CNL-15-192, Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) - Response to NRC Request for Additional Information - Radiation Protection and Consequence Branch
| ML15268A568 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 09/25/2015 |
| From: | James Shea Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CNL-15-192, TAC MF6050, WBN-TS-15-03 | |
| Download: ML15268A568 (219) | |
Text
{{#Wiki_filter:1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-192 September 25, 2015 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License Nos. NFP-90 NRC Docket No. 50-390
Subject:
Application to Revise Technical Specification 4.2.1, "Fuel Assemblies" (WBN-TS-15-03) (TAC No. MF6050) - Response to NRC Request for Additional Information - Radiation Protection and Consequence Branch
Reference:
- 1. Letter From TVA to NRC, CNL-15-001, "Application to Revise Technical Specification 4.2.1, 'Fuel Assemblies,' (WBN-TS-15-03)," dated March 31, 2015 (ADAMS Accession No. ML15098A446)
- 2. Letter from TVA to NRC, CNL-15-077, "Correction to Application to Revise Technical Specification 4.2.1, "Fuel Assemblies" (WBN-TS-15-03)," dated April 28, 2015 (ADAMS Accession No. ML15124A334)
- 3. Letter From NRC to TVA, "Watts Bar Nuclear Plant, Unit 1 - Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Application to Increase Tritium Producing Absorbing Rods (TAC NO. MF6050)," dated May 14, 2015 (ADAMS Accession No. ML15127A250)
- 4. Letter from TVA to NRC, CNL-15-092, "Response to NRC Request to Supplement the Application to Revise Technical Specification 4.2.1,
'Fuel Assemblies' (WBN-TS-15-03)," dated May 27, 2015 (ADAMS Accession No. ML15147A611)
- 5. Letter from TVA to NRC, CNL-15-093, "Response to NRC Request to Supplement Application to Revise Technical Specification 4.2.1, "Fuel Assemblies" (WBN-TS-15-03) - Radiological Protection and Radiological Consequences," dated June 15, 2015 (ADAMS Accession No. ML15167A359)
U. S. Nuclear Regulatory Commission CNL-15-192 Page 2 September 25, 2015
- 6. Electronic Mail from Jeanne Dion (NRC) to Clinton Szabo (TVA), Gordon Arent (TVA), and Edward D. Schrull (TVA), "Watts Bar 1 - FINAL Rad protection RAls for MF6050," dated August 25, 2015 By letter dated March 31, 2015 (Reference 1), Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) to revise Watts Bar Nuclear Plant (WBN) , Unit 1 Technical Specification (TS) 4.2.1, "Fuel Assemblies," to increase the maximum number of Tritium Producing Burnable Absorber Rods (TPBARs) that can be irradiated per cycle from 704 to 1,792. The proposed change also revises TS 3.5.1 , "Accumulators," Surveillance Requirement (SR) 3.5.1.4 and TS 3.5.4, "Refueling Water Storage Tank (RWST)," SR 3.5.4.3 to delete outdated information related to the Tritium Production Program. TVA provided a correction letter on April 28 , 2015 (Reference 2) .
By letter dated May 14, 2015 (Reference 3), the Nuclear Regulatory Commission (NRC) requested that TVA provide additional information to supplement the LAR. TVA provided the requested supplemental information in TVA letters dated May 27, 2015, and June 15, 2015 (References 4 and 5, respectively) . By electronic mail dated August 25, 2015 (Reference 6), the Nuclear Regulatory Commission (NRC) requested that TVA provide additional information to support the NRC review of the LAR. The response to the request for additional information (RAI) is due September 25, 2015. The enclosure to this letter provides TVA's RAI response. Consistent with the standards set forth in Title 10 of the Code of Federal Regulations (10 CFR), Part 50.92(c), TVA has determined that the additional information, as provided in this letter, does not affect the no significant hazards consideration associated with the proposed application previously provided in Reference 1. Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter and the enclosures to the Tennessee Department of Environment and Conservation. There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Mr. Edward D. Schrull at (423) 751-3850. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 25th day of September 2015. ctfully, {;;~
. Shea President, Nuclear Licensing
U. S. Nuclear Regulatory Commission CNL-15-192 Page 3 September 25, 2015
Enclosure:
TVA Response to NRC Request for Additional Information Enclosure cc (Enclosure): NRC Regional Administrator - Region II NRC Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation
ENCLOSURE TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 TVA Response to NRC Request for Additional Information Contents Radiation Protection and Consequence Branch (ARCB) Request for Additional Information (RAI) 1 ______________________________________________________ 2 ARCB RAI 2 ____________________________________________________________ 3 ARCB RAI 3 ____________________________________________________________ 5 ARCB RAI 4 ____________________________________________________________ 8 Attachments: 1 Offsite and Control Room Operator Doses Due to a Main Steam Line Break Calculation 2 Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Calculation 3 Control Room Operator and Offsite Doses Due to a Loss of AC Power Calculation 4 Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Calculation
- 5. Control Room Operator and Offsite Doses From a Fuel Handling Accident Calculation CNL-15-192 Enclosure Page 1 of 8
Radiation Protection and Consequence Branch (ARCB) Request for Additional Information (RAI) 1 In the application supplement dated June 15, 2015 it states: LAR Enclosure 2, Section, "Radiological Consequences of Accidents," was revised to include the inputs and assumptions utilized for each design basis accident (DBA) related to the tritium source term for the current licensing basis and the new licensing basis, if changed.In addition, with one exception, the Main Steam Line Break and Steam Generator Tube Rupture inputs and assumptions are the same as those used to support License Amendment 91 regarding the change to the Dose Equivalent I-131 spike limit, which was approved in the NRC Safety Evaluation dated December 5, 2012 (ADAMS Accession No. ML12279A115). The only exception is the control room isolation delay time, which was increased from 40 seconds to 74 seconds to correct an error in how the delay time was determined. The main steam line break and steam generator tube rupture accidents are being updated with a new tritium concentration and a new control room isolation delay time, therefore provide the dose analysis calculation for both of these accidents, and if not in the calculation provide a tabulation of all analysis inputs and assumptions used in offsite and control room habitability analyses in sufficient detail to enable the staff to evaluate the appropriateness of this data and, if deemed necessary, to perform confirmatory calculations. TVA Response The Main Steam Line Break (MSLB) accident analysis that supports the March 31, 2015, License Amendment Request (LAR) is provided in Attachment 1 to this enclosure. The Steam Generator Tube Rupture (SGTR) accident analysis that supports the March 31, 2015, LAR is provided in to this enclosure. The analysis inputs and assumptions used are provided in the calculations. CNL-15-192 Enclosure Page 2 of 8
ARCB RAI 2 In TVAs letter dated May 21, 2002, Watts Bar Nuclear Plant - Request for Additional Information (RAI) Regarding Tritium Production - Interface Issue Number 5 - Control Room Habitability Systems (TAC No. MB1884), (ADAMS accession number ML021440139) TVA stated: The [total effective dose equivalent] TEDE values were calculated for informational purposes only and do not replace the whole body and thyroid dose guidelines currently in the WBN licensing basis. Future design basis accident radiological analyses, which are intended to demonstrate compliance with regulatory criteria, will continue to assess whole body and thyroid doses and will contain informational data regarding TEDE. The Radiological Consequences of Accidents section in Enclosure 2, Revision 1 submitted for NRC review on June 15, 2015 contains TEDE and beta doses and does not provide whole body and thyroid doses. Provide the current licensing basis and new proposed whole body and thyroid doses for each accident analyzed. TVA Response RAI 2, Tables 1 and 2 below provide the dose results for the SGTR and MSLB accidents, respectively, and include the beta and TEDE doses provided in Table 13 of Enclosure 2, Revision 1 of the LAR. RAI 2, Table 1: SGTR Accident Dose Results Current Licensing Basis Proposed Exclusion Low Area Population Control Room Boundary Zone Pre-Accident Spike (CR) (EAB) (LPZ) CR EAB LPZ Whole Body (rem) 8.56E-02 3.50E-01 1.03E-01 8.64E-02 3.50E-01 1.03E-01 Beta (rem) 9.50E-01 2.03E-01 6.22E-02 9.62E-01 2.04E-01 6.25E-02 Thyroid (rem) 2.18E+01 1.33E+01 3.81E+00 2.29E+01 1.33E+01 3.81E+00 TEDE (rem) 1.15E+00 1.21E+00 3.49E-01 1.28E+00 1.22E+00 3.52E-01 Accident Initiated Spike CR EAB LPZ CR EAB LPZ Whole Body (rem) 8.11E-02 5.03E-01 1.47E-01 8.18E-02 5.05E-01 1.48E-01 Beta (rem) 9.34E-01 2.33E-01 7.15E-02 9.45E-01 2.35E-01 7.19E-02 Thyroid (rem) 3.37E+00 6.33E+00 1.86E+00 3.61E+00 6.37E+00 1.87E+00 TEDE (rem) 5.50E-01 1.06E+00 3.08E-01 6.50E-01 1.08E+00 3.14E-01 CNL-15-192 Enclosure Page 3 of 8
RAI 2, Table 2: MSLB Accident Dose Results Current Licensing Basis Proposed Pre-Accident Spike CR EAB LPZ CR EAB LPZ Whole Body (rem) 7.07E-03 2.92E-02 1.16E-02 7.12E-03 2.92E-02 1.16E-02 Beta (rem) 6.30E-02 9.27E-03 4.34E-03 6.37E-02 9.28E-03 4.35E-03 Thyroid (rem) 1.31E+01 2.63E+00 1.27E+00 1.32E+01 2.63E+00 1.27E+00 TEDE (rem) 4.58E-01 1.92E-01 8.75E-02 4.66E-01 1.92E-01 8.76E-02 Accident Initiated Spike CR EAB LPZ CR EAB LPZ Whole Body (rem) 1.25E-02 1.04E-01 1.23E-01 1.25E-02 1.04E-01 1.23E-01 Beta (rem) 9.93E-02 2.55E-02 2.98E-02 9.98E-02 2.55E-02 2.98E-02 Thyroid (rem) 1.73E+01 3.20E+00 4.59E+00 1.73E+01 3.20E+00 4.59E+00 TEDE (rem) 6.29E-01 3.48E-01 4.69E-01 6.35E-01 3.49E-01 4.69E-01 Note that tritium does not affect the whole body or thyroid doses. The decay emission energy of tritium is insufficient to penetrate the skin and contribute to the whole-body dose and the thyroid dose is explicitly limited to inhalation of radioiodine. The changes to the control room whole body and thyroid doses for both the SGTR and MSLB accidents are due to the change in the control room isolation time. The changes in offsite whole body and thyroid doses for the SGTR are due to the changes in the control room isolation time as well. The release rates from the reactor coolant to the faulted steam generator and environment were modified to account for the different isolation time, but used the same total release. The release rate is input to the model with four significant figures in scientific form (e.g., 1.234E6). Because the magnitude of the release rates are on the order of 1E6 gram/hour and 1E7 gram/hour, rounding of the actual value resulted in a slightly higher total mass being released which resulted in a slightly higher dose. CNL-15-192 Enclosure Page 4 of 8
ARCB RAI 3 Amendment number 40 (ADAMS Accession Number ML022540925) to the WBN Unit 1 Operating License was issued September 23, 2002, and authorized the insertion of up to 2304 TPBARs in the WBN Unit 1 core. In amendment number 40 TVA assessed the following design basis accident analyses affected by the production of 2,304 TPBARs: Loss of offsite power (LOOP) Waste gas decay tank (WGDT) failure Loss of coolant accident (LOCA) Main steam line break (MSLB) Steam generator tube rupture (SGTR) Fuel handling accident (FHA) Failure of small lines carrying primary coolant outside containment Rod ejection accident The LAR dated March 31, 2015 as supplemented by letters dated May 27 and June 15, 2015, TVA provided an analysis of the radiological consequences for the LOCA, FHA, MSLB, and SGTR. However, there is no analysis that provides the impact of increasing the maximum number of TPBARs that can be irradiated per cycle from 704 to 1,792 on the radiological consequences for the LOOP, WGDT failure, failure of small lines carrying primary coolant outside containment, and rod ejection accident. Provide the technical analysis performed to determine that the current licensing basis radiological consequences for the LOOP, WGDT failure, failure of small lines carrying primary coolant outside containment, and rod ejection accident bounds the new radiological consequences for the requested increase to 1,792 TPBARs per cycle. This technical analysis should provide the following: A tabulation of all analysis inputs and assumptions used in offsite and control room habitability analyses in sufficient detail to enable the staff to evaluate the appropriateness of these data and, if deemed necessary, to perform confirmatory calculations as compared to the new inputs and assumptions that reflect the insights gained from Cycles 6 through 12. Explain any differences, or if there are no differences, then it should explain why it is acceptable to remain the same considering the insights gained from Cycles 6 through 12. Provide the current licensing basis and new proposed whole body and thyroid doses for each accident analyzed. TVA Response The loss of offsite power (LOOP) accident and Waste Gas Decay Tank (WGDT) rupture were re-analyzed with the proposed tritium design basis source term. The analyses are included in Attachments 3 and 4 to this enclosure, respectively. The accident-specific inputs and assumptions are included in the calculations. However, some generic information may not be apparent (e.g., atmospheric dispersion and control room parameters) which are given in RAI 3, Tables 1 and 2 below. CNL-15-192 Enclosure Page 5 of 8
RAI 3, Table 1: Atmospheric Dispersion Inputs Atmospheric Dispersion (sec/m3) LOOP WGDT EAB LPZ CR CR 0-2 hr 6.38E-04 1.78E-04 3.85E-03 2.56E-03 2-8 hr 8.84E-05 3.22E-03 1.71E-03 8-24 hr 6.22E-05 2.36E-04 7.26E-04 1-4 day 2.90E-05 1.88E-04 5.21E-04 4-30 day 9.81E-05 1.55E-04 4.30E-04 RAI 3, Table 2: Generic Control Room Parameters Volume 257198 cubic feet (ft3) Emergency pressurization flow 711 cubic feet per minute (cfm) Normal pressurization flow 3200 cfm Emergency recirculation flow 2889 cfm Normal recirculation flow 3200 cfm Unfiltered inleakage 51 cfm Charcoal Filter efficiency first pass 95% 2nd pass 70% HEPA filter efficiency 99% Occupancy Factors 0-24 hr 100% 1-4 days 60% 4-30 days 40% Breathing Rate cubic meters per second 0-8 hr 3.47E-04 (m3/sec) 8- 24 hr 1.75E-04 m3/sec 1-30 days 2.32E-04 m3/sec RAI 3, Tables 3 and 4 below provide the comparison of accident dose results between the current licensing basis and the proposed changes for LOOP and WGDT rupture, respectively. CNL-15-192 Enclosure Page 6 of 8
RAI 3, Table 3: LOOP Accident Dose Results Comparison Current Licensing Basis Proposed Realistic CR EAB LPZ CR EAB LPZ Whole Body (rem) 7.99E-09 1.84E-08 1.05E-08 8.04E-09 1.84E-08 1.05E-08 Beta (rem) 2.83E-04 1.70E-05 9.71E-06 3.58E-04 2.14E-05 1.22E-05 Inhalation (rem) 7.92E-07 1.13E-06 6.46E-07 8.43E-07 1.13E-06 6.46E-07 TEDE (rem) 4.62E-03 2.78E-04 1.59E-04 5.84E-03 3.50E-04 2.00E-04 Conservative CR EAB LPZ CR EAB LPZ Whole Body (rem) 3.32E-04 7.63E-04 4.37E-04 3.34E-04 7.63E-04 4.37E-04 Beta (rem) 3.99E-03 4.60E-04 2.63E-04 4.09E-03 4.64E-04 2.65E-04 Inhalation (rem) 3.29E-02 4.69E-02 2.68E-02 3.50E-02 4.69E-02 2.68E-02 TEDE (rem) 6.18E-03 3.74E-03 2.14E-03 7.47E-03 3.81E-03 2.18E-03 RAI 3, Table 4: WGDT Rupture Accident Dose Results Current Licensing Basis Proposed CR EAB LPZ CR EAB LPZ Whole Body (rem) 7.06E-01 5.00E-01 1.40E-01 7.92E-01 5.00E-01 1.40E-01 Beta (rem) 6.11E+00 1.35E+00 3.79E-01 6.85E+00 1.36E+00 3.79E-01 Inhalation (rem) 6.99E-03 1.29E-02 3.60E-03 1.08E-02 1.29E-02 3.60E-03 TEDE (rem) 9.22E-01 2.93E-01 8.20E-02 1.08E+00 3.02E-01 8.44E-02 It should be noted that the discussion in License Amendment 40 for the failure of small lines carrying primary coolant outside containment was actually referring to ECCS leakage outside containment, which is included as part of the LOCA analysis. As discussed in Enclosure 2, Revision 1 of the proposed LAR, the current licensing basis analysis for the LOCA is bounding. The current WBN licensing basis does not include an analysis for the radiological consequences of the failure of a small line carrying primary coolant outside containment. The staff acknowledged, in NUREG-0847 (Adams Accession No. ML072060490) and subsequent SSER 25 (Adams Accession No. ML12011A024), that the FSAR did not contain this analysis and performed their own confirmatory analysis and found this to be acceptable. As discussed in License Amendment 40 and UFSAR Chapter 15.5.7, the Rod Ejection accident is bounded by the LOCA and is not explicitly analyzed. CNL-15-192 Enclosure Page 7 of 8
ARCB RAI 4 Provide additional detail on the error in control room delay time determination, and explain why the increase in control room isolation delay time is not applicable to the other design basis accidents (DBAs) referenced in question 3 above. If the increase in control room isolation delay time is applicable to the other DBAs then provide the dose analysis calculation for these accidents, and if not in the calculation provide a tabulation of all analysis inputs and assumptions used in offsite and control room habitability analyses in sufficient detail to enable the staff to evaluate the appropriateness of this data. TVA Response The control room radiation monitor loops utilize the RP-30AM analog ratemeter. A time constant of 7.17E-3 minutes was previously used to determine the ratemeter response time, which would be appropriate for a count rate between 1E4 to 1E5 counts per minute (cpm). However, the setpoint for these monitors is 400 cpm; thus a time constant of 4.34E-1 minutes should have been used. This resulted in an increase in the ratemeter response time from 0.86 seconds to 52.08 seconds. Combined with the response times determined for the remainder of the loop, the total loop response time increased from 6.6 seconds to 57.8 seconds. The analyses rounded this to 60 seconds. This error is applicable to the following analyses referenced in ARCB RAI 3, which were reanalyzed with the higher isolation time: Main steam line break (MSLB) Steam generator tube rupture (SGTR) Loss of offsite power (LOOP) Waste gas decay tank (WGDT) failure Fuel handling accident (FHA) The associated calculations are provided in Attachments 1 through 5, respectively, of this enclosure. This error is not applicable to the following analyses referenced in ARCB RAI 3: Loss of coolant accident (LOCA) - Control room isolation is initiated directly on a safety injection signal that is initiated by the ESFAS in response to a LOCA. Failure of small lines carrying primary coolant outside containment - See the TVA response to ARCB RAI 3. Rod ejection accident - See the TVA response to ARCB RAI 3. CNL-15-192 Enclosure Page 8 of 8
Attachment 1 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 Offsite and Control Room Operator Doses Due to a Main Steam Line Break Calculation (56 pages including this cover page)
NPG CALCULATION COVERSHEET I CTS UPDATE Page 1 REV 0 EDMS/RJMS NO. CTSlYPE: EDMSlYPE: EDMS ACCESSIQN NQ {NfA for REV. O} 826931014409 Calculation CALCUl.ATIONS (NUCLEAR) T931 ~09?nnnt. Cale
Title:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break ORO PLANT BRANCH NUMBER CURREY NEW REV CALCID NUC WBN NIB WBNAPS3077 015 016 crs UPDATE ONLY 0 (Verifier and Approval Signatures Not Required) I! NO crs CHANGES O (For calc revision, CfS bas been reviewed and DO CfS changes required) IDfil (check one) SYSTEMS ~ 0181, l 0.2 0,3 0 NIA NIA DCN.EDC,N/A APPLICABLE DESIGN DOCUMEN'T(S} ClASSIE',K;ATION NIA NIA E QUALITY SAFE'IY RELATED? UNVERlFIED SfF.CIAL IWQUIREMENTS AND/OR DESlill! OUTPUT SAR!fS l!l!lm: ISFSI RELATED? (If yes, QR =yes) ASSUMPTION LIMITING CONDIDQNS? ATTACHMENT? SAR/CoC AFFECTED Yes181 NoO Yes 181 NoO YesO No181 YesO No~ YesO Nol'.XI Yes181 NoO CALCUIAI!QN NUMBER R.P,QUESTOR PREPARING DISCIPLINE VERIFICATION METHOD NEW METIIQD OF ANALYSIS Name:N/A PHONE:N/A BecthelMEB Design Review 0Yes 181No d DATE DATE PREPARER (PRINT~~ CHECKER (P~~AND SIGN) ITCHAN ~ 9/(//1 DWWU ,
~/f.J VERIFIER (PmNT NAME AND SIGN)
P.u.rt4-
/U DATE DATE APPROV~(P1uNTNAME ~ ~ ~.,.'
DWWU ~.;J \lulr~ '\11,(t3 STATEMENT QF PROBI.£MJABSTRACT This calculation is performed to show that the offsite and control room operator doses do not exceed the IOCFRlOO and 10CFRSO App.A GDC 19 dose limits due to a Main Steam Line Break at the Watts Bar Nuclear Plant. The computer code STP clctennines the activity releases. The STP output is used as input to computer codes FENCDOSE and COROD, which determine the offsite and control room doses. This calculation also considers the effect of a 74 second unfiltered bypass flow due to the finite closure time of the control room isolation dampers ( 14 sec) and illSbUIDa1t actuation time (60 sec). There are several cases modeled. Each case has two sets. One set has a pre-accident iodine spike where the iodine levd in the reactor coolant is at the 48 hour maYim*un allowable 14 µCi/gm 1-131 equivalent. The other set bas the reactor coolant at the maximum steady state 1-131 equivalent of 0.265 µCi/gm with an accident initiarcd iodine spike consisting of a SOO increase in the rate of iodine release &om the fuel In both sets, the primary to secondary side kale is l SO gpd in the unfaulted loops, and the secondary side activity is at the Technical Specification limit of 0.1 µCi/gm. The Tritium Production Core (TPC) was used. The X/Q values using the ARCON96 methodology were used. The ARCON96 XJQ values give more limiting results. Revision l 0 anaJy:ied the Unit 2 MSLB. The following results should ultimatelvbe mf)ected ill FSAR Tabli< lS.S-17 and Technical :il!a<W!Clli!ml!: Using ICRP-30, the following are the limitations: 14pCl/gm1-131 equlvaleat 48 hour limit 0.265pCl/gm1-131 equlvaleat equlHbriam Umlt 0.1 pCa/gm 1-131 equivalent bl the secondary side The -ximum primary to secondary side leak rate bl the anfaulted steam pneraton is 150 gpd/steam generator The RCS leakage (10 gpm idendOed + 1 gpm unidentified) See Appendix G for other cases. TPC does not affect the limits above, because the limiting doses are the thyroid doses. The tritium affects only the beta dose and TEDE. The TPC obviously bounds the non-TPC configuration. MICROFICHE/EFICHE Yes~ No0 FICHE NUMBERlS) TVA-F-W003233 TVA40S32 Pagel of2 NEDP-2-1 [10-31-2011)
NPG CALCULATION COVERSHEET I CTS UPDATE Page 2 CALC ID 1--
!JiSL::r:~::r:w~===:w-===::ciiX:J ORO PI.ANT BRANJ.1 ~* *~ REV ~~-
NUC WBN NI'B WBNAPS30TI 016 BUIIDING NA I ROOM NIA I ELEVATION NIA I COORD/AZIM NIA I :EIRM Bechtel CATEGORIES NIA KEYWORDS (A-add, D-delete) ACTION KEYWORD ND KEYWQRD (ND) CROSS-REFERENCES (A-add, D-delete) ACTION XREF XREF XREF XREF XREF (ND) CODE PI.ANT 1YPE NUMBER REV A p WBN PER nsss3 CTS ONLY UPDATES: FoUowi.,., are reanired onlv when makine: lcevword/cross reference CTS Ulldatcs and lllll!e 1 of form NEDP-2-1 is n,ot included: PREPARER (PRINT NAME AND SIGN'I DATE CHECKER (PRINT NAME AND SIGNI DATE PREPARER PHONE NO. EDMS ACCESSION NO. TVA40S32 Page2of2 NEDP-2-1 [10-31-2011)
Paqe 3 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNAPS3077 Title Offsite and Control Room Operator Doses Due to a Main Steam Line Break Revision DESCRIPTION OF REVISION No. 0 Initial Issue I Revision I is performed because the XJQ values have changed. RI pages 6 and 7 are new, and all pages are renumbered. Only changed text have revision bars. pages added: 6, 7 pages changed: all pages deleted: none 2 Revision 2 was performed because of new control room makeup flow (711 cfm). pages added: none pages deleted: none pages changed: I -7, 9, 13 3 Revision 3 was performed to incorporate larger Technical Specification primary to secondary side leakage as part of the alternate steam generator tube plugging project. Since the final leakage value has not been established, a 5 gpm and a 10 gpm case was performed. All pages were renumbered, however only pages with text changes will have revision bars and will be listed as having been changed: Pages added: none pages deleted: classification forms pages changed: 1-9, 12 R3 : 13 total pages 4 Revision 4 was performed to change the iodine partition factor for the faulted steam generator from 100 to 1 to account for steam generator dryout. The steady state preaccident leakage is changed to 150 gpd/steam generator. A preaccident iodine spike of 60 uCi/gm is applied to the reactor coolant. An accident initiated spiking factor of 500 increase in iodine release from the fuel to the reactor coolant was incorporated in the STP model. This revision is part of the corrective action for WBN PER 99-017510-000. Pages added: new coversheet (page I) Pages deleted: none Pages changed: la (old coversheet page I), lb (old page la), 2-9, 12 R4: 15 total pages 5 Revision 5 is performed to incorporate new iodine production rates, split the iodine spiking model into two separate cases, and also to perform an additional analysis for a maximum Technical Specification limit of0.35µCi/gm1-131 equivalent (steady state). The non-steady state maximum limit of 60 µCi/gm 1-131
- 0.35 = 21µCi/gm1-131 is also analyzed. Additionally, the maximum allowable primary to secondary side leakage is determined for all cases. Justification for usage of the ANSI-ANS 18.1-1984 spectrum is also provided. Due to the nature of the revision, all pages were changed, with significant additions (Appendix A, and Attachment 1).
Pages added: all pages deleted: all pages changed: all R5: 27 total pages 6 Revision 6 is performed to add the Tritium Production Core (TPC) with a two TPBAR failure, add a 0.265 and 0.177 µCi/gm I-131 equivalent steady state with factor of 500 iodine spike case (and deleted the 0.35 µCi/gm case), add 21µCi/gm1-131 equivalent 48 hour maximum case, change the noble gas inventories to the maximum allowable based on 100/Ebar, add ARCON96 XJQ values (also use Halitsky X/Q values), and to use the latest versions ofCOROD (R5) and FENCDOSE (R4) which now determine thyroid doses based on ICRP-2 and ICRP-30 as well as now determines the TEDE. Add a second actual measurement ofreactor coolant inventories. Incorporated NISYS and Westinghouse 3rd party review comments. Due to the nature of the changes, all pages were renumbered. Actual text changes are indicated with revision bars. The results of this calculation affect the FSAR and Technical Specifications. These changes will be incorporated by the corrective action plan for PER 00-012545-000. Pages added: all Pages deleted: all Pates changed: all R6: 60 total pages 7 Revision 7 is performed to increase the Steam Generator leakage post accident. In order to accommodate a later decision on the actual leakage, several different leakages were analyzed. The calculation methodology from R6 was unchanged. Revision 6 results were preserved by placing them in Appendix G. The results of this calculation affect the FSAR. The SAR change package number is 1770. Pages added: none Pages deleted: all old coversheets Pages changed: 1-8, 13, 14, 18-25, 41, 45, 48, 49 R7: 61 total pages TVA 40709 [10-2008] Page 1 of 1 NEDP-2-2 [10-20-2008]
page 4 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNAPS3077 Title Offsite and Control Room Operator Doses Due to a Main Steam Line Break Revision DESCRIPTION OF REVISION No. 8 Revision 8 is performed to add a 1.4 gpm SG leak case. The results of this calculation may affect the FSAR depending on establishment of final allowable leakage rate. Pages added: none Pages deleted: none Pages changed: 1-3, 5-8, 14, 19, 21, 23 R8: 61 total pages 9 Revision 9 is performed to address the new Replacement Steam Generators (DCN 51754 ). The number of cases analyzed has been reduced to only the preaccident and accident initiated Iodine spike with no additional post accident Steam Generator leakage (alternate repair criteria). The new steam generators have different inventories and mass releases. The previous original steam generator cases used in the FSAR (10 gpm known+ I gpm unknown+ 3 gpm post accident leakage) can be found in Appendix G (COROD results were corrected). In addition, the COROD model recirculation rate was corrected as part of corrective action PER 61493. Two CREVS train operation is addressed in assumption 14. The COROD time increments were corrected (PER 94426). Due to the nature of the revision, all pages were renumbered. Actual text changes are marked with revision bars. The results of this calculation will result in changes to ch.15 of the FSAR. The full impact to the FSAR and TS are discussed in the screening review for DCN 51754. Pages added: 4 Pages changed: I, 2, 5-10, 13-20, 29-35 Pages deleted: design verification form R9: 46 total page 10 Revision 010 of this calculation was created to add/update Unit 2 applicability. This calculation is applicable to Unit 2 based on the following:
- Appendix H of this calculation was added (I) to evaluate the recent Westinghouse steam releases during a Main Steam Line Break (MSLB); and (2) to install Revision 006 (original steam generator) as Appendix H, because it contains more conservative results than Revision 009 which are applicable to Unit 2. This calculation supports Chapter 15 of the FSAR.
Affected design inputs were reviewed and (I) were found to be correct, or (2) were corrected as necessary. The effect of Unit 2 operation on Unit I margins has been reviewed with no impact. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. FSAR AND TECHNICAL SPEClFICA TIO NS HAVE BEEN REVIEWED AND FSAR SECTION 15.5.4 CHANGE PAGES ARE PART OF FSAR AMENDMENT 97. Reviewer: Pages Added: IA,4A,4B,H-I toH-15 Pages Revised: 5A Pages Replaced: 2 Pages Deleted: none Total number of pages in this revision: 64 pages (46 + 18) (Appendix A - 2 pages; Appendix B - 6 pages; Appendix C - 2 pages; Appendix D - 2 pages; Appendix E - I page; Appendix F - I page; Appendix G - I page; Appendix H- 15 pages; Attachment I - 11 pages) II Revision 11 is performed to explicitly evaluate Unit 2, which has the original steam generators. This is performed in Appendix Hand replaces the previous Appendix H added in revision 10. Westinghouse provided revised mass releases. Also, the Unit 2 specific ARCON96 XJQ values were used. The SAR has been reviewed by Marc Berg and this revision of the calculation affects Unit 2 SAR section Chapter 15.5.4. A SAR change shall be processed in accordance with NGDC PP-10 to reflect the calculation results as part of EDCR 54956. Tech Specs have been reviewed and determined not to be affected. Pages added: design verification form (p.6) Pages deleted: none Pages changed: 1, 2, 4-9, 19-21, 37-40 Rl 1: 51 total pages 12 Revision 12 is performed for replacement of the analog ratemeters with digital RMIOOO ratemeters by DCN 52012. The longer response time of the RMIOOO ratemeter in incorporated by increasing the control room isolation time from 20.6 seconds to 40 seconds (14 sec damper closure+ 26 sec instrument response). Revision 12 also corrected typos in the results table of Appendix H. FSAR section 15 .5-4 and Technical Specifications were reviewed by Lynn Cowan and Table 15.5-17 is impacted by the change in isolation time from 20.6 seconds to 40 seconds. See calculation WBNTSR-028 (ref. 28) for 40 second delay time. Pages Added: None Pages Deleted: none Pages Revised/Replaced: 1,2,4-10, 18-20,29-34,39 R12: 51 total pages. TVA 40709 [10-2008] Page 1 of 1 NEDP-2-2 [10-20-2008]
Paqe 4A NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNAPS3077 Title Offsite and Control Room Operator Doses Due to a Main Steam Line Break Revision DESCRIPTION OF REVISION No. 013 Revision 013 is performed to address the PER 327968 and PER 327956 for the following items:
- redeveloped the updated 1-131 equivalent conversion factors based on the RG-1.109 iodine inhalation dose conversion factors,
- recalculated the Unit 1 and Unit 2 offsite doses and control room doses,
- revised the Unit 2 initial mass release from the defective steam generator using the Westinghouse updated data,
- corrected some typographical errors,
- changed a pre-accident iodine spike to 14µCi/gm1-131 equivalent from 21µCi/gm1-131 equivalent to be consistent with WBNTSR008 Rl3,
- reviewed 4 successor calculations to this calculation based on CCRIS reference list, and WBNAPS3 l l OR2 is impacted due to changes made in this calculation.
The effect of Unit 2 operation on Unit 1 margins has been reviewed with no impact. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. FSAR AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND FSAR SECTION 15.5.4 IS IMPACTED BY THE CHANGE OF THE 1-131 EQUIVALENCE FOR THE SOURCE TERM CALCULATION. Reviewer: KW.Peterman 6/3/11 Pages Added: 4A, 8A, Pages Deleted: none Pages Revised/Replaced: 1-51 Rl3: 53 total pages(= 51+2). Appendix A - 2 pages, Appendix B - 6 pages, Appendix C - 2 pages, Appendix D - 2 pages, Appendix E - 2 pages, Appendix F - 1 page, Appendix G - 1 page, Appendix H - 3 pages, and Attachment 1 - 11 pages 14 Revision 14 is performed to upgrade the X/Qs based on the 1991-2010 meteorological data set. FSAR AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND FSAR TABLE 15.5.-17 IS IMP ACTED BY THE CHANGE. There are no successor calculations impacted by this calculation. Reviewer: Marc Berg 7-25-2011 The successors to this calculation are not impacted by this revision. Pages Added: none Pages Deleted: none Pages Revised/Replaced: 1, 2, 4A, 6, 7, 8A, 9, 16, 19-21, 34, 36, 37, 39, 40 Rl4: 53 total pages TVA 40709 [10-2008] Page 1 of 1 NEDP-2-2 [10-20-2008]
Page 4B NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNAPS3077 Title Offsite and Control Room Operator Doses Due to a Main Steam Line Break Revision DESCRIPTION OF REVISION No. OIS Revision 15 is performed to update the Tritium Production Core (TPC) tritium source term to 124 µCi/gm (as revised in predecessor calculation WBNNAL3003). Since the results of this calculation are intended for use in WBN FSAR Section 15.5, a design output attachment was created and added to the calculation in Revision 15. Attachment 2, Form NEDP-2-5, makes the entire Calculation WBNAPS3077 Revision 15 design output. As a result, the calculation classification has been changed to 'EO' from 'E'. Pages replaced in Revision 15 contain an updated page header for Revision 15. All unaffected pages retain their original headers. Changes from the previous revision are marked on the replaced Revision 15 pages with revision bars. CTS was reviewed for successor calculations to Calculation WBNAPS3077, and 3 successor calculations were identified. Calculation WBNTSR080 is affected because it uses the FENCDOSE and COROD models modified in this revision. The other calculations are not affected by this revision because they do not use inputs that were changed as part of this revision. See DCN 61599 for SAR!fech Spec impact determination. Pages Added: 4B and 52 Pages Replaced or Revised: 1, 2, 5 - 7, SA - 10, 19 - 21, 30 - 36, and 40 Pages Deleted: none Total Revision 15 pages: 55 Attachment 1 - Surveillance Test l-SI-68-28 performed on 7/10/00 (5 pages) Surveillance Test l-SI-68-28 performed on 4/9/01 (6 pages) Attachment 2 - NPG Cakulation Design Output (I page) TVA 40709 [I 0-2008] Page I of I NEDP-2-2 (10-20-2008]
NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNAPS3077 Page4C Title Offsite and Control Room Operator Doses Due to a Main Steam Line Break Revision DESCRIPTION OF REVISION No. 016 Revision 016 is issued to address PER n5553. This revision updates the instrument response time for O-RE-90-125 from the currently 26 seconds to 60 seconds. Computer files have been re-run and results have been updated to reflect the changes. This revision also reverts the calwlation classification to "E" as Design Output is only required when information is used by non-engineering departments. Design Output form is also removed. Page 12 Is updated for TS allowable concentration. This does not impact the result or conclusion. Successor calculation WBNTSR080 has been reviewed and is impacted by this revision. The calculation wlU be updated for the cancellation of DCN 52012. Affected engineering judgments and assumptions were reviewed and (1) ware found to be adequate, or (2) were revised as necessary to ensure adequacy. The effect of Unit 2/dual unit operation on Unit 1 margins has been reviewed with no impact. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. FSAR Table 15.5-17 and Technical Specifications have been reviewed and are affected by ttis revision of the calculation. ,.')/ /J 'U Reviewer: K.tJ.~fe,w-M~/~ 'tJn/13 Pages Added: 4C Pages Revised/Replaced: 1, 2, 5, 6, 7, BA. 10, 12, 19, 20, 21, 30-36, 40 Pages Deleted: 52 Total number of pages in this revision includlng Attachments: 55 pages Attachment 1- Surveillance Test 1-SI-68-28 performed on 7/10/00 (5 pages) Surveillance Test 1-SI-68-28 performed on 4/9/01 (6 pages) This page is added by Rev. 016. TVA 40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Page 5 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNAPS3077 I Revision: I O16
!ABLE OF CONTENfS SECTION TITLE PAGE Coversheet/CTS Update 1 Record of Revision 3 Table of Contents 5 Calculation Verification Form 6 Computer Input File Storage Information Sheet 7 Computer Output Microfiche Information Sheet 8 h~~ 9 Introduction 9 Assumptions 10 Special Requirements/Limiting Conditions 11 Calculations 11 Results 19 Discussion and Condusion 20 References 21 Appendix A: Justification for Using ANSI/ANS-18.1-1984 Expected Coolant Spectrum 22 Appendix B: Determination of Letdown Flow Rate Uncertainty 24 Supporting attachments for Appendix B 26 Appendix C: Example of STP run, preaccident iodine spiking case 30 Appendix D: Example of STP run, accident initiated iodine spiking case 32 Appendix E: Example of COROD run 34 Appendix F: Example ofFENCDOSE run 36 Appendix G: Additional Cases 37 Appendix H: Unit 2 MSLB 38 Attachment 1: Surveillance Test 1-SI-68-28 performed on 7/10/00 (5 pages) 41 Surveillance Test 1-81-68-28 performed on 4/9/01 (6 pages) 46 TVA40710 [10-2008) Page I ofl NEDP-2-3 [10-20-2008)
Pae:e 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNAPS3077 Revision 016 Method of verification used: _,£_fr4
- 1. Design Review ~
- 2. Alternate Calculation D Verifier Date ffe~3
~
- 3. Qualification Test D D.W.Wu Comments:
The changes to the calculation described in the Record of Revision for Revision 016 have been reviewed and have been found to be technically adequate in format and content. TVA 40533 [l 0-2008) Pagel ofl NEDP-2-4 [10-20-2008)
p ae;e 7 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNAPS3077 IRev. OI6 I Plant: WBN I
Subject:
Offslte and Control Room Operator Doses Due to a Main Steam Line Break L J Electronic storage of the imut files for this calculation is not reauired. Comments:
~ Input files for this calculation have been stored electronically and sufficient identifying information is provided below for eacli imut file. (Any retrieved file reauires re-verification of its contents before use.)
The RI input files are archived on FILEK.EEPER under reference I.D.# 263447 The R2 input files are archived on FILEK.EEPER under reference I.D.# 2637I5 The R3 input files are permanently stored in FILEKEEPER file # 302577 The R4 input files are permanently stored in FILEK.EEPER file # 302740 The R5 input files are permanently stored in FILEK.EEPER file # 303 I 80 The R6 input files are permanently stored in FILEK.EEPER file# 3036I I The R7 input files are permanently stored in FILEK.EEPER file # 304905 The R8 input files are permanently stored in FILEKEEPER file # 305986 The R9 input files are permanently stored in FILEK.EEPER file # 308282 The RI I input files are permanently stored in eFiche file# TVA-F-WOOI4I2 The RI 2 input files are permanently stored in FILEK.EEPER file # 3 I 4516 The Word file for RI2 is stored in FILEKEEPER file# 3 I45I5 The R13 input files are permanently stored in FILEKEEPER file# 3 I 7573 The Word file for R13 is stored in FILEK.EEPER file# 3 I 7574 The RI4 input files are permanently stored in eFiche file# TVA-F-W002549 The RI5 input files are permanently stored in FILEKEEPER file# 32I942 The RI5 output files are permanently stored in eFiche file# TVA-F-W003I95 The Word file for RI5 is stored in FILEKEEPER file# 32I983 The RI 6 input files are permanently stored in FILEK.EEPER file # 322084 The Word file for RI6 is stored in FILEK.EEPER file# 322432 181 Microfiche/eFiche TVA 40535 (10-2008) Page 1of1 NEDP-2-6 (10-20-2008]
NPG COMPUTER OUTPUT page 8 MICROFICHE INFORMATION SHEET Document WBNAPS3077 I Rev. 014 I Plant: WBN I
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Microfiche Number Description RO:TV A-F-G I 04614 RO Computer Runs see R5 for listing RI :TV A-F-C000078 RI Computer Runs see R5 for listing R2:TV A-F-COOOl 19 R2 Computer Runs see R5 for listing R3 :TV A-F-C000294 R3: Computer Runs see R5 for listing R4:TV A-F-C000301 R4: Computer Runs see R5 for listing R5:TVA-F-C000316 R5: Computer Runs see R5 for listing R6: Computer Runs: R6:TV A-F-W000220 See R9 for listing R7: R7: Computer Runs: TV A-F-W000290 APS77S7# STP source terms APS77C7# COROD control room dose APS77F7# FENCDOSE offsite dose where#=: A. E, I, M, Q, U = 21 µCi/gm with 1O+1 gpm primary to secondary side leak,, Halitsky X/Q B, F, J, N, R, V = 0.265 µCi/gm with 10+1 gpm prim to secondary leak, 500 I spike, Halitsky X/Q C, G, K, 0, S, W = same as A, E, I, M, Q, U but with ARCON96 X!Q D, H, L, P, T, X =same as B, F, J, N, R, V but with ARCON96 XIQ And A, B, C, D = 5.4 gpm SG leak E, F, G, H = 1.0 gpm SG leak I, J, K, L= 2.0 gpm SG leak M, N, 0, P= 3.0 gpm SG leak Q, R, S, T= 4.0 gpm SG leak U, V, W, X= 5.0 gpm SG leak R8: Computer Runs: R8: APS77S8# STP source terms, 1.4 gpm SG leak TV A-F-W000360 APS77C8# COROD control room dose APS77F8# FENCDOSE offsite dose where#=: A= 21 µCi/gm with 10+1 gpm primary to secondary side leak,, Halitsky X/Q B = 0.265 µCi/gm with 10+ I gpm prim to secondary leak, 500 I spike, Halitsky X/Q C = same as A but with ARCON96 X!Q D = same as B but with ARCON96 X!Q R9: Computer Runs: APS77S9# STP source terms R9: APS77C9# COROD control room dose TVA-F-W000611 APS77F9# FENCDOSE offsite dose where#=: A, E =21 µCi/gm with 10+1 gpm primary to secondary side leak ARCON96 X/Q, B,F=0.265 µCi/gm with 10+1 gpm prim to secondary leak, 500 I spike, ARCON96 X!Q A,B=replacement SG. E,F= original SG RI 1: Computer Runs: RI!: APS77Sl0# STP source terms TVA-F-W001412 APS77CIO# COROD control room dose APS77FIO# FENCDOSE offsite dose where#=: A=21 µCi/gm with 10+1 gpm primary to secondary side leak Unit 2 ARCON96 X/Q, B=0.265 µCi/gm with 10+ I gpm prim to secondary leak, 500 I spike, Unit 2 ARCON96 X/Q R12:Computer Runs: Rl2: APS77S12# STP source terms TVA-F-W001578 APS77C12# COROD control room dose where #=: A,B=repl.SG; A = 21 µCi/gm with 1O+1 gpm primary to secondary side leak, ARCON96; B = 0.265 µCi/gm with 10+ 1 gpm primary to secondary leak, 500 I spike, ARCON96 X/Q;
NPG COMPUTER OUTPUT page SA MICROFICHE INFORMATION SHEET Document WBNAPS3077 I Rev. 016 I Plant: WBN I
Subject:
Offslte and Control Room Operator Doses Due to a Main Steam Line Break Microfiche Number Description Rl3: Rl3: Computer files named as: 1VA-F-W002393 $APS77Sl3# STP source terms
$APS77Cl3# COROD control room dose
$APS77Fl3# FENCDOSE offsite dose where
$= 1forUnit1
$ = 2 for Unit 2
# = A for pre-accident iodine spiking, 14 µCi/gm with 10 + 1 gpm primary leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with 10 + 1 gpm primary leak, 500 I spike A total of 12 runs for Rl3.
Rl4: Rl4: Computer files named as: 1VA-F-W002549 $APS77Cl4# COROD control room dose
$APS77Fl4# FENCDOSE offsite dose where
$ = 1 for Unit 1
$ = 2 for Unit 2
# = A for pre-accident iodine spiking, 14 µCi/gm with 10 + 1 gpm primary leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with 10 + 1 gpm primary leak, 500 I spike Rl5: R 15: Computer files named as:
1VA-F-W003195 $APS77Sl5# STP source terms
$APS77Cl5# COROD control room dose
$APS77Fl5# FENCDOSE offsite dose where
$ = 1 for Unit 1
$ = 2 for Unit 2
#=A for pre-accident iodine spiking, 14 µCi/gm with 10 + 1 gpm primary leak
# = B for accident initiated iodine spilcing, 0.265 µCi/gm with 10 + 1 gpm primary leak, 500 I spike Rl6: Rl6: Computerfilesnamedas:
1VA-F-W003233 $APS77Sl6# STP source terms
$APS77Cl6# COROD control room dose
$APS77Fl6# FENCDOSE offsite dose where
$ = 1 for Unit 1
$=2 forUnit2
#=A for pre-accident iodine spiking, 14 µCi/gm with 10 + 1 gpm primary leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with 10 + 1 gpm primary leak, 500 I spike
Calculation No. WBNAPS3077 I Rev: 015 IPlant: WBN IPage: 9
Subject:
Otlsite and Control Room Operator Doses Due to a Main Steam Line Break Purpose The purpose of this calculation is to determine the offsite and control room operator dose due to a Main Steam Line Break (MSLB). The results will be used in FSAR ch. 15.5 to show compliance with 10CFRlOO and IOCFR50 App. A GDC 19. This calculation also establishes the maximum primary to secondary side leakage and the maximum I-131 equivalent concentrations in the primary and secondary side coolant. Introduction A Main Steam Line Break at the Watts Bar Nuclear Plant will result in a significant steam release to the environment. The steam will contain radionuclides if a primary to secondary side leak occurs prior to the MSLB event. This calculation is performed to show that the offsite and control room operator doses do not exceed the 10CFRl00 and 10CFR50 App. A GDC 19 dose limits. This calculation uses the computer code STP (ref. 3) to determine the activity releases. The STP output is used as input to computer codes FENCDOSE and COROD. Computer code FENCDOSE (ref. 4) is used to determine the offsite dose. Computer code COROD (ref. 5) is used to determine the control room operator dose. The base FENCDOSE and CO ROD models are taken from WBNTSR008 (ref. 9). There are 2 cases modeled. The first case has a pre-accident iodine spike where the iodine level in the reactor coolant is at the maximum allowable of 14 µCi/gm I-131 equivalent (assumption 6). The second has the reactor coolant at the maximum steady state I-131 equivalent of 0.265 µCi/gm with an accident initiated iodine spike consisting ofa 500 increase in the rate of iodine release from the fuel. In both cases, the primary to secondary side leak is 150 gpd in the unfaulted loops (ref. 21 ), and the secondary side activity is at the Technical Specification limit of 0.1 µCi/gm (ref. 23). There is additional steam generator leakage post accident of 1 gpm to the faulted steam generator, and the 150 gpd/unfaulted steam generator continues post accident. To establish the Iodine release rate from the fuel, a pre-accident 10 gprn known reactor coolant leak is used with a 1 gpm unknown leak for a total of 11 gprn. Additional cases are performed in Appendix G with other leakage rates and other iodine concentrations. These extra cases were performed so as to give additional information for possible future changes in Technical Specifications. The relative isotopic spectrum is taken from WBNNAL3003 (Reactor Coolant Activities in Accordance with ANSI/ANS-18.1-1984). Justification of the usage of this spectrum as opposed to the historical design spectrum as found in chapter 11 of the FSAR can be found in Appendix A.
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN IPage: 10
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Assumptions I. The primary side to secondary side leakage is 150 gpd/steam generator, steady state, with I gpm in the faulted steam generator (steady state). The I gpm in the faulted steam generator and the 150 gpd/unfaulted steam generators continue following the accident. No additional leakage is assumed. Technical Justification: 150 gpd/steam generator and the 1 gpm are the maximum Technical Specification leakages. Having the 1 gpm in the faulted loop is conservative.
- 2. The maximum letdown of120 gpm (ref. 16)+ 4.39 gpm uncertainty fora total of 124.39 gpm is used.
.Technical Justification: This will maximU.e the removal rate of iodines from the primary coolant, and therefore will maximiz.e the production rate of iodine (production = removal at steady state) and is consistent with NSAL-00-004 (ref. 22). See Calculation section for the formulas used. Note, this value is used for calculation of iodine productioru'removal rates. The letdown is assumed to be isolated at the beginning of the accident to maximiz.e the reactor coolant inventories. The uncertainty of 4.39 gpm is determined in Appendix B.
- 3. The primary to secondary side leak rates and letdown flow rates are based on Standard Temperature and Pressure (STP).
Technical Justification: STP conditions will result in higher densities, therefore higher masses, especially when determining the production rate ofiodines.
- 4. It is assumed that the faulted steam generator dries out at the start of the accident, resulting in an iodine partition factor of 1.0 per ref.
10. Technical Justification: Following an accident, the Main Steam Line will be isolated and the Main and Auxiliary Feedwater will also be isolated. Since the worst case accident occurs with the line associated with a Steam Generator with Technical Specification leakage, that Steam Generator will dry out. In reality, this dry out will not occur until all feedwater has been isolated, and the water boiled off. Assuming dry out conditions at time zero is clearly conservative.
- 5. In the intact steam generators, the iodine partition factor is assumed to be 100.
Technical Justification: The mass of primary to secondary leakage which occurs to the intact steam generators is small relative to the mass of secondary coolant. Therefore none of this leakage is assumed to flash and the release to the environment is through the steaming process. Reference I 0 allows a partition factor of 100 for such cases.
- 6. A pre-accident iodine spike of 14 µCi/gm 1-131 equivalent is assumed at the start of the accident. In other cases, an accident initiated iodine spike of 500 increase in the iodine release rate from the fuel is assumed in the accident initiated case with the reactor coolant starting at 0.265 µCi/gm 1-131 equivalent.
Technical Justification: SRP 15.1.5 subsection 4a specifies the maximum allowable p~accident spike is required. The maximum allowable ~accident iodine spike per Technical Specification is 14 µCi/gm for 48 hours. SRP 15.1.5 subsection 4b specifies that following an accident, the iodine release rate from the fuel to the reactor coolant is increased by a factor of 500.
- 7. The letdown demineraliz.er efficiency is assumed to be 1 for iodines.
Technical Justification: This will maximize iodine removal (=production) rate, and therefore result in larger iodine spiking.
- 8. The control room isolates in 74 seconds due to high radiation in the Control Building Ventilation intake (Ref. 28). This will result in an unfiltered puff into the control room for that 74 seconds.
Technical Justification: This is based on 14 seconds closure time ofthe dampers, plus 60 seconds instrument response time (Ref. 28)
- 9. The tritium inventory in the Tritium Production Core (fPC) assumes 2 TPBAR failures (124 µCi/gm in the reactor coolant, per WBNNAL3003, ref. 2).
Technical Justification: This will maximize the tritium release.
- 10. The iodine production rate is based on 10 gpm identified primary side leakage (all leaks) plus 1 gpm unidentified leak, for a total of 11 gpm.
Technical Justification: This is per Technical Specification 3.4.13 (ref. 21 ), and maximizes the iodine production rates. This methodology is consistent with NSAL-~. ref. 22.
- 11. It is assumed that the secondary side concentrations are at the maximum ofO.l µCi/gm I-131 equivalent Technical Justification: This is the maximum allowed by the Technical Specifications (ref. 23) and is conservative.
- 12. The noble gas inventories are maximi:zed by scaling them up to I 00/Ebar.
Technical Justification: This maximi:zes the noble gas inventories. 100/Ebar is the Technical Specification limit.
- 13. It is assumed that there are no fuel failures associated with the accident.
Technical Justification: This accident will not uncover the core, therefore the core will not see extreme temperatures which would lead to fuel failure.
- 14. Only one train ofCREVS is in operation.
Technical Justification: Normally, each CREVS train takes suction from separate intakes with no cross contamination between trains. This leads to one contaminated train, and one uncontaminated train. The only way a 2 CREVS operation could result in This Page replaced by RI 6
Calculation No. WBNAPS3077 I Rev: OI4 I Plant: WBN I Page: I I
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-20 I I Steam Line Break Checked: JEB Date: 9-1-2011 higher doses would be for both trains to take suction from the same vent. For this to happen, one intake path would require a failed closed intake path AND a fail open of normally close passive manual damper at the beginning of the accident. An active failure of a train plus a failure of a passive component in less than 24 hours is beyond design basis. Special Requirements/Limiting Conditions There are no special requirements or limiting conditions in this calculation. Calculations The STP models consist of a pre-accident iodine spike (see figure 1) model and an accident initiated iodine spike model (see figure 2). The model(s) consist of the following: Volumes:
#1: Reactor Coolant: 5.78E5 lb (ref. 2) = 2.622E8 gm
#2: Steam Generator w/Leak: 5.31E7 gm (ref. 6)
#3: Steam Generators w/out Leak: l.593E8 (ref. 6).
