
From:	Gronberg W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T Office of Nuclear Reactor Regulation
References
CON-NRC-78-36, REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR NUDOCS 8008220385
Download: ML19344A866 (8)

Text

{{#Wiki_filter:. POWER AUTHORITY OF THE STATE OF NEW YORK 10 CoLUMaus CIRCLE NEW YORK, N. Y.10019 (212) 397-6200 GronGs v.msnny r soxlex n.cLann g=,R ,,"l,"l,a,=,",, CM A!RM AN LsWIS M. OsNNETT AsstSTANT GENERAL C"ORGs L. INGALLS 4ANAGER GENERAL VIC3 CM AIRM AN KICHAIS M. rLYNN WILBUR L.GRONBsMG ANA R ENG NEER NG ROSX;r L MILLONZI August 3,1973 "" LEC?N O wibuxM r. Lu ' POWER OPER ATION S a INRC-73-3o 7,,,,,,,,,c,,,,,,,, CONTR OLLE R Director cf Nuclear Feactor Regulation United States Nuclear Regulatory Commission Washington, D. C. 20555 THl' DOCUMENT CONTAINS P0OR QUAUTY PAGES Attention: Mr. Thomas A. Ippolito, Chief Operating Reactars Branch No. 3 l Division of Cperating Reactors

Subject:

Jmas A. Fit: Patrick Nuclear Power Plant USNRC Request for Information Regarding Overhead Handling Systems Used to Lift j Heavy Objects in Vicinity of Spent Fuel ( Docket Nc. 50-333

Dear Sir:

In response to your letter of May 17,1978 on the subject of " Control of Heavy Loads Near Spent Fuel", we are submitting the following inform 9 tion as requested.

1. Question:

Provide a diagram which illustrates the physical relation between the reactor core, the fuel transfer canal, the spent fuel storage pool and the set down, receiving or storage areas for any heavy loads moved on the refueling floor. Resocnse: We have enclosed a copy of the diagram requested. This drawing is 11325-FM-101A, Arrangement Reactor Build-ing, Dismantling and Laydown Plan El. 369' - 6.

2. Question: Provide alist of all objects that are required to be moved over the reactor core (during refueling), or the spent 1

fuel storage pool. Fcr each Object listed, provide its appro:ci-mate weight and sice, a diagram cf th= ~~ ament path utilized (including carrying height) and :he iequency of movement. $QNM3h ~?:2200a; } Resconse: A list of all objects that are required to be moved over the reactor core (during refuelirg) or the spent fuel storage pool is shown on the diagram furnished with the response to Question No.1. The steam dryer, Mark M, shown on the Part List on Drawing 11825-FM-101A weighs 37. 5 tons and is 17. 8 feet in diameter and is 13 feet in height. The movement path and carrying height for a fuel cask is des-Oribed in the FSAR. (See reference below) The final storage location of the listed objects is shcan on the drawing. The movement path is from initia11ocation to final location. The carrying height is kept to a minimum.

Reference:

FSAR - Appendix E, Supplement 11, Section E.2. 6, Fuel Cask Drop FSAR -Information Faquested By AEC Letter of December 21,1972 - Question No.1: Reactor Building Crane Movement Restrictions

3. Question:

What are the dimensions and weights of the spent fuel casks that are or will be used at your facility? Resconse: An evaluation and selection of available fuel casks will be made when it becomes necessary to move spent fuel. The spent fuel casks which were used as the basis for design of the Fitzpatrick plant are a follows: Cask Procerties GE ModelIF-300 GE ModelIF-400 Diameter 64 inches 80 inches Length 209 inches 209 inches Weight (Cask Fully Loaded) 70 ton 107. 5 ton

4. Question:

Identify any heavy load or cask drop analyses performed to date for your facility. Provide a copy of all such analyses not previously submitted to the NRC staf. Resconse: The following heavy load or cask drop analyses have been performed to date for the FitzPatrick plant: a) Fuel Cask Drop - FSAR, Appendix E, Supplement 11, J Section E. 2. 6 b) Abscrbing Pad for the Fuel Cask - FSAR, Sucplement 5, Response to AEC Question of 11/29/71-Question 9.3 c) Energy Absorbing Pad for the Spent Fuel Pool - FSE, Supplement 25, Responses to AEC Questions of 1/12/72 - Question 9.10 d) Reactor Building Crane Movement Pastrictions - FSAR, Supplement 20, Response to AEC Request for Information Letter of 12/21/72 - Question 1 e) Loaded Fuel Cask Tilt / Drop - FSAR, Supplement 20, Response to AEC Request for Information Letter of 12/21/72 - Question 2 f) FitzPatrick Nuclear Power Plant Cask Drop Protection System - Document Dated 10/31/74 Transmitted To the NRC by PASNY Letter Dated 11/12/74 And 12/3/74 g) Free Fall Diagrams for Three Basic Size Casks - PASNY Letter Of 11/19/75 in Response to NRC Letter of Request for Information Dated 10/8/75

