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Category:FUEL CYCLE RELOAD REPORTS
MONTHYEARML20154Q3371998-09-30030 September 1998 Rev 8 to Oyster Creek Cycle 17,COLR ML20134F7381996-09-17017 September 1996 Rev 7 to Topical Rept - 066, Oyster Creek Cycle 16 Colr ML20078E7471994-09-24024 September 1994 Rev 6 to Oyster Creek Cycle 15 COLR Topical Rept 066 ML20069Q4751994-05-27027 May 1994 Rev 5 to Topical Rept - 066, Oyster Creek Cycle 14 Colr ML20128L5891992-12-24024 December 1992 Rev 4 to Oyster Creek Cycle 14 Colr ML20077G8941991-05-21021 May 1991 Rev 3 to Topical Rept TR-066, Oyster Creek Cycle 13 Core Operating Limits Rept ML20072Q3901990-11-20020 November 1990 Rev 2 to Core Operating Limits Rept 066 ML20065D5441990-07-11011 July 1990 Rev 1 to Oyster Creek Cycle 12 Core Operating Limits Rept ML20042G2431990-02-22022 February 1990 Rev 0 to Core Operating Limits Rept. ML20155D0311988-08-30030 August 1988 Reload Info & SAR for Oyster Creek Cycle 12 Reload ML20151A4221988-03-31031 March 1988 Rev 0 to Reload Info & SAR for Oyster Creek Cycle 12 Reload ML20210H6851986-09-17017 September 1986 Rev 1 to App D to, GE Reload Fuel Application for Oyster Creek ML20154S5951986-01-31031 January 1986 Rev 0 to Methods for Analysis of BWR Steady State Physics ML20138J7911985-11-13013 November 1985 Rev 0 to Methods for Analysis of BWRs Lattice Physics ML20084P4501984-05-14014 May 1984 Errata to GE Reload Fuel Application for Oyster Creek ML20081K8231983-08-31031 August 1983 Errata & Addenda for Amend 6 to GE Reload Fuel Application for Oyster Creek ML20081K8221983-07-31031 July 1983 Errata & Addenda for Amend 5 to GE Reload Fuel Application for Oyster Creek ML19305E3121979-08-31031 August 1979 Reload Fuel Application. ML20024D9161979-08-31031 August 1979 Errata & Addenda for Amend 4 to Reload Fuel Application. 1998-09-30
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217K4451999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Oyster Creek Nuclear Generating Station.With ML20211P6731999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Oyster Creek Nuclear Generating Station.With ML20211A7051999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Oyster Creek Nuclear Station.With ML20209G0631999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Oyster Creek Nuclear Generating Station.With ML20212H5491999-06-18018 June 1999 Non-proprietary Rev 4 to HI-981983, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool ML20195E7961999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Oyster Creek Nuclear Generating Station.With ML20206N7431999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Oyster Creek Nuclear Generating Station.With ML20205P5401999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Oyster Creek Nuclear Generating Station.With ML20204C8201999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Oyster Creek Nuclear Generating Station.With ML20199E4671998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Oyster Creek Nuclear Generating Station.With ML20195E8321998-12-31031 December 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Sys & Procedures, for Period of June 1997 to Dec 1998.With ML20198D2091998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Oyster Creek Nuclear Generating Station.With ML20195J8591998-11-12012 November 1998 Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan ML20195C4271998-11-0606 November 1998 Safety Evaluation Supporting Proposed Ocnpp Mod to Install Core Support Plate Wedges to Structurally Replace Lateral Resistance Provided by Rim Hold Down Bolts for One Operating Cycle ML20155J3021998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Oyster Creek Nuclear Generating Station.With ML20154R4981998-10-20020 October 1998 Core Spray Sys Insp Program - 17R ML20154L3051998-10-14014 October 1998 Safety Evaluation Accepting Licensee Request to Defer Insp of 79 Welds from One Fuel Cycle at 17R Outage ML20154Q3371998-09-30030 September 1998 Rev 8 to Oyster Creek Cycle 17,COLR ML20154L5571998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Oyster Creek Nuclear Generating Station.With ML20151V6311998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Oyster Creek Nuclear Generating Station.With ML20237D5691998-08-31031 August 1998 Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification ML20237D5711998-08-18018 August 1998 Rev 0 to SE-000222-002, Core Plate Wedge Installation ML20237B0131998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Oyster Creek Nuclear Generating Station ML20236R0511998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Oyster Creek Nuclear Generating Station ML20249B2981998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Oyster Creek Nuclear Station ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted ML20247F1891998-05-0505 May 1998 Risk Evaluation of Post-LOCA Containment Overpressure Request ML20247G0581998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Oyster Creek Nuclear Generating Station ML20216K0341998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Oyster Creek Nuclear Generating Station ML20151Y4651998-03-31031 March 1998 Non-proprietary Version of Rev 1 to GENE-E21-00143, ECCS Suction Strainer Hydraulic Sizing Rept ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20212E2291998-03-0404 March 1998 Rev 11 to 1000-PLN-7200,01, Gpu Nuclear Operational QAP, Consisting of Revised Pages & Pages for Which Pagination Affected ML20216J0841998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Oyster Creek Nuclear Generating Station ML20203B2781998-02-16016 February 1998 10CFR50.59(b) Rept of Changes to Oyster Creek Systems & Procedures ML20203A3801998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Oyster Creek Nuclear Generation Station ML20198P1791997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Oyster Creek Nuclear Station ML20217C7591997-12-31031 December 1997 1997 Annual Environmental Operating Rept for Oyster Creek Nuclear Generating Station ML20197E9131997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Oyster Creek Nuclear Station ML20199E4561997-11-13013 November 1997 Safety Evaluation Accepting Ampacity Derating Analysis in Response to NRC RAI Re GL-92-08, Thermo-Lag 330-1 Fire Barriers, for Plant ML20199D4381997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Oyster Creek Nuclear Station ML20202E8511997-10-21021 October 1997 Rev 0 to Scenario 47, Gpu Nuclear Oyster Creek Nuclear Generating Station Emergency Preparedness (Nrc/Fema Evaluated) 1997 Biennial Exercise. Pages 49 & 59 of Incoming Submittal Were Not Included ML20211M9481997-10-0303 October 1997 Supplemental Part 21 Rept Re Condition Effected Emergency Svc Water Pumps Supplied by Bw/Ip Intl Inc to Gpu Nuclear, Oyster Creek Nuclear Generation Station.No Other Nuclear Generating Stations Effected by Notification ML20198J7361997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Oyster Creek Nuclear Generating Station ML20211B7461997-09-24024 September 1997 Part 21 Rept Re Failure of Emergency Service Water Pump Due to Threaded Flange Attaching Column to Top Series Case Failure.Caused by Dissimilar Metals.Pumps in High Ion Svc Will Be Upgraded to 316 Stainless Steel Matl ML20210V0181997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Oyster Creek Nuclear Generating Station ML20210L2961997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Oyster Creek Nuclear Station ML20149F9961997-07-18018 July 1997 Safety Evaluation Re Gpu Nuclear Operational Quality Assurance Plan,Rev 10 for Three Mile Island Nuclear Generating Station,Unit 1 & Oyster Creek Nuclear Generating Station ML20196H0111997-07-11011 July 1997 Special Rept 97-001:on 970620,removed High Range Radioactive Noble Gas Effluent Monitor (Stack Ragems) from Service to Allow Secondary Calibr IAW Master Surveillance Schedule. Completed Calibr on 970628 & Returned Stack Ragems to Svc ML20210L3081997-06-30030 June 1997 Corrected Page to MOR for June 1997 for Oyster Creek Nuclear Generating Station ML20141H2051997-06-30030 June 1997 Monthly Operating Rept for June 1997 for Oyster Creek Nuclear Station 1999-09-30
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A . ,
NUCLEAR ENERGY BUSINESS OPER ATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125 C GEN ER AL h ELECTRIC APPLICABLE TO:
PUBUCATION NO. NEDO-24195 T. I. E. NO. 79NED288 TITLE General Electric Reload 7
Fuel Application for Ovster May 1984 DATE Creek NO TE: Correct allcopies of the applicable ISSUE DATE August 1979 publication as specified below.