#4: Environment: 1 gm (arbitrary) (This volume is made into an accumulator through the "A" card to suppress radioactive decay)
Step Sources: The following equation is used to set up the initial activities (in Ci) for each component using the initial ANSUANS-18.1-1984 source modified for WBN operational parameters,,( which is in units of µCi/gm): S =Component Volume [gm] x lE-6 Ci/µCi x I-131 equivalent conversion factor To obtain the I-131 equivalent conversion factor, the ANSI/ANS-18.1-1984 (ref. 2) iodine spectrum must be converted to I-131 equivalence. See Appendix A for justification for using the ANSI/ANS-18.1-1984 spectrum. The method to calculate the I-131 equivalence conversion factor is as follows: I-131 equivalent = dose conversion factor x iodine concentration I I-131 dose conversion factor The dose conversion factor in above equation is the iodine inhalation dose conversion factors for thyroid, which are available based on the different methodologies. The previous versions of this calculation used ICRP-2 methodology. In this revision, RG-1.109 (ref. 29, p. 45) iodine dose conversion factors are used to calculate I-131 equivalent conversion factor for the source term to be consistent with the Technical Specification 1.1 (ref. 33). The results of the I-131 equivalent conversion factors for the RCS are shown in Table 1a. Table la. The I-131 Equivalent Conversion Factor for the RCS Iodine Dose Conversion ANSI-18.1 Coolant 1-131 Equivalent Nuclide Factors Concentration (µCi/gm) (mrem/Ci) (µCi/gm) ref. 29, Table E-7 ref. 2 I-131 1.49E+09 4.77E-02 4.77E-02 I-132 l.43E+07 2.25E-01 2.16E-03 I-133 2.69E+08 1.49E-Ol 2.69E-02 I-134 3.73E+06 3.64E-Ol 9.l lE-04 I-135 5.60E+07 2.78E-01 l.04E-02 total l.06E+OO 8.812E-02 inverse 11.348
Calculation No. WBNAPS3077 I Rev: 016 IPlant: WBN IPage: 12
Subject:
Offsite and Control Room Operator Doses Doe to a Main Steam Line Break The above table shows that the I-131 equivalent concentration of the initial RCS ANSI 18.l source term is 0.08812 µCi/gm from the RG-1. l 09 inhalation dose conversion factors for thyroid. Consequently, to ratio the initial source term up to the TS allowable values, the ANSI 18.l concentrations must be multiplied by 11.348 (1/0.08812), as shown in Table la. This equivalent is dependent on the iodine dose conversion factors. This I-131 equivalent conversion factor developed from the RG-1.109 is about 40% higher than that in the previous versions of this calculation based on the ICRP-2. For the secondary side concentrations from WBNNAL3-003, the same procedure is performed to determine the I-131 equivalence. The results of the I-131 equivalent conversion factor for the secondary side coolant are shown in Table 1b. Table 1b. The I-131 Eauivalent Conversion Factor for the Secondarv Side Coolant Nuclide Iodine Dose Conversion ANSI-18.1 Secondary 1-131 Equivalent Facton Water (µCi/gm) (mrem/Ci) foCi/21n) ref. 29, Table E-7 ref. 2 I-131 1.49E+09 1.41E-06 1.41E-06 I-132 1.43E+o7 3.37E-06 3.23E-08 I-133 2.69E+o8 4.03E-06 7.28E-07 I-134 3.73E+o6 2.93E-06 7.33E-09 I-135 5.60E+o7 6.19E-06 2.33E-07 total 1.79E-05 2.41E-06 inverse 4.15E+o5 To convert to I-131 equivalence, the secondary side I-131 equivalent conversion factor is 4.15E5 (=1/2.41E-6) based on the RG-1.109 dose conversion factors, as shown in Table 1b. (Note: since there is no Technical Specification limit on gross activity for the secondary side like the 100/Ebar for the primary side, this factor is also applied to the secondary side noble gases to retain the proper isotopic ratios). The isotopes other than iodine in the primary coolant must also be scaled up. In NUR.EG-0800 R2 Chapter 15.6.3, section m.5 states *'The reviewer assumes the primary and secondary coolant activity concentrations allowed by the technical specifications." Reference 3 (of the NUREG-0800) states the following "The specific activity of the reactor coolant shall be limited to: a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT I-131, and b. Less than or equal to 100/E microCuries per gram of gross activity." Given the above considerations, the isotopic spectrum found in WBNNAL3-003 was examined. The 100/E values for this particular spectrum are determined in the following Table: This Page replaced by R16
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 13
Subject:
Offsite and Control Room Operator Doses Due to a Main IPrepared: MCB Date: 9-1-2011 Steam Line Break IChecked: JEB Date: 9-1-2011 Table 2: Determination of 100/EBAR A(i) E(i) E(i) E(i) Activity Beta Energy Gamma Energy Total Isotope [uCi/gm] [MeV/dis] [MeV/dis] A(i)*E(i) Kr-85m l.71E-Ol 2.5290E-Ol l.5862E-Ol 4.1152E-Ol 7.04E-02 Kr-85 2.66E-Ol 2.5060E-Ol 2.2102E-03 2.5281E-O 1 6.73E-02 Kr-87 l.6 lE-01 l.3237E+OO 7.9284E-Ol 2.l 165E+OO 3.40E-Ol Kr-88 3.00E-01 3.7500E-Ol l.9629E+OO 2.3379E+OO 7.0IE-01 Xe-13lm 6.54E-Ol l.4280E-Ol 2.0058E-02 l.6286E-Ol l.06E-Ol Xe-133m 7.l 7E-02 l.8980E-Ol 4.1559E-02 2.3136E-Ol l.66E-02 Xe-133 2.53E+OO l.3540E-Ol 4.5385E-02 l.8079E-Ol 4.57E-Ol Xe-135m l.39E-Ol 9.5000E-02 4.3176E-Ol 5.2676E-Ol 7.35E-02 Xe-135 9.04E-Ol 3.1680E-Ol 2.4696E-Ol 5.6376E-Ol 5.IOE-01 Br-84 l.72E-02 l.2842E+OO l.6816E+OO 2.9658E+OO 5.09E-02 Rb-88 2.04E-Ol 2.0617E+OO 6.8631E-Ol 2.7480E+OO 5.60E-Ol Cs-134 7.39E-03 l.5690E-Ol l.0361E+OO l.1930E+OO 8.82E-03 Cs-136 9.08E-04 l.0140E-Ol 2.1985E+OO 2.2999E+OO 2.09E-03 Cs-137 9.79E-03 l.8840E-Ol O.OOOOE+OO l.8840E-Ol l.84E-03 Na-24 4.99E-02 5.5460E-Ol 4.1216E+OO 4.6762E+OO 2.33E-Ol Cr-51 3.26E-03 3.7540E-03 3.2763E-02 3.6517E-02 l.19E-04 Mn-54 l.68E-03 4.1670E-03 8.3592E-Ol 8.4009E-Ol l.41E-03 Fe-55 l.26E-03 4.1920E-03 l.5291E-03 5.721 IE-03 7.22E-06 Fe-59 3.16E-04 l.1800E-Ol l.1923E+OO l.3103E+OO 4.14E-04 Co-58 4.84E-03 2.0490E-Ol 9.7586E-Ol l.1808E+OO 5.72E-03 Co-60 5.58E-04 9.6840E-02 2.5043E+OO 2.6011E+OO l.45E-03 Zn-65 5.37E-04 6.8940E-03 5.8169E-Ol 5.8858E-Ol 3.16E-04 Sr-89 l.47E-04 5.7300E-Ol l.3636E-04 5.7314E-Ol 8.44E-05 Sr-90 l.26E-05 l.9630E-Ol O.OOOOE+OO l.9630E-Ol 2.48E-06 Sr-91 l.02E-03 6.5050E-Ol 6.9508E-Ol l.3456E+OO l.37E-03 Y-90 l.26E-05 9.3610E-Ol O.OOOOE+OO 9.3610E-Ol l.18E-05 Y-9lm 4.93E-04 O.OOOOE+OO 5.5557E-Ol 5.5557E-Ol 2.74E-04 Y-91 5.47E-06 6.0600E-Ol 3.6147E-03 6.0961E-Ol 3.34E-06 Y-93 4.46E-03 l.1721E+OO 8.9414E-02 l.2615E+OO 5.63E-03 Zr-95 4.IOE-04 l.1990E-Ol 7.3474E-Ol 8.5464E-Ol 3.51E-04 Nb-95 2.95E-04 4.4970E-02 7.6430E-Ol 8.0927E-Ol 2.38E-04 Mo-99 6.75E-03 3.9570E-Ol l.6238E-Ol 5.5808E-Ol 3.77E-03 Tc-99m 5.0IE-03 4.8500E-03 l.4263E-Ol l.4748E-Ol 7.38E-04 Ru-103 7.89E-03 6.7400E-02 4.8394E-Ol 5.5134E-Ol 4.35E-03 Ru-106 9.47E-02 l.OIOOE-02 O.OOOOE+OO l.OIOOE-02 9.57E-04
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 14
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Determination of 100/EBAR - continued A(i) E(i) E(i) E(i) Activity Beta Energy Gamma Energy Total Isotope [uCi/gm] [MeV/dis] [MeV/dis] A(i)*E(i) Rh-103m 7.89E-03 3.4620E-02 2.2148E-05 3.4642E-02 2.73E-04 Rh-106 9.47E-02 7.0960E-Ol 2.0348E-Ol 9.1308E-Ol 8.65E-02 Te-129m 2.00E-04 1.9150E-O 1 9.4832E-02 2.8633E-Ol 5.73E-05 Te-129 2.57E-02 5.2260E-Ol 5.9948E-02 5.8255E-Ol l.50E-02 Te-13lm l.59E-03 2.1240E-Ol l.4092E+OO 1.6216E+OO 2.57E-03 Te-131 8.26E-03 7.5970E-Ol 4.1616E-Ol 1.1759E+OO 9.71E-03 Te-132 1.79E-03 1.0020E-Ol 2.0507E-Ol 3.0527E-Ol 5.47E-04 Ba-137m 9.79E-03 6.4260E-02 5.9729E-Ol 6.6155E-Ol 6.48E-03 Ba-140 l.37E-02 3.1500E-Ol 1.9522E-Ol 5.1022E-Ol 6.98E-03 La-140 2.64E-02 5.4050E-Ol 2.3074E+OO 2.8479E+OO 7.52E-02 Ce-141 1.58E-04 l .6930E-Ol 1.0181E-O 1 2.711 IE-Ol 4.28E-05 Ce-143 2.96E-03 3.8420E-Ol 3.4335E-Ol 7.2755E-Ol 2.15E-03 Ce-144 4.21E-03 9.1300E-02 3.2865E-02 1.2417E-Ol 5.23E-04 Pr-143 2.96E-03 3.1430E-Ol O.OOOOE+OO 3.1430E-Ol 9.30E-04 Pr-144 4.21E-03 1.2258E+OO 3.IOIOE-02 l.2568E+OO 5.29E-03 Np-239 2.32E-03 1.2380E-Ol 2.0845E+OO 2.2083E+OO 5.13E-03 Total 5.82E+OO 3.44E+OO EBAR 5.91E-Ol RCS Specific Activity Limit 169.14 The definition of EBAR or Eis as follows: "E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant." The values for Ei in the above table were obtained from reference 9 and the values for Ai are from WBNNAL3-003. The value of E is determined as follows: Ebar = E = (I:Ai~)/ (:LAD The value for E calculated in Table 2 is 0.591 MeV/dis. This results in a non-iodine specific activity limit (100/E) of 169.14
µCi/gm. The total specific activity of the expected coolant is 5.82 µCi/gm.
Therefore, the values for noble gases in the design reactor coolant will have to be increased by a factor of 169.14/5.82 = 29.06. The step sources (Ci/(µCi/gm)) to initialize the reactor coolant and the secondary side activities are: All cases: S=2.622E8 gm x IE-6 Ci/µCi x 29.06 =7.6195E3 (noble gases) S=2.622E8 gm xlE-6 Ci/µCi = 2.622E2 (tritium) Pre-accident iodine spike case (initial concentration= 14 µCi/gm): S=2.622E8 gm x IE-6 Ci/µCi x l l.348[µCi/gm I-l 3Ir 1 x 14µCi/gm 1-131 = 4.166E4 (iodines) Accident initiated iodine spike case (initial concentration= 0.265 µCi/gm):
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 15
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 S=2.622E8 gm x lE-6 Ci/µCi x 11.348 [µCi/gm I-13U 1 x 0.265 µCi/gm I-131= 7.885E2 (iodines) Secondary side, all cases, steam generator w/ leak, release to environment. (concentration= O.lµCi/gm) which is due to dryout (from reference 8, the initial steam from the defective steam generator is 117 ,200 lb): S = 117,200 lb x 453.59 gm/lb x lE-6 Ci/µCi x 4.150E5 [µCi/gm I-13U 1 x O.lµCi/gm I-131 = 2.206E6 (iodines) S = 117,200 lb x 453.59 gm/lb x lE-6 Ci/µCi = 5.3 l 6El (tritium) Secondary side, all cases, steam generators without leak (initial concentration= 0.1 µCi/gm): S = l.593E8 gm x lE-6 Ci/µCi x 4.150E5 [µCi/gm I-13H 1 x O.lµCi/gm I-131=6.610E6 (iodines) S = l.593E8 gm x lE-6 Ci/µCi = l.593E2 (tritium) Continuous Sources: For the accident initiated iodine spike case, the iodine spike is 500 times the iodine release rate from the fuel. At steady state, the iodine release (production) rate is equal to the removal rate. The iodine removal is due to a) radioactive decay, b) removal by the letdown system, and c) removal through reactor coolant leakage. These terms are expressed as: P = Lremoval rates = decay + letdown+ leakage or P = 'A+ fL£1V + p,N where P =production rate [hr. 1]
'A =decay constant for the isotope in question [hr- 1] = ln(2)/T 112 fL =letdown flow rate= 120 gpm + 4.39 gpm = 124.39 gpm
£=letdown demineralizer efficiency= I (assumed so as to maximize removal/production rate)
V =volume of primary coolant= 5.78E5 lb= 2.62E8 gm p, =removal rate of iodine from the primary side due to pre-accident primary side leakage
= 11 gpm (I 0 gpm identified + I gpm unidentified)
T 112 = halflife taken from ref. 15 Note: All the above flow rates are converted to mass flow rates at STP. Removal rate of iodine from primary side to secondary side was not considered above, because of its relatively low rate (3 x 150 gpd = 0.3 gpm) compared with other terms. Production Rates for a Reactor Coolant Leak of 11 gpm (10 gpm identified+ I gpm unidentified) Half Life 'A [l!hr] fLEN [I/hr] PsN [l/hr] Prod rate P [I/hr] 500xP I-131 8.04 d 3.59E-03 l.08E-01 9.53E-03 0.1209 60.43 I-132 2.28h 3.04E-Ol l .08E-Ol 9.53E-03 0.4213 210.64 I-133 20.9 h 3.32E-02 l.08E-Ol 9.53E-03 0.1504 75.22 I-134 52.6m 7.91E-Ol l.08E-01 9.53E-03 0.9079 453.97 I-135 6.61 h l.05E-Ol l.08E-01 9.53E-03 0.2221 111.07 The accident initiated iodine spike of 500 times the increase in the iodine release (production) rate from the fuel is modeled as a continuous source: C =Volume x lE-6 Ci/µCi x Prod Rate x 500 x 1 µCi/gm I-131 equivalent conversion factor x I-131 equiv. where Volume= 2.622E8 gm Prod Rate = see table above 1 µCi/gm I-131 equivalent conversion factor= 11.348 (value determined in Table I a, this is to get the ANSI/ANS-18.1-1984 source into I µCi/gml-131 equivalent I-131equiv.=0.265 µCi/gm I-131 Continuous Source [gm-Ci/µCi-hr] for Accident Initiated Iodine Spike:
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 16
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Reactor Coolant Leak of 11 gpm ( 10 gpm identified + 1 gpm unidentified) Nuclide 0.265 µCi/gm I-131 I-131 4.765E+04 I-132 l.661E+05 I-133 5.931E+04 I-134 3.580E+05 I-135 8.758E+04 Flow Rates: The following flow rates/leakage rates for each component are: Flow from Reactor Coolant# 1 to Environment #4 all classes (consists of 1 gpm and is for leak in the steam generators, however the production rate of iodines is based on a total RCS leakage of 11 gpm (=lOgpm identified+ lgpm unidentified): F = 1.0 gpm x 60 min/hr x 3785.48 cc/gal x 1 gm/cc = 2.27 IE5 gm/hr Flow from Reactor Coolant #1 to Steam Generator without Leak #3 all classes: F = 3 steam generators x 150 gpd x 3785.48 cc/gal I 24 hr/day x 1 gm/cc= 7.098E4 gm/hr From reference 25, the initial steam released from the defective steam generator is 117,200 lb. From the non-defective steam generators (= "steam generators without leak" in this model) the mass release is 442,083 lb (0-2 hr), and 922,918 lb (2-8 hr). The accident releases end at eight hours. To take into account uncovery of the faulted steam generator, there is no iodine partitioning in the release to the environment (iodine partition coefficient= 1). The mass release representing 1 gpm primary leak is a flow directly to the environment. The reactor coolant release to the unfaulted steam generator is small relative to the secondary side mass, therefore partitioning is allowed per the SRP. The iodine partition coefficient due to steaming for the unfaulted steam generators to the environment is 100. These mass releases translate into the following flows: Flow from Steam Generators w/out Leak #3 to Environment #4: F = (442,083 lb)( 453.59 gm/lb )/(2 hr) = l .003E8 gm/hr (0-2 hr, noble gases, tritium) F = (442,083 lb)(453.59 gm/lb)/(100x2 hr)= I .003E6 gm/hr (0-2 hr, iodines) F = (922,918 lb)(453.59 gm/lb)/(6 hr)= 6.977E7 gm/hr (2-8 hr) (noble gases, tritium) F = (922,918 lb)(453.59 gm/lb)/(100x6 hr)= 6.977E5 gm/hr (2-8 hr) (iodines) The STP output is used as input to COROD (which determines control room operator dose) and FENCDOSE (which determines 30-day and 2-hour LPZ offsite dose). Some pertinent information from the COROD and FENCDOSE models used (but not changed) in this analysis are (from ref. 9): 30-day LPZ Offsite X/Q values [sec/cum]: I .784E-4 0-2hr, 8.835E-5 2-8 hr, 6.217E-5 8-24 hr, 2.900E-5 1-4 day, 9.811E-6 4-30 day 2-hr EAB X/Q values: 6.382E-4 Unit 1 Control Room X/Q (ARCON96 method): 3.85E-03 0-2 hr, 3.22-03 2-8 hr Unit 2 Control Room X/Q (ARCON96 method): 2.59E-03 0-2 hr, 2.12E-03 2-8 hr, Control Room volume: 257198 cuft Control Room makeup/pressurization flow: 711 cfrn (3200 cfrn prior to isolation, ref. 24*) Control Room total flow: 3600 cfm Control Room recirculation flow: 2889 cfm Control Room unfiltered intake: 51 cfm Control Room filter efficiency: 95% first pass, 70% second pass, 0% for tritium, 0% all elements prior to isolation Control Room occupancy factors: 100% 0-24 hr, 60% 1-4 days, 40% 4-30 days ICRP-30 dose conversion factors (internal to the codes)
- 3200 cfm has been deleted from l-47W866-4 R39 (ref. 24a), and has been measured to be approximately 2500 cfm (O-SI 31-A). The value comes from l-47W866-4 R20 (ref. 24b). The 3200 cfm will be retained in this calculation revision since this value produces conservative results.
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 17
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Figure 1: STP Model Pre-accident Iodine Spike Step Source 1 Reactor
~ Coolant 14 µCi/g 1-131 3*150 gpd = 450 gpd 1 gpm 3 Steam 2 Steam Step Source Generator Generator w/out Leak w/Leak 0.1 uCi/g 1-131 442,083 lb (0-2 hr) 922,918 lb {2-8 hr) 4 Environment Step Source 0.1 uCi/g 1-131 117,200 lb (0-30 min)
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 18
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Figure 2: STP Model Accident Initiated Iodine Spike Continuous Source Step Source 1 Reactor Coolant =500*1odine Production Rate 0.265 uCi/g 1-131 3* 150 gpd = 450 gpd 1 gpm 3 Steam 2 Steam Step Source Generator Generator w/out Leak w/Leak 0.1 uCi/g 1-131 442,083 lb (0-2 hr) 922,918 lb (2-8 hr) 4 Environment Step Source 0.1uCi/g1-131 117,200 lb (0-30 min)
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN I Page: 19
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Results The following results are based on a Tritimn Production Core (TPC). The results from previous revisions showed that the TPC bounds the conventional core. In the following, the pre-accident reactor coolant leak rate is 11 gpm (l 0 gpm known+ l gpm unknown). The primary to secondazy side steam generator post accident leak rate is l gpm to the faulted steam generator and 150 gpd to each intact steam generators. The results of offsite and control room doses shown in the tables below are based on the 1-131 equivalent conversion factors developed using RG-1. l 09 iodine inhalation dose conversion factors. The Unit 2 results are found in Appendix H. Unit l MSLB Offsite Doses (rem): Pre-accident Iodine soikimz case Accident Initiated Iodine St>ikinll: (500) case 1-131 eQUivalent: 14 uCi/em 1-131 eauivalent: o.265 uci/2n1 2-hrEAB 30-davLPZ limit 2-hrEAB 30-dayl..PZ limit gamma 2.92E-02 l.l6E-02 25 gamma l.04E-Ol l.23E-Ol 2.5 beta 9.28E-03 4.35E-03 300 beta 2.55E-02 2.98E-02 30 Inhalation (ICRP-30) 2.63E+oo l.27E+oo 300 Inhalation (ICRP-30) 3.20E+o0 4.59E+oo 30 TEDE l.92E-01 8.76E-02 25 TEDE 3.49E-Ol 4.69E-Ol 2.5 Unit l MSLB Control Room Doses (rem) Using ARCON96 X/Q values: Pre-accident Iodine Accident Initiated S1>ikin11: Iodine Snilcin11: limit Gamma 7.l2E-03 1.25E-02 5 Beta 6.37E-02 9.98E-02 30 Thyroid (ICRP-30) l.32E+ol l.73E+ol 30 TEDE 4.66E-01 6.35E-Ol 5 The following Unit 1 margins were calculated from the doses in the Tables above. Where: margin= limit- dose, and percent= 100 x (lirnit-dose)/limit Unit 1 MSLB Offsitc Dose Margins: Pre-accident Iodine soikin11: case Accident Initiated Iodine SJJiking {500) case 1~131 1-131 equivalent: 14 UCill!ID eauivalent: 0.265 l.lCi/l!Jll 2-hrEAB 30-davLPZ limit 2-hrEAB 30-davLPZ limit manrin oercent manrin oercent manlin nercent manlin oen:ent stamma 24.97 99.88 24.99 99.95 25 11:amma 2.40 95.83 2.38 95.07 2.5 beta 299.99 100.00 300.00 100.00 300 beta 29.97 99.90 29.97 99.90 30 Inhalation Inhalation OCRP-30) 297.37 99.12 298.73 99.58 300 OCRP-30) 26.80 89.35 25.41 84.69 30 TEDE 24.81 99.24 24.91 99.64 25 TEDE 2.15 86.00 2.03 81.20 2.5 1bis Page replaced by R16
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN IPage: 20
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Unit 1 MSLB Control Room Dose Margins Using ARCON96 X/O values: Pre-accident Iodine Accident Initiated St>ikin2 Iodine Spiking limit manrin nercent maririn t>CrCCnt Gamma 4.99 99.86 4.99 99.75 5 Beta 29.94 99.79 29.90 99.67 30 Thyroid (ICRP-30) 16.77 55.90 12.71 42.37 30 TEDE 4.53 90.68 4.37 87.31 5 Discussion and Conclusion The offsite doses due to a MSLB with pre-accident iodine spiking has 10CFRl00 limits of 25 rem gamma (whole body), 300 rem beta (skin), and 300 rem thyroid The offsite doses due to a MSLB with accident initiated iodine spike (factor of500) has limits of 10% of the 10CFRlOO limits or 2.5 rem gamma, 30 rem beta, and 30 rem thyroid (ref. 10). The control room operator doses limits from 10CFRSO App. A GDC 19 are 5 rem gamma, 30 rem beta, and 30 rem thyroid. With the Technical Specification limits of 0.265 µCi/gm I-131 equivalent steady state (and .14 µCi/gm maximum, see Assumption 6), the control room and offsite doses do not exceed the limits with a 1 gpm leak in the faulted line and 150 gpd in the unfaulted lines. These apply to Unit 2 also. Unit 1 doses bound the Unit 2 doses, except for the 30 day LPZ offsite doses for the accident initiated iodine spike. Note: these limits are based on a maximum 0.1 µCi/gm I-131 limit in the secondary side and using ARCON96 XIQ values. If the secondary side limit were to be reduced, then the primary to secondary side leakage and the primary side I-131 concentrations can increase. The Tritium Production Core (TPC) does not affect the limits above, because the limiting doses are the thyroid doses. The tritium affects only the beta dose and TEDE (Total Effective Dose Equivalent). The TPC obviously bounds the non-TPC configuration. This calculation is conservative because it models the mass releases as linear within each time interval. This allows larger iodine releases for the accident initiated iodine spiking cases because iodine increases over time in the reactor coolant. In reality, the mass releases are greater at the beginning of the accident, and decrease over time. For the pre-existing iodine spike (which is not the limiting case), this has little effect, since the decay of short lived isotopes is compensated by the buildup of iodine in the unfaulted steam generators due to reactor coolant leakage. Note on methodologies used: This calculation determined the offsite and control room operator doses using the RG-1.109 iodine inhalation dose conversion factors for thyroid to determine the 1-131 equivalent conversion factor for the source term. This I-131 equivalence is consistent with the Technical Specification 1.1 (ref. 33). The thyroid dose is reported only based on the ICRP-30 methodology. TheTEDE is calculated from the computer codes, COROD and FENCDOSE using parameters derived from the ICRP-30 data. This Page replaced by Rl6
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN I Page: 21
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break References
- 1. WBN Technical Specification 3.4.16, Amendment 81
- 2. WBNNAL3003 R5 "Reactor Coolant Activities in Accordance with ANSI/ANS-18.1-1984"
- 3. Computer Code STP R7.1, code I.D. 262165 (under dose code program mgr version 1.1)
- 4. Computer Code FENCDOSE R5, code I.D. 262358 (under dose code program mgr version 1.1)
- 5. Computer Code COROD R7 .1, code I.D. 262347 (under dose code program mgr version 1.1)
- 6. WBNAPS3-053 R5 "Steam Generator Leakage Det.ection with the Condenser Vacuum Pump Air Exhaust Monitor (1,2-RM-90-119)"
- 7. WBNAPS3-043 R2 "Shielding Calculation For the Steam Generator Blowdown Demineralizer System" (A search indicates this calculation was not cited)
- 8. Memorandum from J.W. Irons to W.L. Elliott, WAT-D-9489, "Verification of Data in FSAR Table 15.5-16" RIMS#
T33 930927 823
- 9. WBNTSR008 R15 "Control Room Operat.or and Offsite Dose Due to a Steam Generator Tube Rupture"
- 10. NUREG-0800R2section15.1.5
- 11. WAT-D-10690, Nov.9, 1999, Memorandum from John W. Irons to J.E. Maddox "SLB Leak Rates" RIMS# B44 991109 002
- 12. WAT-D-10724, February 10, 2000, Memorandum from John W. Irons to J.E. Maddox "SLB Leak Rates Conversion" RIMS# T71 000217 928
- 13. N3-15-4002 R15 System Description For "Steam Generator Blowdown System"
- 14. FSAR Table 11.1-2 (note: this information is used only for comparison with reference 2, and not used as design input)
- 15. Lederer and Shirley, "Table of Isotopes", seventh ed.
- 16. N3-62-4001 R31 System Description for "Chemical and Volume Control System"
- 17. WBNNAL3-002 R4 "100-Day LOCA-DBA Source Terms for the EGTS and ABGTS Filters, Containment, Sump, and Shield Building Annulus" (A search indicates this calculation was not cited)
- 18. WBNAPS3-050 R7 "Determine the Main Control Room Intake Monitor (O-RE-90-125, -126) Setpoints and Poet Accident Air Intake Concentrations"
- 19. WB-DC-40-70 Rl "Accident Analysis Parameter Checklist (AAPC)", Figures 4.3.2-13 and-25
- 20. WBNAPS3-104 R4 "WBN Control Room 'X/q'
- 21. WBN Technical Specification 3.4.13
- 22. NSAL-00-004 "Nonconservatisme in Iodine Spiking Calculations"
- 23. WBN Technical Specification 3.7.14
- 24. WBN drawing (a) 1-47W866-4 R39, (b) 1-47W866-4 R20
- 25. WCAP 16286-P "Watts Bar Unit 1 Replacement Steam Generator Program NSSS Engineering Report" January 2005 26a. WBT-D-1202 OctDber 22, 2009 "WBS 5.2.11 Revised Steam Releases for Dose" 26b. LTR-CRA~l03 Rev.1 "Watte Bar Unit 2 Completion Project- Results of Steam Releases for Dose Calculations"
- 27. EDCR 54956
- 28. WBNTSR-028, RlO "Main Control Room Emergency and Normal Air Intake Monitors Required Range, Safety Limits, Response Time and Accuracy"
- 29. Regulatory Guide 1.109 Rl
- 30. User's manual for COROD (software ID number 262347) R5
- 31. PER 327968
- 32. PER 327956
- 33. WBN Technical Specification 1.1Amendment81
- 34. DCN 52012
- 35. DCN 61599 1bis Page replaced by Rl6
~~~~~~~~~
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 22
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Appendix A: Justification for Using ANSI/ANS-18.1-1984 Expected Coolant Spectrum The choice of iodine spectrum is fairly important, since several isotopes have short halflives. Noble gas spectrum is not at important because the noble gases contribute only to the gamma and beta doses which are orders of magnitude from the regulatory limits, whereas the limiting doses are thyroid (iodine influenced). Results may be affected when accident times are on the order of the decay of the short lived isotopes. There are several possible spectra available. The spectrum chosen for this analysis is the one that most closely resembles the actual spectrum present at WBN. From the surveillance tests l-SI-68-28 performed on 7/10/00 and 4/9/01 (see Attachment 1), the following concentrations were determined: RCS activities 7110100 RCS activities 4/9/01
µCi/gm µCi/gram µCi/gm µCi/gram RCS RCS RCS RCS Gaseous Degassed Gaseous Degassed Ar-41 l.303E-02 F-18 l.179E-Ol Ar-41 2.696E-03 F-18 1.116E-01 Kr-85M l.915E-04 Na-24 9.169E-04 Kr-85M 2.013E-04 Na-24 2.060E-03 Kr-87 4.575E-04 Mn-56 9.313E-05 Kr-87 4.809E-04 Mn-56 2.088E-04 Xe-133 9.565E-04 Co-58 5.019E-04 Kr-88 4.982E-04 Co-58 6.218E-04 Xe-135 l.429E-03 Nb-95 3.132E-05 Xe-133 1.202E-03 Co-60 2.776E-05 Xe-135M 7.364E-04 I-131 6.070E-05 Xe-135 1.676E-03 Nb-95 2.794E-05 Xe-138 l.796E-03 I-132 l.459E-03 Xe-135M 1.105E-03 1-131 3.881 E-05 I-133 8.208E-04 1-132 1.165E-03 I-134 2.694E-03 1-133 6.105E-04 I-135 l.608E-03 1-134 2.334E-03 Xe-135 8.914E-05 1-135 1.158E-03 Xe-135M l.406E-02 Xe-135 1.380E-04 Cs-138 2.395E-03 Xe-135M 1.972E-02 Cs-138 2.195E-03 Two potential spectra are from WBNNAL3-003 (Reactor Coolant Activities in Accordance with ANSI/ANS-18.1-1984) and from the FSAR Table 11.1-2 (Historical Design Activities). The iodine concentrations and relative concentrations for each spectrum are as follows:
7/10/00 7110/00 419101 419101 WBN actual WBN actual WBN actual WBN actual
µCi/gm relative fraction µCi/gm relative fraction I-131 6.070E-05 0.0091 3.881 E-05 0.0073 I-132 l.459E-03 0.2196 1.165E-03 0.2195 I-133 8.208E-04 0.1236 6.105E-04 0.1151 I-134 2.694E-03 0.4056 2.334E-03 0.4399 I-135 l.608E-03 0.2421 1.158E-03 0.2182 sum: 6.643E-03 5.306E-03 ANS 18.1 ANS 18.1 FSAR 11.1-2 FSAR 11.1-2
µCi/gm relative fraction µCi/gm relative fraction I-131 0.0477 0.0448 2.5 0.2461 I-132 0.225 0.2115 0.9 0.0886 I-133 0.149 0.1401 4 0.3937 I-134 0.364 0.3422 0.56 0.0551 I-135 0.278 0.2614 2.2 0.2165 sum: 1.0637 10.16
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 23
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 As can be seen, the FSAR historical design concentrations do not reflect the actual measured concentrations. The FSAR values are weighted too strongly in favor of I-131 (24.6% of total as opposed to < 1% of the actual total). By comparison, the ANSI/ANS-18.1-1984 fractions are very close to the actual fractions. The worst fit was for I-134 which was 40.1 % actual versus ANSI/ANS-18.1-1984 34.22%. The I-131 is slightly over predicted by ANS-18.1 (0.9% on 7/10/00 and 0.7% on 4/9/01 versus 4.48%), however this difference is not as large compared to the FSAR fraction. The ANSI/ANS-18.1-1984 spectrum overall fit is much better than the FSAR spectrum, therefore it can be concluded that the use of the ANSI/ANS-18.1-1984 spectrum is acceptable.
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 24
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Appendix B: Determination of Letdown Flow Rate Uncertainty The purpose of this appendix is to determine bounding errors for the measurements performed on the orifice restrictor flows using the Letdown Heat Exchanger Flow loop (1-F-62-82) during Preop Test Instruction PTl-062-03 RO. Following these tests, a loop check was performed for the computer point F0134A by injecting a signal into the transmitter and reading the display on the computer. To determine the total loop error, the unmeasurable errors must be combined with the errors present at the time of the loop check. WBN NESSD 1-F-62-1 will be used as a guide for determining the unmeasurable errors for loop 1-F-62-82 since it contains the same model flow element and a similar model transmitter. According to EMPAC, the flow element is a Vickery Simms Model MK-52 and the transmitter is a Foxboro E-13DM. Millers Flow Measurement Engineering Handbook, Third Edition, Chapter 6, Table 6.1 states that Square Edged orifice flowmeters have an accuracy of +/-1-2%URV (upper range value) of the flow rate. A value of +/-2% will be used for the orifice. The loop check performed by WO 94-14264-10 (following pages) gives as found data. The largest error at 50 GPM was 1.36 GPM (50 - 48.64) or 0.68% CS (1.36/200 = 0.68%). The largest error at 100 GPM was 0.48 GPM (100 - 99.52) or 0.24% CS (0.48/200 = 0.24%). The largest error at 150 GPM was 0.06 GPM (150 - 149.94) or 0.03% CS (0.06/200 = 0.03%). Since the plant had not been started at the time of these tests, radiation was not present and need not be considered. Errors for temperature and power supply effect will need to be included. Since there is no data on actual temperature conditions, an enveloping value must be used. Environmental drawing 47E235-46 R5 gives the max abnormal temperature range as 50 - 110 °F for coordinates UA6 I El 737 where the transmitter is located per EMPAC. The transmitter is a model E-13DM per EMPAC. The product specification sheets (following pages) give the ambient temperature effect as +/-1 % per 50 °F for any span between 200 to 850" water. The transmitter will normally be calibrated at room temperature which will be between 60 and 80 °F. A temperature shift of + or - 50 °F will encompass the temperature changes seen by the transmitter. Therefore for a temperature range of +/-50 °F, the temperature effect will be +/-1 % CS d/p. The power supply effect is given as 0.1 % CS for a 10% change in voltage. Thus Power supply effect is 0.1% cs d/p. All errors for the computer should be reflected in the loop check.
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 25
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Utilizing Equation 3-24.8 of W WCAP-12096, Rev. 8 "Westinghouse Setpoint Methodology for Protection Systems, Watts Bar Units 1 and 2, Eagle 21 Version," the unmeasured transmitter errors can be converted from percent error in full scale dip to error in percent full span at a specified point, where Fm is the maximum flow rate of 200 GPM, and Fn is the nominal flow rate (i.e. 50, 100 or 150 GPM). EPFS (Flow) = (Axxx I 2) * (Fm I Fn) TemPerr(Flow)@50GPM = (Temperr (d/p) I 2) * (200 I 50) = +/- 2% CS Flow Temperr(Flow)@100GPM = (Temperr (d/p) /2) * (200I100) = +/- 1% CS Flow TemPerr(Flow)@150GPM = (TemPerr (d/p) I 2) * (200I150) = +/- 0.67% CS Flow pwr SUPPerr(Flow)@50GPM = (pwr SUPPerr (d/p) I 2) * (200 I 50) = +/- 0.2% CS Flow pwr SUPPerr(Flow)@ 1OOGPM = (pwr SUPPerr (d/p) I 2) * (200 I 100) = +/- 0.1 % CS Flow pwr SUPPerr(Flow)@ 150GPM = (pwr SUPPerr (d/p) I 2) * (200 I 150) = +/- 0.067% CS Flow 2 2 2 2 05 Thus total loop error= (FEerr + Loop check err +Temperr (Flow) +pwr SUPPerr (Flow) )
- 2 2 Total loop error @ 50 GPM = (2 + 0.68 + 22 + 0.2 2 ) 0 *5 = +/-2.92% CS = +/-5.84 GPM 2 2 2 Total loop error@ 100 GPM = (2 + 0.24 + 1 + 0.1 2 ) 05 = +/-2.25% CS = +/-4.5 GPM 2 2 2 2 05 Total loop error @ 150 GPM = (2 + 0.03 + 0.67 + 0.067 ) = +/-2.11 % CS = +/-4.22 GPM Total loop error at 120 GPM can be determined by linear interpolation between 100 and 150 GPM. The value will be conservative since the error is nonlinear and is a function of the square root of the d/p values above and the actual loop recorded values which also follow a square root curve.
Total loop error @ 120 GPM =+/-I error @ 100 GPM + 20(error @ 150 GPM - error @ 100 GPM) I (150 - 100) I Total loop error@ 120 GPM = +/-[4.5 GPM + 20(4.22 - 4.5)/50] = +/-[4.5 GPM + (-0.11)] = +/-4.39 GPM The following references were used in preparation of this appendix. Revisions to these references will not impact this appendix; so the references are 'information only' in lieu of 'design input'. 1 WBN NESSD 1-F-62-1 R1 (Methodology & guidance) 2 EMPAC (Manufacturer, Model number and location) 3 Millers Flow Engineering Handbook, Third Edition, Chapter 6, Table 6.1 (Orifice accuracy) 4 WO 94-14264-10 (loop check data) - see next page 5 Drawing 47E235-46 R5 (environmental data) 6 Foxboro product specification sheets (transmitter accuracy data) - see next pages 7 WCAP-12096 R 8 (methodology for converting d/p error to flow error) Prepared Lynn Cowan Date 6/4/01 Checked - - - - - - - - - - - Date _ __
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 26
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Supporting documents for Appendix B:
~**
WORK ORDER FORM WO NO: 94-14264-10 ORIGINATION DATE: 06/22/94 RF NO: C254025 ORIGINATOR: JAMES R RIVERS EXTENSION: 3181 PRIORITY: JC EQUIPMENT IDENTIFIER : WBN-l-LPF-062-0082-EQUIPMENT DESCRIPTION: EXCESS LETDOWN HTX FLOW EQUIPMENT CATEGORY : QR SR lE TYPE OF MAINTENANCE: OTHER MAINTENANCE PROBLEM DESCRIPTION: PERFol(M RIST 7ESTCALIBRRTI(,JJ oF ?YS IC2 ?Of&B 11.JSTRUMfAJTS LISTED ON THE WR CARD FCR PTI-OIP2-0~ l:+CC.EPTf.WCE CRITERIA JOB LOChTION VARIOUS LOCATIONS, SEE SSD LOCATION CODE AlOO - AB ALL AUXILIARY BUILDING GENERAL AREA WORK DES~RIPTION PERFORM POST TEST CALIBRATION OF l-LPF-062-0082, AS REQUIRED, FOR PTI-062-03 ACCEPTANCE CRITERIA LCO: YES [ ) NO [ x l LENGTH: µI.e.. LCO EXPIRES: ,.) fa_ __ SECT XI R/R: YES [ ) NO [X) --m>RDs: YES [ .l NO [XJ RWP REQ YES ( ] NO (X] RWP #: ,.;/IL ALA.RA: YES ( ) NO (X] TAGGING REQ; YES [ SCAFFOLD YES [ NO [X] H*o. I : f
.u e.. SHUTDOWN: YES [ ] NO [X]
NO [X] INSUL: YES [ ) NO [X) PERMITS REQ; NONE DISCIPLINE: MIG EST HOURS ; 4.0 TASK TOT: 8.0 MhN H0URS PRE-MAINT TEST REQ. DURATION: 4.0 HOURS NONE POST-MAINT TEST RtQ.: SEE WORK INSTRUCTIONS SIGNATURES AND DATE / PLANNER: r;rtM!L..l- e:v&.:. ~ -J). -Cf't TECH REVIEW:~~~.tru-9'{: COG SUPV: ~ ~/.,_y"f'y SUT ENGR: 1n&~JA,. ~llL-1! *~
\
1" * -
- Y,
,) * ~
- l
- _,_ ' * .,j
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 27
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 SSD-l~LPF-62.,.82-D PAGE 10 OF i7 PAG~ 4Q (;.~*.~ l0 REVISION 01 * . lNSTRUHENT LOOP CALIBRATION RECORD WID NO: WO 94-142G4-10 _ LOOP COHPONENT S ID~-1:;:r:62:92-----r;;sTRUHENT NO: 1-FI-62-82 HEAO: N/A
~:~: 5 ~g_11;_J__
I 11&TE: VISUAL TEST INPUT *:--R-EQU-IR_E_O_ _ _ _ _A_S_F_OUND AS LEFT POINT ( IN Wt ) ( GPH ) LO LlHIT AS FOUND HI LIHIT LO LIHIT AS LEFT HI LIHIT
---,-- ---,2~-5-- ---2-8.-3-i--Z-3-.3-I---- --3-3-.3- --2-3~J- ------- ----;----
---i--- -j9~,-- ----- 50 .0 1*---45.0
--- 55.0 45.0
---;-- --1;&~3--:- ---,~ --gs.a- - - - ------ --------- 105.0 155.0 200.0 155.0
----- - - - - - -----1----,,..tF
---a-- ---i9.l-- -----:J...-.:;___ -----
7 l!i6.J 100.0 45.o ____ ,___ ---- __ ____ 105.0 55.o 95.0 45.o 105.0 55.o 23.3 33.3 23.3 33.3 ________ ::::=:~~::~:::::::::::=::) I INSTRUHENT NO: LOG Pt F0134A H&TE; VISUAL TEST REQUIRED AS FOUND AS LEFT POINT GPH
------ LO LIHIT AS FOUND HI LIHIT LO LIMIT AS LEFT HI LIMIT
(
)
28.30 23.18 ).5.{,,}__ 23. 18 32.38 .38
~2 SO.DO 52.26 3
-----100.00 47.34 '-li.JL 98.66 ??,i,J 101. 14 -98.66- ----<
52.26 47.34 ~ 101. 14
-----150.00.
--- --- ----- -\.j 1So.6a
------ 149.08 15'0.0/
4 150.68 s ---- ~~- -~ -~- 195,00 194.32 19'1.?2 195.60 194.32 ll.. 195.60 6 150.00 149.08 I '/'I. 'Jf 150.68 149.08 1S0.6e 7 100.00 98.66
'11.5;J. ----
101.14 98.66 4. 101.14 8 50.00 47.3'4 L/8./r,'1_ _52.26 9 28.30 23.18 I .;iS.~J 32.38 --~- 23. l 8 ~:r -~ ---=~~~~- 32.38 R:emar!irr: Non~ Function: Lotdown Heat Exch Flow Reviewed by: Gary L. Hyden Approved by: Ed Hall £:_\'I Date: 03/lil/94
Calculation No. WBNAPS3077 J Rev: 014 I Plant: WBN 1Page:28
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 I ...
-l'OXBORO
- General Specification r
.* .*4 ~ Secitl dip C.CC T,_ltten m * - diff*
....... ,.....,.. ........ of 0.5 to o.sso lndteo (0.127 1D 0.211500 111111) of ...... at mtlc .,.._,,.."" to 6000 p11 c.czo kwcm21. TIMy tratasm1t
- proportlonat 10 to so or 4 1D 20 lllA ck tleMI. cmt onllnaiy ul'llhleld<<l 1Md1,.
~
1D ~ loc8ted up to_.. thouund fff( from th* point of _ a n t . FEATURES
~-
TrouiH-FrM Coclt1n.>c:tion E130L E130M E130tl Hittt P e r l - - Exctll.,.t R~cibolity PERFORMANCE
&. of c.ilibmionAcriurtment-Wide Rangos Copability ~rle'(
Stc~;;;"JJJS mml ::to.sic: of""'"
-n
$t>ibie Foret Balanct Synem 526 to 850-inct> - ;:tQ. 75'1 of - n 113360 to 21590 mm)
Poaitm °"~Pro~ o..d Bond . o.os" of R_...-ity:E130LSer"=< 0.15"of_.,- T~ Housing W~t end MOittu...Proof E130M.E130HSer; 0.101'of-n
~ Q.10% of_,,_
AP\)!~ Venatifrty R..,.-,Cibaity: E 130L Series 0.201.:, * .... .,
. 'E130M.E130H5eues _ _ .* 0.15%ol.P.n (lnctudes effecu of Hvst<<csis. F1..,..tabl!i1y, Oeod Bond *nd
~/Intrinsically Ssfe Oiih q,.;, 1~ periodl BASE TRANSMITTER STANDARD Sl'ECIFfCATIONS
*
- StyloB '
~ Fully ~umblc between range limits of capsule. Output "-*I
~* -
Mmi:lmum Pnicen Teinpemu;., 250 F (120 Cl at cap-sule *
- 0utputg,_.
~~Lood (ellmd N--'--"'
V"'-fnlm c-Ackl MWoeum I Mulm..,. S.--UM Ambient *T~ Omits -40. to +180 .F (-40 to
+82 CL With remote amplifier. -40 to +250 F (~ to
~ ... 20 to..iso 0
.cao1o1 I '60 HO NVok IOVok
....,_..,._.;om-. .
-i:120Cl
- w~ ,_.. ............. deribulioft,...... Include .......
..,....0 tolOO ...... leed Bottint Steel cap screws and nuu thC'Ough body and process c,onnec:tOtS Supply Vol1:1g1 Lkn1tl 24 to 60 volts d-c with 4 to 20 rnA output alld 63 to 100 wits d< wilh 10 to SO mA C-- l1veaded cast aluminum seated on Buna-N O...ing .:KJtput from sieoarate power supply uriit.
seal. Blue textured Wiyt fmish..
~ Vohagt Ef*t Z<<o shift will bl! lea than O. t %
~ Clmdfkldon NEMA4 watertight of SI*\ fOt' a 10% d>ang!! in YOl!ilge within Stated limits.
Eltctnlnic Tllnllllltt9r Md Ampllfltr Solic11111c E!Ktric ctaalflc:stlon Explosionproof ~ t."Groups C 1nd 0, Oivltlon t. . verd1 Elec:lric:lll Cocw-=lloM Two S.foot lead5 from 112-ll\Ch Mounting Direct to process witli *b<ackat for 2-incn hori-
~ mnduit connectiO<'IS zontal or pipe.
-.-.~ .. E13DLS.W. E130M S<<ies E130HSet*
R-..1.Wu Low R.nge Capsule 0-5 to 0.25- w.iter (0-127 to ().{;3Smm) Mediurn ~ C-le - 0.20 to 0:205- wltec 10.508 to 0.5207 mml 0.20 to 0.205- -ter (0.506 to 0.5207 nvnl tf'~ Range Cepsuf!! - ().200 to G-850"' water '().200 to G-850~ W..I<< l().6()80 to ().21 S90 mm I (Q.~O to 0.21590 mml. (
""'-<I In U.SA ;GS 2A-1C1. E N""""'°"'1Q71
Calculation No. WBNAPS3077 I Rev: 014 TPlant: WBN ] Page: 29
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Cd t' C STANDARD SPECIFICATIONS (Continued I E130LS<<la E130MS.iff E130~ s.riet Wm.dPWU: Body and Process Coon Cad ml um plated f orged c;ar. . Cadmium
- I ,. ... . I plated f orged Cir*_ """moum p 1a t.,., __, . ,
rorged ar- I
., bon steel or f0<ged 316 SS bon steel or forged 316 SS bon stool or fOC"ged 316 SS Oiaph1'3111' C-ipsule and Force B a r - - - - 316 suinle*uteel 0
316 suinless steel 316 suioiless steel Foroe e¥ Seal - - - Cobalt nidr;el alloy Cobalt nickel alloy Cobalt nickel alloy Capsule Gasket - - - 316 stainless oteel 316 s*..1:*.*css steel 316 staioiless steel Prociess Conn Gasket - TFE TFE Glass filled TFE Fon:e Bar Seal Gast.et - Silicone elastomer Silicone elastomer Buna-N
~up Pi.le 3113 stainles:uteel 316 suinlcs:s steel I 316 stainleoo steel
-Max.lrnum Stade,.,_,. soO psi 05 kg/c;m2l 2000 psi 1140 kg/cm21 6000 psi {420 kglcm2)
~~- 114 OC" 112 NPT female OC" 112 inch Sch 80 welding 1/4 °' 112 NPT female "'
1/2 inch Sch 80 welding 1/4 or 1/2 NPT or body machined to ~t 9/16-neck. as specif":". nedl:. as specified. - 18 Amil'lCO fittings. as spec-ified.
~Tempwnin .:1:.1.0% per 100 F (55 Cl Medium Range Capsule: Medium Range Capsule; Effect (Z.,,-o shift in changs at 25** (635 mml .:1:.1.0'JQ per 100 *F 155 Cl +/-1-0% per 100 F .<~? q percent of span) water: .t.1.0% per 40 F (22 changeat 100** {2540.mml change at 100.. (2540 niml Cl change at 5.. (127 mmJ water: x 1.0% per 125 F water: .:1:.t.0% per 125 F watef". !69 CJ at 205** (5207 mm) (69 Cl at 205M (5207 mm) water.+/- 1,Q~ per 40 F 122 water; +/-1.0% P'!'" 40 F !22 Cl at 25** 1635 rll!)l water* Cl at 25- (635 mml water.
High Range Capsule: Less
- High Range Capsule: Leo.
than*+/-1% per 50 F (28 Cl than +/-1% per 50 F 128 Cl
. change for' any $pan be- change for any span be-tween 200 to aso** 15080 t~ 200 to 850.. (5080 Potftlon _ _ _ _ _ , to 21590 mml water. to 21590 mml water.
Transmitter should be mounted with -capsule in
~tic;il position.,
. "'*tion J:tt.ct _.__ 1..:. 3%
Maximum of less.than Maximum of 'ies:s *tha~ 31t. ** z~o shirt foe" 90 degree tilt zero shif:t for 90 degree.tilt
- _Vlfndon _ _ __ : of innrumeot in any plane of instrilment in any plane less thatl 1.5" zero shift Less 1han 1% zero shift_foc Less than 1" zero shift for for,,;bration to 1.5G in any vibration to 2G in any vibration to 2G in any
! plane. at frequencies less plane. plane.
lhan 80 Hz. I Siatlc: Pr.._ EH.ct-* Maximum zero shift less Zero shift less dian 0.5" Zero shirt less than 1.S" than 0.5" of ~ for !iOO span for 2000 psi (140 t.g/ span for O.s:lOO psi (o-420 . psi (35 kg/cm2J change. cm2J change at 50 to 850"' t.q/cm2) chaoige at 50 to (1270 to 21 mo nml water. 8SOM 11270 to21590 mml 1.()% spat! for 1000 psi (70 water or 0.5% span for allY t.glan21 change at 20 to 2000 psi (140 kglan21 SOM l508to1270mml- change; 2.0% span for O to ter. 6000 psi IO to 420 t.glcm2) d\ange at 20 to 50" (508 to 1270 nvnl water or 1.°" span for any 2000 psi (140
.kg/crn2) changa*.
- 0-..Dllw ' - - - - 16 118" (410 mini H" 13 114** (337 mnil H IC" 14 1~ 1368 mml H x
** 6 7/8" 1175 mml W. 6 7/8" 1175 mml W.
- j 6 718.. (175 rnml W.
