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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217J3341999-10-19019 October 1999 Forwards Request for Addl Info Re Sale of Portion of Land Part of Oyster Creek Nuclear Generating Station Site Including Portion of Exclusion Area ML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20212J6721999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Oyster Creek Nuclear Generating Station on 990913.No Areas Identified in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Historical Listing of Plant Issues Encl ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20217B2531999-09-24024 September 1999 Informs That on 980903,Region I Field Ofc of NRC Ofc of Investigations Initiated Investigation to Determine Whether Crane Operator Qualification/Training Records Had Been Falsified at Oyster Creek Nuclear Generating Station ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212A7921999-09-13013 September 1999 Forwards Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, Issued on 950817 to Plant ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J9831999-09-0202 September 1999 Discusses 990804 Telcon Re Sale of Portion of Oyster Creek Nuclear Generating Station Land.Requests Info Re Location of All Areas within Property to Be Released Where Licensed Radioactive Matl Present & Disposition of Radioactive Matl ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211C0161999-08-19019 August 1999 Advises That Info Submitted by Ltr,Dtd 990618, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool, Holtec Rept HI-981983,rev 4,will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210U4341999-08-17017 August 1999 Responds to to Chairman Dicus of NRC on Behalf of Fm Massari Concern About Oyster Creek Nuclear Generating Station Not Yet Being Fully Y2K Compliant ML20210Q7331999-08-12012 August 1999 Responds to Re TS Change Request (TSCR)264 from Oyster Creek Nuclear Generating Station.Questions Re Proposed Sale of Property within Site Boundary & Exclusion Area ML20210L6311999-08-0606 August 1999 Discusses Licensee Response to GL 92-01,Rev1,Suppl 1, Rv Structural Integrity, for Plant.Staff Has Revised Info in Rv Integrity Database & Releasing as Rvid Version 2 ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195G6541999-06-0707 June 1999 Discusses 981204 Initiation to Investigate Whether Contract Valve Technician,Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20195G6631999-06-0707 June 1999 Discusses 981204 Intiation to Investigate Whether Contract Valve Technician Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20209B0561999-06-0404 June 1999 Informs That NRR Has Reorganized,Effective 990328.Forwards Organizational Chart ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P5381999-04-14014 April 1999 Ack Receipt of Re Request for Exception to App J. Intended Correction Would Need to Be Submitted as Change to TS as Exceptions to RG 1.163 Must Be Listed in Ts,Per 10CFR50,App J ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record ML20205P0651999-04-0909 April 1999 Discusses 990225 PPR & Forwards Plant Issues Matrix & Insp Plan.Results of PPR Used by NRC Mgt to Facilitate Planning & Allocation of Insp Resources 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205J3281999-04-0101 April 1999 Discusses Arrangements Made on 990323 for NRC to Inspect Licensed Operator Requalification Program at Oyster Creek Nuclear Generating Station During Week of 990524 ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207F0331999-03-0404 March 1999 Forwards Insp Rept 50-219/98-12 During Periods 981214-18, 990106-07 & 20-22.Areas Examined During Insp Included Implementation of GL 89-10 & GL 96-05.No Violations Noted ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny ML20206S2541999-01-20020 January 1999 Confirms Resolution of Thermo-Lag Fire Barriers in Fire Zones OB-FZ-6A & OB-FZ-6B (480 Switchgear Rooms) IAW Previous Commitments Contained in Gpuns Ltrs to NRC & 971001 ML20199J2631999-01-18018 January 1999 Requests That Listed Changes Be Made to Correspondence Distribution List for Oyster Creek Generating Station ML20199D0271999-01-11011 January 1999 Requests Listed Addl Info in Order to Effectively Review TS Change Request 264 Re Ownership of Property within Exclusion Area ML20199A6521999-01-0707 January 1999 Notifies That Reactor Operators G Scienski,License SOP-11319 & D Mcmillan,License SOP-3919-4 Have Terminated Licenses at Oyster Creek