t REFERENCES INSTRUCTIONS ITEM ,jS,ECT g O,N.
pgPAGNE) (CORRECTIONS AND ADDITIONS) 01 Page 5-39 Replace with new page 5-39 02 Page 5-39a Replace with new page 5-39a 03 Page 5-39b Insert new page 5-39b 04 Page 5-51a Replace with new page 5-51a o
p
,J 8405180221 840514 p DR ADOCK 05000219 PAGE 1 Ob I PDR
d * . .
NEDO-24195 ~
_ ~e 3.3.2 Loss-of-Coolant Accident
)
This analysis of the Oyster Creek Nuclear Generating Station loss-of-coolant accident (LOCA) is provided to demonstrate conformance with the ECCS accept-ance criteria of 10CFR50.46. The obj ective of the LOCA analysis contained herein is to provide assurance that the most limiting break size, break loca-tion and single failure combination has been considered for the plant. The documentation contained in this section is intended to satisfy these require-ments.
The general description of the General Electric (GE) LOCA evaluation models is contained in Reference 5-18. Applicability and approvcl for pre-pressurized reload fuel are given in Reference 5-25. Model changes are described in References 5-20 and 5-21, which were approved by the USNRC (Reference 5-19).
The anal.ysis utilizes the short-term thermal-hydraulic model (LAMB) and the transient critical power model (SCAT) in addition to the long-term thermal-hydraulic model (SAFE) and the core heat-up model (CHASTE).
This LOCA analysis differs from previous PR/2 LOCA analyses in the following ways:
- 1. Core flow coastdown and core depressurization are now calculated with the LAMB and SCAT codes, as opposed to no modelling of coastdown.
In order to use the LAMB computer code for a non-jet pump plant, LAMB inputs were developed to execute the code for a recirculation line break. These inputs reflect the geometry of the non-jet pump plant.
The LAMB code only allows for two recirculation pump loops; therefore, the five loops were modeled such that the intact loops are combined into one loop, and the broken loop is modeled as the second loop.
- 2. For the large break region, dryc ut time is now calculated using the SCAT code, as opposed to assuming a set dryout time.
- 3. Increased fission gas release at higher exposures has been included in
( the results.
- 4. As in the previous analyses, convective heat transfer coefficients specified in Appendix K are applied to the fuel rods and channel boxes ,
B-30
o 4 - - . . .
NED0-24195
~
under spray cooling when the reactor pressure blows down to 125 psia, at which time rated core spray system flow is achieved. These convec-(/h s_
tive coefficients were previously substantiated by demonstrating that there is sufficient bundle spray flow to justify the coefficients.
This was based on full sparger spray distribution tests in air.
Results of these tests were considered applicable to core spray per-formance in a reactor steam environment since the steam effect on spray distribution was believed to be small. Recently, a study using currently approved core spray design methodology (Reference 5-38) was performed to calculate in detail the spray distribution performance of ring spargers in steam (Reference 5-39). The results of this study indicate that adequate spray distribution is provided for reactor pressures up to 55 psia. For higher reactor pressures experimental test data are not available to extrapolate the spray distribution methodology to higher pressures in steam. Therefore, there are some uncertainties in the applicability of the spray distribution method-ology for reactor pressures greater than 55 psia. For large break LOCAs, the reactor blows down below 55 psia before credit for core
) spray heat transfer is assumed. Therefore, uncertainties in spray distribution predictions in high pressure steam have no impact on these results. For some small breaks, however, the reactor blows down more slowly and spray heat transfer credit is assumed above 55 psia. For these small break conditions, LOCA sensitivity studies were performed to assess core cooling effects in this pressure range.