Appia.... ~:__ 321b(15kgl 2S1b111k 111 !401b11ai.oi I rm:sur. ...n.. cydially. refer to your nearest Foxboro Sales Office. 'C.u 2A-1C1 E
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN I Page: 30
Subject:
Offsite and Control Room Operator Doses Doe to a Main Steam Line Break Appendix C: Example of Pre-Accident Iodine Spike STP Model //GO.FTOl NV= 4 MS= 2 //GO.FTll $ CLASS DESCRIPTION $ 1 NOBLE GASES $ 2 IODINE $ 3 TRITIUM NI= 23 NK= 7 NG= 0 NL= 3 lKRM 83 1 l.0352E-04 10.0 10.0 10.0 2KRM 85 1 4.2978E-05 10.0 10.0 10.0 3KR 85 1 2.0470E-09 29.8849E-06 10.0 10.0 4KR 87 1 1.5141E-04 10.0 10.0 10.0 SKR 88 1 6.8765E-05 10.0 10.0 10.0 6KR 89 1 3.6328E-03 10.0 10.0 10.0 7XEM 131 1 6.7414E-07 131.3039E-08 181.3039E-08 10.0 8XEM 133 1 3.5656E-06 152.0365E-07 202.0365E-07 10.0 9XE 133 1 1.5165E-06 83.5656E-06 159.0531E-06 209.0531E-06 lOXEM 135 1 7.3818E-04 174.8062E-06 224.8062E-06 10.0 llXE 135 1 2.1043E-05 107.3818E-04 172.4322E-05 222.4322E-05 12XE 138 1 8.1528E-04 10.0 10.0 10.0 13I 131 2 9.9536E-07 10.0 10.0 10.0 14I 132 3 8.4448E-05 10.0 10.0 10.0 15I 133 4 9.2568E-06 10.0 10.0 10.0 16I 134 5 2.1963E-04 10.0 10.0 10.0 17I 135 6 2.9129E-05 10.0 10.0 10.0 18I* 131 2 9.9536E-07 10.0 10.0 10.0 19I* 132 3 8.4448E-05 10.0 10.0 10.0 20I* 133 4 9.2568E-06 10.0 10.0 10.0 21I* 134 5 2.1963E-04 10.0 10.0 10.0 22I* 135 6 2.9129E-05 10.0 10.0 10.0 23H 3 7 1.7785E-09 10.0000E+OO 10.0000E+OO 10.0000E+OO //GO.FT21 1 'REACTOR COOLANT ANS/ANSI-18.1-1984 UCI/GM, WBNNAL3003 RS' 1 0.0 2 1.71E-1 3 2.66E-1 4 1.61E-1 5 3.00E-1 6 0.0 7 6.54E-1 8 7.17E-2 9 2.53EO 10 l.39E-1 11 9.04E-1 12 1.29E-1 13 4. 77E-2 14 2 .25E-1 15 1. 49E-1 16 3. 64E-1 17 2. 78E-1 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 1.24E2 0 2 'SECONDARY COOL ANS/ANSI-18.1-1984 UCI/GM, WBNNAL3003 RS' 1 0.0 2 3.63E-8 3 5.SlE-8 4 3.22E-8 5 6.31E-8 6 0.0 7 1.34E-7 8 1.54E-8 9 5.25E-7 10 2.90E-8 11 1.91E-7 12 2.68E-8 13 l.41E-6 14 3.37E-6 15 4.03E-6 16 2.93E-6 17 6.19E-6 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 1.24E-l 0 T WBN MSLB, 14 UCI/CC INITIAL CONC, lO+lGPM LK SS,1.0 GPM LEAK IN SG NJ= 4 KCONC= 1 1 'REACTOR COOLANT' 2 'STEAM GENERATOR W/LEAK' 3 'STEAM GENERATORS/NO LEAK' 4 'ENVIRONMENT' -1 INITIALIZATION V 1 2.622E8 GM V 2 5.310E7 GM V 3 1.593E8 GM V 4 1.0 GM S 1 1 0 4.166E4 S 1 1 1 7.620E3 S 2 3 0 6.610E6 This Page replaced by Rl6
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN IPage: 31
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break S 1 1 7 2.622E2 S 2 3 7 l.593E2 S 2 4 1 2.206E6 2 4 2 2.206E6 2 4 3 2.206E6 S 2 4 4 2.206E6 2 4 5 2.206E6 2 4 6 2.206E6 S 2 4 7 5.316El A 4 F 1 3 0 7.098E4 GM/HR F 3 4 0 l.003E6 GM/HR F 3 4 1 l.003E8 GM/HR F 3 4 7 l.003E8 GM/HR F 1 4 0 2.271E5 GM/HR 74.0 SEC TIME TO 74.0 SEC N 4 0 p 1 0 4 2.0 HR TIME TO 2.0 HR N 4 0 p 1 0 4 8.0 HR TIME TO 8.0 HR F 3 4 0 6.977E5 GM/HR F 3 4 1 6.977E7 GM/HR F 3 4 7 6.977E7 GM/HR p 1 0 4 T T /* II This Page replaced by RI 6
Calculation No. WBNAPS3077 I Rev: 016 I Plant: WBN IPage: 32
Subject:
Offsite and Control Room Operator Doses Doe to a Main Steam Line Break Appendix D: Example of Accident Initiated Iodine Spike (factor of 500 increase) STP Model //00.FTOl NV= 4 MS= 2 //GO.FTll $ CLASS DESCRIPTION $ 1 NOBLE GASES $ 2 IODINE $ 3 TRITIUM NI= 23 NK= 7 NG= 0 NL= 3 lKRM 83 1 1.03S2E-04 10.0 10.0 10.0 2KRM 8S 1 4.2978E-0S 10.0 10.0 10.0 3KR 8S 1 2.0470E-09 29.8849E-06 10.0 10.0 4KR 87 1 1.Sl41E-04 10.0 10.0 10.0 SKR 88 1 6.876SE-OS 10.0 10.0 10.0 6KR 89 1 3.6328E-03 10.0 10.0 10.0 7XEM 131 1 6.7414E-07 131.3039E-08 181.3039E-08 10.0 8XEM 133 1 3.S6S6E-06 1S2.036SE-07 202.036SE-07 10.0 9XE 133 1 1.Sl6SE-06 83.S6S6E-06 1S9.0S31E-06 209.0S31E-06 lOXEM 13S 1 7.3818E-04 174.8062E-06 224.8062E-06 10.0 llXE 13S 1 2.1043E-OS 107.3818E-04 172.4322E-OS 222.4322E-OS 12XE 138 1 8.1S28E-04 10.0 10.0 10.0 13I 131 2 9.9S36E-07 10.0 10.0 10.0 14I 132 3 8.4448E-OS 10.0 10.0 10.0 lSI 133 4 9.2S68E-06 10.0 10.0 10.0 16I 134 S 2.1963E-04 10.0 10.0 10.0 17! 13S 6 2.9129E-OS 10.0 10.0 10.0 18I* 131 2 9.9S36E-07 10.0 10.0 10.0 19!* 132 3 8.4448E-OS 10.0 10.0 10.0 20I* 133 4 9.2S68E-06 10.0 10.0 10.0 21I* 134 s 2.1963E-04 10.0 10.0 10.0 22I* 13S 6 2.9129E-OS 10.0 10.0 10.0 23H 3 7 1.778SE-09 10.0000E+OO 10.0000E+OO 10.0000E+OO //GO.FT21 1 'REACTOR COOLANT ANS/ANSI-18.1-1984 UGI/GM, WBNNAL3003 RS' 1 0.0 2 l.71E-l 3 2.66E-l 4 l.61E-l S 3.00E-1 6 0.0 7 6.S4E-1 8 7.17E-2 9 2.S3EO 10 l.39E-1 11 9.04E-1 12 1.29E-1 13 4.77E-2 14 2.2SE-l lS 1.49E-l 16 3.64E-1 17 2.78E-l 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E2 0 2 'SECONDARY COOL ANS/ANSI-18.1-1984 UCI/GM, WBNNAL3003 RS' 1 0.0 2 3.63E-8 3 S.SlE-8 4 3.22E-8 S 6.31E-8 6 0.0 7 l.34E-7 8 l.S4E-8 9 S.25E-7 10 2.90E-8 11 l.91E-7 12 2.68E-8 13 l.41E-6 14 3.37E-6 lS 4.03E-6 16 2.93E-6 17 6.19E-6 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E-1 0 T WBN MSLB,.26S UCI/CC INIT CONC,lO+lGPM LK,SOO IODINE SPIKE,1.0 GPM SG LK NJ= 4 KCONC= 1 1 'REACTOR COOLANT' 2 'STEAM GENERATOR W/LEAK' 3 'STEAM GENERATORS/NO LEAK' 4 'ENVIRONMENT' -1 INITIALIZATION V 1 2.622E8 GM V 2 S.310E7 GM V 3 1.S93E8 GM V 4 1.0 GM S 1 1 0 7.88SE2 S 1 1 1 7.620E3 S 2 3 0 6.610E6 This Page replaced by Rl6
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN I Page: 33
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break s 1 1 7 2.622E2 s 2 3 7 1.593E2 s 2 4 1 2.206E6 2 4 2 2.206E6 2 4 3 2.206E6 s 2 4 4 2.206E6 2 4 5 2.206E6 2 4 6 2.206E6 s 2 4 7 5.316El c 1 1 2 4.765E4 c 1 1 3 1. 661ES c 1 1 4 5.931E4 c 1 1 5 3.580E5 c 1 1 6 8.758E4 A 4 F 1 3 0 7.098E4 GM/HR F 3 4 0 1.003E6 GM/HR F 3 4 1 l.003E8 GM/HR F 3 4 7 1. 003E8 GM/HR F 1 4 0 2.271E5 GM/HR 74.0 SEC TIME TO 74.0 SEC N 4 0 p 1 0 4 2.0 HR TIME TO 2.0 HR N 4 0 p 1 0 4 8.0 HR TIME TO 8.0 HR F 3 4 0 6.977E5 GM/HR F 3 4 1 6.977E7 GM/HR F 3 4 7 6.977E7 GM/HR p 1 0 4 T T /* II This Page replaced by Rl 6
Calculation No. WBNAPS3077 I Rev: 016 IPlant: WBN IPage: 34
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Appendix E: Example of COROD Input (ARCON96 X/Q) NIT= 23 NR= 1 ITP= 6 FACT= 1.0 COROD-WBN MSLB s KRM 83 KRM 85 KR 85 KR 87 KR 88 KR 89 XEM 131 XEM 133 XE 133 XEM 135 XE 135 XE 138 I 131 I 132 I 133 I 134 I 135 I* 131 I* 132 I* 133 I* 134 I* 135 H 3 4 'ENVIRONMENT I $ TN= 0.2056E-01 1 O.OOOE+OO 2 1. 064E-01 3 1.624E-01 4 9.552E-02 5 1.852E-01 6 O.OOOE+OO 7 3.959E-01 8 4.503E-02 9 1.547E+OO 10 8.BOlE-02 11 5.609E-01 12 7.835E-02 13 3.147E+OO 14 7.603E+OO 15 9.004E+OO 16 6.734E+OO 17 l.387E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 7.426E+OO 4 'ENVIRONMENT I $ TN= 0.2000E+Ol 1 O.OOOE+OO 2 2.316E+OO 3 4.200E+OO 4 1.520E+OO 5 3.714E+OO 6 O.OOOE+OO 7 1. 029E+Ol 8 l.130E+OO 9 3.992E+Ol 10 8.599E+OO 11 1.704E+Ol 12 3.170E-01 13 3.512E+OO 14 l.221E+Ol 15 1.062E+Ol 16 1.305E+Ol 17 1. 836E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 0.000E+OO 22 O.OOOE+OO 23 7.715E+Ol 4 'ENVIRONMENT ' $ TN= 0.8000E+Ol 1 O.OOOE+OO 2 4.053E+OO 3 1.327E+Ol 4 7.767E-01 5 4.702E+OO 6 O.OOOE+OO 7 3.222E+Ol 8 3.440E+OO 9 1.245E+02 10 1.881E+Ol 11 5.881E+Ol 12 9.790E-04 13 l.039E+Ol 14 1.235E+Ol 15 2.796E+Ol 16 3 .411E+OO 17 3.689E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 2.165E+02
-6 'ENVIRONMENT CURIES I $ TN= 0.2400E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO -6 'ENVIRONMENT CURIES ' $ TN= 0.9600E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO -6 I ENVIRONMENT CURIES ' $TN= 0.7200E+03 1 O.OOOE+OO 2 0.000E+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 0.000E+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO 3.85E-03 3.85E-03 3.22-03 2.36E-04 l.88E-04 l.55E-04 74 7126.0 21600.0 57600.0 259200.0 2246400.0 3200.0 51.0 711.0 51.0 711.0 51.0 711.0 51.0 711.0 51.0 711.0 51.0 0.0 0.0 o.o 0.0 0.0 0.0 3200.0 0.95 0.70 0.95 0.70 0.99 0.0 2889.0 0.95 0.70 0.95 0.70 0.99 0.0 2889.0 0.95 0.70 0.95 0.70 0.99 0.0 2889.0 0.95 0.70 0.95 0.70 0.99 0.0 2889.0 0.95 0.70 0.95 0.70 0.99 0.0 2889.0 100.0 60.0 40.0 1440.0 5760.0 257198.0 1.2492 0.63 0.8352 322.0 45.0 17.75 46.0 9.0 4.0 161.0 22.5 4.0 0.0 ROOFFLUX DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE THROUGH ROOF 1000.0 1000.0 1000.0 20.0 20.0 20.0 500.0 500.0 -16.0 2.25 This Page replaced by RI 6
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN I Page: 35
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM AUX BUILDING 270.0 150.0 148.0 27.0 15.0 14.0 135.0 75.0 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM TURBINE BLDG 322.0 112.0 341.0 32.0 11.0 34.0 161.0 56.0 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM SPREADING ROOM 322.0 45.0 26.0 32.0 9.0 5.0 22.5 161.0 -4.67 0.67 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM CR BLDG END 18.0 45.0 460.0 10.0 10.0 100.0 4.0 22.5 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM CR BLDG END 18.0 45.0 460.0 10.0 10.0 100.0 4.0 22.5 -25.5 3.0 I* II This Page replaced by R16
Calculation No. WBNAPS3077 IRev: 016 IPlant: WBN IPage: 36
Subject:
Offsite and Control Room Operator Doses Due to a Main Steam Line Break Appendix F: Example ofFENCDOSE Model 1 KRM-83 KRM-85 KR-85 KR-87 KR-88 KR-89 XEM-131 XEM-133 XE-133 XEM-135 XE-135 XE-138 I-131 I-132 I-133 I-134 I-135 I*-131 I*-132 I*-133 I*-134 I*-135 H-3 T l.784E-4 8.835E-5 6.217E-5 2.900E-5 9.811E-6 6.382E-4 WBN MSLB, 14 UCI/CC INITIAL CONC, lO+lGPM LK SS,1.0 GPM LEAK IN SG TIME TO 74.0 SEC 4 'ENVIRONMENT I $ TN= 0.2056E-01 1 O.OOOE+OO 2 l.064E-Ol 3 l.624E-01 4 9.552E-02 5 1. 852E-01 6 O.OOOE+OO 7 3.959E-01 8 4.503E-02 9 1.547E+OO 10 8.801E-02 11 5.609E-01 12 7.835E-02 13 3.147E+OO 14 7.603E+OO 15 9 .. 004E+OO 16 6.734E+OO 17 1.387E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 0.000E+OO 23 7.426E+OO WBN MSLB, 14 UCI/CC INITIAL CONC, lO+lGPM LK SS,1.0 GPM LEAK IN SG TIME TO 2.0 HR 4 'ENVIRONMENT I $ TN= 0.2000E+Ol 1 O.OOOE+OO 2 2.316E+OO 3 4.200E+OO 4 1.520E+OO 5 3.714E+OO 6 O.OOOE+OO 7 1.029E+Ol 8 1.130E+OO 9 3.992E+Ol 10 8.599E+OO 11 1.704E+Ol 12 3.170E-01 13 3.512E+OO 14 1.221E+Ol 15 1. 062E+Ol 16 l.305E+Ol 17 1.836E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 7.715E+Ol WBN MSLB, 14 UCI/CC INITIAL CONC, lO+lGPM LK SS,1.0 GPM LEAK IN SG TIME TO 8.0 HR 4 'ENVIRONMENT I $ TN= 0.8000E+Ol 1 O.OOOE+OO 2 4.053E+OO 3 1.327E+Ol 4 7.767E-01 5 4.702E+OO 6 O.OOOE+OO 7 3.222E+Ol 8 3.440E+OO 9 1.245E+02 10 1. 881E+Ol 11 5.881E+Ol 12 9.790E-04 13 1.039E+Ol 14 1.235E+Ol 15 2.796E+Ol 16 3.411E+OO 17 3.689E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 2.165E+02 WBN MSLB TIME TO 1 DAY
-6 'ENVIRONMENT CURIES I $ TN= 0.2400E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 0.000E+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO WBN MSLB TIME TO 4 DAYS -6 'ENVIRONMENT CURIES I $ TN= 0.9600E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 0.000E+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 0.000E+OO 13 0.000E+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO WBN MSLB TIME TO 30 DAYS -6 'ENVIRONMENT CURIES ' $TN= 0.7200E+03 1 O.OOOE+OO 2 O.OOOE+OO 3 0.000E+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 0.000E+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 0.000E+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO This Page replaced byRl6
~~~~~~~~~
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 37
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Appendix G: Additional Cases This appendix documents the original steam generator MSLB results used as input to the FSAR (0.265 µCi/gm 1-131 equivalent or 21 µCi/gm 1-131 equivalent, with a 10 gpm known primary to secondary side leakage and 1 gpm unknown leakage and 150 gpd per steam generator with the 3 gpm post accident leakage). The control room operator dose COROD model recirculation and time increments were corrected. All other cases (with different concentrations and/or leak rates) can be found in revision 8 of this calculation (these were not corrected, as they were not used in the FSAR). Note that all other cases are historical and utilize original steam generator data. Details of these results (excluding the COROD corrections) may be found in revision 8. ALSO: Note, these values were not updated to the 1991-2010 meteorological data set X/Q values Unit 1 MSLB Control Room Doses (rem) Using ARCON96 X/Q values with ORIGINAL STEAM GENERATOR DATA: 10 gpm known + 1 gpm unknown + 3 gpm post accident leakage Pre-accident Iodine Accident Initiated Spiking Iodine Spiking limit Gamma l.340E-02 2.475E-02 5 Beta l.339E-OI 2.052E-Ol 30 Thyroid (ICRP-2) 2.851E+OI 4.730E+OI 30 Thyroid (ICRP-30) l.570E+OI 2.426E+Ol 30 TEDE 5.856E-01 9.425E-OI 5 Unit l MSLB Offsite Doses (rem) with ORIGINAL STEAM GENERATOR DAT A: 10 gpm known + 1 gpm unknown + 3 gpm post accident leakage Pre-accident Iodine spiking case Accident Initiated Iodine Spiking (500) case I-131 equivalent: 21 uCi/cc I-131 equivalent: 0.265 uCi/cc 2-hrEAB 30-dayLPZ limit 2-hrEAB 30-davLPZ limit gamma 6.381E-02 2.233E-02 25 gamma 2.006E-OI l.981E-OI 2.5 beta 2.058E-02 8.335E-03 300 beta 4.981E-02 4.809E-02 30 Inhalation (ICRP-2) l.046E+OI 4.695E+OO 300 Inhalation (ICRP-2) l.296E+OI l.709E+OI 30 Inhalation (ICRP-30) 5.282E+OO 2.469E+OO 300 Inhalation (ICRP-30) 5.251E+OO 7.394E+OO 30 TEDE 3.957E-OI l.698E-OI 25 TEDE 6.212E-OI 7.551E-Ol 2.5
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 38
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Appendix H: Unit 2 MSLB Unit 2 utilizes the original steam generators, however the mass releases were revised by Westinghouse (ref. 26). The main text models apply except for the following changes: Volumes:
#1: Reactor Coolant: 5.4E5 lb (ref. 2) = 2.4494E8 gm
#2: Steam Generator w/Leak: 96,100 lbm = 4.359E7 gm (ref. 26)
#3: Steam Generators w/out Leak: 1.421E8 gm (ref. 2).
Step Sources: The step sources (Ci-gm/µCi) to initialize the reactor coolant and the secondary side activities are: All cases: S=2.4494E8 gm x IE-6 Ci/µCi x 29.06=7.l18E3 (noble gases) S=2.4494E8 gm x IE-6 Ci/µCi = 2.449E2 (tritium) Pre-accident iodine spike case (initial concentration= 14 µCi/gm): S=2.4494E8 gm x IE-6 Ci/µCi x l l.348[µCi/gm I-13U 1 x 14 µCi/gm I-131 = 3.891E4 (iodines) Accident initiated iodine spike case (initial concentration= 0.265 µCi/gm): S=2.4494E8 gm x IE-6 Ci/µCi x 11.348 [µCi/gm I-13U 1 x0.265 µCi/gm I-131= 7.366E2 Secondary side, all cases, steam generator w/ leak, release to environment. (concentration= O.lµCi/gm) which is due to dryout (from reference 26, the initial mass release from the defective steam generator for Unit 2 is 96, 100 lb): S = 96,100 lb x 453.59 gm/lb x IE-6 Ci/µCi x 4.150E5 [µCi/gmI-13U 1 x O.lµCi/gm I-131 =l.809E6 S = 96,100 lb x 453.59 gm/lb x IE-6 Ci/µCi = 4.359El (tritium) Secondary side, all cases, steam generators without leak (initial concentration= 0.1 µCi/gm): S = l .421E8 gm IE-6 Ci/µCi x 4.150E5 [µCi/gm I-13U 1 x O.lµCi/gm I-131 = 5.897E6 S = 1.421E8 gm x IE-6 Ci/µCi = l .421E2 (tritium) Continuous Sources: For the accident initiated iodine spike case, the iodine spike is 500 times the iodine release rate from the fuel. At steady state, the iodine release (production) rate is equal to the removal rate. The iodine removal is due to a) radioactive decay, b) removal by the letdown system, and c) removal through reactor coolant leakage. These terms are expressed as: P = I:removal rates= decay+ letdown +leakage or P =A. + fLeN + p,N where P =production rate [k 1] A.= decay constant for the isotope in question [k 1] = ln(2)/T 112 fL =letdown flow rate= 120 gpm + 4.39 gpm = 124.39 gpm e =letdown demineralizer efficiency= 1 (assumed so as to maximize removal/production rate) V =volume of primary coolant= 5.4E5 lb p, = removal rate of iodine from the primary side due to preaccident primary side leakage
= 11 gpm ( 10 gpm identified + I gpm unidentified)
T 112 = halflife taken from ref. 15 Note: see page 15 Note. pro duct1on Rates f or a R eactor coolant Leak of 11 gpm (10 gpm identified+ I gpm uni"dentI"fi1ed) Half Life 'A [1/hr] fLEIV [l/hr] p,/V [1/hr] Prod rate P [ l/hr] 500xP I-131 8.04 d 3.59E-03 1.15E-01 1.02E-02 0.1291 64.56 I-132 2.28 h 3.04E-01 1.15E-01 1.02E-02 0.4295 214.77 1-133 20.9h 3.32E-02 1.15E-Ol 1.02E-02 0.1587 79.35 I-134 52.6m 7.91E-Ol 1.15E-01 1.02E-02 0.9162 458.10 I-135 6.61 h 1.05E-01 1.15E-01 1.02E-02 0.2304 115.20 The accident initiated iodine spike of 500 times the increase in the iodine release (production) rate from the fuel is modeled as a continuous source:
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 39
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 C =Volume x IE-6 Ci/µCi x Prod Rate x 500 x 1 µCi/gm I-131 equivalent conversion factor x I-131 equiv. where Volume= 2.4494E8 gm Prod Rate = see table above I µCi/gm I-131 equivalent conversion factor= 11.348 (value determined in Table I a , this is to get the ANSI/ANS-18.1-1984 source into I µCi/gm I-131 equivalent I-131equiv.=0.265 µCi/gm I-131 Continuous Source [gm-Ci/µCi-hr] for Accident Initiated Iodine Spike: Reactor Co o Ian t Leak 0 f 11 gpm (IO gpm 1"dentJ"fi1ed + I gpm unidentified) Nuclide 0.265 µCi/gm I-131 I-131 4.756E+04 I-132 1.582E+05 I-133 5.845E+04 I-134 3.374E+05 I-135 8.485E+04 Flow Rates: The following flow rates/leakage rates for each component are (determined by trial and error with the ultimate goal being to find the flow/leak which would lead to the offsite/control room doses reaching the regulatory limits): Flow from Reactor Coolant # 1 to Environment #4 all classes which consists of I gpm and is for leak in the steam generators, the production rate of iodines is based on a total RCS leakage of 11 gpm (=lOgpm identified+ lgpm unidentified): F = 1.0 gpm x 60 min/hr x 3785.48 cc/gal x lgm/cc = 2.271E5 gm/hr Flow from Reactor Coolant# I to Steam Generator w/ no Leak #3 all classes: F = 3 steam generators x 150 gpd x 3785.48 cc/gal I 24 hr/day x lgm/cc = 7.098E4 gm/hr The initial steam released from the defective steam generator is 96,100 lb (entire mass of SG rounded up). From reference 26, the non-defective steam generators(= "steam generators without leak" in this model) the mass release is 433,079 lb (0-2 hr), and 870,754 lb (2-8 hr). The accident releases end at eight hours. To take into account uncovery of the faulted steam generator, there is no iodine partitioning in the release to the environment (iodine partition coefficient= 1). The mass release representing I gpm primary leak is a flow directly to the environment. This is reflected in the flows listed above. For other leak rates, the flow cards will correctly take into account the mass released. The reactor coolant release to the unfaulted steam generator is small relative to the secondary side mass, therefore partitioning is allowed per the SRP. The iodine partition coefficient due to steaming for the unfaulted steam generators to the environment is I 00. These mass releases translate into the following flows: Flow from Steam Generators w/out Leak #3 to Environment #4: F = (433,079 lb)(453.59 gm/lb)/(2 hr)= 9.822E7 gm/hr (0-2 hr, noble gases, tritium) F = (433,079 lb)(453.59 gm/lb)/(100x2 hr)= 9.822E5 gm/hr (0-2 hr, iodines) F = (870,754 lb)(453.59 gm/lb)/(6 hr)= 6.583E7 gm/hr (2-8 hr) (noble gases, tritium) F = (870,754 lb)(453.59 gm/lb)/(100x6 hr)= 6.583E5 gm/hr (2-8 hr) (iodines) The STP output is used as input to COROD (which determines control room operator dose) and FENCDOSE (which determines 30-day and 2-hour LPZ offsite dose). The STP input for Unit 2 is similar to Unit 1 shown in Appendices C and D with the modifications described above. Unit 2 Control Room XJQ (ARCON96 method): 2.59E-03 0-2 hr, 2.12E-03 2-8 hr
Calculation No. WBNAPS3077 I Rev: 016 IPlant: WBN I Page: 40
Subject:
Offsite and Control Room Operator Doses Doe to a Main Steam Line Break Results: The results were (rem): Unit 2 MSLB Offsite Doses frem): Pre-accident Iodine soiking case Accident Initiated Iodine Spiking {500) case 1-131 eauivalent: 14 uCi/l!Ill 1-131 eauivalent: 0.265 uCi/21n 2-hrEAB 30-dayLPZ limit 2-hrEAB 30-dayLPZ limit gamma 2.74E-02 1.llE-02 25 gamma 1.04E-Ol l.25E-Ol 2.5 beta 8.81E-03 4.21E-03 300 beta 2.54E-02 3.02E-02 30 Inhalation (ICRP-30) 2.41E+oo l.21E+oo 300 Inhalation (ICRP-30) 3.09E+oo 4.78E+oo 30 TEDE l.78E-Ol 8.34E-02 25 TEDE 3.42E-Ol 4.82E-Ol 2.5 Unit 2 MSLB Contro es es* Pre-accident Iodine Accident Initiated Spiking Iodine Spiking limit Gamma 4.35E-03 8.02E-03 5 Beta 4.00E-02 6.48E-02 30 Thyroid (ICRP-30) 7.44E+o0 l.03E+ol 30 TEDE 2.65E-Ol 3.84E-Ol 5 The following Unit 2 margins were calculated from the doses in the Tables above. Where: margin= limit- dose, and percent= 100 x (limit-dose)/limit Unit 2 MSLB Offsite Dose .... : Pre-accident Iodine SDikin2 case Accident Initiated Iodine Snikine: (500) case 1-131 1-131 eauivalent: 14 µ.Ci/e:m eauivalent: 0.265 µ.Ci/mn 2-hrEAB 30-dayLPZ limit 2-hrEAB 30-davLPZ limit manrin nercent marein nercent marl!in nercent manrin nercait 2arnma 24.97 99.89 24.99 99.96 25 2anuna 2.40 95.85 2.38 95.00 2.5 beta 299.99 100.00 300.00 100.00 300 beta 29.97 99.90 29.97 99.90 30 Inhalation Inhalation (ICRP-30) 297.59 99.20 298.79 99.60 300 (ICRP-30) 26.91 89.71 25.22 84.08 30 TEDE 24.82 99.28 24.92 99.68 25 TEDE 2.16 86.40 2.02 80.80 2.5 Unit 2 MSLB Control Room Dose ~ Uain
- A'R"'ON96 X/Q values:
Pre-accident Iodine Accident Initiated Soikin2 Iodine Spiking limit manrin nercent manrin nercent Gamma 5.00 99.91 4.99 99.84 5 Beta 29.96 99.87 29.94 99.78 30 Thvroid {ICRP-30) '12.57 75.'12 19.68 65.60 30 TEDE 4.74 94.71 4.62 92.32 5 Discussion: The Unit 2 MSIB doses are less than the 10CFRlOO and 10CFR50 GDC 19 limits. Most of the Unit 1 doses bound the Unit 2 doses, except for the 30 day LPZ offsite doses for the accident initiated iodine spike. This Page replaced by RI 6
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 41
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 : Surveillance tes! l -SI-68-28 performed on 7 /10/00 JUL-24-2000 13:52 T~ PLANT MGRS OFC 423 365 1904 P.02/07 SURVEILLANCE TASk SHI!! CSPP*B.2> PAGE _1_ ol llOAk DROI~; DDD679~DO SI J(IYI P0531 PROCEDURE#1 1*81*68*28 TITLE: PRIMARY RADIOCllENISTRY REQUIREHlNlS PHF SECl: Cl!M _ _ _ _ _ _NIA _ __ TUT REABDIH PElllDDIC PERFlllMANCf DUEi 07/10/00 --AUT-"0-R-IZA~Tl~A._TO-BE_G_IN-:-SR_Q_ DATE TlllE WIN EXT1 OT/11100 KAX EXT: fllll! SPP*8.2*2 FREQr W IQ: N ASME XI: N APP Mme' U! PEAf MDII£: 1;U4 SUISQNT R\1119 s INSTRUCTIONS: Do NOT ctart prior io sc~e&Jled .U. dato
====:;;;;;:;;:;::r.**-**n*emt===!!!l!!D:==~;;:=====-=z£i:rBBma**a:i.*== ::r.:-..ci;illiiD*-&m.m;;;i;;;isa::r.:=:=:===1;1~;;i=;;wwWii;ii;1iU*i;:=======--====i;=====
YEST PErnlllMERS WAS THIS A COMPLETE OR PARTIAL 4$i M_ot144Wf / NAME SlG~TURE lNIT SECT PERFDRMNCE? IE~AIH "PARTIAL" IN REMARlSl PARTIALt L_,1d4. MQit,*W..t tM,( WIRE ALL tECH SPEC/TECH REC/Ol>C*1'1RE / PROT UQ ACCEPTANCE CRIURIA 3ATISFIED7 YES*_ NO;_ NIA:_ WERE ALL OTHER ACCEPtANCE CR!lERIA SAT ISF!lll7 YES1_ NO:_ U/At:::;_ AL!RT SCHIOULING REQUIRED? YES:_ NO:_ N/A:_X_
~:QA~R~E:~ *=~~:~~~:D~E~~:'~~~r:~OT /
D0°"/111 ACTION REQll7 (E~PLAIN IN REMAHBl YES1_ NO;_ NIA:_ TEST DlliECTIJll I ~at:t!R~ ~{\~(4.v NIA _ _H/A_ _ V"'"'"~ulW: s~o DATE I TIME 7/tJtnl Mt il'"!l;iiifiiliiliilllClilae1UUSn11aoi;==.i=====~----*ii----*c*1:z:========== ===~~==u*;;&&**lll*si111111111a:================;;r;;;i;;;:;':ll;;c**********- REMARKS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ COPY OF ns SENT TO SCHEDULI NO: SECl I DNtkNtDUR HRS S&CTIONJllllfEN/DllR HRS SECT IONniiiNIDUR HRS SECl IONJiiiiUIOUR Nd RECOllPS TRANSIUTTAL#:. _ _ _ _ _ __ 111111 H~ DI~ 111111111mn100 I~ IllH
Calculation No. WBNAPS3077 I Rev: 014 1Plant: WBN I Page: 42
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 JUL-24-2000 13:52 TVA PLANT MGRS OFC 423 365 1904 P.03,07 Lv~JuM-iooo 09:23:05.19 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE 01 - RCS - GASEOUS ACTIVTY v' FILE IDENT DKl3600:[TVA.SAMPLE.CHEM.NEW]W0007108796_C401.C~F1l SAMPLE ID W0007108796_C401 .,,,,...,
- OPERATOR LLMANNONE SAMPLE TIME 10-JUL-2000 08:53
- SAMPLE GEOMETRY GMlK
- SHELF HEIGHT 0
- EFFICIENCY FILE GMlKO _..
SAMPLE rYPE ; 1240 CC GAS MARI
- SAMPLE QOANTl'l'Y : 1.00000E+OO CC ACQ DATE & TIME 10-JUL-2000 09:12
- DEADTIME (II!.) : 0.3\ r PRESET LIVE TIME : 0 00:10:00
- SENSITIVITY : 4.00000 ELASPED kEAL TIME : 0 00:10:01
- GAUSSIAN SEN ; 10.00000 ELAPSED LIVE TIME 1 0 OO:lOrOO
- NllR ITERATIONS : 10
/
DETECTO~ DET #3, GSS-3286
- LIBRARY NOBLEGAS EFFIC CAL DATE 29-JUL-1994 13:47
- EFFIC CERT DATE 29-JUL-1994 13:47 DCAL DATE & TIME g-JUL-2000 15:52:
- ENIERGY TOLER 1.25 KEV/CHAN 4,gg928E-Ol
- HALF LIFE RATIO 8.00000 OFFSET -1.48334E-Ol keV
- ABUNDANCE LIMIT 1 80.0%
Q COEFFIC!ENT 3.22120E-OB
- CORRECTION FACTO~ l.OOOOOE+OO PEAk START CHAN : 140
- PEAK END CHAN : 4096 ANALYSES PEAK Vl6.9 NID V3.3 MINACT V2.B WTMEAN/KEY Vl.8
&L 1
COUNTED ON : LION COLLECTED BY : COUNTED :BY : LLMANNONE REVIEWED BY ? _ _\_,.,,,__ _.,...~4al~~.::i......,,c:::._:,.,,.,_...___ _ COMMENTS : <:::=:> Poet-NID Peak Search Report It Energy Area Bkgnd FWHM Channel Left Pw \E:u Fit Nu elides 0 Bl.12 230 92 1.01 162.55 157 12 10.e 0 151. 37 88 53 0.97 303.07 299 8 17.7 XE~133 0 196.38 ICR-85M 70 80 0.88 393.10 388 10 26.7 0 227.79 22 47 1. 35 455.92 448 10 62.9 0 249.81 549 83 0.94 499.98 494 12 S.4 XE-135 0 258.71 72 40 0.89 517,78 514 10 20.2 0 305.21 XE-138 19 34 0.76 610.77 607 8 57.2 KR-95111 0 402.80 48 34 2.39 805.96 801 12 29. 3 ltll-87 0 435.J.9 2'6 16 l.31 870.75 866 11 35. 3 Xl!:-138 0 511. 06 390 49 2.28 1022.51 1014 l8 6.6 0 526.45 39 24 1 ..u 1053.28 1048 12 29,8 XE-13SM 0 898.31 22 16 1.29 1796.96 1791 9 38.8 0 1293.58 904 9 1.60 2587.40 2578 16 3.4 AR-41
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 43
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 JUL-24-2000 13:53 TVA PLANT MGRS OFC 423 355 190'! -- - i:.: 0~1'07
. REPORT DATE lO-JUL-2000 O~:Z3 REQUESTOR ' LI.MANNONE TENNESSEE VALLEY AUTHORITY WATTS Bllll NUCLEAR PLANT POST NID OA ANALYSIS TITLE : Ul - RCS - GASEOUS ACTIVTY SAMP1E No. W0007l06796 C401 OPERATOR NAME LLMANNONE SAMPLE TYPE 1240 CC GAS-MUI SAJllPLE GEOMETRY GMlK COUNT TIME lO-JUL-2000 09:12:51 SAMPLE QUANTITY 1. OOOOOE+OO SAMPLE TIME lO-JUL-2000 08:53:00 DETECTOR D:E:T f13, GSS-3286 LIBRARY NOBLEGAS PEAK ENERGY DECAY CORR ISOTOPE AR-41 ENERGY DIFF (KEV) 1293.64 ----------
-0,06 uci/CC' COMMENTS
----~--------------------
l.303E-02 OA Results OK JCR-85M 151.18 0.19 l.SllSE-04 OA Results OK KR-67 402.58 0.22 4.5751!:-04 QA Re$1,1.lts ox XE-133 81. 00 0.12 9.565E-04 QA Results OK XE-135 249. 79 0.02 l. 4.2PE-OJ OA ResultrJ OR XE-l35M 526.56 -0.11 7, 364E-04 OA Results ox XE-138 258.31 0,40 1. 796!-03 QA Results OK AVG ENERGY DIFF 0.11 l.BSSIE-02
~ = TOTAL GAMMA ACTIVITY O.OOOE+OO = Total I>GL Activity l.SSSIE-02 Total Gas Activity ENERGY NET ....._
AREA ____ FWHM GAMMA/SEC GAMMA/SEC /CC 111 ER: OR !'LAG POTENTIAL IP ACTIVITY 196.38 70, 0.88 4. 728E+OO 4.726E+OO R
--------- ~--------
KR-8~'!5,438£-04 227.79 22. 1.35 1.675E+OO 1. 675!+00 u TE-132 5.163E-05 u CS-138 5 .llOl!:-03 511,06 u NP-23,2_...- 4.252E-04 390. 2.28 7.016E+Ol 7.0l6:E:+Ol u ANN IL O.OOOE+OO 898.31 22. l.29 7.079E+OO 7.0792+00 u RB-BB- l.512E~03 u Y-88 2.049E-04
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 44
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 JUL-24-2000 13:53 TUA PLANT ~S OFC 423 365 1904 P.05/07 10-JUL-2000 09:37:45,71 TENNESSEE VALLEY AUTHORITY WATTS BAlt NUCLEAR PLANT SAMPL.:!!: TITLE Ul - RCS - DEGASSED LIQUID ACTIVITY FILE IDENT DKB600:[TVA.SA1'IPLE.CHEM.NEWJW0007108795_C402.CNF1l SAMPLE ID W0007108795 C402
- OfERATOR LLMANNONE SAli!PLE TIME 10-J'UL-2000-07: 25 _..-*
- SAMPLE GEOMETRY LSV20
- SHELF HEIGHT 1
- EFFICIENCY FILE LSV201 .--
SAMPLE TYPE : RCS 20ML LSV
- SAMPLE QUANTITY : 5.000 0 GRAMS ACO DA'rE & TIME : 10-JUL-2000 08:36
- DEADTIME (%) : 2.2\
PRESET LIVE TIME : 0 01:00:00
- SENSITIVITY 4.00000 ELASfED REAL TIME : 0 01:01:20
- GAUSSIAN SEN : 10.00000 ELAPSED LIVE TIME : 0 Ol:OO:OO
- NBR ITERATIO~ ; 10
*****************************************************r***********************
Dl!:TECl'OR DE'r #4, GSS-3310
- LIBR.AR.Y RCSLIQUUJ*
EFFlC CAL DATE S-AUG-1994 11:11: * !!TIC CERT DATE 5-AOG-1994 11:11: DCAL DAT£ & TIME 1 9-JUL-2000 15:52:
- BNERGY TOLER 1.25 KEV/CHAN 5,00474E-Ol
- HALF LIFE RATIO 8.00000 OFFSET -3.73924E-Ol keV
- ABUNDANCE LIMIT 80,0\
Q COEFFICIENT -l.l4092E-07
- CORAECTION FACTOR 1 l.OOOOOE+OO PEAK START CHAN 140
- fEAIC END CHAN : 4096 k 5L ANALYSES : PEAK V16.9 NID V3.3 MINACT V2.8 W'l'MEAN/l<EY Vl.8 CO"ON'rED ON ; LION COLLECTED BY :
COUNTED B~ : LL.MANNON!! REVIEWED BY ~~->.....,..~--=~'.,c::;ll~:;;;:i....._""'_~..;;..;;;;;;;;;;:._~ COMMENTS <:::::::::=-- Post-NID Peak search Repoi:t It Energy Area Bkqnd FWHM. Chemnei Left PW \Err Fit Nuclides 0 135.60 697 25268 0.92 271. 70 269 7 38.0 I-134 0 249.64 887 20225 0.79 499.61 498 6 25.6 0 287.87 XE-135 540 22521 0.98 576.01 574 7 46.3 I-135 0 364.21 I-135 455 13323 1. 06 728.61 726 8 44.3 I-131 0 405.43 310 6932 1.05 810.99 808 8 46.9 I-134 0 417. 67 485 6009 1,20 835,46 832 8 28 .1 0 462.73 1-135 545 5368 1.28 925.53 922 8 23.7 CS-138 0 511. 00 838084 22601 2.65 1022. 01 1014 19 0.1 F-18 0 522.65 824 999 1.38 1045.31 1043 7 7.3 l 526.58 1048 I-132 823 1.29 1053.15 1050 18 5.2 1,06E+OO I-135 529.88 4510 XE-135M l 1009 1.34 1059.77 1050 18 1. 9 I-133
Calculation No. WBNAPS3077 j Rev: 014 I Plant: WBN I Page: 45
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 JUL-24-2000 13:53 TVA PLANT MGRS DFC 423 365 1904 P.06/07
.K~V.l:CI.' IJA"l'.t: I .LU-JUL-ZOOO O!l: 37 REOUESTOR LLMANNONE TENNESSEE VALL!!:Y AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - DEGASSED LIQUID ACTIV!'l'Y SAMPLE No. W0007108795 C402 OPERATOR. NAME LLMANNONE SAMPt.E TYPE I RCS 20ML LSV SAMPLE GEOMETRY LSV20 COUN'l' TIME 10-JUL-2000 08:36:15 SAPIPLE QUANTITY s.oooOOE+OO SAMPLE TIME 10-JUL-2000 07:25:00 DETECTOR DET i4, GSS-3310 LitlRARY RCSLil)UID PEAR ENERGY DECAY CORR ISOTOl'E ENERGY DIFF (KEV) uCi/GRAM COMMENTS
-------- 511, 00 ---------- ----------
F-18 o.oo 1.1791E-Ol OA Results OK NA-24 l:l6B.53 0.07 9.16!i1E-0.( QA Results OX MN-56 1810.69 -0.63 9. :n:u.:-os OA llesults OK co-sa 810.76 -0.01 5.0l9E-0.( OA Results OK NB-95 765.79 0.89 3. l32E-05 OA Results OK I-131 364.48 -0.27 6.0?0E-05 OA Results OX I-132 667.69 0.03 l.459E-03 OA Results OK I-133 529.87 0.01 8.208E-04 OA Results OIC I-134 847.03 -0.05 2.694E-03 OA 11.esults OK I-135 1260.41 0.03 1.608:1!:-03 QA Results OK XE-135 249.79 -0.16 B.914E-05 QA lesul ts OK X:&:-135M 526.56 0.02 l.406E-02 OA Results OK CS-138 1435.86 -0.11 2,395E-03 QA Results OK AVG ENERGY DIFF = -0.01 1.426E-Ol
- TOTAL GAMMA ACTIVITY l.218E-01 = Total DGL Activity 2.427E-03 = Total FP Activity 1.S12E-03 ~otal AP Activity l.41SE-02 Total Gas Activity 6.643E-03 ~ Total HFP Activity 3.'i~~-J Dose Equivalent Iodine-131 ~ 2.905E-04 Iodine 131/133 Ratio 7.39SE-02 D&~~~
Iodine 133/135 Ratio
- 5.lOSE-01 287.87 xev Peak was used in identifying 2 isotopes 526.58 Kev Pe11.k was used in identifying 2 isotopes 546.88 :keV Peak was used in identifying 2 isotopes 766. 69 KeV Peak was used in identifying 2 isotopes 810,75 l\eV Peak was used in identifyinq 2 isotopes 846. 97 Kev Peak was used in identifying 2 isotopes 857.15 KeV Peak was used in identifying 2 isotopes 1136.29 KeV Peak was used in iilentifyinq 2 hot opes
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 46
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 Surveillance test l-SI-68-28 performed on 4/9/01 SURVEILLANCE TASK SHEET (SPP-8.2) PAGE _1_ OF 2d.
\IORK ORDER: 010028300 SI KEY: P0531 PROCEDURE#: 1-SI-68-28 TITL,E: PRIMARY RADIOCHEMISTRY REQUIREMENTS PERF SECT: CEM TEST REASON; PERIOOIC PERFORMANCE _ _ _ _ _ .N/A._ _ _ _ __ _ _ _ _ _ _ N/A DUE: 04/09/01 AUTHORIZAT!Qlj TO BEGIN; SRO DATE ~
\IBN EXT: 04/10/01 ISTART MAX EXT: FORM SPP-8.2-2 FREQ:
EQ: II N
'i r/MDATE I..!!+/-':il:__ TIME ASME XI: N APP MOOE: 123 PERF MOOE: 1234 61 1 /§0iJ SUBSQNT RVllS: C P ETION DATE ~
INSTRUCTIONS; Do ~OT start prior to scheduled due date TEST PERFORMERS \IAS THIS A COMPLETE OR PARTIAL X NAME PERFORMANCE? (EXPLAIN 11 PARTIAL 11
,_/-ZI: : -il _,s:I1:.G*N:~;ru:R:E:~ *~ ~ ~-~
IN REMARKS) COMPLETE: PARTIAL; _
- \IERE ALL TECH SPEC/TECH REQ/OOCM/FIRE "
PROT REC ACCEPTANCE CRITERIA SATISFIED? YES;~ NO;_ N/A:_ llERF Al L OTHER ACCFPUNCE CRITERIA SATISFIED? YES:_ NO:_ NIAL ALERT SCHEDULING REQUIRED? YES;_ NO: NfA:_X_ c1d~ DATE
===-=====;=-===-=-----======-~===============::::::::;::;;::;::;:;;::;::;;;;;:=; =====::======:==============================================::::==
REMARKS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ COPY OF STS SENT TO SCHEDULING: 57 INITIALS f~ DATE SECTION/#MENfDUR HRS SECTION/#MENfDUR HRS SECTIONf#MENfOUR HRS SECTION/#MEN/OUR HRS RECORDS TRANSMITTAL#: _ _ _ _ _ __ I!IHI 1111111~ llllll IHI lllll lllll lllll lllll l l llllll lllll lllll llll llll
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 47
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 9-APR-2001 14:25:02.67 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - GASEOUS ACTIVTY FILE IDENT : DKB600:[TVA.SAMPLE.CHEM.NEW]W0l04095767_C401.CNF;l SAMPLE ID W0104095767 C401
- OPERATOR DRKERNS SAMPLE TIME 9-APR-2001 l3:25:
- SAMPLE GEOMETRY GMlK
- SHELF HEIGHT 0
- EFFICIENCY FILE GMlKO SAMPLE TYPE : 1240 CC GAS MARI
- SAMPLE QUANTITY ; 2.51000E+OO CC ACQ DATE & TIME 9-APR-2001 14:14:
- DEADTIME (%) 0.1%
PRESET LIVE TIME 0 00:10:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 00:10:00
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 00:10:00
- NBR ITERATIONS 10 DETECTOR EFFIC CAL DATE DET *'* GSS-3310 2-AUG-1994 11:26:
LIBRARY EFFIC CERT DATE NOBLEGAS 2-AUG-1994 11:26: DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1.25 KEV/CHAN 5.00516E-Ol
- HALF LIFE RATIO 8.00000 OFFSET 1.44837E-01 keV
- ABUNDANCE LIMIT 80.0%
0 COEFFICIENT -1.10914E-07
- CORRECTION FACTOR 1. OOOOOE+OO PEAK START CHAN 140
- PEAK END CHAN 4096
££L ANALYSES : PEAK V16.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8 COLLECTED COUIIT*D ONBY COUNTED BY
'ttd:*O
- r ~
REVIEWED BY : COPIMENTS : Post-NID Peak search Report It Energy Area Bkgnd FWHM Channel Left Pw %Err Fit Nuclides 0 80.94 902 374 1.03 161. 44 157 10 5.2 XE-133 0 151.17 294 315 1.07 301. 75 297. 10 12.8 KR-85M 0 166.01 57 226 1.10 331.42 328 8 47.7 KR-88 0 196.08 202 345 1.03 391.51 387 10 18.5 KR-88 0 249.77 2313 340 1.12 498.78 494 12 2.6 XE-135 0 258.57 59 201 1,04 516.37 513 8 43.2 0 402.62 160 so 1.21 804.27 800 8 11. 0 KR-87 0 510.99 7378 174 2.34 1020.88 1013 19 1.2 0 526.86 66 31 1. 08 1052.60 1048 11 20. 7 XE-1351'1 0 609.00 34 35 0.87 1216.79 1209 16 43.6 XE-135 0 677. 82 14 17 1.85 1354.36 1348 9 61.5 0 834.68 38 12 1. 75 1667.96 1662 12 24.5 KR-88 0 897.59 24 13 2.20 1793.76 1789 9 34.3
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN I Page: 48
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 REPORT NAME QA CHECK (Vl0.4) PAGE : 1 REPORT DATE 9=-APR-2001 14:25 REQUEST OR DRKERNS TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - GASEOUS ACTIVTY SAMPLE No. W0104095767 C401 OPERATOR NAME DRKERNS SAMPLE TYPE 1240 CC GAS-MARI SAMPLE GEOMETRY GMlK COUNT TIME 9-APR-2001 14:14:53. SAMPLE QUANTITY 2.51000E+OO SAMPLE TIME 9-APR-2001 13:25:00. DETECTOR DET #4, GSS-3310 LIBRARY NOBLEGAS PEAK ENERGY DECAY CORR ISOTOPE ENERGY DIFF (KEV) uCi/CC COMMENTS AR-41 1293.64 -0.10 2.696E-03 QA Results OK KR-85M 151.18 -0.01 2.013E-04 QA Results OK KR-87 402.58 0.04 4.809E-04 QA Results OK KR-88 196.32 -0.24 4.982E-04 QA Results OK XE-133 81. 00 -0.05 1.202E-03 QA Results OK XE-135 249.79 -0.03 1. 676E-03 QA Results OK XE-135M 526.56 0.30 l.105E-03
....... QA Results OK AVG ENERGY DIFF = -0.01 7.859E-03 TOTAL GAMMA ACTIVITY O.OOOE+OO Total DGL Activity 7.859E-03 Total Gas Activity UNIDENTIFIED/REJECTED PEAKS GAMMA/SEC POTENTIAL ENERGY NET AREA FWHM GAMMA/SEC .-.
/CC________ AG ID ACTIVITY 258.57 ---~-
- 59. 1.04 3.448E+OO 1. 374E+OO R XE-138 1. 724E-03 510.99 7378. 2.34 7.643E+02 3.045E+02 u -1 6.016E-03 u ANN IL 0,000E+OO 677.82 14. 1.85 1. B29E+OO 7.289E-01 u AG-llOM 1. 845E-04 4.949E-04 897.59 1835.84 24.
17. 2.20 2.13 4.089E+OO 1. 629E+OO 5.418E+OO u~ u u U
-8 RB-88 3.932E-04 4.716E-05 3.408E-04 u - 5.872E-05
Calculation No. WBNAPS3077 j Rev: 014 j Plant: WBN I Page: 49
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB Date: 9-1-2011 Steam Line Break Checked: JEB Date: 9-1-2011 9-APR-2001 10:48:39.98 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY FILE IDENT : DKB600:[TVA.SAMPLE.CHEM.NEW]W0104095766_C402.CNF;l SAMPLE ID W0104095766 C402
- OPERATOR WNCLONTZ SAMPLE TIME 9-APR-2001 08:20:
- SAMPLE GEOMETRY 65ML
- SHELF HEIGHT 1
- EFFICIENCY FILE 65ML1 SAMPLE TYPE : RCS 65ML BOTTLE
- SAMPLE QUANTITY 1.58100E+Ol GRAMS ACQ DATE & TIME 9-APR-2001 09:45:
- DEADTIME (%) 4.6%
PRESET LIVE TIME 0 01:00:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 01:02:52
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 01:00:00
- NBR ITERATIONS 10 DETECTOR DET #4, GSS-3310
- LIBRARY RC SL I QUID EFFIC CAL DATE 19-JUL-2000 20:26
- EFFIC CERT DATE 19-JUL-2000 20:26 DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1.25 KEV/CHAN 5.00516E-Ol
- HALF LIFE RATIO 8.00000 OFFSET 1. 44837E-Ol keV
- ABUNDANCE LIMIT 80.0%
0 COEFFICIENT -1.10914E-07
- CORRECTION FACTOR 1.00000E+OO PEAK START CHAN 140
- PEAK END CHAN 4096
;ti /!IL ANALYSES : PEAK V16.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8 coUHTED COLLECTED oNBY ,:
COUNTED BY : LON Z , REVIEWED BY : COMMENTS : ---=- Post-NID Peak Search Report It Energy Area Bkgnd FWHM Channel Left Pw %Err Fit Nuclides 0 134.65 991 52495 0.82 268.75 266 7 38.5 W-187 I-134 0 249.78 2586 43764 1.22 498.80 496 7 13.6 XE-135 0 364.48 569 19365 0.96 728. 04 726 6 39.0 I-131 0 405.26 542 12870 1. 06 809.53 807 7 35.0 I-134 0 433.34 305 11020 0.97 865.66 863 7 57.4 I-134 0 462.88 898 11840 1.08 924.71 921 8 21.3 CS-138 0 478.53 2028 19343 2.93 955.98 948 13 14. 3 BE-7 W-187 0 510.97 1460805 45497 2.66 1020.82 1013 18 0.1 F-18 0 522. 71 1163 2480 1.33 1044.29 1041 8 8.0 I-132 2 526.52 1531 2060 1.17 1051. 91 1048 16 5.5 9.34E-01 I-135 XE-135M
Calculation No. WBNAPS3077 I Rev: 014 I Plant: WBN l Page: 50
Subject:
Offsite and Control Room Operator Doses Due to a Main Prepared: MCB l Date: 9-1-2011 Steam Line Break Checked: JEB l Date: 9-1-2011 9-APR-2001 10:48:39.98 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY FILE IDENT : DKB600:[TVA.SAMPLE.CHEM.NEW]W0104095766_C402.CNF;l SAMPLE ID W0104095766 C402
- OPERATOR WNCLONTZ SAMPLE TIME 9-APR-2001 08:20:
- SAMPLE GEOMETRY 65ML
- SHELF HEIGHT 1
- EFFICIENCY FILE 65ML1 SAMPLE TYPE : RCS 65ML BOTTLE
- SAMPLE QUANTITY : 1.58100E+Ol GRAMS ACQ DATE & TIME 9-APR-2001 09:45:
- DEADTIME (%) 4.6%
PRESET LIVE TIME 0 01:00:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 01:02:52
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 01:00:00
- NBR ITERATIONS 10 DETECTOR DET #4, GSS-3310
- LIBRARY RCSLIQUID EFFIC CAL DATE 19-JUL-2000 20:26
- EFFIC CERT DATE 19-JUL-2000 20:26 DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1. 25 KEV/CHAN 5.00516E-Ol
- HALF LIFE RATIO 8.00000 OFFSET 1. 44837E-Ol keV
- ABUNDANCE LIMIT 80.0%
Q COEFFICIENT -l.10914E-07
- CORRECTION FACTOR 1.00000E+OO PEAK START CHAN 140
- PEAK END CHAN 4096 Pi:tL ANALYSES : PEAK V16.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8
~
~
COLLECTED COUNTED ONBY :' COONTED BY ' Z REVIEWED BY : COMMENTS : -== Post-NID Peak Search Report It Energy Area Bkgnd FWHM Channel Left Pw %Err Fit Nuclides 0 134.65 991 52495 0.82 268.75 266 7 38.5 W-187 I-134 0 249.78 2586 43764 1.22 498.80 496 7 13.6 XE-135 0 364.48 569 19365 0.96 728. 04 726 6 39.0 I-131 0 405.26 542 12870 1. 06 809.53 807 7 35.0 I-134 0 433.34 305 11020 0.97 865.66 863 7 57.4 I-134 0 462.88 898 11840 1. 08 924. 71 921 8 21.3 CS-138 0 478.53 2028 19343 2.93 955.98 948 13 14.3 BE-7 W-187 0 510.97 1460805 45497 2.66 1020.82 1013 18 0.1 F-18 0 522. 7l 1163 2480 1.33 1044.29 1041 8 8.o I-132 2 526.52 1531 2060 1.17 1051. 91 1048 16 5.5 9.34E-Ol I-135 XE-135M
Calculation No. WBNAPS3077 I Rev: 014 l Plant: WBN I Page: 51
Subject:
Offsite and Control Room Operator Doses Due to a Main J Prepared: MCB Date: 9-1-2011 Steam Line Break [Checked: JEB Date: 9-1-2011 REPORT NAME QA CHECK (Vl0.4) PAGE 1 REPORT DATE 9=APR-2001 10:48 REQUEST OR WNCLONTZ TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY SAMPLE No. W0104095766 C402 OPERATOR NAME WNCLONTZ SAMPLE TYPE RCS 65ML BOTTLE SAMPLE GEOMETRY 65ML COUNT TIME 9-APR-2001 09:45:36. SAMPLE QUANTITY 1. 58100E+Ol SAMPLE TIME 9-APR-2001 08:20:00. DETECTOR DET #4, GSS-3310 LIBRARY RCSLIQUID PEAK ENERGY DECAY CORR ISOTOPE ENERGY DIFF (KEV) uCi/GRAM COMMENTS F-18 511. 00 -0.03 1.116E-01 QA Results OK NA-24 1368.53 0.05 2.060E-03 QA Results OK MN-56 1810.69 0.33 2.088E-04 QA Results OK C0-58 810.76 o.oo 6.218E-04 QA Results OK C0-60 1173.22 0.28 2.776E-05 QA Results OK NB-95 765.79 0.46 2.794E-05 QA Results OK I-131 364.48 o.oo 3.881E-05 QA Results OK I-132 667.69 o.oo 1.165E-03 QA Results OK I-133 529.87 0.01 6.105E-04 QA Results OK I-134 847.03 -0.04 2.334E-03 QA Results OK I-135 1260.41 -0.03 1.158E-03 QA Results OK XE-135 249.79 -0.02 :t.380E-04 QA Results OK XE-135M 526.56 -0.04 1.972E-02 QA Results OK CS-138 1435.86 -0.20 2,195E-03 QA Results OK AVG ENERGY DIFF = 0.06
- 1. 419E-01 TOTAL GAMJllA ACTIVITY 1.168E-01 Total DGL Activity 2.223E-03 Total FP Activity 2.918E-03 Total AP Activity
- 1. 986E-02 Total Gas Activity 5.307E-03 Total HFP Activity Dose Equivalent Iodine-131 = 2.098E-04 Iodine 131/133 Ratio 6.357E-02 Iodine 133/135 Ratio = 5.274E-Ol 134.65 KeV Peak was used in identifying 2 isotopes 478.53 Kev Peak was used in identifying 2 isotopes 526.52 Kev Peak was used in identifying 2 isotopes 546.50 KeV Peak was used in identifying 2 isotopes 766.25 KeV Peak was used in identifying 2 isotopes 772. 62 Kev Peak was used in identifying 2 isotopes 810.76 Kev Peak was used in identifying 2 isotopes 835.61 Kev Peak was used in identifying 2 isotopes 846.98 Kev Peak was used in identifying 2 isotopes
Attachment 2 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Calculation (73 pages including this cover page)
NPG CALCULATION COVERSHEET I CTS UPDATE Page 1 REV 0 EDMS/RJMS NO CI'SME* EDMS 'OO'E: EDMS ACCF.SSlON NO (NIA for REV. Ol 826 890410 001 Calculation CALCULATIONS (NUCLEAR) T93l Ca&Tille: Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture CALCID NUC WBN NI'B WBNTSR008 015 . 016 crs tlPDAiB ONLY 0 (Verifier and Approwl Signatures NotRequimi) II No crs CHANGES o (For ew revision, crs bas bcm mvicweii and DO crs ehanges zequimi) Yfilr (check one) SYST£MS ~ 0 181, l 0,2 0,3 0 NIA NIA DCN,EDC NIA APPUCABLE DFSJGN DOClJMENT(Sl cuss!F!CATION NIA NIA B QUAUIY SAFE1Y REIATED? DESIGN OllI'PUT SAR/J'S and/or !SFSI REl'.ATED'l (If ya, QR= yes} A'ITACHMENTI SAR/CoC AJlFEC1'E!) Yes No D Yes No 0 No Yes No S Yes ml No 0 CALCULATION NUMBERREQUESTQR yEJUFICATJQN METHOD tiEWMEilfoD OF ANALVSIS Name: NIA PHONE:N/A Desiga Rmew D Yes 181 No . NAME AND SIGN) STATEMENT OF PRQBLEM/ABSTRACT This analysis de:tennines the oontro1 room operator and otmte dose due to a Steam Generator Tube Rupturc: aceidmt The steam tdeases (primary and seeondary side) are fi'om Westinghouse. The activities of the primary coolant are based on a revised technical specification limit with a preexisting iodine spike of a factor of 14 jlCilgm I-131 equivalent (maximum 48 hoar}. Also analyzed iR reactor coolant activities 0.265
µCi/gm 1-13 l eqaivaI~ with an acc:ident initiated iodine spike of 500 times the release rate from the fuel The seeondary side activities come from WBNNA13-003 and ate 8e\: at 0.1 µCi/gm. Credit is taken for partial flashing of the reactor coolant as it enters the steam gcnerafor. Far conscn:.ttism. no credit is taken for "scrubbing" of iodine in the slam bubb1es es the bubbles rise through the water. The production rate of iodines is based on a pre-accident steady state reactor coolant leakage of 11 gpm (10 gpm known+ l gpm unknown).