5. Question:

Identify any heavy loads that are carried over equipment required for the safe shutdown of a plant that is operating at the time the load is moved. Identify what equipment could be affected in the event of a heavy load handling accident (piping, cabling, pumps, etc.) and discuss the feasibility of such an accident affecting this equipment. Describe the basis for your conclusions. Resconse: There are no heavyloads required to be moved over equipment required for the safe shutdown of the plant while the plant is in operativu.

6. Question:

If heavyloads are required to be carried over the spent fuel storage pool or fuel transfer canal at your facility, discuss the feasibility of handling accident which could result in water leakage severe enough to uncover the spent fuel. Describe the basis for your conclusions. Resconse: This has been discussed in FSAR, Appendix E, Supplement 11, Section E.2.6, Fuel Cask Drop and also in FSAR, Supplement 25, Response to AEC. Questions of 1/12/72 - Question 9.10.

7. Question: Describe any design features of your facility which affect the potential for a heavy load handling accident involving spent fuel, e.g., utilization of a single failure-proof crane.

Pasconse: This has been addressed in the submittal by the Authority, January 11,1973 to the AEC in response to the AEC letter of December 21, 1972. The subjects of the information requested and given are as follows: Reactor Building Crane Movement Restrictions Loaded Fuel Cask Tilt / Drop This also appears in FSAR, Supplement 20, Information Paquested by AEC Letter of December 21, 1972 as Question 1 and Question 2. The Authority is conducting a feasibility study and preparing a con-ceptual design for installing a redundant crane trolley on the existing react,;r building Orane. The Authority has a completely fabricated "MPR" Fuel Cask Drop Protection System stored at the FitzPatrick site for possible future use.

8. Question: Provide copies of all procedures currently in effect at your facility for the movement of heavy loads over the reactor core during refueling, the spent fuel storage pool, or equipment required for the safe shutdown of a plant that is operating at the time the move occurs.

Response: The,following procedures are attached to this letter: L D&

• ~

a) Handling of the Channel Cask b) Paassembly of Reactor Vessel After Refueling, c) Disassembly of Paactor Vessel For Refueling, d) Reactor Building Crane

9. Question: Discuss the degree to which your facility complies with the eight (S) regulatory positions delineated in Regulatory C-uide 1.13 (Revision 1, December 1975) regarding Spent Fuel Storage Facility Design Basis.

Restonse: The degree to which the FitzPatrick plant complies with the eight (3) regulatory positions delineated in Regulatory Guide 1.13 is as follows:

1) The spent fuel storage facility is housed in the reactor building and has been designed to Category I seismic requirements.

Reference:

FSAR - Volume IV, Sections 9.3.3, 12.2.3, and 12. 4. 6

2) The facility has been designed to prevent tornado winds and tornado generated missiles from causing significant loss of watertight integrity and from contacting the fuel within the pool.

Reference:

FSAR - Sections 12,3.1 and 12.4. 5 - Has been Addressed in Question 12.16 of the FSAR

3) The reactor building crane is restricted in its movement by interlocks. This has been addressed in the FSAR -

Information Faquested by AEC Letter of December 21, 1972, Question 1.

4) The spent fuel storage facility including the ventilation and filtration system is based on the assumption that the fuel cladding of the rods in more than one fuel bundle might be breached and that higher inventory of radioactive materials is available for leakage from the building than is assumed in Regulatory Guide 1.2.5.

a) The secondary containment is a con' rolled leakage building enclosing the fuel pool con-sistent with Regulatory Guide 1.13 Position C.4. The secondary containment and standby gas treatment system is seismic Category I. (See FSAR Seetions 5. 3. 3. 2, 5. 3. 3. 3, 5. 3. 3. 4, 5.3.4,12.2.2 and 12. 2.3, respecti' rely). The design of the envelope containing, collecting, and filtering the assumed release from a fuel handling accident is entirely seismic Category I. b) The assumptions related to the release of radioactive material are provided in FSAR Section 14.6.4 which was submitted on June 4, 1971, prior to the issuance of Safety Guide 25 on March 23,1972. The FSAR analysis was based on '?lapter 14, Reference 49, a GE topical report and on AEC recoramendations which included a larger number of fuel rod

._s-failures than those in one bundle. The !.iC reviewed and approved the FSAR and, in par-ticular, the Safety Evaluation Report (SER) issued November _20,1972, which includes AEC assumptions regarding the fuel handling accident. The AEC analysis appears to have ~ been based on the newly issued Safety Guide 25. Based on these independent analyses, the SER concluded (p 10-9) "The calculated radiological doses that might result from any of the potential 4 design basis accidents are well within tra guide-lines given in 10 CFR Part 100". For compari-son purposes, these assumptions are listed in the enclosed Table 4.1. The apparent absence of Kr-85 in the SER will not affect the limiting thyroid dose. The design of the filters is therefore conservative.

