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MONTHYEARML17264B0451997-09-19019 September 1997 Rev 3 to 86-1234820-03, Low Temp Overpressure Analyses Summary. ML17265A6891997-06-30030 June 1997 Rev 1 to Calculation 96190-C-68, Pressure Locking Analysis for Valve(S) 871A/B. ML17264A9261997-06-10010 June 1997 Low Temp Overpressure Analyses Summary Rept, Calculational Summary Sheet 86-1234820-01 ML20132A3921996-12-13013 December 1996 Press Release I-96-81, NRC Staff Proposes to Fine Rg&E $100,000 for Alleged Violations at Ginna Nuclear Plant ML20129B7031996-10-22022 October 1996 Press Release I-96-69,informing That NRC Has Rescheduled Meeting W/Officials of Rg&E to Discuss Latest Systematic Assessment of Licensee Performance at Gnpp ML20128M2981996-10-15015 October 1996 Press Release I-96-68, NRC Staff Rates Ginna Good in Three Areas & Excellent in One Category on Latest SALP Rept ML17264A6131996-08-0202 August 1996 Rev 0 to Verification of Pressure Locking Analysis Program Preslok. ML17264A2651995-11-16016 November 1995 Gamma-Body Offsite Dose Calculation. ML17264A6111993-12-23023 December 1993 Rev 1 to Valve Thrust Calculation for 6x4x6 Class 300# Anchor Darling Gate Valve,857A,B & C & 860A,B,C & D. ML20085M5451991-11-11011 November 1991 Responses to NUMARC Aquatic Resources Survey & Socioeconomic Question in Support of NRC License Renewal Rulemaking ML17262A4511991-04-10010 April 1991 Devonrue Calculation Re Plant Atmospheric Relief Valve (ARV) Area Ambient Temp rise-revised Methodology ML17262A4531990-06-0505 June 1990 Devonrue Calculation Re Plant Turbine Driven Auxiliary Feedwater Pump Area Ambient Temp Rise ML17251A4791989-04-14014 April 1989 Prv Block Valve Thrust Calculation. ML20237D7901987-12-10010 December 1987 Package of Unrelated Documentation Re Listed Plants Concerning NRC Emergency Response Data Sys ML20151M9601987-07-20020 July 1987 Package of NRC Administered Requalification Exam Results Summary IA-86-445, Package of Documents Re Operating Experience W/Westinghouse PWR Steam Generators Through 8201311986-07-28028 July 1986 Package of Documents Re Operating Experience W/Westinghouse PWR Steam Generators Through 820131 ML20203H2851986-07-28028 July 1986 Package of Documents Re Operating Experience W/Westinghouse PWR Steam Generators Through 820131 ML20132D2121985-06-24024 June 1985 Human Factors Engineering Dcrdr In-Progress Audit,Re Ginna Nuclear Power Plant ML20129F4831985-05-0909 May 1985 Package of Compilations of Results of Operator Licensing & Requalification Exams ML17256A8381982-03-23023 March 1982 Slide Presentation from 820323 Meeting Re Steam Generator Repair Activities at Facility ML17261A5941966-09-16016 September 1966 Package of Incomplete Brookwood Project Requirement Outline Repts Re Instruments & Equipment 1997-09-19
[Table view] Category:OPERATIONS SUPPORT-CALCULATIONS
MONTHYEARML17264B0451997-09-19019 September 1997 Rev 3 to 86-1234820-03, Low Temp Overpressure Analyses Summary. ML17265A6891997-06-30030 June 1997 Rev 1 to Calculation 96190-C-68, Pressure Locking Analysis for Valve(S) 871A/B. ML17264A9261997-06-10010 June 1997 Low Temp Overpressure Analyses Summary Rept, Calculational Summary Sheet 86-1234820-01 ML17264A6131996-08-0202 August 1996 Rev 0 to Verification of Pressure Locking Analysis Program Preslok. ML17264A2651995-11-16016 November 1995 Gamma-Body Offsite Dose Calculation. ML17264A6111993-12-23023 December 1993 Rev 1 to Valve Thrust Calculation for 6x4x6 Class 300# Anchor Darling Gate Valve,857A,B & C & 860A,B,C & D. ML17262A4511991-04-10010 April 1991 Devonrue Calculation Re Plant Atmospheric Relief Valve (ARV) Area Ambient Temp rise-revised Methodology ML17262A4531990-06-0505 June 1990 Devonrue Calculation Re Plant Turbine Driven Auxiliary Feedwater Pump Area Ambient Temp Rise ML17251A4791989-04-14014 April 1989 Prv Block Valve Thrust Calculation. 1997-09-19
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Supplement to WCAP-13668 - Gamma-Body CNsite Dose Calculation Anal ical Assum tlons Source Term The noble gas concentrations In the primary coolant prior to the accident are based upon 1% fuel defects. This is approximately equal to the proposed Technical Specification limit for non-iodine nuclides of 100/E pCI/gm, Dose Calculations The following assumptions and parameters were used to calculate the noble gas activity released to the atmosphere and the corresponding gamma-body doses following a SGTR.
The mass of reactor coolant discharged into the secondary system through the rupture is presented in Tables II.5 and ll.6.
- 2. No retention of noble gas in the secondary system is assumed. All noble gas activity transferred to the secondary system, from the primary, is released to the atmosphere.
- 3. No credit was taken for radioactive decay during transport to the site boundary and outer boundary of the low population zone.
Offsite Gamma-Bod Dose Calculation Model Offsite gamma-body doses are calculated using the equation:
D ~ = 0.25 p Eyq g (IAR)IJ (
I J
")J Q
Where:
Dy-body gamma-body dose due to immersion in a semi-infinite cloud of noble gas average gamma energy of noble gas nuclide i in.mev/dis.
(Table Ill.7)
IAR)i integrated activity of noble gas nuclide i released during the time interval j in Ci (x/Q)) atmospheric dispersion factor for time period j in sec/m'5i20iOi32 95iiib 05000244 ..'
PDR ADDCK PD~R ~
Results Doses Rem Gamma-Bod Doses Case 1 Case 2 Guideline Exclusion Area Boundary (0-2 hr) 0.08 0.11 Low Population Zone (0-8 hr) 0.005 0.007
SV 0
Table lll.7 Reactor Coolant Noble Gas Activity (pCi/gm) and Average Gamma Energy (Mev./Dls)
Nuclide RCS Activity E (Mev/Dts)
(pCi/gm)
Kr-85m l.28EO O. l6 Kr-85 5.36EO 0.0023 Kr-87 8.11E-1 0.79 Kr-88 2.38F0 2.2 Xe-131m 2.39EO 0.0029 Xe-133m 3.5EO 0.02 Xe-133 2.03E2 0,03 Xe-135m 3.7E-1 OA3 Xe-135 6.4EO 0.25 Xe-138 4.61E-1 1.2
Attachment 2 RG&E to NRC Letter Dated 11/16/95 Site Plot and Photos
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