The released activities are used es input to oomputm' code cor<.oo which detmmines the control roam operator dose. The base COROD model is taken from TI-RPS-198. Tue control room inlab ventXIQ values are taken from WBNAPSJ-104, and were detcnnined aslngthe ARCON96 code. The activities ere used es input to the computer FENCOOSE, which determines-the olfsitc dose. The control room operator dose case is examined which consi~ the effect of slow closare times ofO-FCV-31-3, -4 (14 seconds) and the radiation monitor response time (60 sec) for a total closure time of 74 seconds. The off'site doses (gamma, beta, thyroid and TEDE) due to a SG'Ill with a preexisting iodine spike de not exceed the 10CFR.100 limits (25 rem gamma (whole body), 300 rc::m beta (skin), 1111d 300 ran thyroid per NUREG-0800). The SOTR with accident initiated iodine spike does not exceed a small &aotion of the lOCfR.l 00 limits (10% of the 10CFR.100 limits of 25 rem gamma, 300 rem beta, and 300 ran thyroid per NUREG-0800). The control room doses dae to a SGTR. do not exceed the 10CFRSO Aw.A GDC 19 limits (5 rem ganuna, 30 rem beta and 30 rem thyroid). This calculation impacts FSAR Table 15.5-18 and ~19 MICkOPICHPJEFJCHE Yes F1CHENUMB S TVA*F-W003232 TVA40532 Pa&o l of2 NEDP-:2:-1 [10-31-2011] LiGIBILrfY EVAUJATEO lliN~
- ACCEPTED FOR ISSUE Inrtiais: M Date~_yJf..<
NPG CALCULATION COVERSHEET I CTS UPDATE BUIIDING NIA I ROOM NIA I ELEVATION NIA I COORD/AZIM NIA I FIRM Bechtel CATEGORIES NIA. SR/LC KEYWORDS (A-add, D-delete) ACTION KEYWORD AID KEYWORD AID CROSS-REFERENCES (A-add, D-delete) ACTION XREF XREF XREF XREF XREF (AID) CODE PLANT TYPE NUMBER REV A p WBN PER 766429 A p WBN PER 775553 CT'S ONLY UPDATES: Followin~ are reauimd onlv when~ keyword/cross reference CT'S utX!atm and :>al!e 1 of form NEDP-2-1 is not included: PREPARER (PRINT NAME AND SIGN) DATE CHECKER (PRINT NAME AND SIGN) DATE PREPARER PHONE NO. EDMS ACCESSION NO. 1VA40532 Page2of2 NEDP-2-1 [10-31-2011]
p aQe 3 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR008 Title Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Revision DESCRIPTION OF REVISION No. 0 Initial Issue Revision I was performed because the base COROD model from TI-RPS-198 changed. I pages changed: 1-6, 8, 9, 24-26, 30, 32-37, 39-60, 62, 63, 73-78 pages added: 24.1-24.3 Revision 2 was performed to incorporate the new steam releases as determined by Westinghouse with WBN specific parameters 2 and to add the offsite dose to the analysis. pages changed: 1-24.2, 25-67 (old coversheet now page I.I) pages deleted: none rnU>es added: new cover (new page I) Revision 3 was performed because of revised steam releases as determined by Westinghouse. The entire calculation was rewritten. 3 All pages were renumbered. Revision bars are shown only on areas of text which actually changed. Text which was reformatted does not show revision bars. pages changed: all pages deleted: none pages added: none Revision 4 was performed because the XJQ values changed. 4 pages added: I (new cover), 1.2 (abstract), 41.1, 41.2 pages deleted: none n:li!es changed: I.I (old cover), 2-8, 17, 18, 20-22, 24-41 Revision 5 was revised because the control room intake flow was changed. 5 pages added: none pages deleted: 41.1, 41.2 pages changed: 1, 1.2, 2-8, 18, 20-22, 24-41, 51-53 Revision 6 was performed because of revised reactor coolant and steam mass releases as determined by Westinghouse. 6 pages added: 6.1 pages deleted: none pages changed: 1-8, 1.2, 11-14, 18, 20, 24-41 Revision 7 was performed as part of the corrective action of WBN PER 98-016506-000. The revision changed the basis of the 7 source terms from the historical design values provided by Westinghouse which are located in the FSAR to the expected source terms based on ANSI/ANS-18.1-1984. No other modifications with respect to methodology were made. Other pending changes (such as alternate XJQ values, new Tech Spec limits, inclusion of the radiation monitor response time in the isolation time of the control room, impact of Tritium Production Core, and iodine spiking) will be dealt with in subsequent revision(s). There is no FSAR impact since there will be more changes in the near future, and the doses in this revision are less than R6. Pages added: 1 (new cover), 13.1, 50.1-50.3 Pages deleted: Attachment 3 Pages changed: Ia (old cover), 2, 3, 6, 6.1, 7-13, 18, 24-40, 46-50 R 7: 75 total pages Revision 8 is performed to increase the delay in the control room isolation time from 14.0 sec to 20.6 sec(= 14 sec damper closure 8
+ 6.6 sec instrument response) as part of the corrective action for WBN PER 01-000080-000. New XJQ values as determined by ARCON96 are incorporated. The Tritium Production Core (TPC) is included. The latest versions of CO ROD (R5) and FENCDOSE (R4) are used, which now determine the thyroid doses based on lCRP-2 and ICRP-30 dose conversion factors as well as the TEDE. Finally, the iodine spiking is treated differently. The preaccident iodine spike is the maximum Technical Specification limit (60 µCi/cc I-131 equivalent. Also 21 µCi/cc and IO µCi/cc are analyzed). An accident initiated iodine spike with a factor of 500 increase in iodine release from the fuel with the initial activity at 1 µCi/cc, 0.265, or 0.177 µCi/cc I-131 equivalent with the baseline iodine production based on either IO gpm, 5.75 gpm or 2.15 gpm, is now included in the analysis.
Cases with the old Halitsky XJQ values are also performed. Additional justification for use of the ANS/ANSI-18.1-1984 spectrum was included. Finally, 3rd party review comments from Westinghouse and NISYS were incorporated. Due to the extent of the revision, all pages were renumbered. Text changes are marked with revision bars. Pages added: all Pages deleted: all Pages changed: all R8: 72 total pages The non-TPC results will affect FSAR section 15.5.5 and Tables 12.5-18 and 12.5-19, and the change will be processed in accordance with NADP-7 to reflect calculation results. A IOCFR50.59 evaluation is needed for these changes. In addition, Technical Specification LCO 3.4.16 and associated bases (RCS Iodine Concentration) will be affected by this calculation revision and a TS change will be required. These actions will be tracked under the corrective actions for PER 00-012545-000. TVA 40709 [12-2000] Page 1 of 1 NEDP-2-2 [12-04-2000]
page 4 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSROOS Title Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Revision DESCRIPTION OF REVISION No. 9 Revision 9 is performed for replacement steam generators (DCN 51754). The mass releases have been changed, resulting in different answers. The original steam generator results are retained. Also, the CREVS recirculation rate and time increments were corrected as part of the corrective action of PER 61493 and 94426. A 2 CREVS train operation for 2 hours is also performed. This calculation impacts FSAR section 15.4 and 15.5. The full impact to the FSAR and Technical Specifications will be addressed in the screening review of DCN 51754. Page numbers were redone, only actual text changes are marked with revision bars. Pages added: 4, Appendix F (p.36-37) Pages deleted: design verification form Pages changed: 1-3, 5-10, 14-29 R9: 63 total pages Revision I 0 is in support of DCN 51754. WBNNAL3003 has been revised to show the change in RCS volume due to the JO Replacement Steam Generators (RSG). R9 of this calculation used a RCS volume from a Westinghouse document (WBJRSG-TR-02) that has been revised since the issue of R9. Also, the 2 train CREVS cases have been deleted as it was determine to be beyond design basis. An assumption has been added to discuss this issue. Impacts to the FSAR and TS's, if any, will be addressed in the screening review ofDCN 51754. Pages Added: None Pages Deleted: None Pages Revised/Replaced: I, 2, 4-7, JO, 14-17, 20, 23, 25-27, 29 RI 0: 63 total pages. Revision 11 is performed to perform the SGTR analysis for Unit 2 (Appendix G). The steam generators are the same as the Ii original Unit I steam generators, however the Westinghouse mass release calculations are different. SAR has been reviewed by _ Marc Berg and this revision of the calculation affects Unit 2 SAR section Chapter 15. A SAR change shall be processed in accordance with NGDC PP-JO to reflect the calculation results as part of EDCR 54956. Tech Specs have been reviewed and determined not to be affected. Pages Added: Appendix G (p. 39,40) Pages Deleted: none Pages Revised/Replaced: I, 2, 4-9, 21, 23 RI I: 66 total pages. Revision 12 is performed to address replacement of the analog ratemeters with digital RMIOOO ratemeters by DCN 52012. The 12 longer response time of the RM 1000 ratemeter is incorporated for the Unit I accident by increasing the control room isolation time from 20.6 seconds to 40 seconds ( 14 sec damper closure+ 26 sec instrument response). The response time for the Unit 2 accident will be revised prior to Unit 2 startup. The only change is an increase in the control room HV AC time to isolation from 20.6 seconds to 40 seconds. Therefore, only the control room dose changes, and the offsite doses are retained from revision I 0. See calculation WBNTSR028, Rev. 7, (ref. 38) for the revised time delay of 40 seconds. FSAR sections 15.5.4 & 15.5.5 and Technical Specifications were reviewed by L):'.nn Cowan and Table 15.5-19 is impacted by the change in isolation time from 20.6 seconds to 40 seconds. Pages Added: 9 Pages Deleted: none Pages Revised/Replaced: I, 2, 4-7, 10 18, 19, 22, 23, 24-31 RI 2: 67 total pages. TVA 40709 [12-2000] Page 1of1 NEDP-2-2 [12-04-2000]
Page 4A NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR008 Title Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Revision DESCRIPTION OF REVISION No. 013 Revision 013 of this calculation corrects the errors in the Uriit 1 and 2 calculations identified in PERs 327956 and 327968:
- corrected some typographical errors,
- changed the Unit 1 isolation time for the offsite doses to 216 s,
- changed the Unit 1 calculated margins for an accident initiated iodine spike,
- changed the gamma and TEDE offsite dose limits to 2.5 rem,
- changed computer code input files,
- changed the Unit 2 steam generator volumes,
- changed the Unit 2 X/Q values for control room,
- changed the method to develop the I-131 equivalent conversion factor,
- added technical specification revision for a preexisting iodine spike of 14 µCi/gm I-131 equivalent,
- reviewed 12 successor calculations to this calculation based on CCRIS reference list, and 9 of them are impacted due to changes made in this calculation. They are: WBNAPS304 7R4, WBNAPS3048R20, WBNAPS3079R3, WBNAPS3 l l OR2, WBNAPS3 l l 8RO, WBNAPS3121RO, WBNTSR041R2, WBNTSR095Rl, and WBNTSR100R8.
The effect of Unit 2 operation on Unit 1 margins has been reviewed with no impact. Successor calculations have been reviewed and are not impacted by this revision. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. FSAR AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND ARE NOT AFFECTED BY THIS REVISION OF THE CALCULATION. Reviewer: _ _ _ _ _ _ _ _ _ __ Pages Added: 4A, 41A, 41B Pages Revised/Replaced: 5- 41 Pages Deleted: None Total number of pages in this revision including Attachments: 67 (Rl2) + 3 (Rl3) = 70 Pages Appendix A (2 pages), Appendix B (2 pages), Appendix C (2 pages), Appendix D (1 page), Appendix E (6 pages), Appendix F (2 pages), Appendix G (4 pages),Attachment 1 (14 pages), and Attachment 2 (12 pages) 14 Revision 14 is performed to upgrade the X/Qs to the 1991-2010 meteorological data set. FSAR sections 15.5.4 & 15.5.5 and Technical Specifications were reviewed by Marc Berg and Table 15.5-18,.is impacted. Calculations impacted by this revision have either already been updated or scheduled to be revised (see WBNAPS3-104). Pages Added: none Pages Revised/Replaced: 1, 2, 4A, 5-7, 9, 10, 19, 22-24, 29, 31, 38, 42, 43 Pages Deleted: None Rl4: 70 total pages TVA 40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Page 4B NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR008 Title Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Revision DESCRIPTION OF REVISION No. 015 Revision 15 is performed to update the Tritium Production Core (TPC) tritium source term to 124 µCi/gm (as revised in predecessor calculation WBNNAL3003). Since the results of this calculation are intended for use in WBN FSAR Section 15.5, a design output attachment was created and added to the calculation in Revision 15. Attachment 3, Form NEDP-2-5, makes the entire Calculation WBNTSR008 Revision 15 design output. As a result, the calculation classification has been changed to 'EO' from 'E'. The special requirement and/or limiting condition are removed in Revision 15 of the calculation. In addition, the efiche number onpage.9 for the Revision 13 computer runs is changed to 'TVA-F-W002394.' The original number
'TV A-F-W002375' is incorrect.
Upon preparation of Revision 15 to the calculation, it was discovered that the Unit 2. STP model for the accident-initiated iodine spike case incorrectly modeled the flow rate governing the 0-153 second time interval from the reactor coolant component to the faulted steam generator component. The flow rate is an order of magnitude low (i.e., l .227E7 gm/hr versus l .227E8 gm/hr). This error is not corrected in this revision. This error will be resolved under SR762528. Pages replaced in Revision 15 contain an updated page header for Revision 15. All unaffected pages retain their original headers. Changes from the previous revision are marked on the replaced Revision 15 pages with revision bars. CTS was reviewed for successor calculations to Calculation WBNTSR008, and several successor calculations were identified. Calculations WBNAPS3077, WBNAPS3Il8, WBNTSR083, and WBNTSR084 are affected because they use the FENCDOSE and CO ROD models modified in this revision. Although Calculation WBNNAL3008 uses the tritium release values from Calculation WBNTSR008, Calculation WBNNAL3008 is not affected because only the gamma contribution is considered in Calculation WBNNAL3008. The other successor calculations are not affected by this revision because they do not use inputs that were changed aS part of this revision. See DCN 61599 for SAR!fech Spec impact determination. Pages Added: 48 and 70 Pages Replaced or Revised: I, 2, 5 - 7, 9 - 12, 22, 24 - 31, 42, 43, and 70 Pages Deleted: none Total Revision 15 pages: 72 Attachment I - Justification for Using ANSI/ANS-18.1-1984 Expected Coolant Spectrum (14 pages) Attachment 2- Maintenance Request Form A-482000 (12 pages) Attachment 3 - NPG Calculation Design Output (I page) TVA 40709 [10-2008] Page I of I NEDP-2-2 [I 0-20-2008]
Paae4C NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSROOB Title Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Revision DESCRIPTION OF REVISION No. 016 Revision 016 of this calculation was created to address the instrument response time based on PER 775553 and to address the errors documented in PER 766429. The longer response time is incorporated by increasing the control room isolation time from 40 seconds to 74 seconds (14 sec damper closure+ 60 sec instrument response). The following changes are made in this revision:
- The primary change is to increase the control room HVAC time to isolation from 40 seconds to 74 seconds. The STP, COROD, and FENCDOSE inputs are affected, and the related cases are rerun.
- This revision corrects the errors in Unit 2 flow rates documented in PER 766429 documents.
- This revision removed design ou1put that was added in RI 5.
- Appendix F is deleted in this revision, because it is out of date.
- R14 used an earlier version of Word file from Rl3. Some editorial comments in the final version ofR13 were not included in Rl4. This revision restores some editorial changes from Rl3.
Successor documents have been reviewed and are not impacted by this revision. However, the following calculations will be updated due to cancellation ofDCN 52012: WBNfSR008, WBNAPS3050, WBNAPS3077; WBNfSR028 and WBNfSR064. Affected engineering judgments and assumptions were reviewed and (1) were found to be adequate, or (2) were revised as.necessary to ensure adequacy. Ultimate heat sink (UBS) temperature was not used as an input to the calcUlation analyses. Therefore, existing calculation results will not be atfected by changing the UHS technical specification temperature. The effect of Unit 2/dual unit operation on Unit I margins has been reviewed with no impact FSAR SECTION 15.5.4 & 15.5.5 AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND TABLE 15.5-18, & 15.5-19 ARE IMPACTED BY THE CHANGE IN ISOLATION TIME FROM 40 SECONDS TO 74 SECONDS./"\ / J/ ~ Reviewer: K..tJ.reierlAt-414 '),_".f4 (JfLUc.--- 9/11/!3 Pages Added: 4C (total 1 page) Pages Revised/Replaced: 1, 2, 5 - 7, 9 - 12, 18 - 19, 22- 31, 38 -43 (total 27) Pages Deleted: 70 Total number of pages in this revision including Attachments: 72 pages (Rev. 15) + 1 page (Rev. 16)-1 page (Rev. 15) = 72 Appendix A 2pages Attachment 1 14pages AppendixB 2pages Attachment 2 12 pages AppendixC 2pages AppendixD 1 pages AppendixE 6pages AppendixF 1 pages AppendixG 4 pages This page added by Revision 016 TVA40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Paae 5 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNTSROOB I Revision: 016 I TABLE OF CONTENTS SECTION TITLE PAGE Calculation Coversheet I CTS update 1 Calculation Record of Revision 3 Calculation Table of Contents 5 Calculation Verification Form 6 Computer Input File Storage Information Sheet 7 Computer Output Microfiche Information Sheet 8 Purpose 10 Introduction 10 Assumptions 11 Special Requirements/Limiting Conditions 12 Calculations 12 Results 22
Conclusions:
23 References 23 Appendix A: Example of STP Model (preaccident Iodine spike) 25 Appendix B: Example ofSTP Model (accident initiated Iodine spike) 27 Appendix C: Example of COROD Model(ARCON96 X/Q) 29 Appendix D: Example of FENCDOSE Model 31 Appendix E: Determination of Letdown Flow Uncertainty 32 Appendix F: Deleted in Rl6 38 Appendix G: Unit 2 SGTR 40 Attachment 1- Justification for Using ANSl/ANS-18.1-1984 Expected Coolant Spectrum (14 44 pages) Attachment 2 - Maintenance Request Form A-482000 (12 pages) 58 TVA 40710 [10-2008) Page 1of1 NEDP-2-3 [10-20-2008) This page replaced by Revision 016
Pae 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNTSR008 Revision 016 Method of verification used:
- 1. Design Review 181
- 2. Alternate Calculation D Verifier
- 3. Qualification Test D Comments:
The changes to the calculation described in the Record of Revision for Revision 016 have been reviewed and have been found to be technically adequate in format and content. TVA 40533 (10-2008] Page 1of1 NEDP-2-4 (10-20-2008) This page replaced by Revision 016
p aae 7 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNTSROOB I Rev. 016 I Plant: WBN I
Subject:
Control Room Operator and Offslte Doses Due to a Steam Generator Tube Rupture I I Electronic storaae of the inout files for this calculation is not reauired. Comments:
~ Input files for this calculation have been stored electronically and sufficient identifying information is orovided below for each inout file. (Anv retrieved file reouires re-verification of its contents before use.)
The computer input for R6 is permanently stored in FILEKEEPER file # 300203. The computer input for R7 is permanently stored in FILEKEEPER file# 303287 The computer input for R8 is pennanently stored in FILEKEEPER file # 30358 I The computer input for R9 is permanently stored in FILEKEEPER file # 30746 I and 308002 The computer input for RI 0 is permanently stored in FILEKEEPER file # 308287 The computer input for RI I is stored in eFiche file TVA-F-WOO 136 I The computer input for RI2 is permanently stored in FILEKEEPER file #3 I4524 The Word file is permanently stored in FILEKEEPER file# 3 I4522 and the pelf for the attachment is stored in FILEKEEPER file #3 I4523 The computer input for Rl3 is permanently stored in FILEKEEPER file# 3 I 7582 The Word file is permanently stored in FILEKEEPER file # 3 I 7583 The computer input for RI4 is permanently stored in eFiche file# TVA-F- W002550 The RI5 input files are permanently stored in FILEKEEPER file# 32 I943 The RI5 output files are permanently stored in eFiche file# TVA-F- W003I96 The Word file is permanently stored in FILEKEEPER file# 32 I982 The computer input for RI 6 is permanently stored in FILEKEEPER file # 322083 The Word file is permanently stored in FILEKEEPER file# 322429 181 Microfiche/eFiche (See next page.) TVA40535 (10-2008] Page 1of1 NEDP-2-6 [10-20-2008) This page replaced by Revision 016
TV AN COMPUTER OUTPUT page8 MICROFICHE INFORMATION SHEET Document WBNTSROOS I Rev. 014 I Plant: WBN I
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Microfiche Number Description R5: TV A-F-C000107 R6: R6: TV A-F-C000244 Name Code Description TSROOSS6 STP I 4sec delay source term TSROOSF6 FENCDOSE offsite dose TSROOSC6 COROD no damper delay control room operator dose R7: R7: TV A-F-C000322 Name Code Description TSROOSS7 STP l 4sec delay source term TSROOSF7 FENCDOSE offsite dose TSROOSC7 COROD no damper delay control room operator dose RS: RS: TV A-F-C000356 Name Code Description TSRSSS# STP 20.6 sec delay source term. TPC TSRSSS#N STP 20.6 sec delay source term. conventional core TSRSFS# FENCDOSE offsite dose. TPC TSRSFS#N FENCDOSE offsite dose. conventional core TSRSCS# COROD control room operator dose, TPC, ARCON96 XJQ TSRSCS#N COROD control room operator dose conventional core, ARCON96 XJQ TSRSCS#X COROD control room operator dose, TPC, Halitsky XJQ TSRSCS#Y COROD control room operator dose, conventional core, Halitsky XJQ where#=: A=60µCi/gm I spike, B=21µCi/gm I spike, C=IOµCi/gm I spike, D=lµCi/gm I w/5001 spike (10 gpm leak basis), E=0.265µCi/gm I with w/5001 spike(lO gpm leak basis), F=0.177µCi/gm(10 gpm leak basis), G=0.265µCi/gm I with w/5001 spike(2.15 gpm leak basis), H=O.l 77µCi/gm(2.15 gpm leak basis), 1=0.265µCi/gm I with w/5001 spike(5.75 gpm leak basis), J=0.177µCi/gm(5.75 gpm leak basis) R9: R9: Name Code Description TVA-F-W000500 and TSRSS9# STP 20.6 sec delay source term, TPC TV A-F-W000573 TSRSF9# FENCDOSE offsite dose, TPC TSRSC9# COROD control room operator dose, TPC, ARCON96 XJQ where #=A,C, E, G=2 l µCi/gm I spike, B,D, F, H=0.265µCi/gm w/500 I spike (IO+ I gpm leak basis); A,B, E, F=ICREVS case, C,D, G, H=2 CREVS case; A,B,C,D=replacement steam generator, E,F,G,H=original steam generators RlO: TSRSS9A STP 20.6 sec delay source term, TPC, 21 µCi/g TV A-F-W000613 TSRSS9B STP 20.6 sec delay source term, TPC, 0.265 µCi/c TSRSF9A FENCDOSE offsite dose, TPC, 21 µCi/g TSRSF9B FENCDOSE offsite dose, TPC, 0.265 µCi/c TSRSC9A CO ROD control room operator dose, TPC, 21 µCi/g TSRSC9B CO ROD control room operator dose, TPC, 0.265 µCi/c TSRSSllA STP 20.6 sec delay source term, TPC, 21 µCi/g Rll: TVA-F-WOOl361 TSRSSllB STP 20.6 sec delay source term, TPC, 0.265 µCi/c TSRSF!lA FENCDOSE offsite dose, TPC, 21 µCi/g TSRSFllB FENCDOSE offsite dose, TPC, 0.265 µCi/c TSRSCllA CO ROD control room operator dose, TPC, 21 µCi/g TSRSCIIB CO ROD control room operator dose, TPC, 0.265 µCi/c (continued on next page)
NPG COMPUTER OUTPUT Page9 MICROFICHE JNFORMATION SHEET Document WBNTSR008 IRev. 16 IPlant WBN I
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Microfiche Number Description Rl2: TSR8Sl2A STP 40 sec delay source term. TPC, 21 µCi/g 1VA-F-W001561 TSR8Sl2B STP 40 sec delay source term, TPC, 0.265 µCi/c TSR8Cl2A COROD control room operator dose, TPC, 21 µCi/g TSR8Cl2B COROD control room operator dose, TPC, 0.265 µCi/c RB: Rl3:Computer Runs: 1VA-F-W002394 $TSR8Sl3# STP source tenns
$TSR8Cl3# COROD control room dose
$TSR8Fl3# FENCDOSE offsite dose where
$= l for Unit l
$=2forUnit2
#=A for pre-accident iodine spiking, 14 µCi/gm with lo+ 1 gpm primacy to secondary side leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with lo+ l gpm primacy to secondary leak, 500 I spike A total of 12 runs for Rl3.
Rl4: Rl 4:Computer Runs: 1VA-F- W002550 $TSR8Cl4# COROD control room dose
$TSR8Fl4# FENCDOSE offsite dose where
$ = I for Unit l
$=2forUnit2
#=A for pre-accident iodine spiking, 14 µCi/gm with lo+l gpm primacy to secondary side leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with lo+ l gpm primacy to secondary leak, 500 I spike Rl5: Rl5: Computer Runs:
1VA-F-W003196 $TSR8S 15# STP source tenns
$TSR8Cl5# COROD control room dose
$TSR8Fl5# FENCDOSE offsite dose where
$ = l for Unit 1
$ = 2 for Unit 2
# = A for pre-accident iodine spiking, 14 µCi/gm with 1o+1 gpm primacy to secondary side leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with 1o+1 gpm primary to secondary leak, 500 I spike Rl6: Rl6: Computer Runs:
1VA-F- W003232 TSR008Rl6S$# STP source terms TSR008Rl6C$# COROD control room dose TSR008Rl 6F$# FENCDOSE offsite dose where
$ = 1 for Unit 1
$ = 2 for Unit 2
#=A for pre-accident iodine spiking, 14 µCi/gm with 1o+1 gpm primary to secondary side leak
# = B for accident initiated iodine spiking, 0.265 µCi/gm with lo+l gpm primary to secondary leak, 500 I spike Total of 16 output files: 4 STP output file. 4 STP punch files, 4 COROD output files, and 4 FENCDOSE output files This page replaced by Revision 016
1 i1
- Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:10
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Purpose This analysis is performed to determine the control room operator and offsite dose following a design basis steam generator tube rupture accident (SGTR). Introduction This analysis determines the control room operator and offsite dose due to a SGTR. The steam releases (primary and secondary side) are taken from reference 45 and 46. The activities of the primary coolant are based on a preexisting iodine spike of 14 µCi/cc 1-131 equivalent for a pre-accident spike (note: all measurements are at STP, therefore 1 g=lcc water). An alternate accident initiated iodine spike case uses initial activity at 0.265 µCi/cc 1-131 equivalent with a factor of 500 increase in iodine release rate from the fuel. The secondary side activities start at 0.1 µCi/cc I-131 equivalent The secondary side activities come from WBNNAL3003 (ref. 29). Credit is taken for partial flashlng of the reactor coolant as it enters the steam generator. For conservatism, no credit is taken for "scrubbing" of iodine in the steam bubbles as the bubbles rise through the water, therefore it is unimportant if the break is above or below the water level at all times. The computer code STP is used to determine the releases. The released activities are used as input to computer code COROD (ref. 15) which determines the control room operator dose. The base COROD model is taken from TI-RPS-198 (ref. 13, ingress and egress dose was not determined because the accident lasts less than 8 hours and the ingress/egress is after 8 hours). The control room operator dose considers the effect of slow closure times ofO-FCV-31-3, -4. The delay is 74 sec (which includes the 14 sec damp~r closure time and the 60 sec monitor response time) (see justification in Calculation section later). This is conservative because the delay in isolation allows a large slug of unfiltered radioisotopes into the
- control room. It is realistic because the isolation of the control room will most likely occur due to a high radiation signal in the control room intake HVAC. The control room intake vent X/Q values are taken from WBNAPS3-104 (ref. 37) which are determined using ARCON96. The activities from STP are also used as input to the computer FENCDOSE (ref. 30) which determines the offsite dose.
This page replaced by Revision 016
1 l l Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:11
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Assumptions
- 1. There is no iodine "scrubbing" by the water in the steam generator when the steam bubbles (formed due to the flashing of the primary water) rise to the surface of the water.
Technical Justification: This is conservative because this increases the amount of iodine released. Since the break may be below the water line, there will actually be some amount of scrubbing (removal) of iodine.
- 2. The maximum reactor coolant activities allowed under WBN Technical Specifications (ref. 3) is assumed, with a distribution found in WBNNAL3-003 (ref. 29), which are the expected source terms from ANSI/ANS-18.1-1984 modified for WBN.
Technical Justification: The maximum concentration is mandated by NUREG-0800 (ref. 7). This assures maximum release of radioisotopes. See Attachment 2 for justification for using expected reactor coolant as the isotope distribution for establishing Technical Specification source terms.
- 3. The primary side to secondary side leakage is 150 gpd/steam generator, steady state.
Technical Justification: This is Technical Specification 3.4.13 (ref. 33)
- 4. The maximum letdown of 120 gpm + 4.39 gpm = 124.39 gpm (ref. 39, 41) is used.
Technical Justification: This value is used for calculation of iodine production/removal rates. This will maximize the removal rate of iodines from the primary coolant, and therefore will maximize the production rate of iodine (production = removal at steady state). See Calculation section for the formulas used. The letdown is assumed to be isolated at the beginning of the accident to maximize the reactor coolant inventories. The uncertainty of 4.39 gpm is determined in Appendix E.
- 5. The primary to secondary side leak rates and letdown flow rates are based on Standard Temperature and Pressure (STP) .
. Technical Justification: This is the method by which the plant measure leakage. Also, this will maximize the releases because the density is higher at STP, therefore more mass (and hence radioisotopes) will be released. For the letdown flow, this will increase the steady state iodine production rate, and therefore increase the iodine releases.
- 6. In the intact steam generators, the iodine partition factor is assumed to be 100. (see also assumption 12).
Technical Justification: The mass of primary to secondary leakage which occurs to the intact steam generators is small relative to the mass of secondary coolant. Therefore none of this leakage is assumed to flash and the release to the environment is through the steaming process. Reference 7 allows a partition factor of 100 for such cases.
- 7. In one case, a pre-accident iodine spike of 14µCi/gm1-131 equivalent is assumed at the start of the accident. In the other case, an accident initiated iodine spike of 500 increase in the iodine release rate from the fuel is assumed in the accident initiated case with the reactor coolant starting at 0.265µCi/gm1-131 equivalent.
Technical Justification: SRP 15.6.3 subsection 6a specifies the maximum allowable pre-accident spike of 14µCi/gm1-131 equivalent for a pre-accident spike. SRP 15.6.3 subsection 6b specifies that following an accident, the iodine release rate from the fuel to the reactor coolant is increased by a factor of 500.
- 8. The letdown demineralizer efficiency is assumed to be 1 for iodines.
Technical Justification: This will maximize iodine removal (=production) rate, and therefore result in larger iodine spiking.
- 9. The tritium inventory in the reactor coolant is assumed to be for the case with 2 TPBAR failures (124 µCi/gm, ref. 29).
Technical Justification: This will give an upper bound for the tritium. Also, a 2 TPBAR failure is considered to be an abnormal event. This will result in conservative doses. 10.It is assumed that there is no additional fuel damage due to the accident. Technical Justification: There is no expected extreme temperatures expected in the core due to the accident, therefore there will not be any fuel damage.
- 11. Only the Tritium Production Core (TPC) inventories are analyzed.
Technical Justification: Except for tritium, the reactor coolant inventories for the conventional and tritium production cores are the same. Therefore using the TPC with the additional tritium in the coolant will be conservative.
- 12. Water that boils in the faulted steam generator has an iodine partition factor of 100.
Technical Justification: Normally, to take into account uncovery of the faulted steam generator, there is no iodine partitioning in the release to the environment (iodine partition coefficient= 1). However, the water that boils is allowed a partition of 100. This is consistent with assumption 6. This page replaced by Revision 016
1 I l Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:12
Subject:
Control Room Operator and Offsite Doses Due to a Stearn Generator Tube Rupture
- 13. Only one train of CREVS is in operation. Normally, each CREVS train talces suction from separate intakes with no cross communication between trains. This leads to one contaminated train, and one uncontaminated train. The only way a 2 CREVS operation could result in higher doses would be for both trains to take suction from the same vent For this to happen, one intake path would require a failed closed intalce path AND a fail open of normally closed passive manual damper at the beginning of the accident An active failure of a train plus a failure of a passive component in less than 24 hours is beyond design basis.
Special ReqnirementslLimiting Conditions There are no special requirements or limiting conditions in this calculation. Calculations The following main text represents the replacement steam generator SGTR. Primary Coolant Activity Releases In NUREG-0800 R2 Chapter 15.6.3 (ret: 7), section ill.5 states "The reviewer assumes the primary and secondary coolant activity concentrations allowed by the technical specifications." Reference 3 ofNUREG-0800 states the following "The specific activity of the reactor coolant shall be limited to: a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT 1-131, and b. Less than or equal to 100/E microCuries per gram of gross activity." Given the above considerations, the isotopic spectrum found in WBNNAL3-003 (ref. 29) was examined. The 1-131 dose equivalent and 100/E values for this particular spectrum are determined in Tables 1 and 2. RG-1.109 (ref. 54) iodine inhalation dose conversion factors are used to calculate 1-131 equivalent conversion factor for the source term to be consistent with the Technical Specification 1.1 (ref. 55). Table 1: Determination ofl-131 Dose Equivalent for Primary Coolant Specific 1-131 DIA Activity equivalent mrem/Ci µCi/gm µCi/gm (ref. 54) (ref. 29) 1-131 l.49E+o9 4.77E-02 4.77E-02 1-132 l.43E+o7 2.25E-01 2.16E-03 1-133 2.69E+o8 l.49E-Ol 2.69E-02 1-134 3.73E+o6 3.64E-01 9.llE-04 1-135 5.60E+07 2.78E-Ol l.04E-02 total l.06E+o0 8.812E-02 inverse 11.348 This page replaced by Revision 016
Calculation No. WBNTSR008 J Rev: 014 J Plant: WBN I Page: 13
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Table 2: Determination of I 00/EBAR A(i) E(i) E(i) E(i) Activity Beta Energy Gamma Energy Total Isotope [uCi/gm] [MeV/dis] [MeV/dis] A(i)*E(i) Kr-85m l.71E-01 2.5290E-01 1.5862E-Ol 4.1152E-01 7.04E-02 Kr-85 2.66E-Ol 2.5060E-Ol 2.2102E-03 2.5281E-Ol 6.73E-02 Kr-87 l.61E-Ol l.3237E+OO 7.9284E-01 2.1165E+OO 3.40E-01 Kr-88 3.00E-01 3.7500E-01 l.9629E+OO 2.3379E+OO 7.0IE-01 Xe-13lm 6.54E-01 l.4280E-Ol 2.0058E-02 l.6286E-Ol l.06E-Ol Xe-133m 7.17E-02 1.8980E-01 4.1559E-02 2.3136E-Ol l.66E-02 Xe-133 2.53E+OO l.3540E-Ol 4.5385E-02 l.8079E-Ol 4.57E-Ol Xe-135m l.39E-Ol 9.5000E-02 4.3176E-01 5.2676E-01 7.35E-02 Xe-135 9.04E-Ol 3.1680E-01 2.4696E-Ol 5.6376E-Ol 5.1 OE-01 Br-84 l.72E-02 l.2842E+OO 1.6816E+OO 2.9658E+OO 5.09E-02 Rb-88 2.04E-Ol 2.0617E+OO 6.8631E-01 2.7480E+OO 5.60E-Ol Cs-134 7.39E-03 1.5690E-01 1.0361E+OO l.1930E+OO 8.82E-03 Cs-136 9.08E-04 1.0140E-Ol 2.1985E+OO 2.2999E+OO 2.09E-03 Cs-137 9.79E-03 1.8840E-Ol O.OOOOE+OO 1.8840E-Ol 1.84E-03 Na-24 4.99E-02 5.5460E-Ol 4.1216E+OO 4.6762E+OO 2.33E-Ol Cr-51 3.26E-03 3.7540E-03 3.2763E-02 3.6517E-02 l.19E-04 Mn-54 1.68E-03 4.1670E-03 8.3592E-Ol 8.4009E-01 l.41E-03 Fe-55 1.26E-03 4.1920E-03 1.5291E-03 5.721 IE-03 7.22E-06 Fe-59 3.16E-04 l.1800E-Ol l.1923E+OO 1.3103E+OO 4.14E-04 Co-58 4.84E-03 2.0490E-01 9.7586E-Ol 1.1808E+OO 5.72E-03 Co-60 5.58E-04 9.6840E-02 2.5043E+OO 2.6011E+OO 1.45E-03 Zn-65 5.37E-04 6.8940E-03 5.8169E-Ol 5.8858E-Ol 3.16E-04 Sr-89 1.47E-04 5.7300E-Ol 1.3636E-04 5.7314E-01 8.44E-05 Sr-90 l.26E-05 l.9630E-01 O.OOOOE+OO 1.9630E-Ol 2.48E-06 Sr-91 l.02E-03 6.5050E-01 6.9508E-01 l.3456E+OO l.37E-03 Y-90 l.26E-05 9.3610E-Ol O.OOOOE+OO 9.3610E-01 1.18E-05 Y-9lm 4.93E-04 O.OOOOE+OO 5.5557E-Ol 5.5557E-01 2.74E-04 Y-91 5.47E-06 6.0600E-01 3.6147E-03 6.0961E-Ol 3.34E-06 Y-93 4.46E-03 l.1721E+OO 8.9414E-02 1.2615E+OO 5.63E-03 Zr-95 4.lOE-04 1.1990E-Ol 7.3474E-01 8.5464E-Ol 3.51E-04 Nb-95 2.95E-04 4.4970E-02 7.6430E-01 8.0927E-01 2.38E-04 Mo-99 6.75E-03 3.9570E-Ol l.6238E-Ol 5.5808E-Ol 3.77E-03 Tc-99m 5.0IE-03 4.8500E-03 l.4263E-Ol 1.4748E-01 7.38E-04 Ru-103 7.89E-03 6.7400E-02 4.8394E-Ol 5.5134E-Ol 4.35E-03 Ru-106 9.47E-02 l.OlOOE-02 O.OOOOE+OO l.OlOOE-02 9.57E-04 Rh-103m 7.89E-03 3.4620E-02 2.2148E-05 3.4642E-02 2.73E-04 Rh-106 9.47E-02 7.0960E-Ol 2.0348E-Ol 9.1308E-Ol 8.65E-02 Te-129m 2.00E-04 l.9150E-Ol 9.4832E-02 2.8633E-01 5.73E-05 Te-129 2.57E-02 5.2260E-01 5.9948E-02 5.8255E-Ol l.50E-02 Te-131m l.59E-03 2.1240E-Ol l.4092E+OO l.6216E+OO 2.57E-03
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 14
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Table 2: Determination of 100/EBAR - continued A(i) E(i) E(i) E(i) Activity Beta Energy Gamma Energy Total Isotope [uCi/gm] [MeV/dis] [MeV/dis] A(i)*E(i) Te-131 8.26E-03 7.5970E-Ol 4.1616E-OI l.1759E+OO 9.71E-03 Te-132 l.79E-03 l.0020E-OI 2.0507E-Ol 3.0527E-01 5.47E-04 Ba-137m 9.79E-03 6.4260E-02 5.9729E-OI 6.6155E-01 6.48E-03 Ba-140 l.37E-02 3.1500E-Ol 1.9522E-Ol 5.1022E-Ol 6.98E-03 La-140 2.64E-02 5.4050E-Ol 2.3074E+OO 2.8479E+OO 7.52E-02 Ce-141 l.58E-04 l.6930E-Ol 1.0181E-Ol 2.711 lE-Ol 4.28E-05 Ce-143 2.96E-03 3.8420E-Ol 3.4335E-Ol 7.2755E-Ol 2.15E-03 Ce-144 4.21E-03 9.1300E-02 3.2865E-02 l.2417E-OI 5.23E-04 Pr-143 2.96E-03 3.1430E-Ol O.OOOOE+OO 3.1430E-01 9.30E-04 Pr-144 4.21E-03 l.2258E+OO 3.lOlOE-02 1.2568E+OO 5.29E-03 Np-239 2.32E-03 1.2380E-Ol 2.0845E+OO 2.2083E+OO 5.13E-03 Total 5.82E+OO 3.44E+OO EBAR 5.91E-Ol RCS Specific Activity Limit 169.14 The DIA values (mrem/Curie) in Table 1 were obtained from reference 54 (p. 45) for each of the iodine isotopes of interest. The 1-131 dose equivalence is calculated as follows: As can be seen in Table 1, the resulting 1-13 I dose equivalency for the expected spectrum is 0.088 I 2 µCi/gm. The definition of EBAR or E is as follows: "E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant." The values for Ei in Table 2 were obtained from reference I 7 and the values for Ai are from WBNNAL3-003. The value of E is determined as follows: The value for E calculated in Table 2 is 0.591 MeV/dis. This results in a non-iodine specific activity limit (100/E) of 169.14 µCi/gm. The total specific activity of the expected coolant is 5.82 µCi/gm. Therefore, the values for noble gases in the design reactor coolant given in reference 29 will have to be increased by a factor of 169.14/5.82 = 29.06 and the values for iodines will have to be increased by a factor of 1/0.08812 = 11.348.
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 15
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 For the secondary side concentrations from WBNNAL3-003, the same procedure is performed to determine the 1-131 equivalence: DIA µCi/gm secondary 1-131 equivalent mrem/Ci side, water ANSI 18.l µCi/gm 1-131 l.49E+09 1.41E-06 l.41E-06 1-132 l.43E+07 3.37E-06 3.23E-08 1-133 2.69E+08 4.03E-06 7.28E-07 1-134 3.73E+06 2.93E-06 7.33E-09 1-135 5.60E+07 6.19E-06 2.33E-07 total l.79E-05 2.410E-06 inverse 4.150E+05 To convert to 1-131 equivalence, the secondary side 1-131 equivalent conversion factor is (l/2.410E-6) = 4.150E5 gm/µCi. Note that this factor has been developed for iodines. There is no limit on noble gases in the secondary side as there is for the primary side (100/Ebar). However, to maintain the proper ratio of isotopes, and for conservatism, the iodine factor will also be applied to the noble gases. Note: the secondary side water does not contain any noble gases. For conservatism, the noble gas inventory is the inventory from the secondary side steam.
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 16
Subject:
Control Room Operator and Offsite Doses Due to a Steam IPrepared: MCB Date: 9-2-2011 Generator Tube Rupture IChecked: JEB Date: 9-2-2011 The STP models consist of a pre-accident iodine spike (see figure 1) model and an accident initiated iodine spike model (see figure 2). The model(s) consist of the following: Volumes:
#1: Reactor Coolant: 5.78E5 lb (ref. 29) = 2.622E8 gm
#2: Steam Generator w/Leak: 5.31E7 gm (ref. 40)
#3: Steam Generators w/out Leak: l .593E8 gm (ref. 40)
#4: Environment: 1 gm (arbitrary) (This volume is made into an accumulator through the "A" card to suppress radioactive decay)
The step sources to initialize the reactor coolant and the secondary side activities are (in units of Ci/(µCi/gm)): S=2.622E8 gm x IE-6 Ci/µCi = 2.622E2 (tritium) S=2.622E8 gm x IE-6 Ci/µCi x 29.06 = 7.620E3 (noble gases) Pre-accident iodine spike case (initial concentration= 14 µCi/gm): S=2.622E8 gm x IE-6 Ci/µCi x 11.348 [µCi/gm 1-13 Ir 1 x 14 µCi/gm= 4. l l 6E4 (iodines) Accident initiated iodine spike case (initial concentration= 0.265 µCi/gm): S=2.622E8 gm x IE-6 Ci/µCi x 11.348 [µCi/gm 1-13 Ir 1 x 0.265 µCi/gm= 7 .885E2 (iodines) Secondary side, all cases, steam generator with leak (initial concentration= 0.1 µCi/gm): S = 5.31E7 gm x IE-6 Ci/µCi = 5.31El (tritium) S = 5.31E7 gm x IE-6 Ci/µCi x 4.150E5 [µCi/gm l-13Ir 1 x0.1µCi/gm=2.203E6 (noble gases, iodines) Secondary side, all cases, steam generators without leak (initial concentration= 0.1 µCi/gm): S = l.593E8 gm x IE-6 Ci/µCi = 1.593E2 (tritium) S = l.593E8 gm x IE-6 Ci/µCi x 4.150E5 [µCi/gm 1-13 U 1 x 0.1 µCi/gm= 6.610E6 (noble gases, iodines) Continuous Sources: For the accident initiated iodine spike case, the iodine spike is 500 times the iodine release rate from the fuel. At steady state conditions, the iodine release (production) rate is equal to the removal rate. The iodine removal is due to a) radioactive decay, b) removal by the letdown system, and c) removal through leakage to the secondary side. These terms are expressed as: P = Lremoval rates = decay+ letdown+ leakage or P = /... + fLfiV + p,N where P =production rate [k 1]
/...=decay constant for the isotope in question [hr- 1] = ln(2)/T 112 fL =letdown flow rate= 120 gpm + 4.39 = 124.39 gpm f =letdown demineralizer efficiency= 1 (assumed so as to maximize removal/production rate)
V =volume of primary coolant= 5.78E5 lb Ps = removal rate of iodine from primary side due to leakage = 11 gpm (= I 0 gpm identified plus 1 gpm unidentified leakage) T 112 =half-life taken from ref. 42 Note: All flow rates are converted to mass flow rates at STP (H20 = 1 gm/cc). Removal rate of iodine from primary side to secondary side was not considered above, because of its relatively low rate (3 x 150 gpd = 0.3 gpm) compared with other terms.
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 17
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Production/Removal Rates for 11 gpm Leakage (=10known+1 unknown) Half Life A, [1/hr] fLEN[l/hr] p,N [I/hr] Prod rate P 500xP I-131 8.04 d 3.59E-03 l.08E-01 9.53E-03 0.1209 60.43 I-132 2.28h 3.04E-Ol l .08E-01 9.53E-03 0.4213 210.64 I-133 20.9h 3.32E-02 1.08E-01 9.53E-03 0.1504 75.22 I-134 52.6m 7.91E-Ol l .08E-Ol 9.53E-03 0.9079 453.97 I-135 6.61 h l.05E-OI l.08E-Ol 9.53E-03 0.2221 111.07 The accident initiated iodine spike of 500 times the increase in the iodine release (production) rate from the fuel is modeled as a continuous source: C =Volume x lE-6 Ci/µCi x Prod Rate x 500 x I µCi/gm I-131 equivalent conversion factor where Volume= 2.622E8 gm Prod Rate = see table above I-131 equiv.= 0.265 µCi/gm I-131 equivalent l µCi/gm I-131 equivalent conversion factor= 11.348 (value determined above, this is to get the ANSUANS-18.1-1984 source into I µCi/gmI-131 equivalent Continuous Source [l/hr] for Accident Initiated Iodine Spike: 11 gpm leak (10 known+lgpm unknown) 0.265 µCi/gm I-131 4.765E+04 I-132 l.661E+05 I-133 5.931E+04 I-134 3.580E+05 I-135 8.758E+04
- ~
11111 I .. l Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 18
Subject:
Control Room Operator and Offslte Doses Due to a Steam Generator Tube Rupture The following table presents the variables that change for each case: Step Source Continuous Case S for iodine SourceC Description A 3.571E4 Not Applicable 14 µCi/gm.1-131 B 7.885E+02 Table above 0.265 µCi/gm 1-131, 500 spiking, 11 gpm leak (10 known+ 1 unknown) Flow Rates: The following is for the replacement steam generators. The amount of secondary side steam released from the ruptured steam generator is 108,200 Ihm from 0-2 hours and 35,500 Ihm from 2-8 hours (ref. 45). The amount of secondary side steam released from the intact steam generators is 539,500 Ihm from 0-2 hours and 925,000 Ihm from 2-8 hours (ref. 45, note that ref. 45b, gives this values as 924,400 Ihm. This is 0.06% less than ref. 45a. It is conservative to use the higher value). The reactor coolant release to the steam generator was a total of 166,200 lb, of which 9189 lb flashed (ref. 45, 46). To account for the release during the 74.0 second interval when the control room is not isolated, the amount of reactor coolant released at 74.0 sec is needed. However, the release from the steam generators does not actually start until 176 sec post accident Therefore, the releases at 176+ 74.0 = 250.0 sec are actually needed for release calculations. Using the releases from reference 46, the reactor coolant release at 250 sec is 12051.425 Ihm (calculated using Break Flow in in ref. 46). The amount of reactor coolant that flashed at 250 sec is 1447.533 Ihm (calculated using "Integrated Flashing Break Flow in Attachment 1 in ref. 46). The mass release rate from the ruptured steam generator is non-linear. However since the time frame for the release is short (74.0 sec), the average release rate can be used. From reference 45, the flashing of the reactor coolant stops at 2208.5 sec, and the break flow stops at 4670 sec. The following flow rates/leakage rates for each component are: Flow from Reactor Coolant #1 to Steam Generator Faulted #2 (non-flashed): 0-250 sec: F = (12051.425 lb- 1447.533 lb) x (3600 sec/hr) I (250 sec)= 1.527E5 lb/hr= 6.926E7 gm/hr 250 ~ 4670 sec: F = [(166,200 lb - 12051.425 lb) - (9189lb - 1447.533 lb)] I (4670 - 250 sec)= 33.124 lb/sec= 5.409E7 gm/hr 467o+ sec: F = 0 gm/hr Flow from Reactor Coolant #1 to Environment #4 (flashed): 176-250 sec: F = (1447.533 lb) x (3600 sec/hr) I (74 sec)= 7.042E4 lb/hr= 3.194E7 gm/hr 250- 2208.5sec: F = (9189 lb - 1447.533 lb) I (2208.5 - 250 sec)= 3.953 lb/sec= 6.455E6 gm/hr 2208.5+ sec: F = 0 gm/hr Flow from Steam Generator Faulted #2 to Environment #4: 176 sec - 2 hr: F = 108,200 lb I (2hr- 176sec I 3600sec/hr) = 5.546E4 lb/hr= 2.515E7 gm/hr (noble gas and tritium) F = 0.01 x (108,200 lb) I (2hr- 176sec I 3600sec/hr) = 5.546E2 lb/hr= 2.515E5gm /hr (iodine) (see note*) 2 - 8 hr: F = 35500 lb I (8hr- 2hr) = 5916.67 lb/hr= 2.684E6 gm/hr (noble gas) F = 0.01 x 35,500 lb I (8hr- 2hr) = 59.1667 lb/hr= 2.684E4 gm/hr (iodine) note* Normally, to take into account uncovery of the faulted steam generator, there is no iodine partitioning in the release to the environment (iodine partition coefficient= 1). For conservatism, no iodine scrubbing of the bubbles in the flashed water is taken into account. However, the water that boils is allowed the iodine partition of 100 (see assumption 6). This page replaced by Revision 016
I l Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:19
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Flow from Steam Generator Unfaulted #3 to Environment #4: 176 sec - 2 hr: F = 539,500 lb I (2hr - l 76sec/3600sec/hr) = 276509 lb/hr= l .254E8 gm/hr (noble gas) F = O.Ql x 539,500 lb I (2hr - l 76sec/3600sec/hr) = 2765.09 lb/hr= I.254E6 gm/hr (iodine) 2 - 8 hr: F = 925,000 lb /(8hr - 2hr) = 1.542E5 lb/hr= 6.993E7 gm/hr (noble gas) F = O.Ql x 925,000 lb I (8hr - 2hr) = 1.542E3 lb/hr= 6.993E5 gm/hr (iodine) Flow from Reactor Coolant# 1 to Steam Generator Unfaulted #3: 0 - 8 hr: F = 3 steam generators x 150 gpd x 3785.48 cc/gal I 24 hr/day x lgm/cc = 7.098E4 gm/hr The STP output is used as input to COROD (which determines control room operator dose) and FENCDOSE (which determines 30-day and 2-hour LPZ offsite dose). Control Room Dose With the exception of the source activities and X/Q's, all of the input and assumptions used in TI-RPS-198 (ref. 13) to calculate the control room operator dose are considered valid for this calculation. The X/Q values are taken from reference 37. Maintenance Request MR-482000 (Attachment 2) gives measured closure times for several flow control valves in the control building ventilation system as measured on 12/8/88. Examination of reference 14 in conjunction with MR-482000 revealed that the worst case involved valves O-FCV-31-3, -4 with closure times of 12.43 sec and 13.15 sec respectively. Therefore, it was conservatively assumed, and per reference 35, that these valves would be full open for 14 sec following the SGTR (This is conservative since in actuality, as the valve closes, the flow decreases). Instrument response time is taken from WBNTSR028 (ref. 38) 60 sec. This leads to a total unisolated control room time of74 seconds. During this time the intake flow is 3200 cfin (reference 14). No filtration is provided for this stream. Offsite Dose The same source terms used in the COROD run are used in the FENCDOSE run. The base FENCDOSE model comes from TI-RPS-197 (ref 34). Some pertinent information from the COROD and FENCDOSE models used in this analysis are (from ref. 34) with the control roomX/Q values from ref. 37: 30-day LPZ Offsite X/Q values [sec/m3]: 1.784E-4 0-2hr, 8.835E-5 2-8 hr, 6.217E-5 8-24 hr, 2.900E-5 5 1-4 day, 9.81 lE-64-30 day 2-hr EAB X/Q values: 6.382E-4 Ul Accident Control Room ARCON96 XJQ [sec/m3]: 3.85E-03 0-2hr, 3.22-03 2-8hr Control Room volume: 257198 cuft Control Room makeup/pressurization flow: 711 cfin, 3200 cfin prior to isolation (ref. 44*) Control Room total flow: 3600 cfin Control Room recirculation flow: 2889, for normal operation (unisolated) each pass is at 3200 cfin (same as intake flow) Control Room unfiltered intake: 51 cfin Control Room filter efficiency: 95% first pass, 70% second pass for iodine, 0% for everything else Control Room occupancy factors: 100% 0-24 hr, 60% 1-4 days, 40% 4-30 days ICRP-30 dose conversion factors, as well as TEDE
- 3200 cfin has been deleted from 1-47W866-4 R39 (re£ 14), and has been measured to be approximately 2500 cfin (O-SI-31-31-A). The value comes from 1-47W866-4 R20 (ref. 44). The 3200 cfin will be retained in this calculation revision since this value produces conservative results.