5) The spent fuel storage facility at FitzPatrick is being modified to include the use of high density storage racks.

This will obviate the moving of fuel casks in the vicinity of the fuel pool until 1989. Item C in Regulatory Position 5 has been fully covered in our response to Question Nos. 6 i and 7, above.

6) The drains and other systems related to the spent fuel pool at FitzPatrick is described in FSAR, Sections 9.3.4.1 and

9. 4.

7) Monitoring equipment is provided to alarm both locally and in the control room if the water level in the storage pool falls below a predetermined level or if high local radiation levels are experienced. The high radiation level instrumen-tation also actuates the filtration system.

Reference:

FSAR - Sections 9. 4 and 9. 4. 3 FSAR - Sections 7.12. 5 and 5.3.3. 4

3) The makeup water for the spent fuel pool is normally pro-vided from the condensate storage tanks to the skimmer surge tan'ts which are components of the Fuel Pool Cooling and Cleanup System. The condensate storage tanks are part of the Condensate Storage System which is classified as seismic Category I.

Reference:

FSAR - Section 9. 4.3

. _ _ _ _ _ _ _ Emergency makeup water may be added through use of the Fire Protection System, a seismic Category I system. Manual fire hose stations - are located on the operatini floor in the vicinity of the fuel pool. ~

Reference:

' FSAR - Section 9. 8 Vecy truly yours, A.

Wilbur L. Gronberg Assistant General Manager-Engineering 1

~, . b e, TABLE 4.1 JAlES A. FITZPATRICK llUCIEAR P0 LIER PLANT COMPAlli 10!! 0F FilMI. Ilar!91,Ill'1 ACCIDMf!T A.I.;ll'D'TIO!!.i SA"ETY GilTDE-?5 AEC.IER 3ECT-~~ .0.3 F;1AR.3ECTION 14.6.4 Time at,which accident Af tor full power at EOL wit.h 24 hrs. art.or shutdown 24 (66) hrs. after occurs a radial. peaking factor of 1.5 op3 ration for 1000d. at full power (1.5 peaking factor) gap act.ivity released 10% of total noble gas other 111 fuel rods (49 rods 111 (O2) fuel rods 1.8% than Kr-85, 305 of Kr-85, per bundle) 0.32%(2(01,)')of nchle gas (compoa ttion) 10j of radioactive lodino 10% of noble gas 10', halogons 10$ of halogena lodine gap inventory 75% inorganic 75% elenental 253 organic 25% organic pool decontanination 100 overall 100 for halogens 100 (10)' for halogens factors 1 for noble gas 1 for noble gas filt.er officiency 90% for inorganic 90% ' for elemental. 90% '/0,', for organic 'M$ for organic 2-hour radiation 2 rom tAvroid (4.38 rem thyroid 1 rom U.B. 0.217 rem W.D.) exposure at nearest, nit.e boundary ?!OTE: (Based on AEC-DIR recommandation) See pige 4.6.26

1. The June 30, 1978 recluost indicates that 1

the clasi. ling in all the fuel rods of one fu31 bunllo abould be nacumed to bo breached. i l

r / J - 4l3 $ U fQf REGULATORY INFORMATION DIGTRIBUTION SYSTEM (RIDS)50-333 DICTRIGUTION FOR INCOMING MATERIAL FEC: IF*CLITO T A - ORG :- GRONBERG W L DOCDATE: 09/0?/79 NRC ~ PWR AUTH OF ST OF NY DATE RCVD: 08/07/79 COPIES RECEIVED 00C TYPE: LETTER NOTARIZED: NO LTR 1 ENCL 0 3U6dEC T: FESPONSE TO NRC'LTP. DTD 05/17/7G. FORWARDING INFO (AS LISTED) RE OVERHEAD HANDLING CYITEMS USED TO LIFT HEAVY OBJECTS IN VICINITY OF SPENT FUEL. W/ATT -INFO.

