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Category:ACRS Summary Report
MONTHYEARML17037C4771974-11-15015 November 1974 Letter Enclosing Supplement No. 1 to the Safety Evaluation Report for the Application for a Full-Term Operating License ML17037C2071974-09-25025 September 1974 Letter Responding to a September 17, 1974 Letter Requesting a Response to Two ACRS Identified Items Concerning Main Steam Isolation Valve Leakage and Inservice Inspection of the Reactor Pressure Vessel and Advising That Inform. ML17037C4801974-09-13013 September 1974 Letter Transmitting ACRS Report Dated September 10, 1974, on the Application for Conversion of the Provisional Operating License to a Full-Term Operating License for the Nine Mile Point Nuclear Station Unit ML17037C4901974-02-20020 February 1974 Letter Submitting Copy of Report of the Advisory Committee on Reactor Safeguards Dated February 12, 1974 on Expected Performance of General Electric 8x8 Fuel Design for Reload Use ML17037C3691971-03-12012 March 1971 Letter Responding to a February 11, 1971 Forwarding an ACRS Report Pertaining to the Request for an Increase from 1538 to 1850 Thermal Megawatts in the Licensed Power Level ML17037C3501970-06-19019 June 1970 Letter Authorizing the Return to Operation of Unit 1 in Accordance with Your Documented Program and Subject to Conditions Defined and Enclosing an ACRS Report Dated June 16, 1970 1974-09-25
[Table view] Category:Letter
MONTHYEARNMP1L3570, Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2024-02-0101 February 2024 Supplemental Information Letter - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation IR 05000220/20230042024-02-0101 February 2024 Integrated Inspection Report 05000220/2023004 and 05000410/2023004 NMP1L3569, CFR 50.46 Annual Report2024-01-26026 January 2024 CFR 50.46 Annual Report ML24004A2122024-01-0808 January 2024 Senior Reactor and Reactor Operator Initial License Examinations ML23354A0012024-01-0404 January 2024 Exemption from Select Requirements of 10 CFR Part 73 (EPID L-2023-LLE-0059 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting)) ML23278A1292023-12-14014 December 2023 Units 1 & 2; Limerick, Units 1 & 2; Nine Mile Point, Units 1 & 2; and Peach Bottom, Units 2 & 3 -Revision to Approved Alternatives to Use Boiling Water Reactor Vessel and Internals Project Guidelines NMP1L3566, Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station2023-12-14014 December 2023 Radiological Emergency Plan Document Revision. Includes EP-AA-1013, Revision 10, Radiological Emergency Plan Annex for Nine Mile Point Station IR 05000410/20243012023-12-14014 December 2023 Initial Operator Licensing Examination Report 05000410/2024301 ML23305A1402023-12-13013 December 2023 Units 1 & 2; Nine Mile Point, Unit 2; Peach Bottom, Units 2 & 3; and Quad Cities, Units 1 and 2 - Issuance of Amendments to Adopt Traveler TSTF-580 NMP1L3564, Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements2023-12-0707 December 2023 Supplemental Response to Part 73 Exemption Request - Withdrawal of Request for Exemption from 10 CFR 73, Subpart B, Preemption Authority Requirements ML23291A4642023-12-0707 December 2023 Issuance of Amendment No. 251 Regarding the Adoption of Title 10 the Code of Federal Regulations Section 50.69, Risk-Informed Categorization and Treatment of SSC for Nuclear Power Plants ML23289A0122023-12-0606 December 2023 Issuance of Amendment No. 250 Regarding the Revision to Technical Specifications to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b NMP1L3563, Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-12-0404 December 2023 Submittal of Relief Request I5R-12, Revision 0, Concerning the Installation of a Full Structural Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) IR 05000220/20234022023-11-28028 November 2023 Security Baseline Inspection Report 05000220/2023402 and 05000410/2023402 NMP1L3557, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-11-22022 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23317A1192023-11-10010 November 2023 Constellation Energy Generation, LLC - 2023 Annual Report - Guarantees of Payment of Deferred Premiums IR 05000220/20234202023-11-0101 November 2023 Security Baseline Inspection Report 05000220/2023420 and 05000410/2023420 ML23305A0052023-11-0101 November 2023 Operator Licensing Examination Approval IR 05000220/20230032023-10-25025 October 2023 Integrated Inspection Report 05000220/2023003 and 05000410/2023003 IR 05000220/20235012023-10-17017 October 2023 Emergency Preparedness Biennial Exercise Inspection Report 05000220/2023501 and 05000410/2023501 IR 05000220/20230112023-10-16016 October 2023 Comprehensive Engineering Team Inspection Report 05000220/2023011 and 05000410/2023011 RS-23-097, Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans2023-10-12012 October 2023 Constellation Energy Generation, LLC, Advisement of Leadership Changes and Submittal of Updated Standard Practice Procedures Plans NMP1L3554, Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases2023-10-0606 October 2023 Submittal