This page replaced by Revision 016
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 20
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Figure 1: STP Model Pre-accident Iodine Spike Steo Source 1 Reactor 14 µCi/g 1-131 Coolant 3*150 gpd=450 gpd Step Source Step Source 2 Steam 3Steam ~ Generator Generator ~~- 1 uCi/ 1_131 0.1uCi/g1-13 9 Faulted Unfaulted
~ Environment
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 21
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Figure 2: STP Model Accident Initiated Iodine Spike Step Source _ 0.265 uCi/g 1-131
~ Reacta.1 Coolant 1
., Continuous Source
=500*1odine Production Rate I
i 3*150 gpd=450 gpd Step Source Step Source 2 Steam 3 Steam Generator Generator 0.1uCi/g1-131 0.1 uCi/g 1-131 Faulted Unfaulted 14 Environment !
l I Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:22
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Results The following Unit 1 doses were calculated for the Tritium Production Core (the Unit 2 results are found in Am>1endix G): Control Room Ooerator Dose (rem) Limit (rem) Pre-Accident Iodine Accident Initiated Iodine Gamma 8.64E-02 8.18E-02 5 Beta 9.62E-01 9.45E-01 30 Thyroid (ICRP-30) 2.29E+01 3.61E+OO 30 TEDE 1.28E+OO 6.50E-01 5 Pre-Accident Iodine Spike Offsite Dose (rem) Limit (rem) 2-hr EAB 30-day LPZ Gamma 3.50E-01 1.03E-01 25 Beta 2.04E-01 6.25E-02 300 Thyroid (ICRP-30) 1.33E+01 3.81E+OO 300 TEDE 1.22E+OO 3.52E-01 25 Accident Initiated Iodine Spike Offsite Dose (rem) Limit (rem) 2-hr EAB 30-dayLPZ Gamma 5.05E-01 1.48E-01 2.5 Beta 2.35E-01 7.19E-02 30 Thyroid (ICRP-30) 6.37E+OO 1.87E+OO 30 TEDE 1.08E+OO 3.14E-01 2.5 The following Unit 1 margins were calculated from the doses in the Table above. Where: margin = limit - dose, and percent = 100 x (limit-dose)!limit Control Room Operator Dose Maroins Pre-Accident Iodine Accident Initiated Iodine Limit (rem) Maroin (rem Percent Maroin (rem Percent Gamma 4.9 98.3% 4.9 98.4% 5 Beta 29.0 96.8% 29.1 96.9% 30 Thyroid (ICRP-30) 7.1 23.7% 26.4 88.0% 30 TEDE 3.7 74.3% 4.3 87.0% 5 Pre-Accident Iodine Soike Offsite Dose Maroins 2hrEAB 30 dav LPZ Limit (rem) Maroin (rem Percent Margin (rem Percent Gamma 24.7 98.6% 24.9 99.6% 25 Beta 299.8 99.9% 299.9 100.0% 300 Thyroid (ICRP-30) 286.7 95.6% 296.2 98.7% 300 TEDE 23.8 95.1% 24.6 98.6% 25 Accident Initiated Iodine Soike Offsite Dose Margins 2hr EAB 30 daf LPZ Limit (rem) Maroin (rem Percent Margin (rem Percent Gamma 2.0 79.8% 2.4 94.1% 2.5 Beta 29.8 99.2% 29.9 99.8% 30 Thyroid (ICRP-30) 23.6 78.8% 28.1 93.8% 30 TEDE 1.4 56.9% 2.2 87.5% 2.5 This page replaced by Revision 016
, i' i Calculation sheet Document: WBNTSR008 I Rev.: 018 I Plant: WBN I Units 1,2 I Page:23
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Conclusion The offsite doses (gamma, beta, thyroid and TEDE) due to a SGTR with a preexisting iodine spike do not exceed the 10CFRlOO limits (25 rem gamma (whole body), 300 rem beta (skin), and 300 rem thyroid per NUREG-0800). The SGTR with accident initiated iodine spike does not exceed a small fraction of the 10CFRlOO limits (10% of the 10CFRlOO limits of 25 rem gamma, 300 rem beta, and 300 rem thyroid per NUREG-0800). The control room doses due to a SGTR do not exceed the 10CFRSO App. A GDC 19 limits (5 rem gamma, 30 rem beta and 30 rem thyroid). The reactor coolant parameters for this conclusion is based on a pre-accident spike of 14 µCi/gm I-131 equivalent and an accident initiated iodine spike with the initial activity at 0.265 µCi/gm I-131 equivalent. The secondary side activity is 0.1 µCi/gm I-131 equivalent. The primary to secondary leak rate (prior to the accident) is a maximum of 150 gpd leak per steam generator. The other RCS operational leakage rate is 11 gpm (10 gpm identified plus 1 gpm unidentified). Note on methodologies used: This calculation determined the offsite and control room operator doses using the RG-1.109 iodine inhalation dose conversion factors to determine the I-131 equivalent conversion factor for the source term. This I-131 equivalence is consistent with the Technical Specification 1.1 (ref. 55). The thyroid dose is reported only based on the ICRP-30 methodology. The TEDE (Total Effective Dose Equivalent) is calculated from the computer codes, COROD and FENCDOSE using parameter derived from the ICRP-30 data. The effect of the increased time delay had been evaluated in this calculation, and it has been shown that adequate margin remains between the calculated doses and regulatory limits to accommodate instrument errors and uncertainties. References
- 1. Westinghouse letter TVA-87-895 from T.A Lordi to H.L. Abercrombie dated 12/28/87, "Increased Radioactivity Release to the Environment Following Reactor Trip" RIMS# SOO 880104 001
- 2a. TI-RPS-14 RO "Containment Activity as a Function of Time for Various Combinations of Primary System Leak Rate and Fuel Damage" p.10-13 RIMS# NEB 810212 326 2b. "Study of Reactor Shutdown Radioactivity 'Spiking' at Three Mile Island Nuclear Power Station During February 20-21, 1976" J.E.Cline and E.D.Barefoot, July 1976
- 3. WBN Unit 1 Technical Specification 3.4.16 "Reactor Coolant System - Specific Activity Amendment 81
- 4. WBN FSAR Table 11.1-2 Amendment 62 (transmitted to TVA by Westinghouse letterWAT-D-2139, March 8, 1976
- 5. WBN FSAR section 15.4.3 Amendment 62 (not used as design input)
- 6. WBN FSAR Table 15.5-18 (Attachment 3) Amendment 62 (not used as design input)
- 7. U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Standard Review Plan, NUREG-0800 R2, Part 15.6.3 "Radiological Consequences of Steam Generator Tube Failure (PWR)" July 1981
- 8. TIO 14844 "Calculation of Distance Factors for Power and Test Reactor Sites", March 23, 1962
- 9. deletedinR7
- 10. N3-68-4001 R3 "System Description for the Reactor Coolant System" RIMS# T29 930225 855
- 11. "Thermodynamic Properties of Steam," 1st Edition, Keenan and Keyes, pp.33, 39, 74-75
- 12. TI-RPS-156 RO "Effect ofz.ero Steam Generator Blowdown on Offsite Dose During Various Events" RIMS# B45 850711235
- 13. TI-RPS-198 R24 "Dose to Control Room Personnel Due to a Regulatory Guide 1.4 Loss of Coolant Accident"
- 14. WBN CCD drawing 1-47W866-4 R39
- 15. Computer code COROD R7 .1, code I.D.262347 (under dose code program mgr version 1.1)
- 16. Regulatory Guide 1.4 R2 "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors"
- 17. NIB Isotope Library, found in GENAPS3-018 RI "NEB Isotope Library Verification"
- 18. WBN drawing 41N712-1 RD
- 19. WBN drawing 41N718-1 RE
- 20. WBN drawing47W415-l RH
- 21. WBN drawing 47W930-2 RP This page replaced by Revision 016
I l Calculation sheet Document: WBNTSROOB IRev.: 016 I Plant: WBN I Units 1,2 I Page:24
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture
- 22. WBN drawing 47W930-3 RP
- 23. WBN drawing 47W930-5 RE
- 24. WBN drawing 47W200-1 Rll
- 25. Halitsky, James et.al, "Wind Tunnel Tests of Gas Diffusion From a Leak in the Shell of a Nuclear Power Reactor and from a Nearby Stack" Department of Meteorology and Oceanography Geophysical Sciences Laboratory Report No.63-2, New York University, April I, 1963
- 26. deleted in R4
- 27. deleted R6
- 28. W AT-D-10336 February 27, 1997 "Draft Data Request for SGTR and SLB Events" RIMS# T25 970306 824
- 29. WBNNAL3003 R5 "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1984"
- 30. Computer code FENCDOSE R5, code I.D.262358 (under dose code program mgr version 1.1)
- 31. Computer code STP R7.l, code I.D.262165 (under dose code program mgr version 1.1)
- 32. Computer code PARINT RI, code I.D.262350
- 33. Technical Specification 3.4.13, Amendment 31
- 34. TI-RPS-197 R22 "Offsite Doses Due to a Regulatory Guide 1.4 Loss of Coolant Accident"
- 35. N3-30CB-4002 R6 "Control Building Heating, Ventilating, Air Conditioning, and Air Cleanup System"
- 36. WBN PER 01-000080-000
- 37. WBNAPS3-104 R3 "WBN Control Room X/Q"
- 38. WBNTSR028 RIO, Main Control Room Emergency and Normal Air Intake Monitors Required Range, Safety Limits, Response Time and Accuracy
- 39. N3-62-4001 R5 "System Description for Chemical and Volume Control System"
- 40. WBNAPS3-053 R3 "Steam Generator Leakage Detection with the Condenser Vacuum Pump Air Exhaust Monitor (l,2-RM-90-119)" RIMS# B45 880620 238
- 41. NSAI.r00-004 "Nonconservatisms in Iodine Spiking Calculations"
- 42. Lederer and Shirley, "Table oflsotopes" seventh ed.
- 43. WBNNAL3-002 R2 "100-Day LOCA-DBA Source Terms for the EGTS and ABGTS Filters, Containment, Sump, and Shield Building Annulus" Note: this calculation is cwrently at R3, however the information is found in R2.
- 44. 1-47W866-4 R20 45a. WCAP-16286-P "Watts Bar Unit 1 Replacement Steam Generator Program NSSS Engineering Report", Jan.2005 45b. WCAP-16286-P RI "Watts Bar Unit 1 Replacement Steam Generator Program NSSS Engineering Report" Sept. 2005
- 46. Westinghouse letter WTV-RSG-05-100 dated May 31, 2005 "Submittal of Steam Generator Tube Rupture Dose Analysis Input from S. Radom.sky to Paul G. Trudel 47a. WBT-D-1015 "Steam Generator Tube Rupture Input to Dose Mass Transfer Data
47b. LTR-CRA-09-153 RI Watts Bar unit 2 Steam Generator Tube Rupture Input to Dose Mass Transfer Data for the Completion Project"
- 48. Unit 2 TS 3.4.13 Rev. A (developmental) "RCS Operational Leakage"
- 49. Unit 2 TS 3.4.16 Rev. A (developmental) "RCS Specific Activity"
- 50. Unit 2 TS 3.4.17 Rev. A (developmental) "Steam Generator (SG) Integrity
- 51. WBNAPS3-053 R2 "Steam Generator Leakage Detection with the Condenser Vacuum Pump Air Exhaust Monitor (1,2-RM-9()..119)" note: this revision is out of date, however the pertinent data, the mass of water in a SG is relevant
- 52. PER 327956
- 53. PER 327968
- 54. Regulatory Guide 1.109 RI
- 55. WBN Technical Specification 1.1 Amendment 81
- 56. PER213412
- 57. DCN 61599
- 58. PER 766429
*This page replaced by Revision 016
l l Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:25
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Appendix A: Example ofSTP Model (pre-accident Iodine spike) II TSR008Rl6SlA JOB 25402-012-572 II ORG.. DWWU.CAG.BECHTEL II EXEC STP7.l; SOUT='*' II Unit 1 SGTR Pre-accident Iodine Spike Case llGO.FTOl NV= 4 MS= 2 I IGO.FTll $ CLASS DESCRIPTION $ 1 NOBLE GASES $ 2 IODINE $ 3 TRITIUM NI= 23 NK= 3 NG= 0 NL= 3 lKRM 83 1 l.0352E-04 10.0 10.0 10.0 2KRM 85 1 4.2978E-05 10.0 10.0 10.0 3KR 8S 1 2.0470E-09 29.8849E-06 10.0 10.0 4KR 87 1 l.5141E-04 10.0 10.0 10.0 5KR 88 1 6.8765E-OS 10.0 10.0 10.0 6KR 89 1 3.6328E-03 10.0 10.0 10.0 7XEM 131 1 6.7414E-07 131.3039E-08 181.3039E-08 10.0 8XEM 133 1 3.56S6E-06 1S2.036SE-07 202.036SE-07 10.0 9XE 133 1 1.Sl6SE-06 83.S6S6E-06 1S9.0S31E-06 209.0S31E-06 lOXEM 13S 1 7.3818E-04 174.8062E-06 224.8062E-06 10.0 llXE 13S 1 2.1043E-OS 107.3818E-04 172. 4322E-OS 222.4322E-OS 12XE 138 1 8;1S28E-04 10.0 10.0 10.0 13I 131 2 9. 9536E-07 10.0 10.0 10.0 14I 132 2 8.4448E-0S 10.0 10.0 10.0 15I 133 2 9.2S68E-06 10.0 10.0 10.0 16I 134 2 2.1963E-04 10.0 10.0 10.0 17I 13S 2 2.9129E-OS 10.0 10.0 10.0 18I* 131 2 9.9S36E-07 10.0 10.0 10.0 19I* 132 2 8.4448E-OS 10.0 10.0 10.0 20I* 133 2 9.2S68E-06 10.0 10.0 10.0 21I* 134 2 2.1963E-04 10.0 10.0 10.0 22I* 135 2 2.9129E-OS 10.0 10.0 10.0 23H 3 3 1. 778SE-09 10.0000E+OO 10.0000E+OO 10.0000E+OO llGO.FT2l 1 'REACTOR COOLANT ANSIANSI-18.1-1984 UCIIGM, WBNNAL3003 RS' 1 0.0 2 l.71E-l 3 2.66E-l 4 l.61E-l S 3.00E-1 6 0.0 7 6.S4E-l 8 7.17E-2 9 2.53EO 10 l.39E-l 11 9.04E-l 12 l.29E~l 13 4.77E-2 14 2.2SE-l lS l.49E-l 16 3.64E-l 17 2.78E-l 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E2 0 2 'SECONDARY COOL ANSIANSI-18.1-1984 UCIIGM, WBNNAL3003 RS' 1 0.0 2 3.63E-8 3 S.SlE-8 4 3.22E-8 S 6.31E-8 6 0.0 7 l.34E-7 8 l.S4E-8 9 S.2SE-7 10 2.90E-8 11 l.91E-7 12 2.68E-8 13 l.41E-6 14 3.37E-6 lS 4.03E-6 16 2.93E-6 17 6.19E-6 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E-l 0 T STEAM GENERATER TUBE RUPTURE ACCIDENT NJ= 4 1 'REACTOR COOLANT' 2 'STEAM GEN FAULTED' 3 'STM GEN UNFAULTED' 4 'ENVIRONMENT' -1 INITIAL ACTIVITY V 1 2.622E8 GM V 2 S.310E7 GM V 3 1. S93E8 GM v 4 1.0 S 1 1 3 2.622E2 S 2 2 0 2.203E6 This page replaced by Revision 016
... L l Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:26
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture S 2 3 0 6.610E6 S 2 2 3 5.310El S 2 3 3 l.593E2 S 1 1 1 7.620E3 S 1 1 2 4.166E4 F 1 2 0 6.926E7 F 1 3 0 7.098E4 A 4 176 SEC TIME TO 176 SEC 250.0 SEC TIME TO 250.0 SEC F 1 4 0 3.194E7 F 2 4 0 2.515E7 F 2 4 2 2.515E5 F 3 4 0 1. 254E8 F 3 4 2 1.254E6 N 4 0 p 1 0 4 2208.5 SEC TIME TO 2208.5 SEC F 1 2 0 5.409E7 F 1 4 0 6.455E6 4670 SEC TIME TO 4670 SEC F 1 4 0 0.0 2 HR TIME TO 2 HOUR F 1 2 0 0.0 N 4 0 p 1 0 4 8 HR TIME TO 8 HOUR F 2 4 0 2.684E6 F 2 4 2 2.684E4 F 3 4 0 6.993E7 F 3 4 2 6.993E5 N 4 0 p 1 0 4 T T /* This page replaced by Revision 016
I i ii Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:27
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Appendix B: Exam.pie of STP Model (Accident Initiated Iodine Spike) II TSR008Rl6SlB JOB 25402-012-572 II ORG=DWWU.CAG.BECHTEL II EXEC STP7.l; SOUT='*' II Unit 1 SGTR Accident Initiated Iodine Spike Case llGO.F'l'Ol NV= 4 MS= 2 I IGO.F'l'll $ CLASS DESCRIPTION $ 1 NOBLE GASES $ 2 IODINE $ 3 TRITIUM NI= 23 NK= 7 NG= 0 NL= 3 lKRM 83 1 l.0352E-04 10.0 10.0 10.0 2KRM 85 1 4.2978E-05 10.0 10.0 10.0 3KR 85 1 2.0470E-09 29.8849E-06 10.0 10.0 4KR 87 1 l.5141E-04 10.0 10.0 10.0 5KR 88 1 6.8765E-05 10.0 10.0 10.0 6KR 89 1 3.6328E-03 10.0 10.0 10.0 7XEM 131 1 6.7414E-07 131.3039E-08 181.3039E-08 10.0 8XEM 133 1 3.5656E-06 152.0365E-07 202.0365E-07 10.0 9XE 133 1 1. 5165E-06 83.5656E-06 159.0531E-06 209.0531E-06 lOXEM 135 1 7.3818E-04 174.8062E-06 224.8062E-06 10.0 llXE 135 1 2.1043E-05 107.3818E-04 172. 4322E-05 222.4322E-05 12XE 138 1 8.1528E-04 10.0 10.0 10.0 13I 131 2 9.9536E-07 10.0 10.0 10.0 14I 132 4 8.4448E-05 10.0 10.0 10.0 lSI 133 5 9.2568E-06 10.0 10.0 10.0 16I 134 6 2.1963E-04 10.0 10.0 10.0 l 7I 135 7 2.9129E-05 10.0 10.0 10.0 18I* 131 2 9.9536E-07 10.0 10.0 10.0 19I* 132 4 8.4448E-05 10.0 10.0 10.0 20I* 133 5 9.2568E-06 iO.O 10.0 10.0 21I* 134 6 2.1963E-04 10.0 10.0 10.0 22I* 135 7 2.9129E-05 10.0 10.0 10.0 23H 3 3 1. 7785E-09 10.0000E+OO 10.0000E+OO 10.0000E+OO llGO.FT2l 1 'REACTOR COOLANT ANSIANSI-18.1-1984 UCIIGM, WBNNAL3003 RS' 1 0.0 2 l.71E-l 3 2.66E-l 4 l.61E-1 5 3.00E-1 6 0.0 7 6.54E-l 8 7.17E-2 9 2.53EO 10 l.39E-l 11 9.04E-l 12 l.29E-l 13 4.77E-2 14 2.25E-l 15 l.49E-l 16 3.64E-l 17 2.78E-l 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E2 0 2 'SECONDARY COOL ANSIANSI-18.1-1984 UCIIGM, WBNNAL3003 RS' 1 0.0 2 3.63E-8 3 5.51E-8 4 3.22E-8 5 6.31E-8 6 0.0 7 l.34E-7 8 l.54E-8 9 5.25E-7 10 2.90E-8 11 l.91E-7 12 2.68E-8 13 l.41E-6 14 3.37E-6 15 4.03E-6 16 2.93E-6 17 6.19E-6 18 0.0 19 0.0 20 0.0 21 0.0 22 0.0 23 l.24E-l 0 T STEAM GENERATER TUBE RUPTURE ACCIDENT NJ= 4 1 'REACTOR COOLANT' 2 'STEAM GEN FAULTED' 3 'STM GEN UNFAULTED' 4 'ENVIRONMENT' -1 INITIAL ACTIVITY V 1 2.622E8 GM V 2 5.310E7 GM V 3 1. 593E8 GM v 4 1.0 S 1 1 3 2.622E2 S 2 2 0 2.203E6 This page replaced by Revision 016
1 1
! Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:28
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture s 2 3 0 6.610E6 s 2 2 3 5.310El s 2 3 3 1. 593E2 s 1 1 1 7.620E3 s 1 1 2 7.885E2 1 1 4 7.885E2 1 1 5 7.885E2 1 1 6 7.885E2 1 1 7 7.885E2 c 1 1 2 4.765E4 1 1 4 1. 661E5 1 1 5 5.931E4 1 1 6 3.580E5 1 1 7 8.758E4 F 1 2 0 6.926E7 F 1 3 0 7.098E4 A 4 176 SEC TIME TO 176 SEC 250.0 SEC TIME TO 250.0 SEC F 1 4 0 3.194E7 F 2 4 0 2.515E5 F 2 4 1 2.515E7 2 4 3 2.515E7 F 3 4 0 1. 254E6 F 3 4 1 1.254E8 3 4 3 1. 254E8 N 4 0 p 1 0 4 2208.5 SEC TIME TO 2208.5 SEC F 1 2 0 5.409E7 F 1 4 0 6.455E6 4670 SEC TIME TO 4670 SEC F 1 4 0 0.0 2 HR TIME TO 2 HOUR F 1 2 0 o.o N 4 0 p 1 0 4 8 HR TIME TO 8 HOUR F 2 4 0 2.684E4 F 2 4 1 2.684E6 2 4 3 2.684E6 F 3 4 0 6.993E5 F 3 4 1 6.993E7 3 4 3 6.993E7 N 4 0 p 1 0 4 T T /* This page replaced by Revision 016
, I l Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:29
Subject:
Control Room Operator and Offslte Doses Due to a Steam Generator Tube Rupture Appendix C: Example ofCOROD Model (ARCON96 X/Q) II TSROOBR16ClA JOB 25402-012-572 II ORG=DWWU.CAG.BECHTEL II EXEC COROD7; SOUT='*' II Unit 1 SGTR Pre-accident Iodine Spike Case llCOROD1.FT05F001 DD* NIT= 23 NR= 1 ITP= 6 FACT= 1.0 COROD-WBN MHA FINAL ABSCE s KRM 83 KRM B5 KR 85 KR B7 KR BB KR B9 XEM 131 XEM 133 XE 133 XEM 135 XE 135 XE 138 I 131 I 132 I 133 I 134 I 135 I* 131 I* 132 I* 133 I* 134 I* 135 H 3 4 'ENVIRONMENT ' $TN= 0.6944E-01 1 O.OOOE+OO 2 3.378E+OO 3 5.302E+OO 4 3.107E+OO 5 5.B93E+OO 6 O.OOOE+OO 7 l.303E+Ol B l.429E+OO 9 5.042E+Ol 10 3. llOE+OO 11 l.807E+Ol 12 2.162E+OO 13 4.B96E+OO 14 2.26BE+Ol 15 l.527E+Ol 16 3.565E+Ol 17 2.B36E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 B.532E+Ol 4 'ENVIRONMENT ' $ TN= 0.2000E+Ol 1 O.OOOE+OO 2 l.369E+02 3 2.514E+02 4 B.725E+Ol 5 2.1BOE+02 6 O.OOOE+OO 7 6.164E+02 8 6.709E+Ol 9 2.3B2E+03 10 2.543E+02 11 9.073E+02 12 l.400E+Ol 13 2.769E+Ol 14 l.147E+02 15 8.532E+Ol 16 l.540E+02 17 l.542E+02 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 4.046E+03 4 'ENVIRONMENT ' $ TN= 0.8000E+Ol 1 O.OOOE+OO 2 2.123E+Ol 3 6.763E+Ol 4 4.315E+OO 5 2.503E+Ol 6 O.OOOE+OO 7 l.644E+02 B l.760E+Ol 9 6.359E+02 10 9.436E+Ol 11 3.1B4E+02 12 5.925E-03 13 l.677E+OO 14 l.B62E+OO 15 4.464E+OO 16 4.954E-01 17 5.764E+OO lB O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 l.OBBE+03
-6 'ENVIRONMENT CURIES ' $ TN= 0.2400E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO
-6 'ENVIRONMENT CURIES ' $TN= 0.9600E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 0.000E+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO
-6 'ENVIRONMENT CURIES ' $TN= 0.7200E+03 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO lB O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO 3.B5E-03 3.85E-03 3.22-03 2.36E-04 l.B8E-04 l.55E-04 74.0 7126.0 21600.0 57600.0 259200.0 2246400.0 3200. 0 51. 0 711. 0 51. 0 711. 0 51. 0 711. 0 51. 0 711. 0 51. 0 711. 0 51. 0 0.0 0.0 0.0 0.0 o.o 0.0 3200.0 0.95 0.70 0.95 0.70 0.99 o.o 2BB9.0 0.95 0.70 0.95 0.70 0.99 o.o 28B9.0 0.95 0.70 0.95 0.70 0.99 0.0 2BB9.0 0.95 0.70 0.95 0.70 0.99 0.0 28B9.0 0.95 0.70 0.95 0.70 0.99 o.o 2889.0 100.0 60.0 40.0 1440.0 5760.0 257198.0 1.2492 0.63 O.B352 This page replaced by Revision 016
I l Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:30
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture 322.0 45.0 17.75 46.0 9.0 4.0 161.0 22.5 4.0 0.0 ROOFFLUX DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE THROUGH ROOF 1000.0 1000.0 1000.0 20.0 20.0 20.0 500.0 500.0 -16.0 2.25 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM AUX BUILDING 270.0 150.0 148.0 27.0 15.0 14.0 135.0 75.0 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM TURBINE BLDG 322.0 112.0 341.0 32.0 11.0 34.0 161.0 56.0 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM SPREADING ROOM 322.0 45.0 26.0 32.0 9.0 5.0 22.5 161.0 -4.67 0.67 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM CR BLDG END 18.0 45.0 460.0 10.0 10.0 100.0 4.0 22.5 -25.5 3.0 ADJACENT DOSE TO CONTROL ROOM PERSONNEL DUE TO SHINE FROM CR BLDG END 18.0 45.0 460.0 10.0 10.0 100.0 4.0 22.5 -25.5 3.0 I* II This page replaced by Revision 016
11
- Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 31
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Appendix D: Example ofFENCDOSE Model II TSR008Rl6F1A JOB 25402-012-572 II ORG=DWWU.CAG.BECHTEL II EXEC FENCDOSE5; SOUT='*' II Unit 1 SGTR Pre-accident Iodine Spike Case llFNCDOS1.FT05F001 DD
- 1 KRM-83 KRM-85 KR-85 KR-87 KR-88 KR-89 XEM-131 XEM-133 XE-133 XEM-135 XE-135 XE-138 I-131 I-132 I-133 I-134 I-135 I*-131 I*-132 I*-133 I*-134 I*-135 H-3 T
1.784E-4 8.835E-5 6.217E-5 2.900E-5 9.811E-6 6.382E-4 STEAM GENERATER TUBE RUPTURE ACCIDENT TIME TO 250.0 SEC 4 'ENVIRONMENT I $ TN= 0.6944E-01 1 O.OOOE+OO 2 3.378E+OO 3 5.302E+OO 4 3.107E+OO 5 5.893E+OO 6 O.OOOE+OO 7 l.303E+Ol 8 1.429E+OO 9 5.042E+Ol 10 3.llOE+OO 11 1. 807E+Ol 12 2.162E+OO 13 4.896E+OO 14 2.268E+Ol 15 1. 527E+Ol 16 3.565E+Ol 17 2.836E+Ol 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 8.532E+Ol STEAM GENERATER TUBE RUPTURE ACCIDENT TIME TO 2 HOUR 4 'ENVIRONMENT I $ TN= 0.2000E+Ol 1 O.OOOE+OO 2 1.369E+02 3 2.514E+02 4 8.725E+Ol 5 2.180E+02 6 O.OOOE+OO 7 6.164E+02 8 6.709E+Ol 9 2.382E+03 10 2.543E+02 11 9.073E+02 12 1. 400E+Ol 13 2.769E+Ol 14 l.147E+02 15 8.532E+Ol 16 1.540E+02 17 1.542E+02 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 4.046E+03 STEAM GENERATER TUBE RUPTURE ACCIDENT TIME TO 8 HOUR 4 'ENVIRONMENT I $ TN= 0.8000E+Ol 1 O.OOOE+OO 2 2.123E+Ol 3 6.763E+Ol 4 4.315E+OO 5 2.503E+Ol 6 O.OOOE+OO 7 1.644E+02 8 1.760E+Ol 9 6.359E+02 10 9.436E+Ol 11 3 .184E+02 12 5.925E-03 13 1.677E+OO 14 1.862E+OO 15 4.464E+OO 16 4.954E-01 17 5.764E+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 1.088E+03 WBN SGTR TIME TO 1 DAy
-6 'ENVIRONMENT CURIES I $ TN= 0.2400E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 0. OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO WBN SGTR TIME TO 4 DAYS -6 'ENVIRONMENT CURIES I $TN= 0.9600E+02 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 0. OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO WBN SGTR TIME TO 30 DAYS -6 'ENVIRONMENT CURIES I $TN= 0.7200E+03 1 O.OOOE+OO 2 O.OOOE+OO 3 O.OOOE+OO 4 O.OOOE+OO 5 O.OOOE+OO 6 O.OOOE+OO 7 O.OOOE+OO 8 O.OOOE+OO 9 O.OOOE+OO 10 O.OOOE+OO 11 O.OOOE+OO 12 O.OOOE+OO 13 O.OOOE+OO 14 O.OOOE+OO 15 O.OOOE+OO 16 O.OOOE+OO 17 O.OOOE+OO 18 O.OOOE+OO 19 O.OOOE+OO 20 O.OOOE+OO 21 O.OOOE+OO 22 O.OOOE+OO 23 O.OOOE+OO This page replaced by Revision 016
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 32
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Appendix E: Determination of Letdown Flow Uncertainty The purpose of this appendix is to determine bounding errors for the measurements performed on the orifice restrictor flows using the Letdown Heat Exchanger Flow loop (1-F-62-82} during Preop Test Instruction PTl-062-03 RO. Following these tests, a loop check was performed for the computer point F0134A by injecting a signal into the transmitter and reading the display on the computer. To determine the total loop error, the unmeasurable errors must be combined with the errors present at the time of the loop check. WBN NESSD 1-F-62-1 will be used as a guide for determining the unmeasurable errors for loop 1-F-62-82 since it contains the same model flow element and a similar model transmitter. According to EMPAC, the flow element is a Vickery Simms Model MK-52 and the transmitter is a Foxboro E-13DM. Millers Flow Measurement Engineering Handbook, Third Edition, Chapter 6, Table 6.1 states that Square Edged orifice flowmeters have an accuracy of +/-1-2%URV (upper range value) of the flow rate. A value of +/-2% will be used for the orifice. The loop check performed by WO 94-14264-1 O (following pages) gives as found data. The largest error at 50 GPM was 1.36 GPM (50 - 48.64) or 0.68% CS (1.36/200 = 0.68%). The largest error at 100 GPM was 0.48 GPM (100 - 99.52) or 0.24% CS (0.48/200 = 0.24%). The largest error at 150 GPM was 0.06 GPM (150 - 149.94) or 0.03% cs (0.06/200 = 0.03%). Since the plant had not been started at the time of these tests, radiation was not present and need not be considered. Errors for temperature and power supply effect will need to be included. Since there is no data on actual temperature conditions, an enveloping value must be used. Environmental drawing 47E235-46 R5 gives the max abnormal temperature range as 50 - 110 °F for coordinates UA6 I El 737 where the transmitter is located per EMPAC. The transmitter is a model E-13DM per EMPAC. The product specification sheets (following pages) give the ambient temperature effect as +/-1 % per 50 °F for any span between 200 to 850" water. The transmitter will normally be calibrated at room temperature which will be between 60 and 80 °F. A temperature shift of +or - 50 °F will encompass the temperature changes seen by the transmitter. Therefore for a temperature range of +/-50 °F, the temperature effect will be +/-1 % CS d/p. The power supply effect is given as 0.1 % CS for a 10% change in voltage. Thus Power supply effect is 0.1 % CS d/p. All errors for the computer should be reflected in the loop check.
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 33
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Utilizing Equation 3-24.8 of W WCAP-12096, Rev. 8 "Westinghouse Setpoint Methodology for Protection Systems, Watts Bar Units 1 and 2, Eagle 21 Version," the unmeasured transmitter errors can be converted from percent error in full scale d/p to error in percent full span at a specified point, where Fm is the maximum flow rate of 200 GPM, and Fn is the nominal flow rate (i.e. 50, 100 or 150 GPM). EPFS (Flow) = (Axxx I 2) * (Fm I Fn) Temp0 ,,(Flow)@50GPM = . (Temp 0 ,, (d/p) I 2) * (200 I 50) = +/- 2% CS Flow Temp0 rr(Flow)@1 OOGPM = (Temperr (d/p) I 2) * (200 I 100) = +/- 1% CS Flow Temp 0 rr(Flow)@150GPM = (Temp 0 rr (d/p) I 2) * (200 I 150) = +/- 0.67% CS Flow pwr SUPPerr(Flow)@50GPM = (pwr SUPPerr (d/p) I 2) * (200 I 50) = +/- 0.2% CS Flow pwr SUPPerr(Flow)@100GPM = (pwr SUPPerr (d/p) I 2) * (200 I 100) = +/- 0.1 % CS Flow pwr SUPPerr(Flow)@150GPM = (pwr SUPPerr (d/p) I 2) * (200 I 150) = +/- 0.067% CS Flow 2 2 2 2 05 Thus total loop error= (FEerr +Loop check err +Temperr (Flow) +pwr SUPPerr (Flow) ) 2 Total loop error @ 50 GPM = (2 2 + 0.68 2 + 2 + 0.2 2 ) 05 = +/-2.92% CS = +/-5.84 GPM 2 2 Total loop error@ 100 GPM = (2 + 0.24 + 1 + 0.1 2 ) 05 = +/-2.25% CS = +/-4.5 GPM 2 Total loop error @ 150 G PM = (2 + 0.03 + 0.672 + 0.067 2 ) 0 *5 = +/-2.11 % CS = +/-4.22 G PM 2 2 Total loop error at 120 GPM can be determined by linear interpolation between 100 and 150 GPM. The value will be conservative since the error is nonlinear and is a function of the square root of the d/p values above and the actual loop recorded values which also follow a square root curve. Total loop error @ 120 GPM =+/-I error @ 100 GPM + 20(error @ 150 GPM - error @ 100 GPM) I (150 - 100) I Total loop error@ 120 GPM = +/-[4.5 GPM + 20(4.22 - 4.5)/50] = +/-[4.5 GPM + (-0.11)] = +/-4.39 GPM The following references were used in preparation of this appendix. Revisions to these references will not impact this appendix; so the references are 'information only' in lieu of 'design input'. 1 WBN NESSD 1-F-62-1 R1 (Methodology & guidance) 2 EMPAC (Manufacturer, Model number and location) 3 Millers Flow Engineering Handbook, Third Edition, Chapter 6, Table 6.1 (Orifice accuracy) 4 WO 94-14264-10 (loop check data) - see next page 5 Drawing 47E235-46 R5 (environmental data) 6 Foxboro product specification sheets (transmitter accuracy data) - see next pages 7 WCAP-12096 R 8 (methodology for converting d/p error to flow error) Prepared Lynn Cowan Date 6/4/01 Checked _D=.L=.K~i""'rb::...iy.___ _ _ _ _ _ _ _ _ _ __ Date 6-28-01 . (original signed in R8)
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 34
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2 2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Supporting documents for Appendix
~*-. '.; :. I WORK ORDER FORM WO NO: 94-14264-10 ORIGINATION DATE; 06/22/94 RF NO: C254025 ORIGINATOR: JAMES R RIVERS EXTENSION: 3181 PRIORITY: JC EQUIPMENT IDENTIFIER : WBN-l-LPF-062-0082-EQUIPMENT DESCRIPTION: EXCESS LETDOWN HTX FLOW EQUIPMENT CATEGORY : QR SR lE TYPE OF MAINTENANCE: OTHER MAINTENANCE PROBLEM DESCRIPTION: PERFol?M RIST TESTCP.LIE'>RRTIUJ oF ':>'IS w2 ?C,ft;,B llJSTRuMEIJT'!:J LISTED ON THE WR CARD FCR PTI-or..l-0~
ACCEPT4~CE CRITERIA JOB LOCATION VARIOUS LOCATIONS, SEE SSD LOCATION CODE AlOO - AB ALL AUXILIARY BUILDING GENERAL AREA WORK DES~RIPTION PERFORM POST TEST CALIBRATION OF l-LPF-062-0082, AS REQUIRED, FOR PTI-062-03 ACCEPTANCE CRITERIA LCO; YES [ ) NO [ x] LENGTH: µI.e... LCO EXPIRES: ,J ~-- SECT XI R/R: YES [ ) NO [X] ---m>RDs: YES [ . ] NO [ )() RWP REQ YES [ ] NO [X] RWP #: .Jfll- ALA.RA: YES [ ) NO [X) TAGGING REQ: YES [ SCAFFOLD YES [ NO [X] H.O. #: ~1.e.. SHUTDOWN: YES ( ] NO [X) NO [X] INSUL: YES [ ] NO [XJ PERMITS REQ: NONE DISCIPLINE: MIG EST HOURS 4.0 TASK TOT: 8.0 MAN HOURS PRE-MAINT TEST REQ. NONE DURATION: 4.0 HOURS POST-MAINT TEST RtQ.: SEE WORK INSTRUCTIONS
.,., * ::: r ~ "- * : "
~1 E:
\ ' ' ' ..
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 35
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
**.(.,: ~ .
SSD-l~LPF-62.,.82-0 PAGE 10 OF i7 REVISION 01 * , INSTRUMENT LOOP CALIBRATION RECORD 1110 NO: WO
-------------------------------------------------------..--------- LOOP COMPONENTS ID~-l=FT:62:92---- INSTRUMENT NO: 1-Fl-62-82 H~o~~-------
~:~2!t9..
TEST 3 {, '1 1 111.TE: VISUAL INPUT--*--R-EQ_U_lR_E_D_ _ _ _ _A_S_F_OU_N_D________________ AS-LEFT _________ _ POINT ( IN WC ) ( GPM ) LO LIMIT AS FOUND HI LIMIT LO LIMIT AS LEFT HI LIMIT
---;-- ---12~-5-- 28.3 23.3 33.3 23.3
--2-- --39~1- -*---- 50.0 t*---
45.0 55.0 95.0 105.0 4 150.0 145.0 155.0
-- --59~1--- 195 _Q 190.0 190.0 200.0
---6""-- --i5i~6-- 150.0 15S.O 145.0 155.0
-- -156~--- --10-0-.0--l---..,....ir:::
105.0
---- 95.0 ---------
105.0
---8-- ---397- -J,..'-4-5-.o- ----i--!\-s-.o ---4s-.o- --------- -------
SS.a
---9-- ---1~-5--1---'--- ----1---- ----1--- ----- -------
------------ 23.3 33.3 23.3 33.3 INSTRUMENT NO: LOG Pt F0134A M&TE; VISUAL TEST REQUIRED AS FOUND AS LEFT POINT GPM
----- LO LIMIT AS FOUND HI LIMIT LO LIMIT AS LEFT
{ ) HI LIMIT 28.30
------ 50.00 23.18 ).S.&3 32.38
- 23. 18 :?2.38 3
------ 100.00 47.34 '-li.l_L 101.14 --ga.55- ----<
52.26 47.34 ~ 52.26 98.66 ??.h 1
--- 150.68 ""149.'Ci8- ""'\.:o; '\)
101. 14
-- 195,DO 150.00.
4 149.08 15"0.0/ 150.6!!
---- ----~ --
5 194.32 19'1,/2 195.60 194.32 ll.. 195.60 6 150.00 149.08 J'/'f.(jf 150.68 149.08 1S0.6e 100.00 98.66 8 50.00 47 .3'4 11.s~ 101.14 52.26
---- 98.66
- 47. 34
~
\lJ 101. 14 52.26 9 28.30 23.18
'"'8.lt,'{_
I .;is.t;, J ---32.38 ---z:i:-iii- -::£' -~ ---32~38-Function: latdown Heat Exch Flow Reviewed by: Gary L. Hyden Approved by: Ed Hall £:_\'I Date: 03/10/94
Calculation No. WBNTSR008 1Rev: 014 I Plant: WBN 1Page: 36
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 I ... -... .... n i ",....~ - JIOXBORO
- General Specification
* *'4 EIKlronlc s.Ms dip C.CI T,_..,ltt.,. ft'llUUft diff*
\nntw ,.....,.. ....... of O.fi to o.aso lndlel (0.127 m 0-21590 mml of watw st lfatic ~- up to 6000 pd (420 kgfcm21. TIMy hlW1tlt
- prapordonal 10 to 50
<< 4 to 20 lllA ck lignlil, onr onll-v unohleld..t kada.
., ..-lvtta ~ up to _... lhOU&llnd ,.., from th*
point o f - e n t . FEATURES Tlftle.l'rov.n Oesi;n Trou~Fne~on Et30l £13t)M Hicih " - ' - - E>cceu ...t R8ptoducibility PERFORMANCE
&. of calibration Ml"~t - Wide Rang<1 C~bility ~ncy Stoibte Force Balance Symm S ;ii[;:;"JJJS mml ::t.0.5~ of - n 526 to 85<J..inch ::t.O. 75% of ,pon Posftm 0.-91nn91 Protec11on 113360 to 21S90mml D...i Band . 0.05% ol _..
R_..i.;lity: E130LSer;.. 0.15" of_.. - EtJOM.ElJOHSer; O.l0%of-n Hyw-c 0.101' of - - A?Plieat_ion v en:atir11y R"P<aducibility: E 130l Series - 0.217*:,
- q>1n l.
.
- ElJOM,ElJOHSene> _ _ ** 0.15'1iol.0.n .:j
~lmrinsically s.fe (l"!"udes effeeu of Hvste<esi1. R..,..,.i.;1i1y. ~od Bond """
Orift .,... 1-ll< 1><riodl BASE TRANSMITTER STANDARD SPECIFICATIONS
* .
- StyloB *
~ Full~ .diumble between range limiu of capsule..
~
Output"'-*' Ma:lmUm Proceis Tempet'afu;._ 250 F (120 Cl at cap- Oul!MS'-'
~l-lood N-lnol'uwfr sule * * (lnAof.d
("""'"' V"'-fwom MWraum I Mulm."' ~u.wc
~ 'T~ -eo* to
..::,.. I "° Umlts +180 .F (-40 to
.... 10
+82 CL With remote amplifier. -40 to +2SO F (~to :MVok 10tclSO MO 80V.M:
+120Cl .
- Bottint Steel cep screws alld nuts through body and w ~o Fod>oro - dCtritmtio<\ per..i. - -
- col~
... soo.....,1oec1....-....-.;--.
- llfQCleSS ~ors Supply Volt1gt Uralts 24 to 60 110IU d< with 4 to 20 rnA OUtPUt *nd 63 to 100 wits d< with 10 to SO rnA C - 11veaded cast aluminum seated on Bum-N CHing :iutput from separate power supply ur\it.
Sllllll. Blue textured Wl'(f fmish.
.Mrppty V~ Effect Zeto shift will be less than 0.1 %
~ Claltlficltion NEMA 4 watertight of span for a 10% d>ange in wltage within sated limits.
Ellctronic TlllMlllltt# .rid AmpllfJtr Solid ltlte Elel;tric Claalflc:stion Explosionproof ~ 1;Groups C ind 0, Oivldon 1. *-* verti1 Elctftc:.f Cocw-tlol* Two S.foot lead$ from \/2-ioch Mounting Direct to process wi1li *b<tcket for 2-inch hor;. female conduit COMeelions zontal or pipe.
~ E1JOLS.W. ElJOM~ E130HSer*
A-..Uioolu Low "->oe Ciipsyle 0-5 to 0.25- w.iter (0.127to~Smml Wum ~ CllPS<Jle 0.:20 to 0.205* w11er 0-20 to 0-205* -ter (Q-508 to 0-5207 mml H9! Ret>ge Capsule - ().200 10 O.SSO- water IG6080 lo 0.21590 mml (().50810 0.5207 mml
-().200 10 ().850 .. W8lllC' (0-~0 to 0-21590 mml,
(
- GS 2A-1C1. E No-nbo<lQ71
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 37
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
-t
>r c re c' c
' - STANDARD SPECIFICATIONS (Continued I
~ E13DLS.ia E130M Se<ift E13014 S<<ie1 Wm.dP..U:
8odv and Process Coon Cadmium plated forged car-bon stool or forged 316 SS Cadmium plated forged en: I C.ldmiJm plated forged car-,I
'* bon steel or forged 316 SS bon steel°' forged 316 SS Diaphragm~ and Force Bar 316 suinlesutl!lll 316 stainlt:i.'SSteel .- Jl 6 stainless Stffl l'orce &(Seal - - - Cobalt nid;el alloy Cobalt nic:kel alloy Cobalt nickel alloy ;
Capsule GMlcet - - - 316 stainless steel 316 s*.i:*-*css steel 316 stainless stffl Prooes:s Conn Gasket - TFE TFE Glass filled TFE FO<QI Bar Seal Gasket - Siliaine elastomer Silicone elastomer Bu!Q-N
&dcupPlale 3l6 stainl= steel 316 stainless steel I 316 stainless steel Maximum Ststic Pnaun soO psi- 05 1:.gfan2J 2000 psi (140 lcgtcm21 6000 psl (420 l:.gtan2J rn.c..~- t/4 or 1/2_ NPT femat., or 1/4 °' 112_ NPT female°' t/4 or 1/2 NPT °' body 1/2 inch Sch eo welding 1/2 inch Sch 80 welding machined to accept 9/1 &.
ned. as specifl<Od. neclc. as specified.
- 18 Amina> fittings.as spec-ified.
AmlMcrt Tempemu,. +/-1.0% P"' 100 F (55 Cl Medium Range Capsule: Medium Range Capsule: Effect (Zero shi~ in chaoga at 25.. 16J5 mml pe<c..nt of spanJ
+/-1.0'lO per 100 F 155 Cl +/-1.0% per 100 F .15?. q *'
water; .t.1.0% per 40 F 122 changut ~oo** 12540.mml c:han!)e at 1oo** (2540 nim) Cl c:har>ge at 5- 1127 mm) water: :t:.1.0% pet 125 F water; +/-1.0% per 12!; F water.. 169 Cl at 205.. 15207 mml
.-... water: +/- 1,Q.'l! per 40 F 122 (69 Cl at 205- 15207 mm) water; +/-1.0" I><< 40 F (22 Cl at 2s** 1635 ""!'I water* Cl at 25- (635 mml water.
High Range Capsule: Less
- High Range Capsule: Less than-+/- 1% per SO F 128 CJ than +/-1% per SO F (28 Cl
. change for: any $pan be- change for any span be-tween 200 to eso- (5080
-- ( to 21590 mml water.
tW<Oen 200 to 850.. (5080 to 21590 mml water. I/ , Potltlon ------1 moc.inted Transmitter should be w;th *Clpsule in
"!'ftial position.:
. ro.ltion !:rt.ct _ . _ . 1..:.. Maximum of le$$. than j% Maximum of *less *Iha~ 3% * -
zei:o shift fOC' 90 degr<0e tilt zero shif:t for 90 degtee"iilt
- of instrume.-.t in any plane of instrilment in any plane
-VlbretJon - - - - - less than 1.5" zero shift l = iha<"l 1% zero shift.for Less than 1% zero shift fOC' fOC'vibration to 1.5G in any vibration to 2G in any vibration to 2G in any
, plane. at frequencies I= plane_ plane.
1han 80 Hz. Maximum zero shift I= Zero shift less than 0.5" Zero shift less than 1.5" than 0.5" of span for 500 span for 2000 psi (140 kg/ span fOC' 0-6000 psi l0-420 . psi (35 1;gtcm21 dlange. °'12J change at SO to 850°" kg/cm2) change at 50 lo (1270 to 21590 mml w.1ter; aso- (1270 to 21590 mml 1.o% span for 1000 psi C10 water or 0.5" span for any -' kg!an2) dlange at 20 to 2000 psi (140 lcg!an21 so- CS08 to l270mml wa- ~ 2.0% s:pan forO to ter. 6000psi (Oto 420 l;g/cm2) d\ange at 20 ta so- .(508 to 1270mml waterOC' 1.~ span fOC' any 2000 psi (140 _l<glcm2J change*.
- 0-..Dll-. 1-~ 16 11a- (410 mmJ H" 13 114.. (337 mnil .H x- 14 ltr 1368 mm! H x
- 6118- (175 mm) W. 6 7/8.. (175 mml W. I 6 718"" (115 mm) W.
f
~~~ 321b(1Sltgl 25 lb (11 kg) I 40 lb (18 l:Ol I
* .._.. -1es cydialfy. refer to your nearest Foxboro Sales Office.
*.; I 7
I~. l Calculation sheet Document: WBNTSROOB l Rev.: 016 I Plant: WBN I Units 1,2 I Page:38
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture AppendixF: Deleted in R16. This page replaced by Revision O16
1l i Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:39
Subject:
Control Room Operator and Offslte Doses Due to a Steam Generator Tube Rupture This page is intentionally left blank. This page replaced by Revision 016
~
*~
l ! Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:40
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Appendix G: Unit 2 SGTR The Unit 2 steam generators are the same as the original Unit 1 steam generators. However, the mass releases were reanalyzed by Westinghouse. The SGTR analyses performed below use the same methodologies as the main text, with the following changes: Volumes:
#1: Reactor Coolant: 5.4E5 lb= 2.4494E8 gm (ref. 29)
#2: Steam Generator w/Leak: Volume/SG = 4.735E7 gm (ref. 51)
#3: Steam Generators w/out Leak: 1.421E8 gm in 3 unfaulted SG's) (ref. 51).