These studies demonstrated that the inherent core cooling predicted during the blow down from 125 to 55 psia due to steam cooling, more than justifies the application of the Appendix K convective heat transfer coefficients. Additionally, there is a further core cooling effect due to core reflooding calculated for these small breaks.
Since heat transfer from steam cooling or reflooding is conservatively neglected for Appendix K applications, the actual heat transfer coef-ficients still exceed those allows by 10CFR50 Appendix K. Therefore, analysis via approved methodology using Appendix K heat transfer coefficients remains a valid licensing basis for Oyster Creek.
/ n x' The Oyster Creek LOCA analysis reported here was performed as an independent
.scif-contained analysis similar to that performed as a lead plant analysis.
5-39a
t 4 -
,,, . i NED0-24195 ;
/s 5.5.2.1 Input to Analysis A list of the significant input parameters to the LOCA analysis is presented
' in Table 5-10.
5.5.2.2 LAMB ANALYSIS This_ code is used to analyze the short-term blowdown phenomena for large ,
postulated pipe breaks (breaks in which nucleate boiling is lost before the water level drops and uncovers the active fuel). The LAMB output (core flow as a function of time) is input to the SCAT code for calculation of blowdown heat transfer.
4 I
The LAMB results presented are:
e Core Average Inlet Flow Rate (normalized to unity at the beginning of the accident) following a Large Break. p 5.5.2.3 SCAT ANALYSIS $
k This code completes the transient short-term thermal-hydraulic calculation for large breaks. The GEXL correlation is used to track the boiling transition in ,
t time and location. The post-critical' heat flux heat transfer correlations are
. built into SCAT, which calculates heat transfer coefficients for input to the ,
core heatup code (CHASTE) .
The SCAT results presented are: [
e Minimum Critical Power Ratio following a Large Break. i o _ Convective Heat Transfer Coefficient following a Large Break.
5.5.2.4 SAFE ANALYSIS ,
()
\% ,/
This code is used primarily to track the vessel inventory and to model ECCS performance during the LOCA. The application of SAFE is identical for all
_ i L
5-39b [
t c - , ,. .c - - - . . ,_..,--, , - - , n- -- -- . . - - -, .
9 - . . ..
NED0-24195 5-30 Letter, J. F. Quirk (GE) to T. P. Speis (NRC), "0DYN Improvements",
October 13, 1931.
5-31 Latter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in GE Methods for Analysis of Cold Water Injection Transients", September 30, 1980.
5-32 Letter, R. C. Tedesco (NRC) to G. G. Sherwood (GE), " Acceptance for g Referencing General Electric Licensing Topical Report NED0-24154/ ;;
NEDE-24154P", February 4, 1981.
5-33 Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Methods for Determination of Operating Limits", January 19, 1981.
5-34 Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), "End of Cycle Coastdown Analyzed With ODYN/TASC", September 1, 1981.
5-35 Letter, H. C. Pfefferlen (GE) to D. G. Eisenhut (NRC), " Correction of ODYN Errors", June 8, 1982. ,
~
5-36 General Electric Company Model for Loss-of-Coolant Analysis in Accordance uith 10CFR50, Appendi K, Amend-tent No. 2 - One Recirculation m
, Loop Out-of-Ser9fce, General Electric Company, Revision 1, July 1978 R (NED0-20566-2).
5-37 Letter, R. E. Engel (GE) to D. B. Vassallo (NRC), " Control Rod Drop Accident," February 24, 1982.
- 5-38 S. A. Sandoz, et al., " Core Spray Design Methodology Confirmation Tests," General Electric Co., March 1983 (NED0-24712-A).
5-39 " Performance Evaluation of the Oyster Creek Core Spray Sparger," General 3; Electric Co., January 1984 (Class III), (NEDE-30010).
k
/
5-51a