PLANT NAMC: F IT:' PATRICK - UNIT 1 REVIEWER INIT I AL:

AJM DISTRIEUTOR INITIAL:$L, .i '+44444 n44+444+*i: DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS

• • • • +44444444*+444 CENERAL DIGTRIEUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.

iDISTRIEUTION CODE AOO1) rGR ACTfON: BR CHIEF ORE #3 GC**LTR ONLY(7) INTEFtIAL REG FILE **LTR ONLY(1) NRC PCR**LTR ONIY(1) I A E44LTR'ONLY(2) OELD**LTR ONLY(1) HANAUER**LTR ONLY(1) CCRE PERFORMANCE ER++LTR ONLYI. AD FOR SYS & PROJ**LTR ONLY(1) CNGINEERING BR**LTR ONLY(1) REACTOR SAFETY BR**LTR ONLY(1) PLANT SYSTEMS ER**LTR ONLY'.1) EEE*4LTR ONLY(1) EFFLUENT TREAT SY5+4LTR ONLY(1 J. MCGOUGH*+LTR ONLY(1) - E.XTERNAL: LPDR'S OSWEGO,'NY44LTR ONLY(1) TERA **LTR ONLY(1) NSIC*4LTR ONLY(1) ACES CAT G**LTR ONLY(16) l "12TRI2UT!CN. LTR -1O ENCL 0 CCNTROL NBR: ~92E10C4? i 5IZE: 7PL1P'iOOP >-t4444+ t 4 + > $4 >44 0644 F444444444444 Ct - THE END 4 4 4 -H -* 4 4 4 4 4 4 4 4 4 4 + 4 + + + 4 4 -* 4 -+ + + $ -* -+

POWER AUTHORITY OF THE STATE OF NEW YORK to CoLUMaus CIRCLE NEW YORK. N. Y.10019 (212) 397 4200 "'h?:"""k:::,",, c'*"" ':c=c' ";N / "'N 7 5,, A, A. ,, AN ccOR3 L. INGALLs ANAGER-GENERAL VIC3 CN AaRM AN KICHAM M. FLYNN WILBUR L. GRONSERG sa'N Ao m g a hata NG RosET,7 L MILLONES JOHN 3 STON WILLIAM F. LUDOY Q J N R C 'iJ -3 6 THOM AS F. MCCR ANN. JR. CONTROLLE A Director of Nuclear Paactor Regulation United S'.ates Nuclent Pagulatory Commission Washington, D. C. 20555 s Attention: Mr. Thomas A. Ippolito, Chief C;xarating Reactors Branch No. 3 Division of Operating Paactors

Subject:

James A. FitzPatrick Nuclear Power Plant USNRC Request for Information Pagarding Overhead Handling Systems Used to Lift Heavy Objects in Vicinity of Spent Fuel Docket No. 50-333

Dear Sir:

In response to your letter of May 17,1978 on the subject of " Control of Heavy Loads Near Spent Fuel", we are submitting the following information as requested.

1. Question:

Provide a diagram which illustrates the physical relation between the reactor core, the fuel transfer canal, the spent fuel storage pool and the set down, receiving or storage areas for any heavyloads moved on the refueling floor. Resconse: We have enclosed a copy of the diagram requested. This drawing is 11825-FM-101A, Arrangement Reactor Build-ing, Dismantling and Laydo7m Plan El. 369' - 6".

2. Question: Provide a list of all objects that are required to be moved over the reactor core (during refueling), or the spent fuel storage pool. For each object listed, provide its approxi-mate weight and size, a diagram of the movement path utiliced (including carrying height) and the frequency of movement.

o\\ 0 12300c J9 Resuonse: A list of all objects that are required to be moved over the reactor core (during refueling) or the spent fuel storage pool is shown on the diagram furnished with the response to Question No.1. The steam dryer, Mark M, shown on the Part List on Drawing 11825-FM-101A weighs 37.5 tons and is 17.8 feet in diameter and is la feet in height. The movement path and carrying height for a fuel cask is des-cribed in the FSAR. (See reference below) The final storage location of the listed objects is shown on the drawing. The movement path is from initial location to final location. The carrying height is kept to a minimum.

Reference:

FSAR - Appendix E, Supplement 11, Section E.2.6, Fuel Cask Drop FSAR -Information Paquested By AEC Letter of December 21, 1972 - Question No.1: Paactor Building Crane Movement Restrictions

3. Question:

What are the dimensions and weights of the spent fuel casks that are or will be used at your facility? Resconse: An evaluation and selection of available fuel casks will Ee made when it becomes necessary to move spent fuel. The spent fuel casks which were used as the basis for design of the Fitzpatrick i plant are as follows: Cask Proterties GE ModelIF-300 GE ModelIF-400 Diameter 64 inches 80 inches Length 209 inches 209 inches Weight (Cask Fully Loaded) 70 ton 107. 5 ton