Nuclear Generating Station, Effective 990101 ML20198T1061999-01-0606 January 1999 Forwards Rev 15 to Gpu Nuclear Corporate Emergency Plan for TMI & Oyster Creek Nuclear Station. with Summary of Changes Which Reflect Use of EALs Approved in NRC Ltr to Gpun on 980908 & Other Changes Not Related to Use of New EALs ML20198K0331998-12-23023 December 1998 Forwards Change Request 268 for Amend to License DPR-16. Amend Would Change TS to Specify Surveillance Frequency of Once Per Three Months ML20198H0181998-12-22022 December 1998 Forwards Attachment Addressing New Info & Modifying 980505 Submittal Re Request for Change to Licensing Bases for ECCS Overpressure,In Response to NRC Bulletin 96-03, Potential Plugging of ECCS by Debris in Bwrs ML20198H8521998-12-16016 December 1998 Dockets Completion of Physical Inventory Performed in July 1997,as Addl Info to Nuclear Matl Balance Rept Submitted on 980416 ML20196H4461998-12-0202 December 1998 Provides Final Response to NRC GL 96-01, Testing of Safety-Related Logic Circuits ML20196B4471998-11-23023 November 1998 Provides Required Response 2 to NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainers in Bwrs. During Recently Completed 17R Refueling Outage,New Strainers Were Installed ML20195J8451998-11-12012 November 1998 Forwards Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan, as Change Previously Made Without Appropriate Notification to NRC ML20195C7201998-11-11011 November 1998 Forwards 120-day Required Response to GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment, ML20195E1221998-11-10010 November 1998 Notifies NRC of First Time Usage of Code Case N-504 & Inclusion Into OCNGS ISI Program,As Accepted by RG 1.147, Inservice Insp Code Case Acceptability ML20155J6851998-11-0505 November 1998 Forwards TS Change Request 266,to Modify Safety Limits & Surveillances of LPRM & APRM Sys & Related Bases to Ensure APRM Channels Respond within Necessary Range & Accuracy & Verify Channel Operability ML20155H5641998-11-0202 November 1998 Informs That Bne Has No Comments on Proposed Change 259 to Ts,Correcting Required Water Level in Condensate Storage Tank So That Design Basis Is Correctly Implemented ML20155G3741998-10-29029 October 1998 Forwards Response to NRC 980619 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions 1999-09-30
[Table view] |
Text
. .
GPU Nuclear Corporation
_ U Nuclear ::'e=8r388 Forked River New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
February 15,1995 C321-95-2070 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Response to Request for AdditionalInformation
References:
- 1) NRC letter, A. W. Dromerick to J. J. Barton, " Oyster Creek Spent Fuel Pool Expansion", dated December 29,1994.
- 2) GPUN letter C321-94-2134, J. J. Barton to Document Control Desk, " Technical Specification Cnange Request No. 222", dated November 11,1994.
The NRC staff requested additionalinformation (Reference 1) regarding a GPU Nuclear request (Reference 2) to utilize forty five additional spent fuel storage locations which currently exist in the fuel storage pool. This existing capacity had been included in the original spent fuel pool expansion project but not reflected in the Technical Specifications. Attachment 1 is our response to the requested information. Attachment 2 provides the Technical Specification pages (5.3-1 and 5.3-2) affected by our Reference 2 request. Page 5.3-1 is revised in response to NRC staff question 9. Both pages reflect a change in font.
This letter is a resubmittal of our letter C321-95-2041 dated February 8,1995. The original submittal inadvertently omitted the attached graph.
Sincerely, x
g . J. BrirtBiT Vice Fresident and Director Oyster Creek 4 Attachments )
cc: Administrator, NRC Region l NRC Senior Resident inspector, Oyster Creek Oyster Creek NRC Project Manager 9502240064 950215 PDR ADOCK 05000219 P PDR I :
GPU Nuclear Corporahon is a subsidiary of the General Pubhc Utihties Corporabon h l 1
Attachment 1 (5 pages)
Response to NRC Request for AdditionalInformation i
References
- 1. " Criticality Safety Evaluation of Oyster Creek Spent Fuel Storage Racks with Fuel up to 4.2% Enrichment", Holtec International (Stanley E. Turner), August 1993.
- 2. EPRI TR-101986, "Boraflex Test Results and Evaluation", February 1993. ,
Question / Responses
- 1. What methods and computer codes were used for the reanalysis of 4 wt.%
fuel and Boraflex shrinkage? What organization performed the calculations and benchmarking?