Step Sources: The step sources (Ci-gm/µCi) to initialize the reactor coolant and the secondary side activities are: All cases: S=2.4494E8 gm x IE-6 Ci/µCi x 29.06=7.l18E3 (noble gases) S=2.4494E8 gm x lE-6 Ci/µCi == 2.4494E2 (tritium) Pre-accident iodine spike case (initial concentration= 14 µCi/gm): S=2.4494E8 gm x IE-6 Ci/µCi x l l.348[µCi/gm l-l31r1 x 14µCi/gm 1-131 = 3.891E4 (iodines) Accident initiated iodine spike case (initial concentration= 0.265 µCi/gm): S=2.4494E8 gm x IE-6 Ci/µCi x 11.348 [µCi/gm l-l31r1 x 0.265µCi/gm1-131= 7.366E2 Secondary side, all cases, steam generator with leak release to environment (concentration = 0.1 µCi/gm) which is due to dryout: S = 4.735E7 gm x lE-6 Ci/µCi x 4.150E5 [µCi/gml-131]" 1 x 0.lµCi/gml-131 =l.965E6 S = 4.735E7 gm x IE-6 Ci/µCi = 4.735El (tritium) Secondary side, all cases, steam generators without leak (initial concentration= 0.1 µCi/gm): S = 1.421E8 gm x IE-6 Ci/µCi x 4.150E5 [µCi/gm l-l31r 1 xO.lµCi/gm1-131=5.897E6 S = 1.421E8 gm x lE-6 Ci/µCi = 1.421E2 (tritium) Continuous Sources: For the accident initiated iodine spike case, the iodine spike is 500 times the iodine release rate from the fuel. At steady state, the iodine release (production) rate is equal to the removal rate. The iodine removal is due to a) radioactive decay, b) removal by the letdown system, and c) removal through reactor coolant leakage. These terms are expressed as: P = l:removal rates = decay+ letdown+ leakage or P = A. + fouV + p,/V where P =production rate [hr' 1] A.= decay constant for the isotope in question [hr'1] = ln(2)/f 112 fL =letdown flow rate= 120 gpm + 4.39 gpm = 124.39 gpm e =letdown demineralizer efficiency= 1 (assumed so as to maximize removal/production rate) V =volume of primary coolant= 5.4E5 lb Ps = removal rate of iodine from primary side due to leakage = 11 gpm ( 10 gpm identified + 1 gpm unidentified) T 112 = halflife taken from ref. 42 Note: all the above flow rates are converted to mass flow rates at STP (H20 = 1 gm/cc). Removal rate of iodine from primary side to secondary side was not considered above, because of its relatively low rate (3 x 150 gpd = 0.3 gpm) compared with other terms. Production Rates for a Reactor Coolant Leak of 11 20m (10 20m identified + 1 l!Dm unidentified) Half Life A. (I/hr] fLe/V (I/hr] p/V [I/hr] ProdrateP [I/hr] 500 xP 1-131 8.04 d 3.59E-03 l.15E-01 l.02E-02 0.1291 60.56 1-132 2.28h 3.04E-01 l.15E-01 l.02E-02 0.4295 214.77 1-133 20.9h 3.32E-02 l.ISE-01 l.02E-02 0.1587 79.35 1-134 52.6m 7.91E-01 l.15E-01 1.02E-02 0.9162 458.10 1-135 6.61 h l.OSE-01 1.15E-01 1.02E-02 0.2304 115.20 Tins page replaced by Revision 016
I I l Calculation sheet Document: WBNTSROOB I Rav.: 016 I Plant: WBN I Units 1,2 I Paga:41
Subject:
Control Room Operator and Offsita Doses Dua to a Steam Generator Tuba Rupture The accident initiated iodine spike of 500 times the increase in the iodine release (production) rate from the fuel is modeled as a continuous source: C =Volume x IE-6 Ci/µCi x Prod Rate x 500 x 1 µCi/gm 1-131 equivalent conversion factor x 1-131 equiv. where Volume= 2.4494E8 gm Prod Rate = see table above 1 µCi/gm 1-131 equivalent conversion factor= 11.348 (value determined above, this is to get the ANSI/ANS-18.1-1984 source into 1 µCi/gm 1-131 equivalent 1-131equiv.=0.265 µCi/gml-131 Continuous Source [gm-Ci/µCi-hr] for Accident Initiated Iodine Spike: Reactor Coolant Leak of 11 2PID. (10 J?Dm identified+ 1 gpm unidentified) Nuclide 0.265µCi/gm1-131 1-131 4.756E+04 1-132 1.582E+05 1-133 5.845E+04 1-134 3.374E+o5 1-135 8.485E+04 Flow Rates: The following is for the Unit 2 steam generators. The amount of secondary side steam released from the ruptured steam generator is 103,300 Ihm from 0-2 hours and 32,800 Ihm from 2-8 hours (ref. 47). The amount of secondary side steam released from the intact steam generators is 492,100 Ihm from 0-2 hours and 900,200 Ihm from 2-8 hours. The reactor coolant release to the steam generator was a total of 191,400 lb, of which 10077.2 (=934.4+9142.8) lb flashed (ref. 47). To account for the release during the 74 second interval when the control room is not isolated, the amount ofreactor coolant released at 74 sec is needed. However, the release from the steam generators does not actually start until 113 sec post accident. Therefore, the releases at 113+ 74 = 187 sec are actually needed for release calculations. Using the releases from reference 47 and adding each time increment release, the reactor coolant release at 187 sec is 9518.550 lb (calculated using "Break Flow" in Attachment 3 in ref. 4 7). The amount that flashed at 187 sec is 1179 .325 lb (calculated using "Total Flashed Break Flow" in Attachment 3 in ref. 47). The mass release rate from the ruptured steam generator is non-linear. However since the time frame for the release is short (74 sec), the average release rate can be used. From reference 47, the flashing of the reactor coolant stops at 2253 sec, and the break flow stops at 5032 sec. The following flow rates/leakage rates for each component are: Flow from Reactor Coolant #1 to Steam Generator Faulted #2 (non-flashed): 0-187 sec: F=(9518.550 lb-1179.325 lb) x (3600sec/br)/ (187sec)=1.605E5 lb/hr=7.282E7 gm/hr 187 - 5032 sec: F = (191,400 lb - 9518.550 lb) - (10077.2 lb - 1179.325 lb) I (5032 - 187 sec)= 35.704 lb/sec=
=5.830E7 gm/hr 5032+ sec: F = 0 gm/hr Flow from Reactor Coolant #1 to Environment #4 (flashed):
113 -187 sec: F = (1179.325 lb) x (3600 sec/hr) I (74 sec)= 5.737E4 lb/hr= 2.602E7 gm/hr 187-2253 sec: F = (10077.2 lb -1179.325 lb)/ (2253 -187 sec)=4.307 lb/sec= 7.033E6 gm/hr 2253+ sec: F = 0 gm/hr This page replaced by Revision 016
,l
- Calculation sheet Document: WBNTSR008 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:42
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture Flow from Steam Generator Faulted #2 to Environment #4: 113 sec - 2 hr: F = 103,300 lb I (2hr- 113sec I 3600sec/hr) = 5.247E4 lb/hr= 2.380E7gm/hr (noble gas and tritium) F = O.Ql x 103,300 lb I (2hr - 113sec I 3600sec/hr) = 5.247E2 lb/hr= 2.380E5gm/hr (iodine) (see note*) 2 - 8 hr: F = 32800 lb I (8hr - 2hr) = 5.467E3 lb/hr= 2.480E6 gm/hr (noble gas) F = O.Ql x (32800 lb) I (8hr-2hr) = 5.467El lb/hr= 2.480E4 gm/hr (iodine) Flow from Steam Generator Unfaulted #3 to Environment #4: 113 sec - 2 hr: F = 492,100 lb I (2hr - 113sec I 3600sec/hr) = 2.500E5 lb/hr= 1.134E8 gm/hr (noble gas) F = O.Ql x 492,100 lb I (2hr - 113sec I 3600sec/hr) = 2.500E3 lb/hr= l.134E6 gm/hr (iodine) 2 - 8 hr: F = 900,200 lb I (8hr - 2hr) = 1.500E5 lb/hr= 6.805E7 gm/hr (noble gas) F = O.Ql x (900,200 lb) I (8hr- 2hr) = l.500E3 lb/hr= 6.805E5 gm/hr (iodine) Flow from Reactor Coolant #1 to Steam Generator Unfaulted #3: 0- 8 hr: F = 3 steam generators x 150 gpd x 3785.48 cc/gal I 24 hr/day x lgm/cc= 7.098E4 gm/hr note* Normally, to take into account uncovery of the faulted steam generator, there is no iodine partitioning in the release to the environment (iodine partition coefficient= 1). For conservatism, no iodine scrubbing of the bubbles in the flashed water is taken into account However, the water that boils is allowed the iodine partition of 100 (see assumption 6). The STP input for Unit 2 is similar to Unit 1 shown in Appendices C and D with the modifications described above. Unit 2 ARCON96 XJQ values (worst case) for Unit 2 (ref. 37): 2.59E-03 sec/m3 0-2 hr, 2.12E-03 sec/m3 2-8 hr. XJQ values after 8 hrs are not used because the release is only considered within 8 hr (see above flow rate calculation). Results: The results (rem) are as follows: Control Room Operator Dose (rem) Limit (rem) Pre-Accident Iodine Accident Initiated Iodine Gamma 6.27E-02 6.05E-02 5 Beta 7.0BE-01 7.06E-01 30 Thyroid (ICRP-30) 1.32E+01 2.20E+OO 30 TEDE 8.30E-01 4.67E-01 5 Pre-Accident Iodine S:>ike Offsite Dose (rem) Limit (rem) 2-hr EAB 30-dayLPZ Gamma 3.87E-01 1.14E-01 25 Beta 2.28E-01 6.99E-02 300 Thyroid (ICRP-30) 1.45E+01 4.15E+OO 300 TEDE 1.34E+OO 3.86E-01 25 Accident Initiated Iodine Soike Offsite Dose (rem) Limit (rem) 2-hr EAB 30-dayLPZ Gamma 5.76E-01 1.69E-01 2.5 Beta 2.66E-01 8.19E-02 30 Thyroid (ICRP-30) 7.56E+OO 2.23E+OO 30 TEDE 1.25E+OO 3.63E-01 2.5 This page replaced by Revision 016
I I l Calculation sheet Document: WBNTSROOB I Rev.: 016 I Plant: WBN I Units 1,2 I Page:43
Subject:
Control Room Operator and Offsite Doses Due to a Steam Generator Tube Rupture The following Unit 2 margins were calculated from the doses in the Table above. Where: margin= limit- dose, and percent= 100 x (limit-dose)/limit Control Room Ooerator Dose Margins Pre-Accident Iodine Accident Initiated Iodine Limit (rem) Margin (rem Percent Margin (rem Percent Gamma 4.9 98.7% 4.9 98.8% 5 Beta 29.3 97.6% 29.3 97.6% 30 Thyroid (ICRP-30) 16.8 55.9% 27.8 92.7% 30
- TEDE 4.2 83.4% 4.5 90.7% 5 Accident Initiated Iodine Spike Offsite Dose Margins 2hr EAB 30 dalf LPZ Limit (rem)
Margin (rem Percent Margin (rem Percent Gamma 24.6 98.5% 24.9 99.5% 25 Beta 299.8 99.9% 299.9 100.0% 300 Thyroid (ICRP-30) 285.5 95.2% 295.9 98.6% 300 TEDE 23.7 94.6% 24.6 98.5% 25 Accident Initiated Iodine Soike Offsite Dose Maroins 2hr EAB 30 da'/ LPZ Limit (rem) Maroin (rem Percent Margin (rem Percent Gamma 1.9 77.0% 2.3 93.2% 2.5 Beta 29.7 99.1% 29.9 99.7% 30 Thyroid (ICRP-30) 22.4 74.8% 27.8 92.6% 30 TEDE 1.3 50.2% 2.1 85.5% 2.5 Discussion and Conclusion The Unit 2 steam generators (same as the original Unit I steam generators) with a Steam Generator Tube Rupture will not exceed the 10CFR50 App. A GDC 19 control room dose limits or the IOCFRIOO offsite dose limits. This page replaced by Revision 016
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 44
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Attachment 1: Justification for Using ANSI/ANS-18.1-1984 Expected Coolant Spectrum The choice of iodine spectrum is fairly important, since several isotopes have short halflives. Results may be affected when accident times are on the order of the decay of the short lived isotopes. There are several possible spectra available. The spectrum chosen for this analysis is the one that most closely resembles the actual spectrum present at WBN. From the surveillance tests l-SI-68-28 performed on 7110100 and 4/9/01 (see following surveillance tests attached), the following concentrations were determined: RCS activities 7110/00 RCS activities 4/9/01
µCi/gm µCi/gram µCi/gm µCi/gram RCS RCS RCS RCS Gaseous Degassed Gaseous Degassed Ar-41 1.303E-02 F-18 l.179E-01 Ar-41 2.696E-03 F-18 1.116E-01 Kr-85M l.915E-04 Na-24 9.169E-04 Kr-85M 2.013E-04 Na-24 2.060E-03 Kr-87 4.575E-04 Mn-56 9.313E-05 Kr-87 4.809E-04 Mn-56 2.088E-04 Xe-133 9.565E-04 Co-58 5.019E-04 Kr-88 4.982E-04 Co-58 6.218E-04 Xe-135 l.429E-03 Nb-95 3.132E-05 Xe-133 1.202E-03 Co-60 2.776E-05 Xe-135M 7.364E-04 I-131 6.070E-05 Xe-135 1.676E-03 Nb-95 2.794E-05 Xe-138 1.796E-03 I-132 1.459E-03 Xe-135M 1.105E-03 1-131 3.881 E-05 I-133 8.208E-04 1-132 1.165E-03 1-134 2.694E-03 1-133 6.105E-04 I-135 l.608E-03 1-134 2.334E-03 Xe-135 8.914E-05 1-135 1.158E-03 Xe-135M l.406E-02 Xe-135 1.380E-04 Cs-138 2.395E-03 Xe-135M 1.972E-02 Cs-138 2.195E-03 Two potential spectra are from WBNNAL3-003 (Reactor Coolant Activities in Accordance with ANSI/ANS-18.1-1984) and from the FSAR Table 11.1-2 (Historical Design Activities). The iodine concentrations and relative concentrations for each spectrum are as follows:
7110/00 7/10/00 419101 4/9/0 l WBN actual WBN actual WBN actual WBN actual
µCi/gm relative fraction µCi/gm relative fraction 1-131 6.070E-05 0.0091 3.881 E-05 0.0073 I-132 l.459E-03 0.2196 1.165E-03 0.2195 I-133 8.208E-04 0.1236 6.105E-04 0.1151 1-134 2.694E-03 0.4056 2.334E-03 0.4399 I-135 l.608E-03 0.2421 1.158E-03 0.2182 sum: 6.643E-03 5.306E-03 ANS 18.1 ANS 18.l FSAR 11.1-2 FSAR 11.1-2
µCi/gm relative fraction µCi/gm relative fraction I-131 0.0477 0.0448 2.5 0.2461 I-132 0.225 0.2115 0.9 0.0886 1-133 0.149 0.1401 4 0.3937 I-134 0.364 0.3422 0.56 0.0551 I-135 0.278 0.2614 2.2 0.2165 sum: 1.0637 10.16
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 45
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 As can be seen, the FSAR historical design concentrations do not reflect the actual measured concentrations. The FSAR values are weighted too strongly in favor ofl-131 (24.6% of total as opposed to < 1% of the actual total). By comparison, the ANSI/ANS-18.1-1984 fractions are very close to the actual fractions. The worst fit was for I-134 which was 40. l % actual versus ANSI/ANS-18.1-1984 34.22%. The I-131 is slightly over predicted by ANS-18.l (0.9% on 7/10/00 and 0.7% on 419101versus4.48%), however this difference is not as large compared to the FSAR fraction. The ANSI/ANS-18.1-1984 spectrum overall fit is much better than the FSAR spectrum, therefore it can be concluded that the use of the ANSI/ANS-18.1-1984 spectrum is acceptable.
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 46
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 (continue~): Surveillance test l-SI-68-28 performed on 7110/00 JUL-24-2000 13: S2 TVA PLANT MGRS OFC 423 365 1904 P.02~7 SURVEILLANCE TASk SHEET (SPP*B.2l PAGf _1_ ot llOllk ORCER: OQ067'1JOO SI tEYI P05l1 PROCIOURE#1 1*t1*611*Z8 TITLE: PRIMARY RADIQCNE.lllSTRY REQUfll!MEHTS PDF HCT: CSH TEST ~&Diii PEllHIDIC PEUORMANCI:
---o""A'""Te,...--'N/A_T_lt11i--
PUEI 07/10/0Q
\lllN EMT: 07/11/00 MAX EKT: f~ llW'*B.2*2
,RElh W IQ: H ASME XI: H APP MOOE: 123 PERF MQOf: 1.D4 "1/(o(ou 1...J..LJ..L SU-T RVW!ls C-LETJOll DATE TllE lNSTRUCTlOlfl: Do NOT Stlilrt prior to Kheduled OJ* dat11
""'~am**-*****m.ee*m==~**-***--*-~aae111111t**-11 -----&GC1;;;;;:a;;;i11:*-*~**~=";;;r;;s**11s*.-**m1111..--=:::=;====
TEST P!RFllRIERS I/AS THIS A CCllPLETE DR PAATIAL 4clb NAM! SIGNATlllE IN!T SECT PERFDJlllANCE? IN REMAUSl
!El(PLAlN "PAllTIAL" / PARTIAL:
l_'1d°= MCH,on"-' M_oMMffl 11.V lol!RI ALL tECH SPEC/TECH REQ/OPCll/FIRE / PROT lfQ ACCEPTANCE CRITrR!A SATl3FIED7 YES:_ NO:_ N{A:_ W!I! ALL OTHER ACC!PtAllCE CRIT!R IA 6ATl5Fll:D7 YEB1_ NO; N/At.::::. ALERT ICllEDUUHG REtlHRED? YU:_ NO:_ N/A:_X_ IF ALL TECH ll'Et/TECH REO/lllr.IVFIRE PROT . REQ WERE NIJT BATIHIED, WAS AN LCO/TR/ / DQ~/DR ACTION REQD7 (EXPLAIN IN AEMARKSl YES1_ NO:_ N/A:_ TEST ~~,~~/Ao DiREClCR I LTEAF 6T AliiiiWiilii:liiilDClila&lllUllll!a1!1aa==:a::ci================1t*E~=========~= ==:;:=~~iilSl'liiii:ilill'B*lllll~ei=:i==~===========:i--~=li'************ REMARKSo _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I COPY OF ST$ SEHT TO 8CNEDULINO: s~ct I ON/iflllalfD\111 HRS sel:floti/Rli!oUI! m SECT !ON/-NIOUR HRS si;moN/liMfNi6UR lid RECOllP$ TRANSMITTAL#: _ _ _ _ _ __ 111111111~ 11111111mu11111~ 1111~
Calculation No. WBNTSR008 l Rev: 014 j Plant: WBN I Page: 47
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 JUL-24-2000 13;52 TUA PLANT MGRS OFC 423 365 1904 P.03/07 Lv~JuM-iooo 09:23:05.19 TENNESS~ VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE 01 - RCS - GASEOUS ACTIVTY V"' FILE IDENT DKl3600:[TVA,SAMPLE,CHEM.NEW)W0007108796_C401.CNF1l SAMPLE ID W0007108796_C401 .,/"
- OPERATOR LI.MANNON!:
SAMPLE.TIME 10-JUL-2000 08:53
- SAMPLE GEOMETRY GMlK
- SHli:l.F HE!GHT 0
- EFFICIENCY FILE GMlKO _..
SAMPLE rYPE ; 1240 CC GAS MAAI
- SAMPLE QOANTITY : 1,00000E+OO CC ACO DATE & TIME 10-JUL-2000 09:12
- DEADTIME (%) : 0. 3\ ,,,..,.
PRESET LIVE TIME : 0 00:10:00 ELASPED k~AL TIME : 0 00:10:01
- SENSITIVITY : 4.00000
- GAUSSIAN SEN : 10.00000 ELJl.PSED LIVE TIME 1 0 OO:lOrOO
- NllR ITERATIONS : 10
/
DETECTO~ DET #3, GSS-3286
- LIBRARY NOBLEGAS EFFIC CAL DATE 29-JUL-1994 13:47
- EFFIC CERT DATE 29-JUL-1994 13:47 DCAL DATE ~ TIME 9-JUL-2000 15:52:
- ENERGY TOLER 1.25 KEV/CHAN 4.99928!-0l
- HALF LIFE RATIO 8.00000 OFFSET -1,48334E-Ol keV
- ABUNDANCE LIMIT 1 80.0\
Q COEFFICl~NT 3.22120E-OB
- CORRECTION FACTO~ l.OOOOOE+OO PEAk START CHAN 140
- PEAK END CHAN : 4096
***********************************~*******************************************
ANALYSES I PEAR Vl6.9 NID V3.3 MINACT v2.e WTMEAN/KEY Vl.8
&L COON'l'ED ON : LION COLLECTED BY :
COlJHTED BY : LLMANN"ONE REVIEWED BY : _ _\......__..,,....._.....,,....,.__.,,c:::._...-. COMMENTS : o::::::::=>" Post-NID Peak Search Report It Energy Area Bkgnd FWHM. Channel Left Pw \En Fit Nuclides 0 Sl.12 230 92 1. 01 162.55 157 12 10.8 XE-133 0 151.37 88 53 0.97 303.07 299 8 17.7 0 196.38 KR-85M 70 80 0.88 393.10 388 10 26.7 0 227.79 22 47 1.35 455.92 448 10 62.9 0 249.81 549 83 0.94 499.98 494 12 S.4 XE-135 0 258. 71 72 40 0.89 517.78 514 10 20.2 0 305.21 XE-138 19 34 0.76 610.77 607 8 57.2 XR-85M 0 402.80 48 34 2.39 805.96 801 12 29.3 ltR-87 0 435.l.9 2'6 16 l.31 870.75 866 11 35. 3 Xl!:-138 0 511. 06 390 49 2.28 1022.51 1014 lS 6.6 0 526.45 39 24 1.21 1053.28 1048 12 29,8 XE-135M 0 898.31 22 16 1.29 1796.96 1791 9 38.8 0 1293.58 904 9 1.60 2587.40 2578 16 3.4 AR-41
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 48
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 JUL-24-2000 13:53 TVA PLANT MGRS OFC
. REPORT DA'I'E 10-JUL-2000 09:23 REOUESTOR ' LLMANNONE TENNESSEE VALLEY AUTHORITY WATTS B.llll NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - GASEOUS ACTIVTY SAMPLE No. W0007l08796 C401 OPERATOR NAME LLMANNONE SAMPLE TYPE 1240 CC GAS~l SAJllPLE GEOMETRY GMlK COUNT TIME 10-JUL-2000 09:12:51 SAMPLE QUANTITY l.OOOOOE+OO SAMPLE TIME lO-JUL-2000 08:53:00 DETECTOR tlli:T fl3 I GSS-3286 LllJRARY NOBLEGAS PEAK ENERGY DECAY CORR ISOTOPE AR-41 ENERGY 1293,64 DIFF (KEV)
-0.06 uci/CC l.JOJE-02 COMMENTS OA Results OK JCR-85M 151.18 0.19 l.SllSE-04 0~ Results OK KR-87 402.56 0.22 4.5751!:-04 QA Rl'fi!n,1.l ts OK XE-133 81. 00 0.12 9.565E-04 OA llesults OK XE-135 249.79 0.02 l. 4291-03 OA ResultrJ OK XE-l35M 526.56 -0.11 7. 364E-04 OA Results OR XE-138 258.31 0,40 l . 796!:-03 QA Results OK AVG DIFF 0.11 ENE~GY ~ l.BS!ilE-02 = TOTAL GAMMA ACTIVITY O.OODE+OO = Total IlGL Activity l.859E-02 'l'otal Gas Activity
_.._ ______ NET ENERGY AREA FWHM GAMMA/SEC GAMMA/SEC
/CC POTENTIAL IP ACTIVITY 196.38 70, 0.88 ---------
......... -~--
- 4. 728!!:+00 4.72SE+OO R KR-8~'5,438E-04
~--------
227.79 22. 1.35 1.675E+OO 1.675!+00 U TE-132 5.163E-OS U CS-138 5.llOE-03 U NP-232,..-- 4.252E-04 511. 06 390. 2.28 7.016E+01 7. 016E:+Ol U ANNIL O.OOOE+OO 898.31 22. 1.29 7.079E+OO 7.079E+OO u n>>-se- i.s12E-03 U Y-88 2.049E-04
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 49
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 JUL-24-2000 13:53 TVA PLANT M~S OFC 423 365 1904 P.05/07 10-JUL-2000 09:37:45,71 TENNESSEE VALLEY AUTHORITY WATTS BAll NUCLEAR PLANT SAMPL}!; TITLE Ul - RCS - DEGASSED LIQUID ACTIVITY FILE !DENT DKB600:[TVA.SAPIPLE.CHEM.NEWJW0007108795_C402.CNF1l SAMPLE ID W0007108795 C402
- OP!l!.ATOR LLMANNONE SAli!PLE TIME 10-JUL-2000-07: 25 _..
- SAMPLE GEOMETRY LSV20
- SHELF HEIGH~ 1
- EFFICIENCY FILE LSV201 .---
SAMPLE TYPE : RCS 20ML LSV
- SAMPLE QUANTITY 5.00000E+OO GRAMS
**********~********-************************************************~*~********
ACQ DATE & TIME : l0-JUL~2000 08:36
- DEADTIME (%) 2.2\
PRESET LIVE TIME : 0 01:00:00
- SENSITIVITY 4.00000 ELASPED REAL TIME : 0 01:01:20
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME : 0 Ol:OO:oo
- NBR ITERATIONS ; 10 Dl!:TE:CTOR DET #4, GSS-3310
- LIBR.AR.Y RCSLigu:r.a-*
EFFlC CAL DATE S-AUG-1994 ll:ll: * !!'!'IC CERT DATE S-AOC-1994 ll:ll: DCAL DAT£~ rIME 1 9-JUL-2000 15:52:
- KNERGY TOLER 1.25 KEV/CHAN 5.00474E-Ol
- HALF LIFE RATIO 8.00000 OFFSET -3.73924E-01 keV
- ABONDANCE LIMIT 80,0\
Q COEFFICIENT -l.14092!-07
- CORllECTION FACTOR 1 l.OOOOOE+OO PEAK START CHAN 140
- PEAK END CHAN : 4096 ANALYSES : PEAK V16.9 NID V3.3 MINACT V2.8 WTMEAN/X'.E:Y Vl.8 k
***************************************************************w******w********
COON'rED ON : LION ~ COLLEC'l'ED BY : COUNT!!!D BY : LLMANNONE REVIEWED COMMENTS BY ~~-~.:.....::~--=~~~~Qo...--'-~..;;.;;;==;;;:_~ c::::::::=;::oo
*****-*********************************************~***************************
Post-NID Peak Search Report It Energy Area Bkgnd FWHM. Chemnei Left PW \Eri: Fit Nuclides 0 135.60 697 25268 0.92 271. 70 269 7 38.0 I-134 0 249.64 887 20225 0.79 499.61 498 6 25.6 0 287.87 XE-135 540 22521 0.98 576.01 574 7 46.3 I-135 I-135 0 0 364.21 405.43 455 310 13323 6932
- 1. 06 728. 61 726 a 44.3 I-131 1.05 810.99 808 8 46,9 I-134 0 417.67 485 6009 1,20 835,46 832 8 28.1 I-135 0 462. 73 545 5368 1.28 925.53 922 8 23.7 0 511. 00 CS-138 638064 22601 2.65 1022.01 1014 19 0.1 F-18 0 522.65 824 9PP 1.38 1045.31 1043 7 7.3 l 526.58 1048 I-132 823 l.2'il 1053.15 1050 18 5.2 1-. 06E+OO I-135 l 529.88 4510 XE-135N 1009 1.34 1059.77 1050 18 l.9 I-133
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN j Page: 50
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 JUL-24-2000 13:53 TVA PLANT MGRS OFC 423 365 1904 P.06/07
.Kta:'V.K".I." LIA"l".t!< I J.U-JUL-ZOOO 0!1: 37 REOUESTOR LLMANNONE TENNESSEE VALLEY AUTHO~ITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - DEGASSEP LIQUID ACTIVITY SAMPLE No. W0007108795 C402 O!'DATOR. NAlllE LLMANNONE SAMPLE TYPE 1 RCS 20ML LSV SAMPLE GEOMETRY LSV20 COUNT TIME 10-JUL-2000 08:36:15 SAMPLE QUANTITY S,OOOOOE+OO SAMPLE Tl.ME lO-JUL-2000 07:25:00 DETECTOR DET #4, GSS-3310 Lil3RARY RCSLIOUID ISOTO:PE PEAK ENERG! _______ __
ENERGY DIFF (KEV) DECA! CORR uCi/GRAM COMMENTS
----YT""-- ----------
F-18 511. 00 o.oo 1. l79E-Ol. QA Results OK NA-24 l36B.53 0.07 9 .169E-04 OA Results OK MN-56 1810.69 -0.63 9.313E-05 OA !iesults OK C0-58 810.76 -0.01 S.Ol.9E-04 OA Results OK NB-95 765.79 0.89 3.l32E-05 OA Results OK I-131 364.48 -0.27 6.070E-05 OA Results OK I-132 667.69 0.03 1,459E-03 OA Results OK I-133 529.87 0.01 8.208E-04 QA Results OK I-134 847.03 -0.05 2.694.E-03 OA Results OK I-135 1260.41 0.03 1. 608E-03 OA Results OK XE-13S 249.79 -0.16 9.914E-OS QA Results OR XE-135.K 526.56 0.02 1. 406E-02 OA Results ox CS-138 1435.86 -0.ll 2..395E-03 QA Results OK AVG ENERG! DIFF =
-0.01 ---------
l.426E-Ol TOTAL GAMMA ACTIVITY 1.218£-01 = Total DGL Activity 2.427E-03 = Total FP Activity 1.Sl2E-03 ~ctal AP Activity
- 1. 415E-02 Total Gas Activity 6.643E-03 ~ Total HFP Activity
. :s,t.;~G-.l Dose Equivalent Iodine-131 ~ 2.905E-04 Iodine 131/133 Ratio 7.395E-02 D&~;I ~
Iodine 133/135 Ratio ~ 5.lOSE-Ol 287.87 xev Paak wae used in identifyinq 2 isotopes S26.58 KQV Pel!.k was used in identifying 2 isotopes 546.88 Rev Peak was used in identifying 2 isotopes 766.69 KeV Peak was used in identifying 2 isotopes 810.75 KeV Peak was used in identifyin9 2 isotopes 846.97 Kev Peak was used in identifying 2 isotopes 857.15 KeV Peak was used in identifying 2 isotopes 1136.29 Kev Peak was used in identifying 2 isotopes
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 51
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 JUL-24-2000 13:53 TVA PLFINT MGRS OFC 423 365 1904 P.07/07
-lt.C.J:"U~'.l. UA'.1:.1!> ~V-JUL-~OUU U~:37 REQUESTOR LLMANNONE TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS ONIDENTIFIED/REJ'ECTEn PEAKS POTENTIAL NET AREA
-------.... ----- .....-- ------ l.1431::+0 FWfil'r
\\ ERROR FLAG ACTIVITY E~G'l ID 6Bl.90 334. 2.71 2.285E+OO 25.7 u 1:-;a.1.--~--~,;--------
684.87 217. 2.04 7.446E+OO ,469E+00 31.1 R W-187 f-6" l.448E-04 1288.56 38. 1.81 2.239E+OO .478E-01 36. ~a,,.)#,_) ...... - 1291.07 30. 1. 03 l.812E+OO 3.624E-Ol 41. R FE-5i 2.244E-05 1566.76 75. 2.12 5.204E+OO 1. 41E+OO 25.3 u 1-1!>~ 1835.61 31. 1.08 2. 411E+OO 4.8 2E-01 36.4 R llB-8~" 9,196E-Ol'i u Y-88 l.312E-05 Total Unidentified/Rejected Peak ~ 6
\ Unidentified/Rejected Peaks ~ 0.17
~lags: u - Unknown Line R - Rejected During Analysis P Positively Identified (lin not in analysis libiaty)
TOTAL P.07
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 52
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 (continued) Surveillance test l-SI-68-28 performed on 4/9/01 SURVEILLANCE TASK SHEET (SPP*8_2) PAGE _1_ OF 2:1 WORK ORDER: 010028300 SI KEY: P0531 PROCEDURE#: 1-SI-68-28 TITLE: PRIMARY RADIOCHEMISTRY REQUIREMENTS PERF SECT: CEM TEST REASON: PERIODIC PERFORMANCE _ _ _ _ _ N/A.~~~--- _ _ _ _ _ _ N/A QUE: 04/09/01 AUTHORIZATtON TO BEGIN: SRO DATE ~ UBN EXT: 04/1 D/01 d? l*1 I MAX EXT: FORM SPP-8_2-2 FREQ: U I -2J:.':(__£_
/
EQ: N START CATE TIME ASME XI: N c'IP~/ APP MODE : 123 PERF MOOE: 1234 /500 SUBSQNT RVWS: PETION DATE ~ INSTRUCTIONS: Do NOT l;Otart prlor to scheduled due date TEST PERFORMERS x UAS THIS A COMPLETE OR PARTIAL SIGNATURE !NIT SECT PERFORMANCE? (EXPLAIN 11 PARTIAL 11 JN REMARICS) COMPLETE: PARTIAL: _
-~T-+~~.,,.'9'~'-"N'-----~. ~
~/RF~~'//£......, _ _ IX a"'
UERE ALL OTHER ACCEPTANCE CRITERIA SATISFIED? YES:_ NO:_ NIAL ALERT SCHEDULING REQUIRED? YES:_ NO:_ N/A:_X_ CHEM VINDliEV
<td~ DATE
====================~**-*--****======;;:======-====~
REMARKS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ COPY OF STS SENT TO SCHEDULING: 57 INITIALS
/~ OATE SECTION/#HEN/DUR HRS SECTION/#l!EN/DUR HRS SECTION/#HEN/DUR HRS SECTION/#HEN/DUR HRS RECORDS TRANSMITTAL#:. _ _ _ _ _ __
Illllll lllll llll llllll lllll lllll llll lllll 111111111111111111111111111111111
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 53
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 9-APR-2001 14:25:02.67 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - GASEOUS ACTIVTY FILE IDENT : DKB600:[TVA.SAMPLE.CHEM.NEW]WOl04095767_C401.CNF;l SAMPLE ID W0104095767 C401
- OPERATOR DRKERNS SAMPLE TIME 9-APR-2001 l3:25:
- SAMPLE GEOMETRY GMlK
- SHELF HEIGHT 0
- EFFICIENCY FILE GMlKO SAMPLE TYPE : 1240 CC GAS MARI
- SAMPLE QUANTITY : 2.51000E+OO CC ACQ DATE & TIME 9-APR-2001 14:14:
- DEADTIME (%) 0.1%
PRESET LIVE TIME 0 00:10:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 00:10:00
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 00:10:00
- NBR ITERATIONS 10 DETECTOR DET #4, GSS-3310
- LIBRARY NOBLEGAS EFFIC CAL DATE 2-AUG-1994 11:26:
- EFFIC CERT DATE 2-AUG-1994 11:26:
DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1.25 KEV/CHAN 5.00516E-01
- HALF LIFE RATIO 8.00000 OFFSET 1. 44837E-01 keV
- ABUNDANCE LIMIT 80.0%
0 COEFFICIENT PEAK START CHAN . 140
-1.10914E-07
- CORRECTION FACTOR
- PEAK END CHAN 1.00000E+OO 4096
'ti UL ANALYSES : PEAK Vl6.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8 COLLECTED COUNTED ONBY : .
COUNTED BY : RN , ~ REVIEWED BY : COMMENTS : Post-NID Peak search Report It Energy Area Bkgnd FWHM Channel Left Pw %Err Fit Nuclides 0 80.94 902 374 1.03 161.44 157 10 5.2 XE-133 0 151.17 294 315 1.07 301.75 297. 10 12.8 KR-85M 0 166.01 57 226 1.10 331.42 328 8 47.7 KR-88 0 196.08 202 345 1.03 391. 51 387 10 18.5 KR-88 0 249.77 2313 340 1.12 498.78 494 12 2.6 XE-135 0 258.57 59 201 1.04 516.37 513 8 43.2 0 402.62 160 50 1.21 804.27 800 8 11.0 KR-87 0 510.99 7378 174 2.34 1020.88 1013 19 1.2 0 526.86 66 31 1. 08 1052.60 1048 11 20.7 XE-135M 0 609.00 34 35 0.87 1216.79 1209 16 43.6 XE-135 0 677. 82 14 17 1.85 1354.36 1348 9 61.5 0 834.68 38 12 1. 75 1667.96 1662 12 24.5 KR-88 0 897.59 24 13 2.20 1793.76 1789 9 34.3
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 54
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 REPORT NAME QA CHECK (Vl0.4) PAGE : 1 REPORT DATE 9=APR-2001 14:25 REQUEST OR DRKERNS TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - GASEOUS ACTIVTY SAMPLE No. W0104095767 C401 OPERATOR NAME DRKERNS SAMPLE TYPE 1240 CC GAS-MARI SAMPLE GEOMETRY GMlK COUNT TIME 9-APR-2001 14:14:53. SAMPLE QUANTITY 2.51000E+OO SAMPLE TIME 9-APR-2001 13:25:00. DETECTOR DET #4, GSS-3310 LIBRARY NOBLEGAS PEAK ENERGY DECAY CORR ISOTOPE ENERGY DIFF (KEV) uCi/CC COMMENTS AR-41 1293.64
-0.10 2.696E-03 QA Results OK KR-85M 151.18 -0.01 2.013E-04 QA Results OK KR-87 402.58 0.04 4.809E-04 QA Results OK KR-88 196.32 -0.24 4.982E-04 QA Results OK XE-133 81.00 -0.05 1. 202E-03 QA Results OK XE-135 249.79 -0.03 1. 676E-03 QA Results OK XE-135M 526.56 0.30 l.105E-03 QA Results OK AVG ENERGY DIFF =
-0.01 7.859E-03 TOTAL GAMMA ACTIVITY O.OOOE+OO Total DGL Activity 7.859E-03 Total Gas Activity
_______ ------ GAMMA/SEC POTENTIAL ENERGY NET
..... AREA FWHM GAMMA/SEC /CC AG ID ACTIVITY 258.57 59. 1.04 3.448E+OO l.374E+OO
---~-
R XE-138
- 1. 724E-03 510.99 7378. 2.34 7.643E+02 3.045E+02 u -1 6.016E-03 u ANN IL O.OOOE+OO 677.82 14. 1.85 l.829E+OO 7.289E-01 u AG-llOM 1. 845E-04 4.949E-04 897. 59 1835.84 24.
17. 2.20 2.13 4.089E+OO 1. 629E+OO 5.418E+OO u~ u u U
-8 RB-88 3.932E-04
- 4. 716E-05 3.408E-04 u - 5. 872E-05
Calculation No. WBNTSR008 1Rev: 014 J Plant: WBN J Page: 55
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 9-APR-2001 10:48:39.98 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY FILE !DENT : DKB600:(TVA.SAMPLE.CHEM,NEW]W0104095766_C402.CNF;l SAMPLE ID W0104095766 C402
- OPERATOR WNCLONTZ SAMPLE TIME 9-APR-2001 08:20:
- SAMPLE GEOMETRY 65ML
- SHELF HEIGHT 1
- EFFICIENCY FILE 65ML1 SAMPLE TYPE : RCS 65.ML BOTTLE
- SAMPLE QUANTITY l.58100E+Ol GRAMS ACQ DATE & TIME 9-APR-2001 09:45:
- DEADTIME (%) 4.6%
PRESET LIVE TIME 0 01:00:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 01:02:52
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 01:00:00
- NBR ITERATIONS 10 DETECTOR DET #4, GSS-3310
- LIBRARY RCSLIQUID EFFIC CAL DATE 19-JUL-2000 20:26
- EFFIC CERT DATE 19-JUL-2000 20: 26 DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1.25 KEV/CHAN 5.00516E-Ol
- HALF LIFE RATIO 8.00000 OFFSET 1. 44837E-01 keV
- ABUNDANCE LIMIT 80.0%
0 COEFFICIENT -1.10914E-07
- CORRECTION FACTOR 1. OOOOOE+OO PEAK START CHAN 140
- PEAK END CHAN 4096
/!IL ANALYSES : PEAK Vl6.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8 COLLECTED COUNTED ONBY COUNTED BY
~. LON Z ,
REVIEWED BY : COMMENTS : Post-NID Peak Search Report It Energy Area Bkgnd FWHM Channel Left Pw %EII Fit Nuclides 0 134.65 991 52495 0.82 268.75 266 7 38.5 W-187 I-134 0 249.78 2586 43764 1.22 498.80 496 7 13.6 XE-135 0 364.48 569 19365 0.96 728. 04 726 6 39.0 I-131 0 405.26 542 12870 1.06 809.53 807 7 35.0 I-134 0 433,34 305 11020 0.97 865.66 863 7 57.4 I-134 0 462.88 898 11840 1. 08 924.71 921 8 21.3 CS-138 0 478.53 2028 19343 2.93 955.98 948 13 14.3 BE-7 W-187 0 510.97 1460805 45497 2.66 1020.82 1013 18 0.1 F-18 2 0 522. 71 1163 2480 1.33 1044.29 1041 8 a.a I-132 526.52 1531 2060 1.17 1051. 91 1048 16 5.5 9.34E-01 I-135 XE-135M
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 56
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 9-APR-2001 10:48:39.98 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT SAMPLE TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY FILE !DENT : DKB600:[TVA.SAMPLE.CHEM,NEW]W0104095766_C402.CNF;l SAMPLE ID W0104095766 C402
- OPERATOR WNCLONTZ SAMPLE TIME 9-APR-2001 08:20:
- SAMPLE GEOMETRY 65ML
- SHELF HEIGHT 1
- EFFICIENCY FILE 65ML1 SAMPLE TYPE : RCS 65ML BOTTLE
- SAMPLE QUANTITY : l.58100E+Ol GRAMS ACQ DATE & TIME 9-APR-2001 09:45:
- DEADTIME (%) 4.6%
PRESET LIVE TIME 0 01:00:00
- SENSITIVITY 4.00000 ELASPED REAL TIME 0 01:02:52
- GAUSSIAN SEN 10.00000 ELAPSED LIVE TIME 0 01:00:00
- NBR ITERATIONS 10 DETECTOR DET #4, GSS-3310
- LIBRARY RCSLIQUID EFFIC CAL DATE 19-JUL-2000 20:26
- EFFIC CERT DATE 19-JUL-2000 20:26 DCAL DATE & TIME 9-APR-2001 02:40:
- ENERGY TOLER 1.25 KEV/CHAN 5.00516E-Ol
- HALF LIFE RATIO 8.00000 OFFSET 1. 44837E-01 keV
- ABUNDANCE LIMIT 80.0%
Q COEFFICIENT -1.10914E-07
- CORRECTION FACTOR 1.00000E+OO PEAK START CHAN 140
- PEAK END CHAN 4096
;IL ANALYSES : PEAK Vl6.9 NID V3.3 MINACT V2.8 WTMEAN/KEY Vl.8 coUN**n oNBY COLLECTED COUNTED BY
~. LO Z ,
REVIEWED BY : COMMENTS : --= Post-NID Peak Search Report It Energy Area Bkgnd FWHM Channel Left Pw %Err Fit Nuclides 0 134.65 991 52495 0.82 268.75 266 7 38.5 W-187 I-134 0 249.78 2586 43764 1.22 498.80 496 7 13.6 XE-135 0 364.48 569 19365 0.96 728. 04 726 6 39.0 I-131 0 405.26 542 12870 1.06 809.53 807 7 35.0 I-134 0 433.34 305 11020 0.97 865.66 863 7 57.4 I-134 0 462.88 898 11840 1. 08 924. 71 921 8 21.3 CS-138 0 478.53 2028 19343 2.93 955.98 948 13 14.3 BE-7 W-187 0 510.97 1460805 45497 2.66 1020.82 1013 18 0.1 F-18 0 522. 71 1163 2480 1.33 1044.29 1041 a 8,0 I-132 2 526.52 1531 2060 1.17 1051,91 1048 16 5.5 9.34E-Ol I-135 XE-135M
Calculation No. WBNTSR008 \Rev: 014 I Plant: WBN I Page: 57
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 REPORT NAME QA CHECK (VlO. 4) PAGE 1 REPORT DATE 9=APR-2001 10:48 REQUEST OR WNCLONTZ TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT POST NID QA ANALYSIS TITLE : Ul - RCS - DEGASSED LIQUID ACTIVITY SAMPLE No. W0104095766 C402 OPERATOR NAME WNCLONTZ SAMPLE TYPE RCS 65ML BOTTLE SAMPLE GEOMETRY 65ML COUNT TIME 9-APR-2001 09:45:36. SAMPLE QUANTITY l.58100E+Ol SAMPLE TIME 9-APR-2001 08:20:00. DETECTOR DET #4, GSS-3310 LIBRARY RCSLIQUID PEAK ENERGY DECAY CORR ISOTOPE ENERGY DIFF (KEV) uCi/GRAM COMMENTS 511. 00
-0.03 l.116E-01
-----------------------~--
QA Results OK F-18 NA-24 1368.53 0.05 2.060E-03 QA Results OK MN-56 1810.69 0.33 2.0SSE-04 QA Results OK C0-58 810.76 o.oo 6 .218E-04 QA Results OK C0-60 1173.22 0.28 2. 776E-05 QA Results OK NB-95 765.79 0.46 2.794E-05 QA Results OK I-131 364.48 o.oo 3.881E-05 QA Results OK I-132 667.69 0.00 1.165E-03 QA Results OK I-133 529.87 0.01 6.105E-04 QA Results OK I-134 847.03 -0.04 2.334E-03 QA Results OK I-135 1260.41 -0.03 l.158E-03 QA Results OK XE-135 249.79 -0.02 l.380E-04 QA Results OK XE-135M 526.56 -0.04 1. 972E-02 QA Results OK CS-138 1435.86 -0.20 2.195E-03 QA Results OK AVG ENERGY DIFF = 0.06 1.419E-01 TOTAL GAMMA ACTIVITY 1.168E-Ol = Total DGL Activity 2,223E-03 Total FP Activity 2.918E-03 Total AP Activity
- 1. 986E-02 = Total Gas Activity 5.307E-03 = Total HFP Activity Dose Equivalent Iodine-131 = 2.098E-04 Iodine 131/133 Ratio 6.357E-02 Iodine 133/135 Ratio = 5.274E-01 134.65 KeV Peak was used in identifying 2 isotopes 478.53 KeV Peak was used in identifying 2 isotopes 526.52 Kev Peak was used in identifying 2 isotopes 546.50 xev Peak was used in identifying 2 isotopes 766.25 KeV Peak was used in identifying 2 isotopes 772.62 Kev Peak was used in identifying 2 isotopes 810.76 KeV Peak was used in identifying 2 isotopes 835.61 KeV Peak was used in identifying 2 isotopes 846.98 KeV Peak was used in identifying 2 isotopes
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN l Page: 58
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 AU. WOIU(/TESTfNQ COMl'l..ITE.
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 59
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 tf:- ~~:~~~~~~~~~ ir~~~~~~~~~~::::::iiiiii::B~ f~~
- f
¥. -
p ...... '* *='*n t.. I
~
'rJ..-*./f!A-
....,r
~,
;l.
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 60
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
~
*A l/8ZD.2Q
" * --'-- of _j_
1lll'% JrllC'?tOll nsm
...,~
.ADIUSS .
~
Ra\11. ...... --.,. -
- - - . -- ..- - UICA%IOM IQml'!mr!
-o
~
!1...: A. !JfJe.~ Jt;a ""_Tc. t .-: ...... "A ;..e.
- _ _ _ ,.., , _ .. t4w
~FCl/:!t. ~/!=;
a-.~ Frv*--- I ::"t .. 4-A ...... JI If ~-.r--- --;:,-'1
** 1,!-;:I_
-~a ~~ ..-
~
0 CJ-6* JI I} Ft.fl 31 ft Jt
~o Fen .31 10-A ti
- 11 .
.31 1/,-A n
() II Fr:n (j ~Fen ~ 1 17-A 0 iJ If
'F-rn .25 II 0 Frt"'J ?A 11 II I(
~I 0 Frv ?/ -:s J.-11 I{
"n
(') Frv ?I tr II
~7-A 11 0 FCV ~I ?OJ./. cc I\
$1
Calculation No. WBNTSROOS I Rev: 014 l Plant: WBN I Page: 61
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 Calculation No WSNTSR-008 Rev 12 Plant WBN Page 59
+ff;
,....111**
Fcv.:31 -'I OPE/II FCO ...JI .. 'l OP£N Fco- JJ* JO OPSN FCo- JI* JI,, OP£N F-Co- JJ*J? ()PIN Fco . . 31 ... z.s {)/1£11
- FCO-Jl -z /, ()ft#
- FCV*.31-3& . __ oPVJ Fc\l*JI* 31 _ OfEN
- - FC'l-Jl-ZOI/ ___ OPEN-
Calculation No. WBNTSR008 j Rev: 014 I Plant: WBN I Page: 62
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
;.v II * .Ar -
.ll 1- . ,, ESTA6'1su- _i:o,.._.,.CATlo.,->>>'1'1<-HAS~~
t~~.~---::--~--~----*----*------- I
---~7-~C'f:~CZO'f- - ~ . 0 ( 111~11"%tzlfi_
- -:TL-- -
/1~ -
_8) /?rr:tJlt!) ..EC.V-31:.ZOJf-_CLO.SUttf, ...:r1~6.:,,, M 4iE. _/)A7A.- -~
----.STDA.t.mK -ZD .*4t qa -i.s- 't , _ *
-----1CAt. o. . , _D.tra_. 7-/f -S't _ _ , _____ _
-----Citlstt&I ..TIM!._ a, !f 6 SE~ . _ - - -
_ _ _ _ _ _ _ _ _ ___.i:.>..,r.t 7Att.£H_Br Bbl- "5./L+...M'- /1-i.ls/114'
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 63
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 A-A"a s-a- DtV-~,..----."""" ON __~~---'-.p,f:.¥29.. OJI -~-n----lllllU=-
- .:*1-::..:_-s;~~~ --s~; F~" *-:::..:.. ... ?2 OAJ
_...-.,-1 ;;i5u~; i_-:. >- ~ (Ja. __.. ~ ~ -..r.
- _, ~W."' A/ --
..... ~~~1,....rM ,.....,. RS*.Jl*ZD'l.,1 ~v.1r --ia~-----.~""""
.:r ... .. * * -
- -**~rC.'/*.31-3 _. . :.,* ** .. -;;....=.....~ riiJT...~-"Zr-""'--~~---""""-'...,.-..
I*' ~
~..._~7~;;-,.~oa-a.1
..:.--......:......-:~-='"'.. ~ ......~ --- ~.....
FC0-J1-"'zs cPEN
... _,, f#!'\!:,..-.....tr: .., ~ ........ "\. - ... *~ * "'~ .,._ ~
- J:c.o*..31*Z6 O/Jtl/ ---41~---..iU!:!.~
Fco-31-10 OP£N Fc.o-.;1 . . 11 Ol'aJJ Fcf-Jl-..37 OPEN Fer* J1-z.01t OPEJ.J II) f3m1JLJ.SJI . co'"'""""'"'1170N IJITH
. .tT FCV-.31-~0Jf.
/tJ.la/s'IJ J1z-/Ja _
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 64
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
~
/II) V£i.~~ 7H£ FDl:.L.OWllJG FCV-31 ...-'f Fco-31 ... '1
. FC0-31-/h
- ......... 'I... - .. ' ...... """'
lo) REJ:oUJ Fey- 31-.2011 CJ.o.suu. 711"1s. ~D /11 i TE. ~rA
- - - *
- S7WJ.Mrcil ID '*--24,;o::;_:"Z-.~S~'...J,t..--
- _CAL~ /Mrs. *7 -{Cf.* ett __ _
- IJ.as~_:n/'14 8. 2Z see
- - _ _ _ _ _ _ _ _ _ _ JJ,,,, ~.Br iij;J-:.;,&~1'blr1tJ
_ _ _ j,) ./l..bir_J-:H5:..3}:./77E
.** w
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 65
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
*~ -
\~ .
~. , _______.......tt-r:.--r1..11<0.--.,--....
- - :_ 7U _6£ - At.J~O A'7" _ TR~JIL _DIMIU:t'J*Jfil-
- .. ~-=-=- t - ---- - ---.....;,,:,~'11:.o'"'~""'rf.._11_.,.,.
* - ,, -- /
' -~- --- "":'_ .... --- .
l') Ks.t:IJlf.O 7'f£. CUI.Si.LC n""ts t:o~ TJ& FOU.oww& MM~ll.S AS JfltLU-..UD &t' 7Hll. .$11C1' e'.MAll.7.$. (E.r. *.,,,.. J:cV"-JJ*l/)l/. WUltHI '-'A.S M<4¥ttAD d)" $~A.Jtlf>Tt:N) A.. r ...,.i eta:.. s- 7Atiru cA..I
-- * .;.. Fc.v 3 ~ 12 If3 ,., .:~ Rr-31-__'if 1 l!S
_:1.... src.
... .,.,... 1 - - --.. w ""' - ,,._ - ..,._ - -
~ ~-:.St~t~~~ 9,6,'1_ ~~c:i:tff~~t;'J;.-10,9if -* .u1: L~ ~;:rr::
~~~~," ~F.._c.o::.31-:J?** 71h3 . .s6;,~rccr.3l*/b ~Z69 >-s,8(!-~-~ l"-
~.r:..--~FcY~ii~#-~73,Sh ;;C.n.T!--;.;;.=-.iJ-.3b -LZ,.3Z. S4~-- * -
................Fco-.31_.~,..... 8,qO sic *-->:cD=.Ji-H"V . . S,S7 sic..,,. - . .
;:.co-JJ--26 .. J,st) s*c. 1=co-J1~.1' 6,11'1- s8c.
FC~-.3/-.20'/ - fJ, ¥8 .. s.sc. rCV*3r~t:PI 8,22 $4C. DtrA k<Jl'Dtto 13r ~ /1r9!6s
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 66
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 t::. Page 64
""" 1-f'I ...
*: 'i'i;
- fetr: FOttll\* ~;
.~
'5fGNlhvfl!' IN.TrrtM.. *
)~
;f;I_
Calculation No. WBNTSROOS I Rev: 014 I Plant: WBN I Page: 67
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
-P/ Ktu!,
~ ~
YkM ! ' ., V.KbL~ :*~ £ I~ 3 ,_,,..
~ <<#-1ull 3 9 .. l)af
~ S' /tllS 11-/t{dt
Calculation No. WBNTSR008 I Rev: 014 I Plant: WBN I Page: 68
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011 A1r11tllltf8~11 ,1. ' if1' .3 ,f 'f el ~Mt GJ,l.l
,,,.,.,Id
~JIJ Vkll2f 3 3*,N
~? J-J~
ti> ,...itl'i 3 ,..,.... fWlt t..t.1-~ .. .)*3M
,_.,.nlll fllfli I
I '
Calculation No. WBNTSROOS j Rev: 014 I Plant: WBN I Page: 69
Subject:
Control Room Operator and Offsite Doses Due to a Steam Prepared: MCB Date: 9-2-2011 Generator Tube Rupture Checked: JEB Date: 9-2-2011
'1"111* IJI /JM.4 t.JA. 't l.f/) ~
(A i/1/11 1:.
~ lllAU 3
- b0 "lo fOllL 3
Attachment 3 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 Control Room Operator and Offsite Doses Due to a Loss of AC Power Calculation (20 pages including this cover page)
NPG CALCULATION COVERSHEET I CTS UPDATE Page I REV 0 EDMS/RIMS NO. Cl'STYPE: EDMSTYPE: EDMS ACCESSION NO (NIA for REV. 0) B26 910626 20 I Calculation CALCULATIONS (NUCLEAR) T9 ~ 1 ~ 1 n? A n f\ L Cale
Title:
Control Room Operator and Offsite Doses Due to a Loss of AC Power ORG Pl.ANT BRANCH NUMBER CUR REV NEW REV CAI.CID NOC WBN NTB WBNfSR080 011 012 Cl'S UPDATE ONLY 0 (Verifier and Approval Signatures Not Required) II NO Cl'S CHANGES D (For caJc revision, Cl'S bas been reviewed and no Cl'S changes required) UNIT (check one) SYSTEMS UNIDS 0 181. 1 D. 2 D. 3 D NIA NIA DCN,EDC,NLA APPUCABl.E DESIGN DOCUMENT<S) CLASSifl~ATION NIA NIA E OUAUTY SAFETY REIATED? UNVERlFIED SPECIAL Rf.QUIREMENTS AND/OR DESIGN OUTPUT SAR/TS ll!lll2r ~fSI REIATED? (If yes, QR= yes) ASSUMPTIQN YMmNQ OONUITIQNS? ATTACHMENT? SAR/CoC AFFECTED YesOO No0 Yes 181 NoD Yesn No181 YesO No181 YesD No181 Yes~ Nor l CALCULATION NUMBER REOUESTOR PREPARING DISCIPLINE VERIFICATION MEI'HOD NEW METH~ QF ,AHALY!.!~ Name:N/A PHONE: NIA BechtelMEB Design Review DYes 181No Jd PREPARER (PRINT NAME AND SIGN) DATE CH13CKER (PRINT NAME AND SIGN) DATE DWWu lt>/IJ/,3 MJBramer /11..ot-~~-*~ 10/..,/1~
\.' NW'mtt \r.IW'l l NAME AND SIGN) DATE APPROVAL ~*.n .v.i.. SIGN)lfJV DATE MJBn:nner ~~ ro/'1/r3 ~w11' Fie~ ~I . f
~,(t..t'2A4--
lO/q_/f 3 STATEMENT OF PBPBLEMIABSTRACT A loss of AC power to the Watts Bar Nuclear Plant will result in a significant amount of steam release to the environment. This steam will contain radionuclides if a primary to secondary side leak occurs prior to the steam dump. This calculation determined the control room operator and offsite doses following a Loss of AC Powa-. The inventory of radionuclides released to the environment was determined using the secondary steam inventories in WBNNAI..3003. The amount of steam dumped was obtained from Westinghouse for the new Steam Generators. The computer code COROD was used to determine the control room operator doses. Computer code FENCDOSE was used to determine the offsite doses. The Technical Specification Limiting Case was detennined by multiplying the realistic secondary steam inventory by 41500. The results are provided in the results section. The calculated offsite doses are substantially below (<10%) the regulatory limits of 25 rem whole body, 300 rem beta, 300 rem thyroid, and 25 rem TEDE. The control room operator doses are substantially below the regulatory limits of 5 rem gamma, 30 rem beta, 30 rem thyroid, and 5 rem TEDE. The results of this calculation are direct input to FSAR Table 15.5-2 LEGIBILITY EVALUATED AND AW!l!-!:.OR -
/-fl( i?.4P~
ISSUE. t 0 (J-/t3 t$1GNATURf ~~J ,,.__ nATE MICROFICHF/EFICHE Yes181 Noll FICHE NUMBERfS) TVA-F-W003269 TVA40S32 Page 1 of2 NEDP-2-1 [10-31-2011)
NPG CALCULATION COVERSHEET I CTS UPDATE Page 2
~ ORG PLANT RVA~CH NUMBER REV NUC WBN NIB WBNfSR080 012 JlUilDING NA I ROOM NIA I ELEVATION NIA I COORD/AZIM NIA I FIRM Bechtel CATEGORJF.S NIA KEYWORDS (A-add, D-delete) ~ KEYWORD Ml KEYWORD (ND)
CROSS-REFERENCES (A-add D-delete) ACTION XREF XREF XREF
' XREF XREF (ND) CODE PLANT TYPE NUMBER REV A p WBN PER 775553 CTS QNLY UPDATES:
Followine: are reauired onlv when makin2 kevwonl/cross reference CTS undatrs and ,,...e I of form NEDP-2-l is not included: PREPARER (PRINT NAME AND SIGN) DATE CHF.CKER. (PRINT NAME AND SIGN'l DATE PREPARER PHONE NO. EDMS ACCESSION NO. TVA40532 Page2 of2 NEDP-2-1 [10-31-2011)
p*aae 3 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER: WBNTSR080 Title Control Room Operator and Offsite Doses Due to a Loss of AC Power Revision DESCRIPTION OF REVISION No. 0 Initial issue. 1 This revision was prepared to remove the FSAR as a reference and to apply the offsite dose limit of 10% of 10CFRlOO per ANSI/ANS 51.1. Also, per OICP 92-lc, the results were compared to 10CFR20 eJq>OSUre limits. The dose results as previously calculated in revision 0 of this calculation satisfy these limits. Therefore, the conclusion of the calculation remains the same. CCRIS was checked on 02/09/93, and no changes which impact this calculation were found. As no drawings were used, a OCCM review was not required. pages added: 3, 6a pages deleted: 3 pages changed: 1,2,4,9, 10, 13, 14 total pages: 15 As this calculation is not used as a design input by any other discipli.De, an impact review is not required. The calculation results have not chan1>ed. 2 Revision 2 of this calculation was performed because the X/Q values changed. All pages were rewritten for legibility and renumbered. Only actual text changes are marked with a revision bar. pages changed: all pages added: oone Da2es deleted: none 3 Revision 3 was performed because the control room makeup flow changed from 325 to 711 cfm. pages changed: 1-7, 9, 14, 15 pages added: none pages deleted: none R3total1>82es= 15 4 Revision 4 implements EDC E50629A, which implements the use of a Tritium Production Core. The calculation was rewritten and renumbered, actual text changes are marlc:ed by a revision bar. The revision revised the methodology for determining the source tenns for the steam relcme; the use of STP was replaced by using the source terms for secondary steam provided in WBNNAL3-003. The evaluations in this revision utili7.e the latest version of COROD and FENCDOSE, which calculate dose using ICRP-2 and the new ICRP-30 methodology as well as the TEDE. New x/Q's are being used from the ARCON96 methodology in addition to the Halitslcy values. Applicable changes to the FSAR are being handled via EDC E50629A Pages changed : 1-11, including 2a (old cover sheet) Pages added: all Pages deleted: all R4 total Pages = 22 5 Revision 5 is in support of the Steam Generator Replacement Project (DCN 51754). The mass releases have been changed and are thus updated in this calculation. PER 61493 was addressed in regards to operation with 2 trains of CREVS. An assumption was added that discussed the 2 trains of CREVS is beyond design basis. The FSAR and Technical Specifications impacts, if any, are addressed in the screening review for DCN 51754. FSAR Table 15.5-2 is directly affected by the revision. Pages changed : 1-10 Pages added: none Pages deleted: 2a, old page 5 (Design Verification Form), old page 7 (Output Info Sheet), Appendices A and B (2 pages), and Attachment 1 (8 pages) R4 total Pages= 10 6 Revision 6 is perfonned for a Unit 2 accident. The original steam generators are used, however Westinghouse has provided revised mass relcmes. The Unit 2 ARCON96 values were also used. The SAR has been reviewed by Marc Berg and this revision of the calculation affects Unit 2 SAR section Chapter 15 . A SAR change shall be processed in accordance with NGOC PP-10 to reflect the calculation results as part ofEDCR 54956. Tech Specs have been reviewed and determined not to be affected. Pages added: design verification form (p.S), Appendix B (p 13, 14) Pages deleted: none Pages changed: 1-8, 10-12 R6: 14 total DalZes TVA 40709 (10-20081 Page 1of1 NEDP-2-2 [10-20-2008} This page replaced by Revision 008
Page4 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER: WBNTSR080 TiUe Control Room Operator and Offsite Doses Due to a Loss of AC Power Revision DESCRIPTION OF REVISION No. 7 Revision 7 is performed to address replacement of the analog ratemeters with digital RM1000 ratemeters by DCN 52012. The longer response time of the RM1000 raterneter is incorporated for the Unit 1 accident by Increasing the control room isolation time from 20.6 seconds to 40 seconds (14 sec damper closure+ 26 sec instrument response). The response time for the Unit 2 accident wiU be revised prior to Unit 2 startup. The only change is an inaease in the control room HVAC time to isolation from 20.6 seconds to 40 seconds. Therefore, only the control room dose changes, and the offsite doses are retained from revision 5. FSAR section 15.5 and Technical Specifications were reviewed by Lynn Cowan and Table 15.5-2 is impacted by this change. See calculation WBNTSR028, Rev. 7, (ref. 19) for the revised time delay of 40 seconds. Pages Added: 4, 11 Pages Deleted: none Pages Revised/Replaced: 1, 2, 5-14, 16 R7: 16 total naaes. 008 Revision 008 is performed to address the PER 327968 and PER 327956 for the following items:
- The COROD input files in the previous revisions contained the filter removal during the first 4-0 seconds.