4. Question:

Identify any hea'ry load or cask drop analyses performed to date for your facility. Provide a copy of all such analyses not previously submitted to the NRC staff. Resoonse: The following heavy lead or cask drop analyses have been perforn:ed to date for the FitzPatrick plant: a) Fuel Cask Drcp - FSAR, Appendix E, Supplement 11, Section E. 2. 6 b) Absorbing Pad for the Fuel Cask - FSAR, Surplement 5, Response to AEC Question of 11/29/71-Question 9.3 c) Energy Absorbing Pad for the Spent Fuel Pool - FSAR, Supplement 25, Responses to AEC Questions of 1/12/72 - Question 9.10 d) Reactor Building Crane Movement Restrictions - FSAR, Supplement 20, Response to AEC Request for Information Letter of 12/21/72 - Question 1 e) Loaded Fuel Cask Tilt / Drop - FSAR, Supplement 20, Response to AEC Request for Information Letter of 12/21/72 - Question 2 f) FitzPatrick Nuclear Power Plant Cask Drop Protection System - Document Dated 10/31/74 Transmitted To the NRC by PASNY Letter Dated 11/12/74 And 12/3/74 g) Free Fall Diagrams for Three Basic Size Casks - PASNY Letter Of 11/19/75 in Response to NRC Letter of Request for Information Dated 10/8/75

5. Question:

Identify any heavy loads that are carried over equipment required for the safe shutdovm of a plant that is operating at the time the load is moved. Identify what equipment could be affected in the event of a heavy load handling accident (piping, cabling, pumps, etc.) and discuss the feasibility of such an accident affecting this equipment. Describe the basis for your conclusions. Resconse: There are no heavyloads required to be moved over equipment required for the safe shutdown of the plant while the plant is in operation.

6. Question: If heavy loads are required to be carried over the spent fuel storage pool or fuel transfer canal at your facility, discuss the feasibility of handling accident which could result in water leakage severe enough to uncover the spent fuel. Describe the basis for your conclusions.

Resconse: This has been discussed in FSAR, Appendix E, Supplement 11, Section E.2.6, Fuel Cask Drop and also in FSAR, Supplement 25, Response to AEC Questions of 1/12/72 - Question 9.10. 8

7. Question: Describe any design features of your facility which affect the potential for a heavy load handling accident involving spent fuel, e.g., utilization of a single failure-proof crane.

Fasponse: This has been addressed in the submittal by the Authority, January 11, 1973 to the AEC in response to the AEC letter of December 21, 1972. The subjects of the information requested and given are as follows: Reactor Building Crane Movement Restrictiom. Loaded Fuel Cask Tilt / Drop This also appears in FSAR, Supplement 20, Information Requested 1 by AEC Letter of December 21, 1972 as Question 1 and Question 2. 1 The Authority is conducting a feasibility study and preparing a con-ceptual design for installing a redundant crane trolley on the existing reactor building crane. The Authority has a completely fabricated "MPR" Fuel Cask Drop Protection System stored at the FitzPatrick site for possible future use.

8. Question: Provide copies af all procedures currently in effect at your facility for the movement of heavy loads over the reactor core during refueling, the spent fuel storage pool, or equipment required for the safe shutdown of a plant that is operating at the time the move occurs.

Response: The following procedures are attached to this letter: a) HanT.ing of the Channel Cask b) Reassembly of Reactor Vessel After Refueling c) Disassembly of Reactor Vessel For Refueling d) Faactor Building Crane

9. Question: Discuss the degree to which your facility complies with the eight (8) regulatory positions delineated in Regulatory Guide 1.13 (Revision 1,' December 1975) regarding Spent Fuel Storage Facility Design Basis.

Restonse: The degree to which the FitzPatrick plant complies with the eight (8) regulatory positions delineated in Regulatory Guide 1.13 is as follows:

. 1) The spent fuel storage facility is housed in the reactor building and has been designed to Category I seismic requirements.

Reference:

FSAR - Volume IV, Sections 9.3.3, 12.2.3, and 12. 4. 6

2) The facility has been designed to prevent tornado winds and tornado generated missiles from causing significant loss of watertight integrity and from contacting the fuel within the pool.

Reference:

FSAR - Sections 12.3.1 and 12.4.5 - Has been Addressed in Question 12.16 of the FSAR

3) The reactor building crane is restricted in its movement by interlocks. This has been addressed in the FSAR -

Information Requested by AEC Letter of December 21, 1972, Question 1.