of Revision 28 to the Final Safety Analysis Report (Updated), Fire Protection Design Criteria Document, 10CFR50.59 Evaluation Summary Report, 10CFR54.37(b) Aging Management Review, and Technical Specifications with Revised Bases C IR 05000220/20233032023-09-20020 September 2023 Retake Operator Licensing Examination Report 05000220/2023303 ML23250A0822023-09-19019 September 2023 Regulatory Audit Summary Regarding LARs to Adopt TSTF-505, Rev. 2, and 10 CFR 50.69 ML23257A1732023-09-14014 September 2023 Requalification Program Inspection IR 05000220/20230052023-08-31031 August 2023 Updated Inspection Plan for Nine Mile Point Nuclear Station, Units 1 and 2 (Report 05000220/2023005 and 05000410/2023005) RS-23-080, Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs2023-08-30030 August 2023 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-264-A, Revision 0, 3.3.9 and 3.3.10 - Delete Flux Monitors Specific Overlap Requirement SRs NMP2L2851, Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations2023-08-25025 August 2023 Relief Request Associated with Successive Inspections for Generic Letter 88-01 / BWRVIP-75-A Augmented Examinations ML23151A3472023-08-21021 August 2023 Issuance of Amendments to Adopt TSTF-295-A, Modify Note 2 to Actions of PAM Table to Allow Separate Condition Entry for Each Penetration NMP1L3534, License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation2023-08-18018 August 2023 License Amendment Request - Revision to the Technical Specifications Design Features Sections to Remove the Nine Mile 3 Nuclear Project, LLC, Designation ML23220A0262023-08-0808 August 2023 Licensed Operator Positive Fitness-for-Duty Test IR 05000220/20234012023-08-0808 August 2023 Cyber Security Inspection Report 05000220/2023401 and 05000410/2023401 (Cover Letter Only) NMP1L3545, Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems .2023-08-0404 August 2023 Supplemental Information Letter to Adopt TSTF-505, Provide Risk- Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 and 10 CFR 50.69, Risk-informed Categorization and Treatment of Structures, Systems . RS-23-087, Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor2023-08-0404 August 2023 Revision to Approved Alternatives Associated with the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor IR 05000220/20230022023-08-0101 August 2023 Integrated Inspection Report 05000220/2023002 and 05000410/2023002 ML23207A0762023-07-14014 July 2023 EN 56557 - Update to Part 21 Report Re Potential Defect with Trane External Auto/Stop Emergency Stop Relay Card Pn: XI2650728-06 NMP1L3544, Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations2023-07-14014 July 2023 Fifth Inservice Inspection Interval, First Inservice Inspection Period 2023 Owner'S Activity Report for RFO-27 Inservice Examinations ML23186A1642023-07-0606 July 2023 Operator Licensing Retake Examination Approval NMP2L2846, Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications2023-07-0505 July 2023 Nine Mire Point Nuclear Station, Units 1 and 2, General License 30-day Cask Registration Notifications ML23192A0622023-06-30030 June 2023 Engine Systems, Inc., 10CFR21 Reporting of Defects and Non-Compliance, Report No. 10CFR21-0136, Rev. 0 IR 05000220/20230102023-06-29029 June 2023 Biennial Problem Identification and Resolution Inspection Report 05000220/2023010 and 05000410/2023010 ML23131A4242023-06-23023 June 2023 Issuance of Amendment No. 249 Regarding the Revision to Technical Specification 3.3.1 to Adopt Technical Specifications Task Force Traveler TSTF-568 RS-23-077, Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations2023-06-16016 June 2023 Response to NRC Regulatory Issue Summary 2023-01, Preparation and Scheduling of Operator Licensing Examinations NMP1L3539, Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208)2023-06-0909 June 2023 Day Commitment Response - Relief Request I5R-11 Concerning the Installation of a Weld Overlay on Reactor Pressure Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle Dissimilar Metal Weld (32-WD-208) ML23159A0052023-06-0505 June 2023 56557-EN 56557 - Paragon - Redlined RS-23-042, Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling2023-05-25025 May 2023 Application to Revise Technical Specifications to Adopt TSTF-580, Provide Exception from Entering Mode 4 with No Operable RHR Shutdown Cooling IR 05000410/20233022023-05-15015 May 2023 Initial Operator Licensing Examination Report 05000410/2023302 2024-02-01
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v'EB Docket No. 50-220 20 1974 M.agar a R)hawk Power Corporation ATlM: R. Philip D. Raymond Vice President sneering 300 Erie Boulevard West Syracuse, New York 13202 Gentlel1)en:
A copy of the report of the Advisory Committee on Reactor Safeguards dated February 12, 1974, is enclosed for your information. The report reflects the Coamhtee's review of the design and expected perfora)ance of Genera1 E1ectric 8 x 8 fuel bundles to be used in part;ial and flQ1 core reloads for boilirg water reactors such as your Nine M.le Point Nuclear Station Urd.t l.