- Source terms in the COROD input file for Xe-l 33m and Xe-13 7 for realistic cases were 2-3 orders of magnitude higher during the first 40 seconds.
- The Technical Specification limiting case replaces the 1% failed fuel case in the previous revisions using the 0.1
µCi/g 1-131 equivalent factor based on the RG-1.109 iodine inhaJation dose conversion factors for thyroid (Ref.
23).
- Unit 2 control room doses in the realistic case were higher than those in the 1% failed fuel case mRevisions 6 and 7.
- Control room operator and offsite doses are recalculated for Unit l and Unit 2 as well as for TPC and non-TPC.
- There are no successor calculations to this calculation based on CCRIS reference list FSAR AND TF.CHNICAL SPOCIFICATIONS HAVE BEEN REVIEWED AND FSAR TABLE lS.S-2 IS IMPACTED BY nm CHANGE.
Reviewer:--------- Pages Added: none Pages Deleted: none Pages Revised/Replaced: 1-16 RS: 16 total pages.
- ** A 1 oa2e, and Appendix B 2 pages.
9 Revision 9 is performed to update the XJQ values to the 1991-2010 meteorological data set. FSAR AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND FSAR TABLE 15.5-2 IS IMPACTED BY THE CHANGE. There are no successor calculations impacted by this revision. Reviewer: Marc Berg Pages Added: none Pages Deleted: none Pages Revised/Replaced: 1, 2, 4, 6-9, 12-16 R9: 16 total Dlll!:es TVA 40709 [1~2008) Page 1of1 NEDP-2-2 [1~20-2008)
page 4a NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR080 Title Control Room Operator and Offsite Doses Due to a Loss of AC Power Revision DESCRIPTION OF REVISION No. 10 Revision 10 is perfonned under DCN 61599 to revise WBN calculations for increased average tritium permeation rates and for changes in the number of tritium-producing burnable absorber rods (fPBARs) analyzed for TPC use. This calculation is revised to update the TPC tritium source tenn. Specifically, the secondary steam design basis tritium source tenn from WBNNAL3003 is utiliud, which is based on 2 TPBAR failures in a TPC containing 2500 TPBARs with the design basis average annual tritium permeation rate (10 Ci/TPBAR/yr). The concentration of tritium in the secondary steam is used to develop the tritium soW"Ce tenns for input into STP. Control room and offsite beta doses and TEDE increase due to the larger tritium soW"Ce tenn; however, the increases do not affect the conclusions of the calculation. Control room and offsite gamma and thyroid doses are unaffected by the larger tritium source term. Since the results of this calculation are intended for use in WBN FSAR Section 15.5, a design output attachment was created and added to the calculation in Revision 10. Attachment A, Fonn NEDP-2-5, makes the entire Calculation WBNTSR080 Revision 10 design output. As a result, the calculation classification has been changed to 'EO' from 'E'. Since the TPC is bounding and is the only core configuration analyzed in this revision, statements were revised in the Introduction section and Appendix B to remove reference to control room operator and offsite dose for the standard/conventional core. Additionally, references utiliud in DCN 61599 calculations were updated to their most recent revision. Discussions of specific revisions have been removed from the Purpose section because the changes for each revision are discussed in the Record of Revision. An editorial change was made to the Introduction section to define the TPC acronym. Lastly, the title and number of appendix pages were changed/added to Appendix B to keep current with calculation fonnat guidelines. The header of all pages has been changed to Revision 10, and changes to the calculation are indicated by revision bars. CTS was reviewed for successor calculations to WBNTSR080, and no successor calculations were found. Therefore, there are no impacts due to this revision. See DCN 61599 for SAR/Tech Spec impact detennination. Pages added: 4a, 17 (Attachment A) Pages deleted: none Pages replaced: 1, 2, 6, 7 Pages changed: 5, 8-13, 15, 16 RIO: 18 total pages Attachment A- NPG Calculation Design Output (1 page) TVA 40709 [10-2008) Page 1of1 NEDP-2-2 [10-20-2008)
p'aa& 4b NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR080 Tltle Control Room Operator and Off8lte Doses Due to a Lon of AC Power Revision DESCRIPTION OF REVISION No. II Revision H is performed uoder DCN 61599 to revise WBN calculations for ioc:rased average tritium permeation rates and for changes in lhe namer of tritium-producing burnable absorber rods (TPBARs) analyzed for TPC use; Revision 11 of WBNTSR080 is an administrative revision to remove the calculation design output attachment and to reclusify lhc calculation as E.asential ("E'). Then: ue no tllclmical changes to the calculation as as result of Revision 11. Pages replaced/changed in Revision 11 contain 1111 updated page header for Revision 11 with any changes dt.noted by revision bars. All unaffected pages retain their original headers. crs was reviewed for successor calculaUons to WBNTSROBO. and no successor calculations were found. Therefore, there are no impacts due to this nsvision. See DCN 61599 for SAR/Tech Spec impact determination. Pages added: 4b Pages delemd: 17 Paaes replaced/changed: 1. 2. S-8
. llll; l 8 &o~ pages iVl\40709 (10.2008) Page1of1 NEDP"""" [1().20.2008)
Paae4c NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR080 Title Control Room Operator and Offsite Doses Due to a Loss of AC Power Revision DESCRIPTION OF REVISION No. 12 Revision 012 of this calculation was created to address the instrument response time based on PER 775553. The longer response time is incorporated by increasing the control room isolation time from 40 seconds to 74 seconds (14 sec damper closure+ 60 sec instrument response). Successor documents have been reviewed and are not impacted by this revision. The effect of Unit 2 operation on Unit 1 margins has been reviewed with no impact. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification tern perature. IMP~CTED i:J FSAR AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND TABLE 15.5-2 IS TH?,CHANGE IN!?Jtt410~1 Reviewer: ~ * ~~""'"""" ....... .. E FROM 40 SECONDS TO 74 SECONDS.
,4~,,
Pages Added: 4c (1 page) Pages Revised/Replaced: entire document has been updated with change bars added Pages Deleted: none Total number of pages in this revision including Attachments: 18 + 1 = 19 pages AppendixA(1 page) Appendix B (2 pages) This page added by Revision 012 TVA 40709 [10-2008) Page 1 of1 NEDP-2-2 [10-20-2008)
Page 5 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNTSR080 I Revision: 12 I TABLE OF CONTENTS SECTION TITLE PAGE Coversheet I crs Update 2 Record of Revision 3 Table of Contents 5 Calculation Verification Form 6 Computer Input File Storage Information Sheet 7 Computer Output Microfiche Information Sheet 8 Purpose 9 Introduction 9 Design Input 9 Assumptions IO Special Requirements/Limiting Conditions 10 Calculations IO Results 12 Discussion and Conclusion 13 References 13 Appendix A- Original Steam Generator Results, Unit 1 (1 page) 14 Appendix B - Unit 2 Loss of AC Power (2 pages) 15 TVA 40710 [10-2008) Page 1of1 NEDP-2-3 [10-20-2008)
Page 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNTSR080 Revision 12 Method of verification used:
- 1. Design Review 181
- 2. Alternate Calculation D Verifier ~Date 10/...,/1~
- 3. Qualification Test D M.J. Brenner Comments:
The changes to the calculation described in the Record of Revision for Revision 012 have been reviewed and have been found to be technically adequate in format and content. TVA 40533 [10-2008) Page 1of1 NEDP-2-4 [10-20-2008]
Page 7 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNTSR080 I Rev. 12 I Plant: WBN I
Subject:
Control Room Operator and Offsite Doses Due to a Loss of AC Power I I Electronic storaae of the input files for this calculation is not reQuired. Comments:
~ Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file. (Any retrieved file requires re-verification of its contents before use.)
The R3 input files are stored in FILEKEEPER file# 263720 The R4 input files are stored in FILEKEEPER file # 303460 (FENCDOSE runs and ARCON96 COROD)
# 303556 (Halitsky COROD runs)
The R5 input files are stored in Fll..EK.EEPER file # 30S245 The R6 input files are stored in eFiche file TVA-F-W001413 The R7 input files are stored in FILEKEEPER file# 314530 (The word file for R7 is stored in Fll..EKEEPER file #314529) The RS input files are stored in Fll..EK.EEPER file # 3176S6 (The word file for RS is stored in FILEKEEPER file # 3176S7) The input files are stored in eFiche file# TVA-F- W002551 The RIO Word, Excel, and computer input files are permanently stored in FILEKEEPER file# 3219Sl The Rl 1 Word and Excel files are permanently stored in FILEKEEPER file# 322415 The R12 input files are permanently stored in FILEKEEPER file # 322563 (The Word and Excel files for R12 are stored in FILEKEEPER file# 322565 and 322564, respectively) 181 Microfiche/eFiche (See next page) TVA 40535 [10-2008) Page 1of1 NEDP-2-6 [10-20-2008)
*~ ~
I l NPG COMPUTER OUTPUT page 8 MICROFICHE INFORMATION SHEET Document WBNTSR080 I Rev. 12 I Plant:
Subject:
Control Room Operator and Offslte Doses Due to a Loss of AC Power Microfiche Number Description RO :WRAD-27 R2 : TVA-F-C000079 R3 : TVA-F-C000118 R4: TVA-F-C-000344 TVA-F-C000351 R5 : TVA-F-W000606 Filename Code Description TSR080FAS FENCDOSE Offsite Dose, Realistic Case - TPC TSR080FBS FENCDOSE Offsite Dose, l % Case - TPC TSR080FCS FENCDOSE Offsite Dose, Realistic Case - Non-TPC TSR080FDS FENCDOSE Offsite Dose, 1% - Non-TPC TSR080CAS COROD Control Room operator dose, Realistic Case - TPC TSR080CBS COROD Control Room operator dose, l % Case - TPC TSR080CCS COROD Control Room operator dose, Realistic Case - Non-TPC TSR080CDS COROD Control Room operator dose, l % Case - Non-TPC R6: TVA-F-W001413 TSR080FA6 FENCDOSE Offsite Dose, Realistic Case- TPC TSR080FB6 FENCDOSE Offsite Dose, l % Case - TPC TSR080CA6 COROD Control Room operator dose, Realistic Case - TPC TSR080CB6 COROD Control Room operator dose, l % Case - TPC R7: TVA-F-W0015n TSR080CA7 COROD Control Room operator dose, Realistic Case - TPC TSR080CB7 COROD Control Room operator dose, l % Case - TPC TSR080CC7 COROD Control Room operator dose, Realistic Case - Non-TPC TSR080CD7 COROD Control Room operator dose, l % Case - Non-TPC RS: TVA-F-W002415 Computer file name as $$R80#&8 where $$ = TS for Unit l; $$ = U2 for Unit 2
# = C for COROD code;#= F for FENCDOSE code
& =A for realistic TPC; & = B for TS limiting case TPC; & = C for realistic non-TPC; & =D for TS limiting case non-TPC A total of 16 runs for RS R9: TVA-F- W002551 Computer file name as $TSR80#&9 where $ = 1 for Unit l; $ = 2 for Unit 2
# = C for COROD code;#= F for FENCDOSE code
& = A for realistic TPC; & = B for TS limiting case TPC; RlO: TVA-F-W003192 Computer file name as $TSR80#&10_out where $ = l for Unit l; $ = 2 for Unit 2
# = C for COROD code; # = F for FENCDOSE code
& = A for realistic TPC; & = B for TS limiting case TPC; Rl2: TVA-F-W003269 Computer file name as TSR080Rl2#$&
where # = C for COROD code; # = F for FENCDOSE code; $ = l for Unit l; $ = 2 for Unit 2
& = A for realistic TPC; & = B for TS limiting case TPC.
Total of8 output files: 4 COROD output files, 4 FENCDOSE output files
-~~~~~~I r--~~ Calculation No. WBNTSR080 Rev: 12 I Plant: WBN I Page: 9
Subject:
Control Room Operator and Offsite Doses Due to a Loss of AC Power Purpose The purpose of this calculation is to determine the control room operator and offsite dose due to a Loss of AC Power. This calculation supports FSAR chapter 15.5. Introduction A Loss of AC Power to the Watt's Bar Nuclear Plant will result in a significant steam release to the environment. The steam will contain radionuclides if a primary to secondary side leak occurs prior to the event The secondary steam inventory from WBNNAL3-003 consists of expected radionuclide activity levels (ANS/ANSI-18.1-1984, ref.2). Computer code COROD (ref.4) will be used to determine the control room operator dose using the secondary steam inventory. Computer code FENCDOSE (ref.5) will be used to determine the offsite dose using the secondary steam inventory. The calculation will provide the control room operator and offsite dose for the tritium production core (TPC) for both the realistic case and Technical Specification limitiiig case. It should be noted that there is no standard review plan or regulatory guide for this accident This is a simple best estimate analysis. The Technical Specification limiting case is calculated utilizing a factor of 41500 as a multiplier to the realistic case. This factor is developed based on the specific activity of the secondary coolant being less than 0.1 µCi/g ofl-131 Dose Equivalent at the Technical Specification (ref. 24). The secondary coolant 1-131 Equivalent is developed based on Regulatory Guide 1.109 to be 4.15E+05 g/µCi as shown in ref. 12 (Table lb). The offsite dose limits are 10% (ref. 16) of the following regulatory limits: 25 rem gamma (whole body) (10CFRI00.11 ), 300 rem thyroid (IOCFRI00.11), 300 rem beta (skin), and 25 rem TEDE (10CFR50.67). SRP 6.4 in NUREG 0800 shows that the thyroid dose and beta skin dose limits are equivalent for the control room, therefore the offsite beta dose limit can be assumed the same as the offsite thyroid dose limit, 300 rem. The control room dose limits are 5 rem gamma (IOCFR50 Appendix A GDC19), 30 rem thyroid (SRP 6.4), 30 rem beta (SRP 6.4), and 5 rem TEDE (IOCFR50.67). Design Input The amount of steam released to the environment due to the loss of AC power is provided below as given in ref. I 0. 0-2 hours 455,718 lbs. 2- 8 hours 962,213 lbs. For the purpose of estimating the effect of the 74 second delay in control room isolation (ref. 19), it is assumed that the flow rate is linear as a function of time (Assumption 4). The flow for the initial 74 seconds is 4684 lbs (455,718
- 14n200 = 4683.77), and the flow for the remaining 7126 seconds is 451,034 lbs.
The following are the 'X)Q values (sec/m3) used in the computer code models: 30-day LPZ Offsite XJQ values [sec/cum] (ref.12): 1.784E-4 0-2 hr, 8.835E-5 2-8 hr, 6.217E-5 8-24 hr, 2.900E-5 1-4 day, 9.811E-6 4-30 day 2-hr EAB XJQ values: 6.382E-4 Unit 1 Control Room XJQ (ARCON96 method, ref.13): 3.85E-03 0-2 hr, 3.22E-03 2-8 hr, 2.36E-04 8-24 hr, 1.88E-04 1-4 day, 1.55E-04 4-30 day Unit 2 Control Room XJQ (ARCON96 method, ref.13): 2.59E-03 0-2 hr, 2.12E-03 2-8 hr, l.77E-04 8-24 hr, l.33E-04 1-4 day, 1.12E-04 4-30 day
-~~~~~~- Calculation No. WBNTSR080 I ---~ Rev: 12 I Plant: WBN I ~ Page: Io
Subject:
Control Room Operator and Offsite Doses Due to a Loss of AC Power Assumptions
- 1. The secondary side source consists of expected/realistic radionuclide activity levels for a reactor based on ANSI/ANS 18.1-1984, as calculated in WBNNAL3-003 (ref.2).
- 2. WBNNAL3-003 (ref.2) provides the inventory for tritium in a TPC. Only the design basis 2 TPBAR failure source term is used for each case, as the tritium has only a small impact on the result and using the design basis 2 TPBAR failure source term is bounding and conservative since additional failed fuel has no impact on tritium from a failed TPBAR.
- 3. Only one train ofCREVS is in operation. Normally, each CREVS train takes suction from separate intakes with no cross communication between trains. This leads to one contaminated train, and one uncontaminated train. The only way a 2 CREVS operation could result in higher doses would be for both trains to take suction from the same vent For this to happen, one intake path would require a failed closed intake path AND a fail open of normally closed passive manual damper at the beginning of the accident An active failure of a train plus a failure of a passive component in less than 24 hours is beyond design basis.
- 4. For the purpose of modeling the effect of a 74 second isolation delay, it is assumed that the steam release following a loss of power is constant over the first 2 hows. This is consistent with the model used for the main steam line break (ref. 12).
Special Requirements/Limiting Conditions There are no special requirements or limiting conditions in this calculation. Calculations The radionuclide inventory is secondary side steam released to the environment due to the loss of AC power. Its concentration is provided in µCi/g in WBNNAL3-003 (ref. 2). The releases, in Ci, are determined for each isotope per the following equation and are provided in the Table 1 below. The table also provides the Technical Specification limiting case by multiplying the realistic values by 41500 (= 4.15E+05 g/µCi x 0.1 µCi/g), except for tritium. Ci (isotope)= µCi/g (isotope) * (Ci/1E6 µCi)
- 453.59 g/lb.
- steam released lbs.
[Ill IRev: 12 I Plant: WBN I Page: 11 ~ ~~~~~~~~~~~~~~~~~........-~~~--.~~~~~--~~~~---. Calculation No. WBNTSR080
Subject:
Control Room Operator and Offslte Doses Due to a Loss of AC Power Table I. Source Terms for the Unit 1 Steam Released to the Environment due to the Loss of AC Power Realistic Inventory Technical S :>ecification Limiting Case Time Period 0-74s 74s-2 hr 2-8 hr 0-74s 74s-2 hr 2-8 hr Steam released (lb) 4.684E+03 4.510E+05 9.622E+05 4.684E+03 4.510E+05 9.622E+05 Secondary Cone Isotope Ci Ci Ci Ci Ci Ci uCi/a Kr-83m O.OOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO Kr-85m 3.63E-08 7.712E-08 7.426E-06 1.584E-05 3.200E-03 3.082* .575E-01 Kr-85 5.51E-08 1.171E-07 1.127~~5E-05 4.858E-03 4.678E 9.980E-01 I Kr-87 3.22E-08 6.841E-08 6.588E 5E-05 2.839E-03 2.734E-01 5.832E-01 Kr-88 6.31E-08 1.341E-07 1.291E-05 2.754E-05 5.563E-03 5.357E-O 1.143E+OO Kr-89 O.OOE+OO O.OOOE+OO O.OOOE O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO Xe-131m 1.34E-07 2.847E-07 2.741E 5.848E-05 1.181E-02 1.138E+OO 2.427E+OO Xe-133m 1.54E-08 3.272E-08 3.151E .721E-06 1.358E-03 1.308E-01 2.789E-01 Xe-133 5.25E-07 1.115E-06 1.074E-04 2.291E-04 4.629E-02 4.457E+OO 9.509E+OO Xe-135m 2.90E-08 6.161E-O .933E-06 1.266E-05 2.557E-03 2.462E-01 5.253E-01 Xe-135 1.91E-07 4.058E-07 3.908E-05 8.336E-05 1.684E-02 1.622E+OO 3.460E+OO Xe-137 7.62E-09 1.619E-08 1.559E-06 3.326E-06 6.718E-04 6.470E-02 1.380E-01 Xe-138 2.68E-08 5.694E-08 5.483E-06 1.170E-05 2.363E-03 2.275E-01 4.854E-01 1-131 1.41E-08 2.996E-08 2.885E-06 6.154E-06 1.243E-03 1.197E-01 2.554E-01 1-132 3.37E-08 7.160E-08 6.895E-06 1.471E-05 2.971E-03 2.861E-01 6.104E-01 1-133 4.03E-08 8.562E-08 8.245E-06 1.759E-05 3.553E-03 3.422E-01 7.299E-01 1-134 2.93E-08 6.225E-08 5.994E-06 1.279E-05 2.583E-03 2.488E-01 5.307E-01 1-135 6.19E-08 1.315E-07 1.266E-05 2.702E-05 5.458E-03 5.255E-01 1.121E+OO H-3 2-Rod 1.24E-01 2.634E-01 2.537E+01 5.412E+01 2.634E-01 2.537E+01 5.412E+01 For each case the released radionuclides are input into computer code FENCOOSE (ref. 5) to calculate the Low Population Zone (LPZ) offsite dose. The FENCOOSE model is taken from WBNAPS3-077 (ref. 12). For each case the released radionuclides are also input into computer code COROD (ref. 4) to determine the control room operator dose. The COROD model is taken from WBNAPS3-077 (ref. 12). The 'X/Q values used are from WBNAPS3-104 (ref.13) for the SGTR accident, because the steam release points are the same.
-~*~~~----I ~~--- Calculation No. WBNTSR080 Rev: 12 I Plant: WBN I -~ Page: 12 Subject Control Room Operator and Offsite Doses Due to a Loss of AC Power Results Unit 1 Doses Due to Loss of AC Power (Rem) TPC Offsite Control Room Comouter file TSR080R12F1A TSR080R12C1A 2 hrEAB 30dayLPZ Operator Dose Type (rem) (rem) (rem) Gamma 1.84E-08 1.05E-08 8.04E-09 Beta 2.14E-05 1.22E-05 3.58E-04 Thvroid (ICRP-30) 1.13E-06 6.46E-07 8.43E-07 TEDE 3.50E-04 2.00E-04 5.84E-03 TS Limiting Case Offsite Control Room Comouter file TSR080R12F1 B TSR080R12C1 B 2 hrEAB 30dayLPZ Operator Dose Type (rem) (rem) (rem) Gamma 7.63E-04 4.37E-04 3.34E-04 Beta 4.64E-04 2.65E-04 4.09E-03 Thyroid (ICRP-30) 4.69E-02 2.68E-02 3.50E-02 TEDE 3.81E-03 2.18E-03 7.47E-03
-~~~~--T~~-- Calculation No. WBNTSR080 I Rev: 12 I Plant: WBN
--~ I Page: 13
Subject:
Control Room Operator and Offsite Doses Due to a Loss of AC Power Discussion and Conclusion The calculated offsite doses are substantially below(< I 0%, ref.16) the regulatory limits of 25 rem whole body, 300 rem beta, 300 rem thyroid, and 25 rem TEDE. The control room operator doses are substantially below the regulatory limits of 5 rem gamma, 30 rem beta, 30 rem thyroid, and 5 rem TEDE. The calculated offsite TEDE dose is also less than the IOCFR20.130I (ref.14) limit ofO. I rem. The Unit I accident bounds the Unit 2 accident (see Appendix B). Notes on margin Because all calculated doses are less than O.OI rem, the margin for all doses is at least 99%. References I. DCN 5I754, Steam Generator Replacement (1/0)
- 2. WBNNAL3003 R5 " Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-I8. I-I984"
- 3. Deleted in revision 4
- 4. Computer code COROD R7.I, code ID 262347 (under dose code program mgr version 1.1)
- 5. Computer code FENCDOSE R5, code ID 262358 (under dose code program mgr version 1.1)
- 6. Deleted in revision 4
- 7. Deleted in revision 4
- 8. Deleted in revision 4
- 9. Deleted in revision 4
- 10. WCAP-I6286-P, January 2005, Watts Bar Unit I Replacement Steam Generator Program NSSS Engineering Report
- 11. Deleted in revision 4
- 12. WBNAPS3077 RI 6, "Offsite and Control Room Operator Doses Due to a Main Steam Line Break"
- 13. WBNAPS3-I04 R3, WBN Control Room '1./Q" I4. IOCFR20, section 20.130I, Dose Limits for Individual Members of the Public" I5. EDC E50629A I 6. ANS/ANSI 51.1-I983, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants" I7a WBT-D-I202 October 22, 2009 WBS 5.2.11 Revised Steam Releases for Dose" I 7b. LTR-CRA-09-I03 Rev.I Watts Bar Unit 2 Completion Project-Results of Steam Releases for Dose Calculations" I8. EDCR 54956 I9. WBNTSR028, RIO Main Control Room Emergency and Normal Air Intake Monitors Required Range, Safety Limits, Response Time and Accuracy
- 20. PER 327968
- 21. PER 327956
- 22. DCN 520I2
- 23. Regulatory Guide I.I09 RI
- 24. Technical Specification 3.7.I4, Amendment 8I
- 25. DCN 6I599
- 26. PER 775553
-~'~~~--r~~-r-~ Calculation No. WBNTSR080 I Rev: 12 I Plant: WBN I Page: 14 Subject Control Room Operator and Offsite Doses Due to a Loss of AC Power Appendix A Original Steam Generator Results, Unit 1. The following are the results for the original steam generators (R4 of this calculation). THIS IS KEPT FOR IDSTORICAL PURPOSES AND HAS NOT BEEN UPDATED FOR THE 1991-2010 X/Q METEOROLOGICAL DATA SET Realistic Case - Non-TPC Control Room Operator 30daylPZ 2 hr BAB ARCON96 Halitsky Gamma (whole body) 9.625E-09 2.399E-08 9.344E-09 8.012E-09 Beta 5.583E-09 l.392E-08 1.046E-07 8.967E-08 Iodine (thyroid)-ICRP-2 1.130E-06 2.818E-06 1.499E-06 l.301E-06 Iodine (thyroid)-ICRP-30 5.908E-07 1.473E-06 8.227E-07 7.144E-07 TEDE* 4.365E-08 l.088E-07 4.004E-08 3.443E-08 1% Failed Fuel Case - Non-TPC Control Room Operator 30daylPZ 2 hr EAB ARCON96 Halitsky Gamma (whole body) 7.696E-08 l.919E-07 7.447E-08 6.389E-08 Beta 4.452E-08 1.l BE-07 8.301E-07 7.126E-07 Iodine (thyroid)-ICRP-2 9.045E-06 2.255E-05 l.199E-05 1.041E-05 Iodine (thyroid)-ICRP-30 4.727E-06 l.178E-05 6.583E-06 5.716E-06 TEDE* 3.492E-07 8.705E-07 3.200E-07 2.752E-07 Realistic Case -TPC (2 TPBAR Failure) Control Room Operator 30 daylPZ 2hrEAB ARCON96 Halitsky Gamma (whole body) 9.625E-09 2.399E-08 9.344E-09 8.012E-09 Beta 8.880E-06 2.214E-05 3.314E-04 2.842E-04 Iodine (thyroid)-ICRP-2 1.130E-06 2.818E-06 l.499E-06 1.301E-06 Iodine (thyroid)-ICRP-30 5.908E-07 l.473E-06 8.227E-07 7.144E-07 TEDP l.453E-04 3.623E-04 5.415E-03 4.644E-03 1% Failed Fuel Case - TPC (2 TPBAR Failure) Control Room Operator 30 daylPZ 2 hr BAB ARCON96 Halitsky Gamma (whole body) 7.696E-08 1.919E-07 7.447E-08 6.389E-08 Beta 8.919E-06 2.224E-05 3.321E-04 2.849E-04 Iodine (thyroid)-ICRP-2 9.045E-06 2.255E-05 1.199E-05 1.041E-05 Iodine (thyroid)-ICRP-30 4.727E-06 l.178E-05 6.583E-06 5.716E-06 TEDE* 1.456E-04 3.631E-04 5.415E-03 4.644E-03
_llE'~~~~~~~----r--~-r-----, Calculation No. WBNTSR080 I Rev: 12 I Plant: WBN I Page: 15
Subject:
Control Room Operator and Offslte Doses Due to a Loss of AC Power AppendixB Unit 2 Loss of AC Power (pg. 1 of 2) This appendix evaluates the Unit 2 Loss of AC Power. The steam generators are the original steam generators; however Westinghouse has revised the mass releases. The same methodology as in the main text is applicable to this appendix. Although Unit 2 may not use a TPC, the Unit 2 doses are only calculated for a TPC because it bounds the conventional core. Design Inputs Unit 2 Control Room X/Q {ARCON96 method, ref.13): 2.59E-03 0-2 hr, 2.12E-03 2-8 hr, 1.77E-04 8-24 hr, l.33E-04 1-4 day, l.12E-04 4-30 day The amount of steam released to the environment due to the loss of AC power is provided below as given in ref.17. 0 - 2 hours 444,875 lbs. 2 - 8 hours 903,530 lbs. Calculations The releases, in Ci, are determined for each isotope per the equation on page 10 and are provided in the Table B-1 below. 1be table also provides the Technical Specification limiting case by multiplying the realistic values of 41500 except for tritium. The flow for the initial 74 sec is 4572 lb, and for the remaining 7126 sec is 440,303 lb. Table B-1. Source Terms for the Unit 2 Steam Released to the Environment due to the Loss of AC Power Seconda Cone ification Limitin Case Time Period 2-8 hr Steam released (lb) 9.035E+05 lsoto Cit Ci Ci Ci Ci Kr-83m O.OOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO Kr-85m 3.63E-08 7.528E 1.488E-05 3.124E 6.174E-01 Kr-85 5.51E-08 1.143E-07 2.258E-05 4.742E-03 9.371E-01 Kr-87 3.22E-08 6.678E-08 1.320E-05 2.771E-O 5.477E-01 Kr-88 6.31E-08 1.309E-07 2.586E-05 5.431E-03 1.073E+OO Kr-89 O.OOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO Xe-131m 1.34E-07 2.779E 2E-05 1.153E-02 2.279E+OO Xe-133m 1.54E-08 3.194E 6.311E-06 1.325E-03 2.619E-01 Xe-133 5.25E-07 1.089E-06 2.152E-04 4.519E 8.929E+OO Xe-135m 2.90E-08 6.014E-08 1.189E-05 2.496E-03 4.932E-01 Xe-135 1.91E-07 3.961E-07 7.828E-05 1.644E-02 3.249E+OO Xe-137 7.62E-09 1.580E-08 3.123E-06 6.558E-04 1.296E-Ot Xe-138 2.68E-08 5.558E-08 1.098E-05 2.307E-03 4.558E-01 1-131 1.41E-08 2.924E-08 5.779E-06 1.214E-03 2.398E-01 1-132 3.37E-08 6.989E 1.381E-05 2.901E-03 5.732E-01 1-133 4.03E-08 8.358E 1.652E-05 3.469E-03 6.854E-01 1-134 2.93E-08 6.077E 1.201E-05 2.522E-O 4.983E-01 1-135 6.19E-08 1.284E 2.537E-05 5.328E-03 1.053E+OO H-3 2-Rod 1.24E-01 2.572E-01 5.082E+01 2.572E-01 5.082E+01
-~'~~~~~~~~ Calculation No. WBNTSR080 I Rev: 12 I Plant: WBN I Page: 16
Subject:
Control Room Operator and Offsite Doses Due to a Loss of AC Power AppendixB Unit 2 Loss of AC Power (pg. 2 of2) Results Unit 2 Doses Due to Loss of AC Power (Rem) TPC Offsite Control Room Computer file TSR080R12F2B TSR080R12C2A 2 hr EAB 30 dayLPZ Operator Dose Type (rem) (rem) (rem) Gamma 1.80E-08 1.01E-08 5.09E-09 Beta 2.09E-05 1.17E-05 2.26E-04 Thyroid (ICRP-30) 1.10E-06 6.18E-07 5.37E-07 TEDE 3.42E-04 1.92E-04 3.70E-03 TS Limitina Case Offsite Control Room Computer file TSR080R12F2A TSR080R12C2B 2 hrEAB 30dayLPZ Operator Dose Type (rem) (rem) (rem) Gamma 7.45E-04 4.18E-04 2.11E-04 Beta 4.53E-04 2.54E-04 2.58E-03 Thyroid CICRP-30) 4.58E-02 2.57E-02 2.23E-02 TEDE 3.72E-03 2.09E-03 4.73E-03 Discussion and Conclusion The Unit 1 doses in the main body bound the Unit 2 doses above. 1be calculated offsite doses are substantially below (< 10%, re£16) the regulatory limits of25 rem whole body, 300 rem beta, 300 rem thyroid, and 25 rem TEDE. The control room operator doses are substantially below the regulatory limits of 5 rem gamma, 30 rem beta, 30 rem thyroid, and 5 rem TEDE. The calculated offsite TEDE dose is also less than the 10CFR20.1301 (ref.14) limit of0.1 rem. The Unit 2 accident is bounded by the Unit 1 accident Notes on margin Because all calculated doses are less than 0.01 rem, the margin for all doses is at least 99%.
Attachment 4 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Calculation (19 pages including this cover page)
NPG CALCULATION COVERSHEET I CTS UPDATE Page I REV 0 EDMS/RIMS NO CTSTYPE* EDMSTYPE: EDMS ACCESSION NO lli/A {gr REV. 0) B26 910624 200 Calculation CALCULATIONS (NUCLEAR) T 9 31 3 0 92 7 0 2 5 Cale
Title:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture ORG PLANT BRANCH NUMBER CUR REV NEW REV CALCID NUC WBN NIB WBNTSR064 011 012 crs UPDATE ONLY 0 (Verifier and Approval Signatures Not Required) II NO CTS CHANGES D (For calc revision, CTS has been reviewed and no CTS changes required) i UNIT (check one) SYSTEMS UNIDS o~.10,20,30 NIA NIA DCN,EDC,N/A APPI.K;ABIE DESIGN IXJCUMENT{S} CLASSIDCATION NIA NIA E QUAUIY SAFE'IY RELATEDZ UNVERIFIED SPECIAL RF.QUIREMENTS AND/OR DESK!N OUIPUT SAR/TS illl!Jlor ISFSI RELATED? (If yes, QR= yes) AS§UMPTION LIMITING CONDITIONSZ ATIACHMENTI ::!AR!CoC AFFECTED Yes~ NoO YeslS<I No0 Yesn No DC!! Yes~ No0 YesO No~ Yes~ NoO (PRINT~~ CALCUIATION NUMBER RF.QUESTOR PREPARING DISCIPLINE VERIFICATION METHOD NEW METHOD OF ANALYSIS Name: NIA PHONE: NIA BechtelMEB Design Review 0Yes ~No PREPARER NAME AND SIGN) DATE DATE
~KER(Pz_y DWWu tUJ
- D..<TE 1f2j1 DATE VERIFIER ITChan (P~k~ ~ #;/J ;~;1r;RINT~} l/L~~~ °tj20/(J
, (J STATEMENTOFPRQBIEM/ABSI.BAQI:
This calculation determined the offsite and control room operator doses due to a Waste Gas Decay Tanlc (WGD1) rupture. The calculation supports FSAR section 155.1 and Table 155-5. Three cases were calculated. One case assumed realistic gas source terms (from WBNNAL3-006) were present in the WGDT, the other assumed design basis source terms (lo/o failed fuel) and the third determined source tenns which results in an offsite gamma dose S0.5 rem. Computer code STP was used to determine the inventory released to the environment. Decay was allowed only for the realistic case and only for the time to fill the tan1c (55.75 hrs to fill the 600 cuft tanlc). The tan1c was assumed to release everything instantaneously and nonmechanistically to the environment after it was filled. The STP results were used as input to computer code COROD. COROD was used to determine the control room operator doses. The model was the same as that found in Tl-RPS-198, except for the 1}Q values and adding a 74 sec control room isolation delay. The '1.IQ values were determined using ARCON96 in WBNAPS3-104. The STP results were also used as input to computer code FENCDOSE, which determined the offsite dose. The FENCOOSE model was the same found in TI-RPS-197. The results are provided in Tables 3 and 5 in the Results section.s The 30 day LPZ offsite doses for a realistic and Regulatory Guide 1.24 Wasre Gas Decay Tanlc rupture accident were calculated to be substantially below the regulatory limitS of25 rem gamma, 300 rem beta, 300 rem .thyroid, and 25 rem TEDE. The control room operator doses for a realistic and Regulatory Guide 1.24 WGDT rupture were calculated to be below the regulatory limits of 5 rem gamma, 30 rem beta, 30 rem thyroid, and 5 rem TEDE. The Regulatory Guide 1.24 control room beta dose is relatively high, but this is due to the very conservative (high) noble gas inventory for that case. For the realistic case, the 2-hr EAB/Site Boundary offsite dose was less than the 500 mrem criterion set forth in NUREGOSOO section 11.3. ne Reg. Gmde 1.24 cue exceeds the 500 mrem Hmlt. Thus a case was performed to determine the source term based on Xe-133 equivalency, which results in the Site Boundary dose being just less than 500 mrem. For this case the Xe-133 equivalent is 7.12E+o4 Ci. Malntainlng the Xe-133 eqmvaleacy to less than 7.12E+o4 Ci, a rupture of a single Waste Gas Decay Tank will meet the intent of Regulatory Gulde 1.24. This Is a special reqairementllimltlng conclltlon for this calculation. MICROFICHFJEFICHE Yes~ Nol I FICHE NUMBERfS) TVA-F-W003248 TVA40S32 Page 1 of2 LEGIBILITY EVALUA*~.... 1 c:n... ::,.. 11 ACCEPTED FOR ISSUE. I J:;L;~ ~?-!\s. q 1' I Js
~1~S ~f~TE 1
§JGNATURE ALL f A
NPG CALCULATION COVERSHEET I CTS UPDATE Page 2 CAL.CID ORG PLANT BRANn:r NUMBER REV NUC WBN NIB WBNTSR064 012 BUllDINy NA I ROOM NIA I ELEVATION NIA I COORD/AZIM NIA I FIRM Bechtel CA1EGORIES SR/LC KEYWORDS (A-add, D-delete) ACTION KEYWORD AID KEYWORD (A/D) CROSS-REFERENCES (A-add, D-delete) ACTION XREF XREF XREF XREF XREF (A/D) CODE PLANT TYPE NUMBER REV A p WBN PER 775553 CT'S ONLY UPllATES: Followin11: are reauircd only when making keyword/cross reference CT'S undates and mwe I of fonn NEDP-2-1 is not included: PREPARER (PRINT NAME AND SIGN) DA1E CHECKER (PRINT NAME AND SIGN) DA1E PREPARER PHONE NO. EDMS ACCESSION NO. 1VA40532 Page2of2 NEDP-2-1 [10-31-2011]
p aae 3 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR-064 Title Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Revision DESCRIPTION OF REVISION No. 0 Initial issue. I Revision* I was performed to update references and validate/justify an FSAR reference. pages added: 5.5 pages deleted: none pages changed: 1-6, 9, 10, 15 2 Revision 2 was performed because the X/Q values changed and to update references. All pages were renumbered. Only text with actual changes will have revision bars. pages added: 7 paged deleted: none pages changed: all 3 Revision 3 was performed because of new control building makeup flow. pages added: none pages deleted: none pages changed: 1-7, 9, 15, 17 4 Revision 4 was performed because the Regulatory Guide 1.24 source terms as supplied by Westinghouse changed due to the extended fuel cycle (18 month, 1000 EFPD, 5% enrichment). These source tenns are an unverified assumption that need to be verified later. pages changed: all (all pages replaced since original R3 lost. Actual text changes have revision bars). pages added: none pages deleted: none 5 Revision 5 was performed to eliminate the unverified assumption by using the latest Westinghouse information. pages changed: 1-7, 9-11, 15, 17 pages added: none pages deleted: none RS total pages = 18 6 ReviSion 6 implements EDC E50629A, which implements the use of a Tritium Production Core. This revision also supports the corrective action for PER WBN 01-000395-000. The calculation was reformatted and renumbered. The revision added an evaluation to determirie the source inventory tising Xe-133 equivalency that results in a Site boundary dose of slightly less than 500 mrem. The evaluations in this revision utilize latest revision of COROD and FENCDOSE, which now calculate an ICRP-30 thyroid dose in addition to the ICRP-2, and a 1EDE dose. The results for the Non-TPC core are not impacted. Further, the discussion related to the administrative limit in PAJ-15.01 was deleted. An independent 3rd party was performed on this calculation by Wesdyne and the comments were incorporated into the calculation, with no technical impact. Pages changed: 1-11, 14-18 Pages added: All Pages deleted: All R6 total Pages: 22 7 Revision 7 is performed to incorporate a control room 20.6 second isolation time. Also incorporated ARCON96 control room X/Q values. Fixed minor typographical errors. All pages were renumbered, only pages with actual text changes are marked with revision bars. This revision is part of corrective actions for WBN PERs 03-012566-000 and 03-014473-000. The Technical Specifications and FSAR section 15.5 were reviewed and are not affected by this revision. Pages changed: 1-9, 11-13, 15-17 Pages deleted: the following page numbers correspond to the R6 pagination: 2, 12, 13, 21 Pages added: all R7: 18 total pages 8 Revision 8 is performed as corrective action for PER 61493 and PER 94426. PER 61493 documents that the recirculation flow rate for the control room was incorrectly modeled. PER 94426 documents that the time increments do not accurately reflei:t the 20.6 second delay. This revision corrects both modeling errors. This calculation directly impacts FSAR Table 15.5-5. The Technical Specifications are not impacted. Successor calculation WBNAPS3-074 is impacted by this revision. WBNAPS3104 is listed as a successor calculation, but a review of this calculation determined it does not use any information from this calculation is being deleted as a successor. Successors WBNOSG4243 and WBNTSR040 do not use any information that was revised in this revision and are thus not impacted. Pages Revised/Replaced: 1-3, 6-9, 11, 12, 16, 17 Pages Added: none Pages Deleted: none RS: 18 total mu*es TVA 40709 (12-2000] Page 1 of 1 NEDP-2-2 (12-04-2000]
Page4 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR064 Title Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Revision DESCRIPTION OF REVISION No. 9 Revision 9 is performed for replacement of the analog ratemeters with digital RMlOOO ratemeters by DCN 52012. The longer response time of the RMlOOO ratemeter in incorporated by increasing the control room isolation time from 20.6 seconds to 40 seconds (14 sec damper closure+ 20 sec instrument response). The only change is an increase in the control room HVAC time to isolation from 20.6 seconds to 40 seconds. Therefore, only the control room doses change, and the offsite doses are retained from revision 8. See calculation WBNTSR028, Rev. 7, (ref. 32) for the revised time delay of 40 seconds. FSAR section 15.5.2 and Technical Specifications were reviewed by Lynn Cowan and Table 15.5-5 is impacted by the change in isolation time from 20.6 seconds to 40 seconds. Pages Added: 4, 20, 21 Pages Deleted: none Pages Revised/Replaced: 1, 2, 5-9, 12, 13, 15-18 R9: 21 total pages. 10 Revision 10 is performed to upgrade the X/Qs to the 1991-2010 meteorological data set. FSAR section 15.5.2 and Technical Specifications were reviewed by Marc Berg and Table 15.5-4 and Table 15.5-5 are impacted. Calculations WBNTSR-074 and WBNTSR-040 are impacted by this calculation. These revisions will be tracked by WITEL item PL-11-3858 Pages Added: none Pages Deleted: R9 p.15. All pages renumbered. Pages Revised/Replaced: 1, 2, 4-10, 13-18 RIO: 20 total pages. 11 Revision 11 is performed under DCN 61599 to revise WBN calculations for increased average tritium permeation rates and changes in the number of tritium-producing burnable absorber rods (TPBARs) analyzed for TPC use. This calculation is revised to change the tritium source term for the three tritium production core (TPC) WGDT release cases. The revised tritium source term is based on a TPC containing 2500 TPBARs with the design basis average annual tritium permeation rate (10 Ci/TPBAR/yr). The STP model was updated to include the 55.75 hr decay time for the realistic source term as described in the body of the calculation. The change in the STP model resulted in insignificant changes in the amount of isotopic quantities of activity for each release case due to slight rounding differences. These impacts are deemed negligible; therefore, only the TPC release cases were affected, due to the change in tritium activity. For the release cases affected by the change in the tritium activity, only beta dose and TEDE are affected (gamma and thyroid doses are unaffected by the change in tritium activity). The special requirement/limiting condition regarding the Tech Spec limit on Xe-133 equivalency in a single tank is not affected by this revision. The first sentence in the Discussion and Conclusion section was modified to make it generally applicable for TPC operation. The Notes on Margin section was updated to correctly reflect the Tech Spec limit of7.12E+04 Ci Xe-133 equivalent and associated whole body gamma dose of 0.4997 rem. The section also was updated to reflect the current licensing basis of the plant (ICRP-30). Lastly, editorial improvements were made to remove discussion of revision-specific changes from the calculation body. CTS was reviewed for successor calculations to WBNTSR064, and 4 successor calculations were identified. WBNAPS307 4 is affected because it uses the RG 1.24, TPC CO ROD file modified in this revision. The other calculations are not affected by this revision because they are either inactive or do not use values that were changed as part of this revision. The header of all pages has been changed to Revision 11, and changes to the calculation are indicated by revision bars. See DCN 61599 for SAR/Tech Spec determination. Pages Added: Sa Pages Deleted: 17-20 Pages Revised/Replaced: 1, 2, 4-7, 9-16 R 11: 17 total pages TVA 40709 [10-2008] Page 1 of 1 NEDP-2-2 [10-20-2008]
NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR064 Page 4a Title Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Revision DESCRIPTION OF REVISION No. 012 Revision 012 of this calculation was created to address the instrument response time based on PER 775553. The longer response time is incorporated by increasing the control room isolation time from 40 seconds to 74 seconds ( 14 sec dam per closure + 60 sec instrument response). The primary change is to increase the control room HVAC time to isolation from 40 seconds to 74 seconds. The STP, COROD, and FENCDOSE inputs are affected, and the related cases are rerun. Successor documents have been reviewed and WBNAPS3074 is impacted by this revision. The effect of Unit 2 operation on Unit 1 margins has been reviewed with no impact. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. FSAR SECTION 15.5.2 AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND TABLE 15.5-.5 IS IMPACTED BY TH~A~ G I OLATION TIME FROM 40 SECONDS TO 74 SECONQS. n '
- Reviewer: Ji, , 'fc fer
- 41. wt"f '1 L I 9//3I1"3 Pages Added: 4a (1 page)
Pages Revised/Replaced: 1, 2, 5, 6, 7, 8a, 9, 12, 13,14, 16(total11 pages) Pages Deleted: none Total number of pages in this revision including Attachments: 17 (R11) + 1 (R12) =18 pages (R12) This page added by revision 012 TVA 40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Page 5 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNTSR064 I Revision: I 12 TABLE OF CONTENTS SECTION TITLE PAGE Coversheet I CTS Update 1 Record of Revision 3 Table of Contents 5 Calculation Verification Form 6 Computer Input File Storage Information Sheet 7 Computer Output Microfiche lnfonnation Sheet 8 Purpose 9 Introduction 9 Design Input 9 Assumptions 10 Special Requirements/Limiting Conditions 10 Calculations 11 Results 13 Discussion and Conclusion 15 Notes on Margin 15 References 16 TVA 40710 [10-2008] Page 1of1 NEDP-2-3 [10-20-2008) This page replaced by Revision 012
Page 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNTSR064 Revision 12 Method of verification used:
~Z
- 1. Design Review 181
- 2. Alternate Calculation D Verifier Date efr/'
- 3. Qualification Test D /JiCtlall Comments:
The changes to the calculation described in the Record of Revision for Revision 012 have been reviewed and have been found to be technically adequate in format and content. TVA 40533 [10-2008] Page 1of1 NEDP-2-4 [10-20-2008] This page replaced by Revision 012
Paqe 7 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNTSR064 IRev. 12 IPlant: WBN I
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture D Electronic storaoe of the inout files for this calculation is not required. Comments:
~ Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file. (Any retrieved file requires re-verification of its contents before use.)