4) The spent fuel storage facility including the ventilation and filtration system is based on the assumption that the fuel cladding of the rods in more than one fuel bundle might be breached and that higher inventory of radioactiva materials is available for leakage from the building than is assumed in Regulatory Guide 1.2.5.

a) The secondary containment is a controlled leakage building enclosing the fuel pool con-sistent with Regulatory Guide 1.13 Position C.4. The secondary containment and standby gas i treatment system is seismic Category I. (See i FSAR Seetions 5. 3. 3. 2, 5. 3. 3. 3, 5. 3. 3. 4, 5.3.4,12,2.2 and 12. 2.3, respeetively). The design of the envelope containing, collecting, and filtering the assumed release from a fuel handling accident is entirely seismic Category I. b) The assumptions related to the release of radioactive material are provided in FSAR Section 14. 6.-4~which was submitted on June 4, 1971, prior to the issuance of Safety Guide 25 on March 23,1972. The FSAR analysis was based on Ohapter 14, Reference 49, a GE topical report and on AEC recommendations which included a larger number of fuel rod

-6_ failures than those in one bundle. The AEC reviewed and approved the FSAR and, in par-ticular, the Safety Evaluation Report (SER) issued November 20, 1972, which includes AEC assumptions regarding the fuel handling accident. The AEC analysis appears to have been based on the newly issued Safety Guide 25. Based on these independent analyses, the SER concluded (p 10-9) "The calculated radiological doses that might result from any of the potential design basis accidents are well within the guide-lines given in 10 CFR Part 100". For compari-son purposes, these assumptions are listed in the enclosed Table 4.1. The apparent absence of Kr-85 in the SER will not affect the limiting thyroid dose. The design of the filters is therefore conservative.

5) The spent fuel storage facility at FitzPatrick is being modified to include the use of high density storage racks.

This will obviate the moving of fuel casks in the vicinity of the fuel pool until 1989. Item C in Regulatory Position 5 l has been fully covered in our response to Question Nos. 6 and 7, above.

6) The drains and other systems related :o the spent fuel pool at FitzPatrick is described in FSAR, Sections 9.3.4.1 and 9.4.

7) Monitoring equipment is provided to alarm both locally and in the control room if the water level in the storage pool falls below a predetermMed level or if high local raxiiation i

l 1evels are experienced. The high radiation level instrumen-l tation also actuates the filtration system. l

Reference:

FSAR - Sections 9.4 and 9.4.3 FSAR - Sections 7.12. 5 and 5.3.3. 4

8) The makeup water for the spent fuel pool is normally pro-vided from the condensate storage tanks to the skimmer l

surge tanks which are components of the Fuel Pool Cooling and Cleanup System. The condensate storage tanks are part of the Condensate Storage System which is classified as seismic Category I.

Reference:

FSAR - Section 9.4.3 ~ l

, i Emergency makeup water may be added through use of the Fire Protection System, a seismic Category I system. Manual fire hose stations are located on the operating floor in the vicinity of the fuel pool.

Reference:

FSAR - Section 9.3 ~ Very truly yours, 4 A Z. Wilbur L. Gronberg 1 Assistant General Mana.ger-Engineering 1 l l t

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UNITED STATES y* NUCLEAR REGULATORY COMMISSION / q 0 WASHINGTON, D. C. 20666 ) I \\ 6 L (.....,/ SEPTEllBER 2 71978 I Docket No. 50-333 Power Authority of the State of New York ATTN: Mr. George T. Berry General Manager and Chief Engineer 10 Columbus Circle New York, New York 10019 Gentl emen: By letters dated June 20, 1978 and June 22, 1978, the Power Authority of the State of New York (PASNY) provided proposed programs for the inspection and/or modification of the feedwater system and control rod hydraulic drive system. In specific, three items.were proposed for FitzPatrick: (1) inspection of botn the feedwater and control rod drive return line nozzles, (2) installation of a new feedwater sparger/ thermal sleeve, and (3) cutting and capping the CRD return line. ) The staff has reviewed the feedwater and control rod drive nozzle inspection and has found this program acceptable subject to the below limitations. Likewise, the staff has reviewed and found acceptable the new feedwater sparger/themal sleeve installation. The portion of your submittal regarding cutting and capping the CRD return line is pending staff review. General Electric Company has been requested to provide additional infomation necessary to justify such a modification. l The feedwater and control rod drive return line nozzle inspection program should proceed based on the following requirements: 4 1. Prior to commencement of work, the FitzPatrick procedure and safety evaluation should be approved by the Onsite Safety Review Committee; 2. At any time during the procedure that the shield plug or any other heavy object is suspended above or in the "eactor vessel, secondary containment integrity and standby gas treatment system operability shall be maintained in accordance with Technical Specification Sections 3.7/4.7B and 3.7/4.7C; 3. Provide documentation of overhead crane handling system reliability transmitted verbally on September 22, 1978; i.e., reactor building crane redundancy and safety margins while handling the shield plug; and 30e 70 oo56102 Tg