You will be advised when we have conyleted our review of your request for operation with 8 x 8 fuel.
Sincerely, Donald J. Skovholt Assistant Director for Operating Reactors Directorate of Licensing
Enclosure:
Distribution ACHS Report RNDiggs Docket File RO (3)
NDube AEC PDR OGC cc: See next; page l4Jinks Local PDR VStello Bsch~f (ig) RP Reading JGallo SKari Branch Reading JSaltzman SVarga JRBuchanan, OHNL FLIngram HBoyd 93Abernathy, DTIE HINueller ECase WNoore, L JWagner, OC DTSkovholt, L:OR WOM.lier, DRA TJCarter, L:OR PCollins, L:OLB DLZiemann, L:ORB g2
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Form hFC-318 (Rev. 9.53) LCM 0240 CRI3 C@3 13 11423 I 320 204
II i;,'iagarn lfohawk Power Corporation February 20, 1974 cc w/enclosure:
J. Bruce HacDonald, Esquire Deputy Commissioner and Counsel
.New York State Department of Commerce and Counsel to the Atomic Energy Council 99 Washington Avenue Albany, New York 12210 Arvin E. Upton, Esquire LeBoeuf, Larry, Leiby 6 HacRae .
1757 N Street', N. W.
Washington, D. C. 20036 Anthony Z. Roisman, Esquire Berlin, Roisman and Kessler 1712 N Street, N. W.
Washington, D. C. 20036 Hr. Donald X. Gleason, Chairman County Office Building
/>6 East Bridge Street Oswego, New York 13126 Hr."Robert P. Jones, Supervisor Town of Scriba R. D. II4 Oswego, New York 13126 Oswego City Library
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNlTED STATES ATOMlC ENERGY COMMlSSlON WASHINGTON, D.C. R054b February 12, 1974 Hanorable Dixy Lee Ray Chaiznsn U. S. Ataoia Emmgy Gmni'ssion 54545 8dxJect: HEX'ORC CN GENEBAL ELECXREC 8XS PGEL XESX(Ã K)R HEZQAD USE'ear Dx'. Bays At its 166th meeting, XhhemLxy 7<<9, 1974t the Advisory CGHlAittee an Reactor Safeguazds campleted a gexxeric review of the design and ex.-
in partial and
'te pand peeiarxtlaxxce of General Electric 8xS fuel bunclles to be splayed full care reloads topics were discussed at the 165th and 555 in Denver, in boiling water reaclurs.
ACRS
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These meeting on January 10-12, 1974t Colorado, an January 24, 1974. During its ravines the Ommi.ttee had the bermCit. of discussians with representatives of the 5574, Genaml ZXa~ic Gxnpxxzy, the AEC Regu3atozy Staff, and of the docu-maxts listed beld.
The General Electric SxS fuel assalhly consists of 63 fuel xocls plus one unfueled; water-f le@, spacex"~ptxxre xod in a sctuare Sx8 bundle array within a- squaxm channel box. The design of the fuel xcds and fuel xcd bundle in the SxS reload fuel assembly is, except far diffex ences in the length. of the fuel and gas plenam, the same as in the 8x8 to in the Caanittee's n~~6 fuel BSSHllbly used 3Il the bundles are interchangeable bo113LIlg water xitor concept referred of Sex~her 21, 1972. The SxS fuel in General Electric boiling water reactors with the peeviously used 7x7 bundles.