CASMO-3 was used for the 4.0 wt.% enriched fuel analysis and was verified using the NITAWL-KENO 5a code with the 27 group SCALE cross section library.
The analysis method involves determining the k, of an infinite bundle lattice in the fuel pool geometry. Adjustments for mechanical, manufacturing, fuel burnup and computer code uncertainties are then applied to establish a fuel pool k, at a 95/95 confidence level. Previous analysis and benchmarking of the codes was performed by Southern Science and the new analysis was performed by Holtec International (Reference 1). Both analyses use the same methodology and were supervised by Dr. Stanley E. Turner. Additional benchmarking and verification was performed by GPU Nuclear on the 27 group (SCALE) NITAWL-KENO 5a code package which was used to determine the ak associated with Boraflex gaps.
- 2. Since credit was taken for the Gadolinia burnable poison pins, at what point in core life does the fuel attain its peak reactivity?
Peak reactivity for 4.0 wt.% fuel occurs at 10.5 GWD/MT. ;
- 3. What uncertainty was included to account for burnup calculations?
Burnup calculations include an uncertainty of 0.01 ak. 1
- 4. How have the other 95/95 calculational uncertainties (e.g. lattice spacing, channel bulge, etc.) changed as a result of the higher enrichment?
The 95/95 calculational uncertainties were not changed. The calculational uncertainties were calculated in the original criticality analysis based on a fuel enrichment of 3.01 wt.% without Gadolinia. The 3.01 wt.% enriched lattice 1
l
Y without Gadolinia is more reactive (k = 0.9295) than 4.0% fuel enriched with seven 3.0 wt.% Gadolinia pins at peak reactivity (k. = 0.8938). Hence, the 95/95 '
uncertainties for the 4.0% analysis are bounded by the original analysis.
- 5. Describe the Boraflex gap assumptions in more detail. For example, what does the assumption that the gaps are coplanar mean in terms of axial location and what percentage of shrinkage does a 3.9 inch gap correspond to?
The coplanar assumption is that a 3.9 inch gap (2.8% shrinkage) occurs in a single gap at the same axial level in all Boraflex panels. This is the maximum shrinkage expected by EPRI studies for the type of racks installed at Oyster Creek. A gap size of 3.9 inches bounds all gaps found to date as a result of blackness testing and is expected to bound all future gap formation due to Boraflex shrinkage reaching a saturation point at exposures greater than 1x10 rads. The coplanar assumption represents the most conservative gap configuration since reactivity is compounded by neutron coupling effects. Actual measurement data has shown that less than half of the Boraflex panels have gaps and have an axial distribution which will reduce neutron coupling.
- 6. Recent EPRI data indicates that, in addition to shrinkage, additional Boraflex degradation may occur due to long-term exposure to pool water flow. Has this been considered in the reanalysis?
No. While GPUN is awara of the loss of silica due to long term exposure to fuel pool water flow in combination with gamma exposure, it has not been considered in the reanalysis because, currently, the phenomenon is not well understood (see question 12). Due to the rack designs at Oyster Creek the Boraflex panels are not directly exposed to fuel pool water and therefore are resistant to this form of degradation. Although small holes exist on each panel for the purposes of gas release it does not provide sufficient communication with the fuel pool water to cause degradation due to water impingement. A surveillance program is in place to measure any materialloss from Boraflex test coupons. These coupons are expected to bound any degradation that would occur within the rack structure due to more direct fuel pool water contact. At this time no significant ;
weight loss in the test coupons is indicated. The next coupon measurement is scheduled for March 1995.
- 7. How is assurance provided that no fuel assemblies enriched to 3.8 wl.% or higher have less than 7 Gadolinia pins of 3 wt.% loading?
The minimum poison loading criterion of 7 Gadolinia pins of 3 wt.% is based on an analysis for a fuel assembly that was utilized at Oyster Creek that had 3.19 2
i l
wt.% fuel with 7 Gadolinia pins of 3 wt.%. As fuel enrichments increase, it has necessitated going to a higher number of fuel rods containing Gadolinia and to higher wt.% Gadolinia. A fuel lattice enriched to 3.6% is the highest enriched i lattice at Oyster Creek containing 3 wt.% Gadolinia and has 9 Gadolinia rods.