Power Authority of the State of New York 4. Provide analysis of impact on the fuel in the unforeseen event of crane malfunctin while removing vessel internals and/or installing the shield plug. Should you have any questions please contact us. Sincerely, e Th Chief Operating Reactors Branch #3 Division of Operating Reactors

Enclosure:

Review of Feedwater and CRD Return Line Nozzle Programs Final Safety Analysis e v

Power Authority of the State of New York cc: Lewis R. Bennett, Assistant General Manager / General Counsel Power Authority of the State of New York 10 Columbus Circle New York, New York 10019 Mr. Peter W. Lyon Manager-Nuclear Operations Power Authority of the State of New York 10 Columbus Circle New York, New York 10019 l Mr. J. D. Leonard, Jr. Resident Manager I James A. Fitzpatrick Nuclear Power Plant P. O. Box 41 1ycoming, New York 13093 Director, Technical Development Programs State of New York Energy Office Agency Building 2 Empire State Plaza ~ Albany, New York 12223 Oswego County Office Building 46 E. Bridge Street Oswego, New York 13126 George M. Wilverding, Licensing Supervisor Power Authority of the State of New York 10 Columbus Circle New York, New York 10019

i 4 j i FITZPATRICK REVIEW OF FW AND CR0 RL N0ZZLE PROGRAMS FINAL SAFETY ANALYSIS The Power Aughority of the State of New York (PASNY), in accordance with the requirements and recommendations published in NUREG-0312, submitted plans for nozzle inspections at the FitzPatrick Nuclear the feedwater (FW) program proposed for inspection and modification of Power Plant. The nozzles was given in a letter from P. J. Early, PASNY, to G. E. Lear, NRC, dated June 22, 1978, and identified by PASNY as JNRC-78-28. The program proposed for inspection and modifica- -tion of the control rod drive return line (CRD RL) and its nozzle was given in a letter from P. J. Early, PASNY, to G. E. Lear, NRC, dated June 20, 1978, and identified by PASNY as JNRC-78-27. It was concluded that the prograns proposed for FitzPatrick-1 by PASNY are acceptable. However, the review was limited to those aspects which bear on the problem of the initiation and propagation of i fatigue cracks in BWR FW and CRD RL nozzles. More specifically, the review followed the lean of NUREG-0312, July,1977, to define the i problem and point to acceptable solutions. The goal was' to assure the continued safe operation of the FitzPatrick huclear reactor pressure vessel. Feedwater Nozzles The program proposed for the FitzPatrick FW nozzles during the 1978 outage can be summarized as follows. The stainless steel cladding 4 will be removed using established machining techniques. The nozzles will be nondestructively examined before and after machining to assure that any and all cracks are removed. New spargers/ thermal sleeves, designed by General Electric, will be installed in each nozzle. i Two of the principal factors in the FW nozzle cracking problem are: (1) the' differential thermal expansion between the stainless steel weld-deposited cladding and the underlying mild steel nozzle forging, i and (2) bypass cold water flow in the annulus between the thermal sleeve and the nozzle. At FitzPatrick the former will be corrected by removing the cladding and the latter by the new three-tube, double interference fit, sparger thermal sleeves. The machining procedure will be essentially the same as was used during the most recent refueling outages at Monticello and Browns Ferry Units Nos.1 and 2. The sparger/ thermal sleeve will be of the same design and workmanship as those recently installed at Browns Ferry Unit No. 2. The clad removal and the new sparger/ thermal sleeve installation are acceptable as proper steps to correct any damage done to date and prevent recurrences. t +-,---.-n .-.,-s r ,,,,,--a

. In addition to executing the plan for inspection, clad removal and sparger installation in an efficient, professional, manner, the licensee must comply with the following. The applicable requirements and recomendations published in NUREG-0312 must be followed. In particular, PASNY should follow the requirements on pages 10 and 11 of NUREG-0312 relative to reparting promptly the inspection results. However, the request to delay sparger assembly until the inspection results have been discussed with the NRC is unnecessary unless the extent on the depth of fatigue cracking is unusually severe. PASNY is expected to prepare and submit to the NRC a final report which must include the inspection results, the clad removal details, reanalysis of the nozzles in accordance with ASME Code requirements (reinforcement, fatigue, etc.) and sparger installation details. A formal report should be prepared and submitted within 90 days following the end of the refueling outage. Control Rod Drive Return Line Nozzle The program proposed for the FitzPatrick CRD RL nozzle during the 1978 outage can be sumarized as follows. The line will be cut and capped and the system will be operated in the isolated mode in accord with the GE SIL 200, Supplement 2. The surface of the nozzle (bore and blend radius) and of the apron area on the vessel wall beneath the nozzle will be prepared by flapper wheel grinding then a liquid penetrant (PT) inspection will be performed. Any crack-like indications will be removed by local hand grinding followed by additional PT inspection to assure that no cracks remain. The problem of cracking in CRD RL nozzles is similar in many respects to that in the FW nozzles. Differential thermal expansion between weld deposited stainless steel cladding and the mild steel nozzle and the thermal transients induced by the influx of cold water are important factors. The inspection program proposed by PASNY relative to the CRD RL nozzle is acceptable. The PT inspection must include the nozzle bore, blend radius and the " apron" region of the vessel shell below the nozzle. PASNY must include the details of the inspection and modification in the final report mentioned in conjunction with the FW nozzle work, above. The cutting and capping of the CRD return line is pending staff review. l