Xh general, the thexual margins to fuel damage design bases are greater far SxS fuel than far'x7 fuel. %he design value af the linear heat gemmation rate far normal opexmtion is 33.4 kw/ft far SxS fuel and 17.5 to 18.5 lee/ft far.7x7 fuel specific power is slightly greater far the 8xS fuel than for the 7x7 fuel. The Genera. Electric Catapany believes the 3xeer linear heat generation rate and slightly greater ratio of clad tlxxc1cness to xad dieter shouM result in fewee failures in SxS ftxel than in 7x7 fuel. Although the hyQraulic resistaxxce of Sx8 bunDes is slightly greater than that of 7x7 buxxllm, the thecal-hydraulic pex-faxmance of boxes either partially or fully loaded with SxS assemblies is not chsgnlded relative to cares loaded with 7x7 assemblies.
icy Lee. Ray l~
Emeable February 12, 1974 Since the U-235 enrichnents for the indivMmQ. fuel rods, the t lh ratio are similax'n the SxS
~l and llggtlhd, Ith 7x7 number designs, the neutnnic behavior of the tsm designs is not significantly diffanmt. The intexnally located water xod in the SxS design reduces xod-to-xod power peaking.
7x7 t '\PlbWbl. Ih~ fl'hd Since the nextmania and fuel huxG.es are themal~rauLic characteristics of similar, thorax behavior under ahnoxnsl SxS and operatianal cantxol xod drop,'uel handling, and steam line break acx68ents are nat exited to be significantly different for the two fuel designs.
II ~51 'WWW, 'l kU'tV~
water reactors, and the Regulatory Staff have perfamtect analyses of the h~
behavior of 8x8 fuel in several mixed and fully reloaded cca~ under txansient and accident cxmditions. These analyses ymBict that,, far boiling water mectxuw which have jet pumps, the peak clad temperatures during a po.~~ted large-break IQCA using the Xnterim Acceptance Criteria axe less for SxS than for 7x7 assariblies. ERwmrer, for large postulated breaks in nan-)et pump plants and far small and intexaectiate size breaks far all plants, the pr~x8 peak clad temperatures are in the same range far both SxS a+i 7x7 fuel. Consequently, indivtitual reviews by the Regu-latory Staff of the equi perfarmmce of SxS reload fuel, incluKLng plantmpecific system effects azxl any significant core fuel loadLing il1b I If h I f w H 8xS fuel to deteexnine limits an reactor operating conditiorm. The Cce-th mittee wishes to be irdhraed concerning the results of these reviews for the initial SxS fuel load:tngs in each of the several General Electric boiling water reactar product lines.
IlE!fS16 N ~~~
The Regulatory 8 ~~~I Staff plans to use the results of recently canducted
'll analytical nedels far SxB fuel. The Qanaittee wishes to be kept infcemd.
and hun31es whose designs brac]set the dimensians of the SxS fuel and an the perfarmance of cares cantaining mixbmes of assaablies with different rmobers and sizes of fuel rcds. Although mechanical tests have been perfaxned'an 8xS fuel assemblies anR campanents to daaanstrate their integri.ty, additional tests to vex'ify spacer grid strength have been requested by the Regulatory Staff. This matter should be resolved in a niner satisfactary to the RecpQatoxy Staff.
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ly RIMmable Dixy Lee Hay Fehrum~ 12, 1974 The Aneeal program and wishes The Hegulatory
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Electric Caqxmy has planned a precgam of both pre- and post-inmHatian enauninatians to monitor the performance of 8xS fuel to be kept informed.
!.1l Staff is currently revhm9xg the treatment. of uncex tainties in the establishment and mmi;toring of operating limits far boiling water reactors. The Ccxanittee wishes to be kept infarneQ.
He-evaluation of care operating limits will be necessary as a result of the recently gcanulgated Accephmce Criteria for Emergency Core Cooling Systems. The Ganmittee vrishes to be kept infarrneci.
The ~ttee believes that, with due reganl to the abave comments, incluQiLng the individual reload reviews by the Hegulatory Staff which will address specific plant features, the Genera. Electric 8x8 fuel assanblies are acceptable for use in the reload of General Electric boiling water xeactxam.
St
&E.
- m. H. Stratton Chaixaen Befezences attached.