Higher enriched fuel lattices have both higher wt.% Gadolinia and more than 7 ;
Gadolinia rods. Criticality analyses at higher fuel enrichments have continued to assume 7 rods of 3 wt.% Gadolinia for conservatism.
From an administrative perspective, new fuel designs are required to underoo design review by GPUN procedures. A New Fuel Design Checklist incorporated in the review process ensures that new fuel complies with enrichment, poison loading and other criteria established by analysis.
- 8. Has the dry storage vault been analyzed for storage of 4.0 wt.% fuel under accident conditions of full flooding and optimum moderation?
No. The dry storage vault design meets the design requirements of 10CFR general design criteria 61 and 62 for the initial fuel loading at Oyster Creek for unflooded and fully flooded conditions. No subsequent analysis has been performed to update the dry storage vault analysis for new fuel and/or for optimum moderation. For this reason, Oyster Creek maintains restrictions on the uncovering of the dry storage vault to prevent a spray type moderation from the refueling floor fire system. The initial fuelloading at Oyster Creek did not contain Gadolinia and is more reactive at beginning of life than any subsequent i I
fuel designs loaded at Oyster Creek. Although enrichment loadings have increased substantially, the Gadolinia loading have kept the beginning of life reactivity at much lower reactivity than bundles initially loaded in the dry storage 1 vault. The initial Oyster Creek bundles at beginning of life have a k_ of 1.227 and the highest reactivity in any current new bundle lattice has a k. of only 1.120. Therefore, the current fuel designs are bounded by previous analysis on the dry storage vault.
- 9. Is the phrase "as performed on the poison racks" in the third line of Basis 5.3.1 misplaced?
l This sentence was intended to remain unchanged "as performed on the poison racks" should be removed. A revised page 5.3-1 is included in Attachment 2.
- 10. How was the maximum gap of 3.9 inches in Boraflex panels determined?
What was the corresponding gamma exposure (rads)?
3.9 inches is the maximum expected gap size that would result from 2.8%
Boraflex shrinkage based on EPRI studies (Reference 2) for the rack type 3
I 1
s .
installed at Oyster Creek. The enclosed plot shows Oyster Creek's blackness testing results and coupon data as it compares to EPRI maximum expected shrinkage levels. The 2.8% is from the upper maximum shrinkage curve. This is the shrinkage associated with gap formation. Additional shrinkage will occur at j the top and bottom ends of the Boraflex panels which has a very small effect on reactivity due to leakage effects. The largest total gap size found at OysterCreek l was 2.42 inches at an exposure of 3.0x10' rads (2 gaps of 1.07 and 1.35 '
inches). The shrinkage is expected to saturate at exposure levels greater than ,
1.0x10'" rads and gap growth is expected to be bounded by the 3.9 inch gap )
assumed in the coplanar gap analysis. I i
- 11. What kind of evaluations were performed on the surveillance coupons?
Coupons are visually inspected as well as measurements recorded of thickness, length, width, weight and hardness. Neutron attenuation testing is planned if j problems are suspected from the weight and dimensional checks. Coupons are inspected at predesignated intervals depending on if the samples are from the long term test or the accelerated test assembly. High gamma dose bundles are !
placed near the accelerated coupons to insure that they are receiving a higher l than average gamma dose rate than the rack panels. j l
- 12. Was the concentration of silica measured in the spent fuel pool? If this i measurement was made, was the data used to predict degradation (loss of boron) of Boraflex panels?
Silica concentration measurements are made on a continual basis as part of normal fuel pool chemistry monitoring. The peak silica level was found to be 3.0 ppm. However, fuel pool clean-up system is placed in service periodically to remove the silica. Starting this cycle (Cycle 15), the silica will not be removed until end of cycle. This will allow for a better understanding of silica buildup which will be needed to evaluate Boraflex degradation in the future.
At this time no significant weight loss has been indicated based on the Boraflex sample coupons. The sample coupons are expected to bound degradation within the racks due to more direct contact with fuel pool water. GPUN is participating in an EPRI study currently underway to develop a methodology for pred:cting the degradation of Boraflex. Results of the study should be available by mid 1995. When a methodology is available, GPUN will evaluate the need to include Boraflex degradation into the fuel pool criticality analysis.