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.....s G wlL.UR, L.,GRON t t flCHARD M.'PLYNN 013RRT L MILLONIl a' aria== wituaM r. Luooy January 9, 1979 JPN-7 9 -2 m owA;, g y,a, ANN.JR. Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Thomas A. Ippolito Operating Reactor Branch No. 3 Division of Operating Reactors

Subject:

James A. FitzPatrick Nuclear Power Plant Responses to NRC Questions Regarding Feedwater Sparger Modification Docket No. 50-333

Dear Sir:

Transmitted herewith are responses to your letter dated September 27, 1978 which forwarded four questions regarding the feedwater sparger modification work. The feedwater sparger mod'fication work was performed during the refueling outage which commenced in September 1978. very truly yours, ~ Paul J. Early ^ Assistant Chief Engineer-Projects l O% 790117015 ? l

QUESTION 1: Prior to commenccmsnt of worx, ths Fit-Patrick procedure and safety evaluation should be approved by the Onsite Safety Review Committee.

RESPONSE

The FitzPatrick procedure and safety evaluation were approved by the Plant Operating Review Committee prior to commencement of feedwater sparger modification work. QUESTION 2: At any time during the procedure that the shield l plug or any other heavy object is suspended above or in the reactor vessel, secondary containment integrity and standby gas treatment system operability shall be maintained in accordance with Technical Specification Section 3.7/4.7B and 3.7/4.7C.

RESPONSE

Secondary containment integrity and standby gas treatment system operability were maintained while the shield plug or Peavy object (i.e. reactor inte rnals) was suspended above or in the reactor vessel. j QUESTION 3: Provide documentation of overhead crane handling system reliability transmitted verbally on September 22, 1978; i.e., reactor building crane redundancy and safety margins while handling the shield plug.

RESPONSE

The FitzPatrick Plant overhead crane has a rated capacity of 125 tons with one 125 ton main hook and one 20 ton auxiliary hook. Preventive Maintenance (P.M.) on the crane was carried out just prior to the commencement of the outage. The P.M. included visual inspection for brake wear, rope wear, limit switches actuation, etc., complete electrical checkout and lubrication. The main hook of the crane was magna fluxed prior to the outage. The main hook is a sister hook (Two Js back to back)- and is supported by 14 individual cables. Each of these cables is 1 1/8" Diam. and has 6 x 46 IWRC* strands; each cable has a certified breaking crength of 122,000 lbs. To each side of the hook is attached a 5" Diam. pear shaped steel link with a rated capacity of 77 tons. Each paar link has a minimum breaking strength of 198 tons and is certified to be tested for 148,000 lbs. each. To each pear shaped link are attached two wire rope slings. These wire rope slings are 2" Diam. 5x19 class wire rope with a 198 ton minimum breaking strength and with heavy wire rops thimbles and torpedo collars at each end. The free end of the sling is attached to a 12"x2 3/4" Jaw and Jaw Turnbuckle rated at 187 ton minimum breaking strength. Attached to each turnbuckle is a 3" i

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anchor shackle with bolt pins and nuts each having a 375 ton minimum breaking strength. Each shackle Attached has a safe working capacity of 75 tons. to each of these shackles is a 1 3/4" size anchor Each of these shackle with a screw pin arra..:-ment. shackles, which has a minimum breaking strength of 150 tens and is proof tested for 50 tons, is attached to one of the four (4) lifting eyes of the concrete The platform is estimated to weigh 51 tons. platform. It was determined that the reactor building crane safety margin is greater than six (6). The lifting sling safety margin is five (5). Provide analysis of impact on the fuel in the unforeseen QUESTION 4: event of crane malfunction while removing vessel internals and/or installing the shield plug.

RESPONSE

Vessel internals (mositure separater, dryer, etc.) are routinely removed during refueling outage at every BWR plant. Therefore, the Authority believes that the analysis of impact on the fuel in the unforeseen event of crane malfunction while removing heavy objects should be performed on a generic basis. As per our telephone conversation with Mr. Philip Polk of your office,the Authority requests further direction from the Commission for this item. h - _ _,}}