Honorable DDT Lee Ray Pebruary 12, 1974 "Dresden 3 Nuclear Power Station, Second Reload License Suhnittal",
General Electric Co., Nuclear Puel D@lartment, SepteInber 1973@ and SuppleIRRIt Ag RR%6her 17 ~ 1973~ Supplement B~ DeceH58K' p 1973 Supplatent C, Decanber 6, 1973; Supplement D, Deliber 17, 1973; Supplement Ec De~Mr 17'973'upplement Pr Januazp's G, Januaxy 9~ 1974) Supplement H, January'23, 1974 1974'upplement 20 -
"Nine Mile Point Unit 1 Semnd Refueling", P. D. RayIImd to A.
Giambusso, September 14, 1973
'Nine Mi1e Point Unit 1 Safety Analysis for Type 5 and Type 6 Reload Puel g Niagara Hat'uvr8c Power ~aratiang October 15, 1973 "Nine Mile Point Unit lr Part lc Nan<<Proprietary Response and Part 2, Proprietary Respanse," Jamlary 15, 1974 3Q Monticello Nuclear Generating Plalltg PexlIBnent Plant Changes to Accamncdate F~ilibrium Core Scram Reactivity Insertian Charac-teristics", January 23, 1974 4 NED0-20103, "Qerumal Design InfoxIoatian for General Electric Boiling Water Reactor Reload Fuel Camellcing in SpriIlg, '74", September 1973
- 5. H. E. Wi11iamsan aIld D. C. Ditmore, "Experience with.HWR Fuel 9hrough Septarher 1971", NED0-10505, May 1972
- 6. GEAP-4059, "Vibration of Puel Rods in P~el Plow", E. P. Quinn, July 1962 70 Letter, J. A. Hinds to V. Nmre,'ebruary 4, 1974, "P~lrietary Infoxmation, SxS Puel Wear Tests" X~. J A, Hinds to V veillance PIugram Moore Pehnlary 4 1974 SXS Pue1 Sur-8 NEM-10735, "Densification Cansiderations in HNR Fuel Design and Per foxmallce", D. C. Ditmare and R B. E13dns, Decaaber 1972/ Supplement 2, "Response to AEC Questions, NEW&-10735 Apri1 1973 "(Proprietary) g Supplement 2, "Response to AK'. Questions, NEDDY-10735 Supplement 1",
May 1973'Proprietary), Supplement 3, "Respanse to AEC Questions, NEDH 10735@ Supplement 1 g June, 1973 (Proprietary), Supplement 4 g "Responses to AEC Questions NEXT-10735", tuly 1973, (Proprietary) f Supplement 5'g Densif xcatiaIl. Consideratians 1n HNR Fuel 'g July 1973 (Proprietary).g SUpplements 6j 7g aIKl Sg Fuel DensificatiaIl Effects an GeneIzL Electric HoiUing Water Ppm~ Fuel" ~
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Hceozab3a Dixy'ee Ray February 12, 1974, References 'contend)
NEDO 20181'uppl6fMlt "A J4xM.
1g December 3t 1973 (Pxoprj~~) ~GAP III g
for the Prediction of Pellet Conductance in BNR Fuel Rods"
- 10. "Technical Report an Densification of Gm~ Electric Reat'~ Fuels",
U. S~ Atomic Entry Camaission, August 23, 1973, and "Supple>rent 1 to the Technical Report on Densification of Canexal Electric Reactor Fuels", December 14, 1973
- 11. "Sensitivity Stay on ~6 Fuel Bunts Response C. M. Maser and R. W. Grume, Deaeaher 1973 to a PoshQated IDGE, 12 .NEIPr 10801, "Mod~ the E+/6 Los~&Dcelant Accident" Core Spryly and Bottxxn Floodix@ Heat Transfer Effectiveness", J.. D. Duncan and J. E. Leenaxd, 'Maxch 1973, and "Response to AEC Ret:gmst for Additional Information an NEDE-10801", May 1973 9'mpa5.etary),
13, . ME0-10993, "Core Spray and Bottcm Flooding Effectiveness J. D. Durban and J. E. Leceard, Segteaher 1973 in the ~6" g
- 14. "Core Tkmenal'Analyses of a Stainless Steel Clad Heater Rod Bundle",
C. M. user and R. W. Griebe, December 1973
- 15. "Status RepI~'n the General Electric Company Sx8 Fuel Assembly",
January 18, 1974, USAEC Regu1atory Staff
- 16. "Technical Report on the General Electric Oarttpany SxS Fuel Assembly",
February 5, 1974, USMC Regulatory Staff
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