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Attachment 2 (2 pages)
Replacement Technical Specification Pages
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i l
5.3 AUXILIARY E0VIPMENT 5.3.1 Fuel Storage A. The fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.
B. Loads greater than weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility.
C. The spent fuel shipping cask shall not be lifted more than six inches above the top plate of the cask drop protection system.
Vertical limit switches shall be operable to assure the six inch vertical limit is met when the cask is above the top plate of the cask drop protection system.
D. The temperature of the water in the spent fuel storage pool, measured at or near the surface, shall not exceed 125"F.
E. The maximum amount of spent fuel assemblies stored in the spent fuel storage pool shall be 2645. I BASIS The specification of a K-effectiva less than or equal to 0.95 in fuel storage facilities assures an ample margin from criticality. This limit applies to unirradiated fuel in both the dry storage vault and the spent fuel racks as well as irradiated fuel in the spent fuel racks. Criticality analyses were performed on the poison racks to ensure that a K-effective of 0.95 would not be exceeded.
The analyses took credit for burnable poisons in the fuel and included manufacturing tolerances and uncertainties as described in Section 9.1 of the FSAR. Calculational uncertainties described in 5.3.1.A are explicitly defined in FSAR Section 9.1.2.3.9. Any fuel stored in the fuel storage facilities shall I be bounded by the analyses in these reference documents.
The effects of a dropped fuel bundle onto stored fuel in the spent fuel storage facility has been analyzed. This analysis shows that the fuel bundle drop would not cause doses resulting from ruptured fuel pins that exceed 10 CFR 100 limits (1,2,3) and that dropped waste cans will not damage the pool liner.
The elevation limitation of the spent fuel shipping cask to no more than 6 inches above the top plate of the cask drop protection system prevents loss of the pool integrity resulting from postulated drop accidents. An analysis of the effects of a 100-ton cask drop from 6 inches has been done (4) which showed that the pool structure is capable of sustaining the loads imposed during such a drop. Limit switches on the crane restrict the elevation of the cask to less than or equal to 6 inches when it is above the top plate.
0YSTER CREEK 5.3-1 Amendment No.: 22,76,77,121
Detailed structural analysis of the spent fuel pool was performed using loads resulting from the dead weight of the structural elements, the building loads, I hydrostatic loads from the pool water, the weight of fuel and racks stored in the pool, seismic loads, loads due to thermal gradients in the pool floor and the walls, and dynamic load from the cask drop accident. ihermal gradients result in two loading conditions; normal operating and the accident conditions with the loss of spent fuel pool cooling. For the normal condition, the containment air temperature was assumed to vary between 65 F and 110*F while the pool water temperature varied between 85"F and 125 F. The most severe loading from the normal operating thermal gradient results with containment air temperatures at 65'F and the water temperature at 125*F. Air temperature measurements made during all phases of plant operation in the shutdown heat exchanger room, which is directly beneath part of the spent fuel pool floor slab, show that 65 F is the appropriate minimum air temperature. The spent fuel . pool water temperature will alarm control room before the water temperature reaches 120"F.
Results of the structural analysis show that the pool structure is structurally adequate for the loadings associated with the normal operation and the condition resulting from the postulated cask drop accident (5) (6).
The floor framing was also found to be capable of withstanding the steady state thermal gradient conditions with the pool water temperature at 150"F without exceeding ACI Code requirements. The walls are also capable of operation at a steady state condition with the pool water temperature at 140 F (5).
Since the cooled fuel pool water returns at the bottom of the pool and the heated water is removed from the surface, the average of the surface temperature and the fuel pool cooling return water is an appropriate estimate of the average bulk temperature; alternately the pool surface temperature could be conservatively used.
References
- 1. Amendment No. 78 to FDSAR (Section 7)
- 2. Supplement No. , to Amendment No. 78 to the FDSAR (Question 12)
- 3. Supplement No. I to Amendment 78 of the FDSAR (Question 40)
- 4. Supplement No. I to Amendment 68 of the FDSAR
- 5. Revision No. I to Addendum 2 to Supplement No. I to Amendment No.
78 of FDSAR (Questions 5 and 10)
- 6. FDSAR Amendment No. 79 -
- 7. Deleted l OYSTER CREEK 5.3-2 Amendment No. 121