The R2 input files are stored in FILEKEEPER file# 263465 The R3 input files are stored in FILEKEEPER file# 263714 The R4 input files are stored in FILEKEEPER file# 300426 The R5 input files are stored in FILEKEEPER file# 300588 The R6 input files are stored in FILEKEEPER file# 303398 The R7 input files are stored in FILEKEEPER file# 305736 The RB input files are stored in FILEKEEPER file # 308013 The R9 input files are stored in FILEKEEPER file# 314528 The WORD file for R9 is stored in FILEKEEPER file# 314527 The R10 input files are stored in eFiche file# TVA-F-W002481 The R11 Word, Excel spreadsheet, and computer input files are stored in FILEKEEPER file# 322075 The R12 input files are stored in FILEKEEPER file# 322436 The WORD file for R12 is stored in FILEKEEPER file# 322477
~ Microfiche/eFiche (See next page)
TVA 40535 [10-2008) Page 1 of 1 NEDP-2-6 [10-20-2008) This page replaced by Revision 012
page8 TV AN COMPUTER OUTPUT MICROFICHE INFORMATION SHEET Document WBNTSR-064 IRev. 10 I Plant: WBN I
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Microfiche Number Description RO: TSR064S STP Activity in WGDT WRAD-27 (ufilm) TSR064Fl FENCDOSE Offsite Dose, Realistic Case TSR064F2 FENCDOSE Offsite Dose, RG l.24 case TSR064Cl CO ROD Cont. Room Operator Dose, Realistic TSR064C2 COROD Control Room Operator Dose, RG 1.24 R2: TVA-F-C000081 TSR64FA2 FENCDOSE Offsite Dose, Realistic Case TSR64FB2 FENCDOSE Offsite Dose, RG 1.24 case TSR64CA2 COROD Cont. Room Operator Dose, Realistic TSR64CB2 COROD Control Room Operator Dose, RG 1.24 R3: TVA-F-COOOI 18 TSR64CA3 COROD Cont. Room Operator Dose, Realistic TSR64CB3 CO ROD Control Room Operator Dose, RG 1.24 R4: TSR64FB4 FENCDOSE Offsite Dose, RG 1.24 case TV A-F-C000273 TSR64CB4 CORODControl Room Operator Dose, RG 1.24 RS: TSR64FB5 FENCDOSE Offsite Dose, RG 1.24 case TV A-F-C000282 TSR64CB5 CORODControl Room Operator Dose, RG 1.24 R6: TSR64FA6 FENCDOSE Offsite Dose, Realistic case -TPC TV A-F-C000337 TSR64FB6 FENCDOSE Offsite Dose, RG 1.24 case -TPC TSR64CA6 COROD Control Room Operator Dose, Realistic -TPC TSR64CB6 CO ROD Control Room Operator Dose, RG 1.24 -TPC TSR64FC6 FENCDOSE Offsite Dose, Reduced case -TPC TSR64FD6 FENCDOSE Offsite Dose, Reduced case -Non-TPC R7: TSR64IS7 STPISOTP set up STP isotopic input TV A-F-W000351 TSR64S7 STP source tenns releases (eFiche) TSR64CA7 COROD RG 1.24, conventional core, control room dose TSR64CB7 COROD realistic, conventional core, control room dose TSR64CC7 COROD RG 1.24, TPC, control room dose TSR64CD7 COROD realistic, TPC, control room dose R8: TSR64C8A COROD RG 1.24, conventional core, control room dose TV A-F-W000574 TSR64C8B COROD realistic, conventional core, control room dose TSR64C8C COROD RG 1.24, TPC, control room dose TSR64C8D COROD realistic, TPC, control room dose TSR64C8E COROD RG 1.24, conventional core, control room dose - 2 CREV* TSR64C8F COROD realistic, conventional core, control room dose - 2 CREV* TSR64C8G COROD RG 1.24, TPC, control room dose,..- 2 CREV* TSR64C8H COROD realistic, TPC, control room dose - 2 CREV*
- note these are for information only and the results are not given, see assumptions #9 R9:
TV A-F-W001569 TSR64C9A COROD RG 1.24, conventional core, control room dose TSR64C9B COROD realistic, conventional core, control room dose TSR64C9C COROD RG 1.24, TPC, control room dose TSR64C9D COROD realistic, TPC, control room dose TSR64S9 STP source terms releases TSR64C10A COROD RG 1.24, conventional core, control room dose RIO: TSR64C10B COROD realistic, conventional core, control room dose TVA-F-W002481 TSR64C10C COROD RG 1.24, TPC, control room dose TSR64C10D COROD realistic, TPC, control room dose TSR64C10E COROD Tech Spec limited, TPC, control room dose TSR64F10A FENCDOSE Offsite Dose, RG 1.24 case -conventional core TSR64F10B FENCDOSE Offsite Dose, Realistic case -conventional core TSR64F10C FENCDOSE Offsite Dose, RG 1.24 case -TPC TSR64F10D FENCDOSE Offsite Dose, Realistic case -TPC TSR64F10E FENCDOSE Offsite Dose, Tech Spec Limited-TPC TSR64S10 STP source terms releases
TVAN COMPUTER OUTPUf page Ba MICROFICHE INFORMATION SHEET Document WBNTSR084
- 1 Rev. 12 I Plant: WBN I
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Microfiche Number Description Rll: TSR64S 11 out STP source term releases 1VA-F-W003229 TSR64Cl 1C_out COROD RG 1.24, TPC, control room dose TSR64Cl ID_out COROD realistic, TPC, control room dose TSR64Cl lE_out COROD Tech Spec limited, TPC, control room dose TSR64Fl1C_out FENCOOSE Offsite Dose, RG 1.24 case - TPC TSR64Fl lD out FENCOOSE Offsite Dose, Realistic Case - TPC TSR64Fl lE out FENCOO SE Offsite Dose, Tech Spec Limited - TPC Rl2: 1VA-F-W003248 TSR064Rl2S STP source term releases TSR064Rl2CA COROD RG 1.24, Standard Core, control room dose TSR064R12CB COROD realistic, Standard Core, control room dose TSR064Rl2CC COROD RG l.24, TPC, control room dose TSR064Rl2CD COROD realistic, TPC, control room dose TSR064Rl2CE COROD Tech Spec limited, TPC, control room dose TSR064Rl2FA FENCOOSE Offsite Dose, RG 1.24, Standard Core TSR064Rl2FB FENCOOSE Offsite Dose , realistic, Standard Core TSR064Rl2FC FENCOOSE Offsite Dose, RG 1.24 case - TPC TSR064Rl2FD FENCOOSE Offsite Dose, Realistic Case - TPC TSR064Rl2FE FENCOOSE Offsite Dose, Tech Spec Limited - TPC Total of 12 output tiles: l STP output file, l STP punch file, 5 COROD output files, 5 FENCDOSE output files This page replaced by Revision 012
Calculation No. WBNTSR064 I Rev: 12 I Plant: WBN I Page: 9
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture The pwpose of this calculation is to detennine the offsite and control room operator doses due to a Waste Gas Decay Tank rupture. This calculation will support Watts Bar Safety Analysis Report (FSAR) section 15.5.2 and Table 15.5-5. Introduction From reference 4, the gaseous waste disposal system (WDS) contains nine (9) gas decay tanks. These tanks receive gaseous waste from the Chemical Volume and Control System (CVCS) Holdup Tank, CVCS Volume Control Tank, WDS Spent Resin Tank, CVCS Boric Acid Evaporator, and WDS Reactor Drain Tank. The probability of a gas decay tank rupturing is low. However, the probability of an accidental release resulting from such things as operator error or malfunction of a valve or the overpressure relief system is considered to be sufficiently high that the calculated offsite whole body exposure that might result from a single failure during normal operation should be substantially below the guidelines of 10CFR Part 100," (ref.5). This calculation will determine the offsite and control room operator doses due to a Waste Gas Decay Tank rupture. The calculation will be done for both a Regulatory Guide 1.24 (1 % failed fuel, ref.5) accident and for a realistic case (ANS/ANSI-18.1-1984, ref.6) and one which results in an EAB/Site Boundary dose of just less than 500 mrem gamma. The maximum content of the failed decay tank is assumed to be released non-mechanistically to the environment over a two hour time period (ref.5). Radioactive decay is only taken into account for the time required to transfer the gasses to the decay tank (ref.5). Computer code S1P (ref. I) will be used to detennine the inventory of the radioactive gasses in the tank for the realistic case. The R.G. 1.24 case uses the inventory provided by Westinghouse in reference 11. Computer code FENCOOSE (ref.3) will be used to determine offsite doses utilizing SlP results as input. The FENCOOSE model parameters, other than the releases activity are taken from Tl-RPS-197 (ref.8). Computer code COROD (ref.7) will be used to determine the control room operator dose utilizing the S1P results as input. The COROD parameters, other than released activity, and x/Q values, are taken from Tl-RPS-198 (ref.9). The ARCON96 x/Q values were determined in WBNAPS3-104 (ref.24). The model sections pertaining to the 74 sec isolation delay are taken from WBNTSR028 (ref.32). The main dose limit for this calculation is an offsite gamma dose of 500 mrem from NUREG-0800 section 11.3. The following are standard limits for accidents. The offsite ganuna dose limit is 25 rem (10CFRl00.11), the thyroid dose limit is 300 rem (10CFRl00.11), and the TEDE dose limit is 25 rem (10CFRS0.67). SRP 6.4 in NUREG 0800 (ref. 23) shows that the thyroid dose and beta dose limits are equivalent for the control room, therefore the offsite beta dose limit can be assumed the same as the offsite thyroid dose limit, 300 rem. 10CFR20.1201 also states that the organ (thyroid) dose and skin (beta) dose are equivalent The control room dose limits are 5 rem gamma (10CFRSO Appendix A GDC19), 30 rem thyroid(SRP 6.4), 30 rem beta ( SRP 6.4), and 5 rem TEDE(10CFRS0.67). Design Input The Following are the Regulatory Guide 1.24 WGDT activities from WAT-D-10436. Kr-83m 17 Ci Kr-85 4200 Ci Kr-85m 130Ci Kr-87 29 Ci Kr-88 160 Ci Kr-89 0.1 Ci Xe-13lm 890 Ci Xe-133 68000 Ci Xe-133m lOOOCi Xe-135 940 Ci Xe-135m 48 Ci Xel37 0.27Ci Xe-138 3.2 Ci 1-131 0.048 Ci 1-133 0.033 Ci 1-135 0.012 Ci Total= 7.54E4 Ci ARCON96 X/Q values (ref.24): 0-2hr=2.56E-3 sec/m3 , 2-8hFl .71E-3, 8-24hr=7.26E-4, l-4day=5.21E-4, 4-30day=4.30E-4 WGDT Room volume= 11269 cuft (room 692-AS, ref.29, which is smaller volume than WGDT room A3=11503 cuft) Flow out ofWGDT room= 944 cfm (=largest measured value, ref.30). This page replaced by Revision 012
Calculation No. WBNTSR064 I Rev: 11 I Plant: WBN I Page: 10
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Assumptions
- 1. The tank is assumed to be filled with the highest concentration for each isotope from all sources into the WGDT. This will ensure maximum concentration of all isotopes.
a) The realistic source terms come from WBNNAL3-006 (ref.10). The WBNNAL3-006 concentrations correspond to the realistic inventory (ANSI/ANS 18.1-1980 4). The Regulatory Guide 1.24 (design basis, 1% failed fuel) source terms are provided by Westinghouse in WAT-D-10436 (ref.11). b) WBNNAL3-003 (ref. 6) provides the total design basis inventory of tritium for a TPC with 2500 tritium-producing burnable absorber rods (TPBARs) and an average tritium permeation rate of 10 Ci/TPBAR/yr. In accordance with NUREG-0017(ref. 25) 10% of the tritium is released as gas, thus the tritium source terms are: TPC - Normal Operation: 7301.0 Ci (total)* 10% = 730.1 Ci TPC (1 TPBAR Failure): (11,600 Ci+ 7301.0 Ci)* 10% = 1890.1 Ci TPC (2 TPBARFailure): (2*11,600 Ci+ 7301.0 Ci)* 10% = 3050.1 Ci Only the 2 TPBAR Failure case is run, as the tritium has only a small impact on the results and using the 2 TPBAR failure source term is conservative.
- 2. Radioactive decay is only taken into account for the time period required to transfer the gasses to the tank (ref.5),
except for tritium. The maximum content of the failed decay tank is assumed to be released non-mechanistically to the environment over a two hour time period (ref.5). For tritium, due to its 12.3 year halflife, it is considered that no decay occurs.
- 3. The tank failure is assumed to occur immediately upon completion of the waste gas transfer (ref.5).
- 4. Only one tank is assumed to fail, as all decay tanks are isolated from each other whenever they are in use (ref.4).
- 5. deleted in RS
- 6. The release path of the radioisotopes from the ruptured tank is through the Auxiliary Building vent (ref.12).
Technical Justification: A rupture of a Waste Gas Decay Tank will lead to release into the Auxiliary Building and hence into the normal ventilation. Planned releases of the WGDT inventory to the environment will be through the Shield Building Vent. The Shield Building Vent release path is monitored by radiation monitors such that if excessive releases occur, the vent will isolate (ref. 4). The Auxiliary Building Vent is monitored and alarmed (1-RE-90-101), but no automatic isolation occurs (ref. 27). Also, the xJQ values from the Auxiliary Building Vent are worse for the control room than the Shield Building Vent (see calculation WBNAPS3-104). Therefore, this release path is the most likely path and the most conservative path.
- 7. deleted in revision 7
- 8. All assumptions from TI-RPS-198 (ref.9) regarding the COROD model hold, except as denoted above.
- 9. Only one train ofCREVS is in operation.
Technical Justification: Normally, each CREVS train takes suction from separate intakes with no cross communication between trains. This leads to one contaminated train, and one uncontaminated train. The only way a 2 CREVS operation could result in higher doses would be for both trains to take suction from the same vent. For this to happen, one intake path would require a failed closed intake path AND a fail open of normally closed passive manual damper at the beginning of the accident. An active failure of a train plus a failure of a passive component in less than 24 hours is beyond design basis. Special Requirements/Limiting Conditions In order to meet the intent of NUREG0800 section 11.3, a Waste Gas Decay Tank must be limited to the Xe-133 equivalency to less than 7.12E+04 Ci.
Calculation No. WBNTSR064 I Rev: 11 I Plant: WBN I Page: 11
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Calculations A waste gas decay tank has a volume of600 cuft and is filled at 1.4 scfm (from 2 volume control tanks at 0.7 scfm each) until it reaches 100 psig (114.696 psia, ref.4). Once full, the flow is diverted to another tank. Therefore the volume of the gas at STP and the time to fill the tank is: or V2 = P 1Vi/P2 = (114.696 psia(600 cuft))/(14.696 psia)
= 4682.7 cuft at STP then time= (4682.7 cuft)/ (1.4 cuft/min
- 60 min/hr)
= 55.75 hr The realistic case (ANSI/ANS 18.1-1984) isotope concentration is taken from WBNNAL3-006 (ref.10). The Regulatory Guide 1.24 case (1% failed fuel) isotope concentration is taken from WAT-D-10436 (ref.11). For each case the tritium concentration for a TPC is taken from WBNNAL3-003 (ref.6). Using computer code STP (ref.I), a continuous realistic source flows at 1.4 cfm for 55.75 hr into a component labeled either "Real Room" or "Real TPC Room". Because a total inventory is given in reference 11, the Regulatory Guide 1.24 source is stepped into a component labeled either "RG 1.24 Room" or "RG 1.24 TPC Room". No decay is assumed for the R.G. 1.24 case and no decay is assumed for tritium. Since the STP realistic case input values are in µCi/cc, and the FENCDOSE/COROD codes require input values in Curies, the realistic continuous source flow rate into either the "Real Room" or "Real TPC Room" is:
F = (x µCi/cc)(l.4 cuft/min)(60 min/hr)(28317 cc/cuft)(l0-6 Ci/µCi)
= (x µCi/cc)(2.3786 Ci-cc/ µCi-hr)= x
- 2.3786 Ci/hr At the end of the 55.75 hr time period, the inventory of the tank is assumed to be released into the atmosphere. The inventory as calculated by STP is used as input into computer code FENCDOSE (ref.3) to calculate the Low Population Zone (LPZ) and the site boundary (SB), which is the same as the Exclusion Area Boundary (EAB), offsite dose. The FENCDOSE model is taken from TI-RPS-197 (ref.8).
Calculation No. WBNTSR064 I Rev: 12 I Plant: WBN I Page: 12
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture The inventory as calculated by STP is also used as input into computer code COROD to determine the control room operator dose. The COROD models (less the containment shine) are taken from TI-RPS-198 (ref. 9) with a 74 second delay in isolation. The closure time of dampers is 14 seconds, with a signal response time of 60 seconds, which gives a total closure time of74 seconds (ref. 32). The X/Q values used are different than the TI-RPS-198 model because the release points are different. There are 2 cases ofX/Q values for the control room with a release from the Auxiliary Building exhaust vent. Per TI-RPS-198, the worst case X/Q values for the first 8 hours are used, with the better X/Q values used after 8 hours due to operator action to select a better intake. This page replaced by Revision 012
Calculation No. WBNTSR064 I Rev: 12 I Plant: WBN I Page: 13
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Results Table 3 Doses Due to Waste Gas Decay Tank Rupture (Rem) Doses Due to Waste Gas Decav Tank Rupture (rem) Offsite Offsite Control Room 2-hr EAB 30-dayLPZ Case A - RG 1.24 Conventional Core Gamma 5.961E-01 1.666E-01 9.443E-01 Beta 1.613E+OO 4.509E-01 8.152E+OO Thyroid (ICRP-30) 1.286E-02 3.595E-03 1.079E-02 TEDE 3.104E-01 8.676E-02 1.026E+OO Case B - Realistic Conventional Core Gamma 2.880E-02 8.050E-03 4.272E-02 Beta 1.102E-01 3.0BOE-02 5.615E-01 Thyroid (ICRP-30) 1.205E-02 3.368E-03 1.003E-02 TEDE 1.487E-02 4.158E-03 4.865E-02 Case C - RG 1.24 Tritium Production Core Gamma 5.9&1E-01 1.666E-01 9.443E-01 Beta 1.615E+OO 4.516E-01 8.166E+OO Thyroid (ICRP-30) 1.286E-02 3.595E-03 1.079E-02 TEDE 3.520E-01 9.840E-02 1.245E+OO Case D - Realistic Tritium Production Core Gamma 2.BBOE-02 8.050E-03 4.272E-02 Beta 1.127E-01 3.151E-02 5.748E-01 Thyroid (ICRP-30) 1.205E-02 3.368E-03 1.003E-02 TEDE 5.650E-02 1.579E-02 2.675E-01 This page replaced by Revision 012
Calculation No. WBNTSR064 I Rev: 12 I Plant: WBN I Page: 14
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture The R.G. 1.24 EAB gamma dose presented in Table 3 exceeds the limit of 500 mrem gamma. In order to define a limiting condition for the WGDT that would, if released, not exceed the limits, a case is developed to determine the source inventory of the WGDT that would result in the EAB being just less than 500 mrem. The limiting condition for the WGDT is addressed as a total Xe-133 equivalency. To determine the limiting source terms a Xe-133 equivalent was determined for the R.G. 1.24 source inventory using dose factors for y-body in Table B-1 ofR.G. 1.109 (ref.26). A reduction factor value was determined by trial and error (by using various reduced source inventories as input to FENCDOSE) to be 0.838898 so that the resulting o:ffsite dose is just below the limit of 500 mrem. The R.G. 1.24 inventories were reduced by this factor. This calculation is provided in Table 4. The reduced isotopic values were entered into FENCDOSE to validate the reduced source inventory and yielded the results provided in Table 5 for TPC core (limiting case). Table 4 - R.G. 1.24 case Dose factors based on R.G. 1.109 RI Table B-1 (DFBJ Reduction Factor (RF)= 0.838898 Nuclide DFB; Ratio Reduced Reduced mrem-m3/ i- DFB- FB RG-1.24 Ci Xe-133 v Ci Kr-83m 7.56E-08 2.57E-04 l.70E+ol 4.37E-03 1.43E+ol 3.67E-03 Kr-85m 1.17E-03 3.98E+o0 l.30E+o2 5.17E+o2 l.09E+o2 4.34E+02 Kr-85 l.61E-05 5.48E-02 4.20E+o3 2.30E+o2 3.52E+o3 l.93E+02 Kr-87 5.92E-03 2.0IE+ol 2.90E+ol 5.84E+o2 2.43E+ol 4.90E+o2 Kr-88 l.47E-02 5.00E+ol l.60E+o2 8.00E+o3 1.34E+02 6.71E+o3 Kr-89 l.66E-02 5.65E+Ol l.OOE-01 5.65E+o0 8.39E-02 4.74E+o0 Xe-131m 9.15E-05 3. llE-01 8.90E+02 2.77E+o2 7.47E+02 2.32E+o2 Xe-133m 2.51E-04 8.54E-01 l.OOE+03 8.54E+02 8.39E+02 7.16E+o2 Xe-133 2.94E-04 1.00E+OO 6.80E+04 6.80E+04 5.70E+04 5.70E+04 Xe-135m 3.12E-03 l.06E+ol 4.80E+ol 5.09E+o2 4.03E+Ol 4.27E+02 Xe-135 l.81E-03 6.16E+o0 9.40E+o2 5.79E+03 7.89E+o2 4.85E+03 Xe-137 l.42E-03 4.83E+o0 2.70E-01 l.30E+o0 2.27E-Ol l.09E+OO Xe-138 8.83E-03 3.00E+ol 3.20E+OO 9.61E+ol 2.68E+o0 8.06E+Ol Total Xe-133 Equivalent 8.49E+04 7.12E+04 Table 5 Technical Specification Limited* Inventory, Tritium Production Core Doses Due to Waste Gas Decay Tank Rupture (rem) Offsite Offsite Control Room 2-hr EAB 30-dayLPZ Case E -Technical Soecification Limited TPC Gamma 4.997E-01 1.397E-01 7.915E-01 Beta 1.355E+OO 3.786E-01 6.847E+OO Thyroid (ICRP-30) 1.286E-02 3.595E-03 1.079E-02 TEDE 3.019E-01 8.439E-02 1.079E+OO
*Reduced= Limited to Xe-133 equivalent of7.12E4 Ci This page replaced by Revision 012
Calculation No. WBNTSR064 I Rev: 11 I Plant: WBN I Page: 15
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture Discussion and Conclusion The implementation of a TPC core impacts the beta doses by a small amount; however, the tritium does have a significant impact on the TEDE. Thus the following conclusions are applicable to both the TPC and Non-TPC cores. The 30 day LPZ offsite doses for a realistic and Regulatory Guide 1.24 Waste Gas Decay Tank rupture accident were calculated to be substantially below the regulatory limits of25 rem gamma, 300 rem beta, 300 rem thyroid, and 25 rem TEDE. The control room operator doses for a realistic and Regulatory Guide 1.24 WGDT rupture were calculated to be below the regulatory limits of 5 rem gamma, 30 rem beta, 30 rem thyroid, and 5 rem TEDE. The Regulatory Guide 1.24 control room beta dose is relatively high, but this is due to the very conservative (high) noble gas inventory for that case. For the realistic case, the 2-hr EAB/Site Boundary offsite dose was less than the 500 mrem criterion set forth in NUREG0800 section 11.3. The Reg. Guide 1.24 case exceeds the 500 mrem limit. In order to not exceed the 500 mrem limit, the Xe-133 equivalent for a WGDT must be maintained less than 7.12E+04 Ci. By maintaining the Xe-133 equivalency to less than 7.12E+04 Ci, a rupture ofa single Waste Gas Decay Tank will meet the intent of NUREG0800 section 11.3. This requirement is implemented by section 4.2 of the Gaseous Waste Disposal System Description, N3-77A-4001. Notes on Margin Because the approach used develops the maximum WGDT inventory that corresponds to the dose limit (500 mrem WB at the EAB), there is no margin in either the calculated inventory of7.12E+04 Ci Xe-133 equivalent, or the corresponding calculated dose of0.4997 rem. The margin is in the difference between the actual WGDT inventory and the inventory limit of7.12E+04 Ci Xe-133 equivalent. Note on Methodologies used for Doses: This calculation determined the doses using 3 different methodologies which are in revision 7 ofCOROD. The gamma, beta and Thyroid (ICRP-2) doses are all based on TID-14844 methodologies utilizing the ICRP-2 iodine dose conversion factors found in TID-14844. The second methodology is the Thyroid (ICRP-30) dose, which is also based on TID-14844, but uses the ICRP-30 iodine dose conversion factors. The ICRP-30 iodine dose conversion factors are less conservative than the ICRP-2 factors. Finally, the third methodology used is the TEDE (Total Effective Dose Equivalent). The TEDE presents an overall weighted dose and is more representative of the impact of all isotopes on the body as a whole. The TEDE dose is presented for potential future use. It is important to note that tritium does not impact the thyroid doses utilizing the TID-14844 methodology, because only iodine is applied to the thyroid dose. However, in fact tritium does contribute to the thyroid dose, as well as other organs of the body. This is why the TEDE is a more representative dose when discussing the impact of tritium. It is up to the end user to choose the dose which is to be used, with the understanding that each methodology has a different meaning.
Calculation No. WBNTSR064 I Rev: 12 I Plant: WBN I Page: 16
Subject:
Offsite and Control Room Operator Doses Due to a Waste Gas Decay Tank Rupture References
- 1. Computer Code STP R7, code I. D. 262165, controlled user's manual# 2
- 2. Halitsky, James et. al., "Wind Tunnel Tests of Gas Diffusion From a Leak in the Shell of a Nuclear Power Reactor and from a Nearby Stack" Department of Meteorology and Oceanography Geophysical Sciences Laboratory Report No.63-2, New York University, April 1, 1963
- 3. Computer Code FENCDOSE R4, code I. D. 262358
- 4. System Description N3-77A-4001 R4 "Gaseous Waste Disposal System"
- 5. Safety Guide 24 (Regulatory Guide 1.24) "Assumptions Used For Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Storage Tank Failure" 3/23/72
- 6. WBNNAL3003 R5 "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1984"
- 7. Computer Code COROD R7, code I. D. 262347
- 8. TI-RPS-197 R22 "Offsite Doses Due to a Regulatory Guide 1.4 Loss of Coolant Accident"
- 9. TI-RPS-198 R24 "Dose to Control Room Personnel Due to a Regulatory Guide 1.4 Loss of Coolant Accident"
- 10. WBNNAL3-006 RI "Gaseous Source Term Study" RIMS# B26 940916 362
- 11. WAT-D-10436, September 12, 1997 RIMS#T28 970912 803
- 12. WBN CCD drawing 1-47W866-2 RIO
- 13. WBN drawing 47W930-5 RE
- 14. WBN drawing 47W200-1 Rll
- 15. WBN drawing41N712-1 RD
- 16. WBN drawing 41N718-1 RE
- 17. WBN drawing47W200-13 R5
- 18. WBN drawing 47W930-2 RP
- 19. WBN drawing 47W930-3 RP
- 20. WBN drawing 47W200-8 R7
- 21. deleted in R4
- 22. deleted in R6
- 23. NUREG-0800 R2 section 11.3
- 24. WBNAPS3- l 04, R3 "WBN Control Room X/Q"
- 25. NUREG-0017 RI, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Efiluents from Pressurized Water Reactors', USNRC, April 1985
- 26. Regulatory Guide 1.109 RI, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effiuents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I.
- 27. Design Criteria WB-DC-40-24 RIO, Radiation Monitoring"
- 28. EDC E50629A
- 29. Calculation EPM-PFS-072991 R3 "Air Volumes, Penetrations and Door Opening Data for Auxiliary Building"
- 30. WBN CCD drawing 1-47W866-2 R24
- 31. Calculation WBNTSR-009 RI 5 "Control Room Operator and Offsite Doses From a Fuel Handling Accident"
- 32. WBNTSR028, RI 0, "Main Control Room Emergency and Normal Air Intake Monitors Required Range, Safety Limits, Response Time and Accuracy *
- 33. EDC 58273-A
- 34. DCN 61599 This page replaced by Revision 012
Attachment 5 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 Control Room Operator and Offsite Doses From a Fuel Handling Accident Calculation (40 pages including this cover page)
NPG CALCULATION COVERSHEET I CTS UPDATE Page 1 REV 0 EDMSLRIMS NQ, CTSTYP!;!; EDMSTYPE; EDM§ A~~SIQH NQ (N(A tile BBY !ll B26 890302 008 Calculation CALCULATIONS (NUCLEAR)
'Tl I l 4 04 0 4 10 Cale
Title:
Control Room Operator and Offslte Doses From a Fuel Handling Accident QIW lUfil ~ ~ CURBBY NEWimy: CAI.CID NUC WBN NTB WBNTSR009 015 016 crs UPDAIB ONLY D NO crs CHANGES D (Verifier and Approval Signatures Not Required) (For calo revision, CTS has been rcviowcd and no CTS changes required) lJNIT (check one) SYSTEMS .uN.!I2S. 0181, l 0.20,30 I NIA NIA D!J.11 ImQ NlA APPLIC~Ul I2!.ll!I!J:N DQCUMENTIS) CWSlflCAilQN - NIA OUALIIY RELATED? I SAFE'fYRELA1ED7 (If yes, QR~ N/A yes) UNVERIFIED ASSUMPTION I SfECIAL REOUIRBMEJ:ITS AND/OR LIMITING CQNDmQNS? I D~Yl:i OUI'PUI' ATIACHMBNT? SA'R/f'n(' B SAR/fS and/or IllFSI R I Yes 181 No 0 Yes 181 No 0 YesD No~ Yes181 No0 YesO No181 Yes181 No0 CALCULATIQNNUMBER REOUBSIQR PREPARINQ 121.'!CIPLINI! YIIB.IEICATIQN MBTiiQI! NEW METiiQI! Q.f'. Ali&IlIS Nemc:N/A . PHONE:N/A N(AREVA) Design Review 0Yes 181No a;:;. d .,, .., DATE DAlE PREPARE~N~ bAThomes
- rn{r.i/l.t>t~
CHECKER (P,NAME LGN) GDSeeborger £. .A ~ I ,,/ 3/2.8/ 2-0/U c2.. .... O... ,..... VERIFIER (PRINT NAME AND SIGNl DATE AP~WNAMEANDSIGN) DATE Or tevt 5<< lo""V"9*_1 )
~"'
"l.ol<<.6 ~~ ~t~fl~
l!IATEMENT QI fRO!lLl3MlABSTRACI This calculation was performed to determine the dose to control room operators following a. design basis fuel handling accident (FHA), In addition, the offilite doses resulting from a. design basis FHA were also calculated. Base assumptions utilize either Regulatory Guide 1.2S (Safety Guide 2S) or Regulatory Guide 1.183 (Alternate Source Tenn). A TPBAR only accident is also analyzed. The calculation considers a FHA occurring in containment with activity passing directly to the environment (no Purge Filters) until isolation in 12.7 seconds, a. FHA in containment exhausting to the environment via Purge Filters, a FHA in containment with contamlnation migrating due to open penetrations to the AB with release through the Auxiliary Building vent (no filters). Finally, a FHA in the spent fuel pit (Auxiliary Building) with releases through the Auxiliary Building vent (no ABI and no filters). The FHA is assumed to occur at 100 hours after shutdown. All of the other assumptions used to calculate the activity released are in accordance with Safety Guide 2S and NUREG/CR-5009 or Regulatory Guide 1.183 (Alternate Source Term/AS I). All of the activity is assumed to be released overa two hour time period per Safuty Guide 2S or RG 1.183. The computer code STP was used to calculate the activlty released a.lier a FHA. The activity released to the environment as determined by STP was input Into the computer code COROD. The control room model used is identical to that described in TI-RPS-198 except that the containment shine Is not included This calculation also considers the effect ofa 74 second unfiltered bypass flow due to the finite closure time of the control room isolation dampers (14 sec) and instrument actuation time (60 seo). The activity released to the environment as determined by STP was also used as input to computer code FENCDOSE to calculate the doses at the Site Boundary (SB)/Exclusion Area Boundary after 2 hours* and at the Low PopuiationZone (LPZ) boundary after 30 days. The FENCDOSB model came from TI-RPS-197. The control room operator doses are below the 10CFRSOAppendixA, GDC 19 limits ofS rem gamma, 30 rem beta, 30 rem thyroid, and 10CFRS0.67 limit ofS rem TEDE. The of!Site doses ere less than2S% of the 10CFR.100 limits of2S rem gamma, 300 rem beta, 300 rem thyroid, end 10CFRS0.67 limit of2S rem TEDE (= 6.2S rem gamma, 7S rem beta/thyroid, and 6,2S rem TEDE). If the design basis of the plant Is Regulatory Guide 1.183, then there are no Special Requlrements/Llmitlng Conditions. See Page 10 for Special Requirements for a Regulatory GuJde 1.25 accident. This calculation directly impacts FSAR Table 15.5-23 MICROFICHE/EFICHE Yes181 NoD FICHE NUMBER(S) TVA-F-W00320S 1VA40S32 Pagel of2 NEDP-2-1 [10-31-2011]
NPG CALCULATION COVERSHEET I CTS UPDATE NUC WBN NTB WBNTSR009 016 BUILDING NIA I ROOM NIA I ELEVATION NIA I COORDIAZIM NIA I FIRM AREVA CATEGORIES NIA KEYWORDS (A-add, D-delete) ACTION KEYWORD AID KEYWORD (A/D) CROSS-REFERENCES (A-add, D-delete) ACTION XREF XREF XREF XREF XREF (AID) CODE PLANT TYPE NUMBER REV D p WBN PER 429145 A p WBN PER 428401 A p WBN PER 114028 CTS QNLY UPDATES* Following are reauired onlv when making kevword/cross reference CTS updates and Jage 1 offonn NEDP-2-1 is not included: PREPARER (PRINT NAME AND SIGN) DATE CHECKER (PRINT NAME AND SIGN) DATE PREPARER PHONE NO. EDMS ACCESSION NO. TVA40532 Page 2 of2 NEDP-2-1 [10-31-2011]
p aqe 3 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR009 Title Control Room Operator and Offsite Doses From a Fuel Handling Accident Revision DESCRIPTION OF REVISION No. 0 Initial Issue I This revision was performed because the key references changed, resulting in COROD and FENCDOSE models needing revision. Also, the case of a FHA occurring in the Auxiliary Building was added Pages added: 2.1-2.4, 4.1 Pae:es chane:ed: 1-12 2 This revision was performed because the previous analysis oversimplified the dilution process. The error has been corrected analytically without running the STP code. Ref. 12, FSAR chapter 15 was deleted as it was sufficient to refer to ref. 4. Safety Guide 25. Ref. 11 MR 482000 was no longer valid and replaced by WB-DC-36.1 R4. The control room operator doses are slightly reduced. There is no impact on the conclusions of the calculation. Pages added: la, 2.1, 2.5 Pages deleted: 2.1 Pages changed: 1,2,2.2,4, 7,8,9, 10 Total pages: 35 CCRIS and DCCM were checked on 01.04.93 and no changes which impact this calculation were found. This calculation does not require imoact review as no other discioline uses it as design inout. 3 Revision 3 was performed to take into account a single failure of the Auxiliary Building General Ventilation Exhaust Fan in the "on" position concurrent with a single isolation damper failing to close resulting in ABGTS filter bypass. All pages were rewritten for legibility and renumbered. Only areas with changed text are identified with revision bars. Pages added: all Pages deleted: all Pae:es changed: all 4 Revision 4 was performed because the ABGTS bypass was fixed by DCN M-29141-A. R4 reinstated the R2 models and results. Pages added: !(new cover) Pages changed: 1.l(old cover), 2-6, 8, 9, 11, 13-15, 17 Pa.e;es deleted: none 5 Revision 5 was performed because the X/Q values changed Pages added: 1.2 Pages deleted: none Pages changed: 1-8, 10-14, 16, 17 6 Revision 6 was performed because the control room makeup flow was changed. Pages changed: 1, 1.2, 2-6, 8, 12-14, 16, I 7 Pages added: none Pages deleted: none TVA 40709 (12-2000] Page 1 of 1 NEDP-2-2 (12-04-2000]
paae 4 TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR009 Title Control Room Operator and Offsite Doses From a Fuel Handling Accident Revision DESCRIPTION OF REVISION No. 7 Revision 7 was performed to upgrade the calculation to the cycle 2 1000 EFPD, extended burnup (18 month) fuel. pages added: 1, 2.1 pages changed: 1a (old cover), 1.2, 3-17 pages deleted: none 8 Revision 8 was performed to change the FHA source terms from 1000 EFPD to 1500 EFPD as part of the corrective action of WBPER960798. pages added: none pages changed: l, 2.1, 3-7, 9, 10, 12, 14-17 pages deleted: 1a, 1.1, 1.2 9 Revision 9 is performed to discuss the impact ofD-50378-A which allow containment penetrations to be open during fuel movement. There was no impact on the final answers. Pages changed: 1, 2.1, 3, 17 Pages added: 9 .1 Pages deleted: none R9: 36 total pa11;es 10 Revision 10 is performed to incorporate NUREG/CR-5009 gap inventories, increased isolation time (WBNPER 01-000080-000), incorporate X/Q values as determined by ARCON96, and incorporate the Tritium Production Core (TPC). The latest versions ofCOROD (RS) and FENCDOSE (R4) were used, which determine thyroid doses based on both ICRP-2 and ICRP-30 dose conversion factors, as well as determine the TEDE. Also, independent third party review comments by Westdyne (Westinghouse) and NYSIS were incorporated where appropriate. Due to the extent of the changes, all pages were renumbered. Actual text changes are marked with revision bars. Changes in this revision will be screened for 50.59 applicability via the EDC referenced on the coversheet. Pages added: all Pages deleted all Pages changed: all RIO: 47 total pages 11 Revision 11 is performed in support of PERs 61493 (control room recirculation rate modeling), 94426 (control room time increment modeling), 95217 (potential for 15 minute unfiltered releases and migration of contamination to other un-isolated areas), and 96939 (failure to evaluate FHA in the transfer canal or cask loading area) and EDC 51930 (downgrade Purge Filters) and also to add Alternate Source Tenn (AST) cases. EDC 51930 downgrades the Reactor Building Purge filters to non-safety related and thus credit cannot be taken for them to mitigate the FHA. The design basis for a FHA in containment is to take credit for containment isolation, which occurs in 12. 7 seconds once EDC 51930 is implemented. PER 61493 documents that the wrong control room recirculation flow rate is used. The recirculation flow rate is 3600 cfin- makeup flow (711) = 2889 cfin. 2 trains of CREVS in operation for the first 2 hours is addressed in an assumption. Another case was added to analyze a FHA in containment with migration to the AB after isolation through open penetrations. The discussion about penetrations on page 11 ofRlO was deleted and a Special Requirement was added to require a CVI with an ABI and vice versa. It was also required that the ABSCE be established within 4 minutes even if there are other penetrations to outside the ABSCE. AST cases were performed with and without Purge/ABGTS filters. This calculation directly impacts FSAR Table 15.5-23. EDC 51930 contains a Technical Specification change for the Reactor Building Purge filters Pages Revised/Replaced:l, 2, 4-29, renamed Attachment 4 as Attachment l, renamed Attachment 5 as Attachment 2, renamed Attachment 6 as Attachment 3 Pages Added: 15, 23 Pages Deleted: old cover (2.1), Design Verification form, 11,13, old attachments 1-3 Total Pages = 29 TVA 40709 [12-2000) Page 1of1 NEDP-2-2 [12-04-2000)
PaQe 5 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR009 Title Control Room Operator and Offsite Doses From a Fuel Handling Accident Revision DESCRIPTION OF REVISION No. 12 Revision 12 is performed to address replacement of the analog ratemeters with digital RMIOOO ratemeters by DCN 52012. The longer response time of the RMI 000 ratemeter is incorporated by increasing the control room isolation time from 20.6 seconds to 40 seconds ( 14 sec damper closure + 26 sec instrument response) The primary change is to increase the control room HVAC time to isolation from 20.6 seconds to 40 seconds. A secondary change is to bring the total tritium release in the RG 1.25 cases into agreement with the RG 1.183 (AST) cases. In revision 11, 100% of the spent fuel pool (SFP) tritium inventory was released in the RG 1.25 cases and 25% was released in the RG 1.183 (AST) cases. The 100% value is unduly conservative, and as shown in Assumption 13, the 25% release is also very conservative. Therefore, using 25% of the SFP tritium inventory for all cases is both consistent and conservative. The STP, COROD, and FENCDOSE inputs are affected. However, because not all calculations are affected, only the STP/COROD/FENCDOSE runs that are affected are rerun. The results from those that are not rerun are retained from revision 11. See calculation WBNTSR028, Rev. 7, (ref. 13) for the revised time delay of 40 seconds. FSAR sections 15.5.6 and Technical Specifications were reviewed by Lynn Cowan and Table 15.5-23 is impacted by the change in isolation time from 20.6 seconds to 40 seconds. Pages Added: 5, 6, 22-25 Pages Deleted: none Pages Revised/Replaced: I, 2, 7-10, 12-14, 16-21, 23, 24, 26, 27 Rl2: 34 total pages. 13 Revision 13 is performed to upgrade the X/Qs to the 1991-2010 meteorological data set. Cases were performed for a 7 second unfiltered release or a 17 second unfiltered release from the Auxiliary Building as documented in PER 252012, with the control room isolating in 40 seconds. Other cases were performed with this 7 second or 17 second unfiltered release, but with the control room already isolated. A TPBAR only accident is also analyzed. All cases with unfiltered release from the Auxiliary Building now use the normal ventilation flow for the 7 or 17 seconds. FSAR sections 15.5.6 and Technical Specifications were reviewed by Marc Berg and Table 15.5-20 and -23 are impacted by the change. The following calculations are impacted by this revision: WBNSPA3-050, -085, WBNTSR-020, -023. The revision of these successors will be tracked by WITEL item PL-11-3861. Pages Added: none Pages Deleted: none Pages Revised/Replaced: I, 2, 5, 6, 8-28 Rl3: 34 total pages. 14 Revision 14 is performed to delete RG 1.25 cases (except for containment FHA with purge or containment FHA with 12. 7 sec release, no filters). Added a RG 1.183 case for containment FHA with 12.7 sec release through the Shield Building (no filters), with remainder release through the Auxiliary Building vent (no AB and no filters). Appendix Bis added to address PER 429145 issue of nonconforming TEDE formulas in COROD/FENCDOSE with RG 1.183. Corrected input for STP run case series 4 and 11 for initial H3 inventory (25% factor incorrectly included in initial inventory and also in "S" card). FSAR sections 15.5.6 and Technical Specifications were reviewed by Marc Berg and Table 15.5-20 and -23 are impacted by the change. The following calculations are impacted by this revision: WBNSPA3-050, -085, WBNTSR-020, -023. Pages Added: Appendix A (page 23), Appendix B (pages 24-29) Pages Deleted: none Pages Revised/Replaced: I, 2, 5-22 Rl4: 36 total pages. TVA 40709 (10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
PaQe SA NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR009 Title Control Room Operator and Offsite Doses From a Fuel Handling Accident Revision DESCRIPTION OF REVISION No. 15 Revision 15 of this calculation was created to address the instrument response time based on PER 775553. The longer response time is incorporated by increasing the control room isolation time from 40 seconds to 74 seconds (14 sec damper closure+ 60 sec instrument response). The primary change is to increase the control room HVAC time to isolation from 40 seconds to 74 seconds. The STP, COROD, and FENCDOSE inputs are affected, and the related cases are rerun. TPBAR only case is not changed, because its model is not related to the isolation time. Successor documents have been reviewed and are not impacted by this revision. However, the following calculations will be updated due to cancellation of DCN 52012: WBNTSR008, WBNAPS3050, WBNAPS3077, WBNTSR028 and WBNTSR064. Affected engineering judgments and assumptions were reviewed and (1) were found to be adequate, or (2) were revised as necessary to ensure adequacy. Ultimate heat sink (UHS) temperature was not used as an input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature. The effect of Unit 2/dual unit operation on Unit 1 margins has been reviewed with no impact. FSAR SECTION 15.5.6 AND TECHNICAL SPECIFICATIONS HAVE BEEN REVIEWED AND TABLE 15.5-23 IS IMPACTED BY THE CHANGE IN ISOLATION TIME FROM 40 SECONDS TO 74 SECONDS. Reviewer: originally signed by K.W. Peterman 10/3/13 Pages Added: 5A, 9A (total 2 pages) Pages Revised/Replaced: entire document has been updated with change bar added Pages Deleted: none Total number of pages in this revision including Attachments: 36 pages (Rev. 14) + 2 pages (Rev. 15) = 38 Appendix A 1 page Attachment 1 5 pages Appendix B 6 pages Attachment 2 1 page Attachment 3 1 page This page is added by Revision 015 TVA 40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Page 58 NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSR009 Title Control Room Operator and Offsite Doses From a Fuel Handling Accident Revision DESCRIPTION OF REVISION No. 16 Revision 16 of this calculation was created to address an incorrect callout/reference to PER 429145 (made for Revision 14) in Appendix B. The reference to PER 429145, which discusses BFN, is changed to PER 114028 Action 10 and PER 428401, which discuss WBN. This will allow the closure of PER 114028 Action 10. In addition, some reference revisions will be updated and minor editorial changes made. No conclusions or results are changed as result of this revision. Therefore, there are no calculations impacted by this revision of WBNTSR009. Pages Added: 58 Pages Revised/Replaced: entire document has been updated with change bar added Pages Deleted: none Total number of pages in this revision including Attachments: 38 pages (Rev. 15) + 1 page (Rev. 16) = 39 pages Appendix A 1 page Attachment 1 5 pages Appendix B 6 pages Attachment 2 1 page Attachment 3 1 page Note the Table on page 26 (Appendix B) shows a DCF value for Kr-83m that is said to come from Table 5-1 of EPA-400-R-92-001. The value listed is suspect, since Table 5-1 of EPA-400-R-92-001 does not contain a DCF for Kr-83m and the value listed is too large, but the value is not used for the results in this calculation an so does not change the results and conclusions of WBNTSR009. This page is added by Revision 016 TVA 40709 [10-2008] Page 1of1 NEDP-2-2 [10-20-2008]
Paoe 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNTSR009 Revision 016 Method of verification used: 1. 2. 3. Design Review Alternate Calculation Qualification Test 181 D D Verifier ...AJ,4 Date >pr-:/ 3, 1.Dll/ Comments: . The changes to the calculation described in the Record of Revision for Revision 016 have been reviewed and have been found to be technically adequate in format and content. TVA 40533 [10-2008] Page 1of1 NEDP-2-4 (10-20-2008]
Page 7 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNTSR009 I Revision: I 016 TABLE OF CONTENTS SECTION TITLE PAGE Calculation Cover Sheet I CTS Update .. .. .... ... .. .. .. ......... .... .. ... .. ....... ....... ............. 1 Record of Revision .................. .................................................... .. ....................... 3 Calculation Verification Form ....... .............. ....... ...... ....... ............ ....... ..... ............... 6 Table of Contents ................................................................................................. 7 Computer Input File Storage Information Sheet ................................................... 8 Computer Output Microfiche Storage Information Sheet....................................... 9 Purpose ................................................................................................................ 10 Special RequiremenULimiting Conditions ............................................................. 10 Introduction ... .............. ............ .. ........................ ....... ........... ... ............ ............. ...... 10 Assumptions ............... ................................... .............. ....... ..... ............ .. .. ....... ...... 11 Calculations ... ........... ... ................. ......... .... ... ......... ... ................... ............. ............. 13 Results.................................................................................................................. 17 Conclusions........................................................................................................... 21 References ........................................................................................................... 21 Appendix A RG 1.25 Containment FHA with 12. 7 sec unfiltered release, containment closed to Auxiliary Building ............................................................................................... 23 Appendix B TEDE Calculation per Regulatory Guide 1.183 ...... ......... ........... ....... ....... ............. 24 Attachment 1 Tritium Target Qualification Project Procedure, TTQP-1-091 ................................ 30 Attachment 2 memorandum T35 001023 938 ............................................................................ 35 Attachment 3 reference 28b memo TTQP-00-175....................................................................... 36 TVA40710 [10-2008) Page 1 of 1 NEDP-2-3 [10-20-2008)
PaQe 8 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNTSR009 I Rev. 16 I Plant: WBN I
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident I I Electronic storaQe of the input files for this calculation is not required. Comments: n Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file. (Any retrieved file reQuires re-verification of its contents before use.) R6: The computer input is permanently stored in FILEKEEPER file# 263662 R7: The computer input is permanently stored in FILEKEEPER file # 292579 RS: The computer input is permanently stored in FILEKEEPER file # 300126 R10: The computer input is permanently stored in FILEKEEPER file# 303621 R11: The computer input Is permanently stored in FILEKEEPER file# 308333,308360 R12: The computer Input is permanently stored in FILEKEEPER file# 31 4526 The WORD file for R12 is stored in FILEKEEPER file# 314525 R13: The computer input Is permanently stored in eFiche file# TVA-F-W002485 and TVA-F-W002502 R14: The computer input is permanently stored in eFiche file# TVA-F-W002566 R15: The computer input Is permanently stored in FILEKEEPER file# 321990 The WORD file for R15 is stored in FILEKEEPER file# 322560 R16: The WORD file for R16 is stored In FILEKEEPER file# ###### 181 Microfiche/eFiche See next page TVA 40535 (10-2008) Page 1of 1 NEDP-2-S (10-20-2008)
page 9 TVAN COMPUTER OUTPUT MICROFICHE INFORMATION SHEET Document WBNTSR009 I Rev. 16 I Plant: WBN I
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Microfiche Description R3:TVA-F-Gl 04672 RIO: Name Code Description RS:TVA-F-C000074 TS9Sl0$ STP source term R6:TVA-F-C000108 TS9C10#$ COROD control room operator dose R7:TVA-F-C000138 TS9Fl0#$ FENCODSE Offsite dose R8:TVA-F-C000219 where
$= A = standard core, instant control room isolation RIO: B =standard core, 20.6 sec control room isolation TVA-F-W000221 C =Tritium Production Core, once burned assembly, instant control room isolation D =Tritium Production Core, once burned assembly, 20.6 sec control room isolation E =Tritium Production Core, twice burned assembly, instant control room isolation F =Tritium Production Core, twice burned assembly, 20.6 sec control room isolation G =Tritium Production Core, 3X burned assembly, instant control room isolation H =Tritium Production Core, 3X burned assembly, 20.6 sec control room isolation X =standard core, 20.6 sec isolation time, revision 9 (old Halitsky) X/Q values
#= A= Spent Fuel Pit/Auxiliary Building/ABGTS FHA P = Containment/PAE FHA RI!:
Name Code Description TS9S I I$# STP release models TS9Cl 1$# COROD control room dose with I train ofCREVS, 20.6 sec control room isol TS9Fll$# FENCDOSE offsite dose where
$= A = standard core B = Tritium Production Core, once burned assembly C = Tritium Production Core, twice burned assembly Rll: D = Tritium Production Core, thrice burned assembly TVA-F-W000575 #=I= RG 1.25 Contain FHA w/ 12.7 sec contain isolation, containment closed to AB, no Purge Filters TVA-F-W000622 2= RG l .25Spent Fuel Pit/Auxiliary Building FHA, AB open or closed to containment TVA-F-W000624 3 = RG 1.25 Containment FHA with Purge Filters (no containment isolation) 4=RG 1.183 AST Auxiliary Building FHA with no ABI 5=RG 1.183 AST Auxiliary Building FHA with ABI (ABGTS)
R12: Name Code Description TS9Sl2$# STP Release models R12: TS9Cl2$# COROD Control room dose with I train of CREVS, 40.0 second control room isolation TVA-F-W001564 TS9Fl2$# FENCDOSE Offsite dose Where:
$ = A = standard core B =Tritium Production Core, once burned assembly C = Tritium Production Core, twice burned assembly D = Tritium Production Core, thrice burned assembly E = TPBAR only
#=I= RG 1.25 Contain FHA w/ 12.7 sec contain isolation, containment closed to AB, no Purge Filters 2 = RG l .25Spent Fuel Pit/Auxiliary Building FHA, AB open or closed to containment Rl3: 3 = RG 1.25 Containment FHA with Purge Filters (no containment isolation)
TVA-F-W002485 4 = RG 1.183 AST Auxiliary Building FHA with no AB! And 5 = RG 1.183 AST Auxiliary Building FHA with ABI (ABGTS) TVA-F-W002502 R13: Name Code Description TS9C13$# COROD Control room dose with I train ofCREVS, 40.0 second control room isolation TS9F13$# FENCDOSE Offsite dose Where:$ =see R12 above; #=see R12 above and 6=7 sec unfiltered release from AB, 7=7 sec unfiltered AB release & CR already isolated, 8=TPBAR only. R14: R14: TVA-F-W02566 TS9S14$# STP source terms TS9Cl4$# COROD Control room dose with I train ofCREVS, 40.0 second control room isolation Where:$ =see R12 above; #=I I =case 11 for RG 1.183 containment FHA, 12.7 sec SB unfiltered release, remainder unfiltered release through AB vent
page 9A TVAN COMPUTER OUTPUT MICROFICHE INFORMATION SHEET Document WBNTSR009 IRev. 16 IPlant: WBN I
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Microfiche Description Rl5: Rl5: TVA-F-W003205 Name Code Description TSR009Rl 5S##$ STP source term TSR009Rl 5C##$ CO ROD control room operator dose TSR009Rl 5F##$ FENCDOSE offsite dose where
## = 3 = RG 1.25 Containment FHA with Purge Filters (no containment isolation) 4 = RG 1.183 AST Auxiliary Building FHA with no ABI 11= RG 1.183 Containment FHA, 12.7 sec SB unfiltered release, remainder unfiltered release through AB vent
$ = A = standard core B = Tritium Production Core, once burned assembly C = Tritium Production Core, twice burned assembly D =Tritium Production Core, thrice burned assembly Total of 44 output files for Rl5: 12 STP output files, 12 STP punch files, 12 COROD output files, and 8 FENCDOSE output files This page is added by Revision 015
,1
- Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page: 10
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Purpose The purpose of this calculation is to determine the dose to the control room operators following a design basis Fuel Handling Accident (FHA). In addition, the offsite doses resulting from a FHA is also to be determined. This calculation is to address concerns raised during the vertical slice review program as to whether the Loss of Coolant Accident (LOCA) actually produces the bounding control room operator doses (ref. I). Special Requirements/Limiting Conditions If the design basis for WBN is RG 1.25, then ifthe equipment hatch or any penetration between the Auxiliary Building and Containment is open, the containment purge system shall be operational during fuel movement and an Auxiliary Building Isolation (ABI) due to a high radiation signal shall initiate a Containment Ventilation Isolation (CVI) and a CVI due to a high radiation signal must initiate an ABI. If other penetrations are open to the outside of the ABSCE, the ABGTS system must be able to draw down within 4 minutes of the initiating event. Also, for RG 1.25, the HVAC intake vent in the transfer canal must be blocked, and the -103 monitor must be raised so that it has a line of sight across the 757' floor. The HVAC intake vents for the cask loading area shall be blocked when handling irradiated fuel in this area. The -102 monitor is far enough away so that it will see very close to the floor at the canal/cask loading area, therefore it will not have to be raised (see assumption 17 for further discussion). This requirement is to prevent radioisotopes from entering the HVAC ductwork in the transfer canal (and ultimately released via the Auxiliary Building Vent without filtration) and therefore bypassing the isolation function of the -102 and -103 radiation monitors. If the design basis for WBN is RG 1.183 (AST), then there are no special requirements or limiting conditions. Based on the results of this analysis, no isolation of either containment or the auxiliary building is required following a Fuel Handling Accident for AST. Introduction The computer code STP is used to determine the releases. Using the STP output, the computer code FENCDOSE determines the offsite doses, and the computer code COROD determines the control room doses. The FHA accident is analyzed for both the Auxiliary Building and the Containment. Also, 4 types of assemblies are analyzed: the 1500 EFPD end of life assembly for a standard core, a once burned TPC assembly with 24 TPBAR rods (which contain the tritium), a twice burned TPC assembly with 24 TPBAR rods, and a three times burned TPC assembly (no TPBARs).
1 1 Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page: 11
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Assumptions
- 1. The FHA occurs at 100 hours after shutdown, consistent with the FSAR and the Technical Specifications (ref.4 and 18).
- 2. All of the rods in the worst fuel assembly are assumed to be damaged.
Technical Justification: Safety Guide 25, ref.4, implies that the activity from the worst peak assembly is released. It is conservative to assume that all rods will break, thereby maximizing the release. Regulatory Guide 1.183 (AST), ref.36, section 3.6 requires that the case with the highest radioactivity release should be analyzed.
- 3. For all cases except the 12.7 containment isolation case, it is assumed that everything is released to the environment within 2 hours (ref.4, 36). To assure this, at 2 hours all remaining isotopes above the spent fuel pool (or in containment) are stepped into the environment (using the appropriate filter efficiency as a multiplication factor).
- 4. deleted in Rl4
- 5. For the RG.1.25 containment FHA cases (case series 1, found in Appendix A), all of the gap activity in the damaged rods is released which consists of 10% of the inventory in the rods at the time of the accident (ref.4), except for the following (per NUREG/CR-5009 for 60 GWd/t, note for lesser burnups the releases are less, therefore use of these 60 GWd/t values for all burnups is conservative):
Kr-85 = 14% Kr-87 = 10% Note: The NUREG/CR-5009 value is actually 0.7%. Since STP is limited to 9 classes, and the halflife ofKr-87 is 76 min (ref33), after 100 hours of decay there will be exp(-100*ln(2)/(76/60)) = l.7E-24 or l.7E-22% left. Therefore the increase in the gap percentage does not affect the results. Kr-88 = 10% Note: The NUREG/CR-5009 value is actually 1%. Since STP is limited to 9 classes, and the halflife ofKr-88 is 2.84 hr (ref.33), after 100 hours of decay there will be exp(-1 OO*ln(2)/2.84) = 2.5E-l l or 2.5E-9% left. Therefore the increase in the gap percentage does not affect the results. Kr-89 = 10% Xe-133 = 5% Xe-135 = 2% 1-131=12% For the RG 1.183 AST cases, all of the gap activity in the damaged rods is released which consists of8% 1-131, 10% Kr-85, 5% other noble gasses and other halogens. Note that RG 1.183 also specifies 12% of Alkali metals (Cs, Rb), however since particulates have essentially an infinite partition factor, no alkali metals will be released and therefore are not included in this analysis.
- 6. The values assumed for individual fission product inventories are calculated assuming full power operation at the end of core life immediately preceding shutdown with a radial peaking factor of 1.65 (ref.4, 36) for the standard core assembly. For the TPC assemblies, the inventories are taken at the end of cycle, with the factor of 1.65 applied to all isotopes except tritium.
Also, the factor of 1.65 is the maximum peaking factor allowed by the COLR. The factor of 1.65 is not applied to the tritium isotope because the maximum inventory of tritium is used already at a maximum (see assumptions #13, and ref.29) at l.2g tritium/rod with 24 rods/assembly. It would be too conservative to apply the 1.65 to a value which is already the maximum inventory which can occur.
- 7. From RG 1.25 (ref.4), the iodine gap inventory is composed of inorganic species (99.75%) and organic species (0.25%).
From RG 1.183, the inorganic species is 99.85% and the organic species is 0.15%. An overall decontamination factor is utilized in the RG 1.183 cases (see assumption 8), therefore the makeup of the species is not utilized in AST.
- 8. From RG 1.25, the pool decontamination factors for the inorganic iodine is assumed to be 133, and organic iodine is assumed to be 1 (ref.4). From RG 1.183 (AST) the decontamination factors are specified to be 500 for elemental (inorganic) iodine, and 1 for organic iodine. Doing the math, this leads to an overall decontamination factor of 286
=1/(0.9985/500+0.0015/l). However, RG 1.183 also specifies an overall decontamination factor of200. The use of the 200 factor is more conservative (also, BFN was asked by the NRC to use the overall factor instead of the species specific factors),
and therefore the overall factor of 200 for AST will be used in this analysis.
- 9. The retention ofnoble gasses in the pool is negligible (ref.4, 36).
l Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page: 12
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident
- 10. For FHA in containment, with isolation (case series 1 and 11 ), it is assumed that the Purge Air Exhaust (PAE) System isolates in 12.7 seconds (ref. 2). This includes instrument loop response time (6.7 sec) and containment purge valve closure time (6 sec). This should be noted to be a very conservative value. The instrument loop response time contains very conservative assumptions and rounding. In the event that containment needs to be purged (for instance if entry is required into containment), then it is possible to defeat the isolation.
- 11. For RG 1.25 cases, the filter efficiencies for the PAE filter are 90% for inorganic iodines and 30% for organic iodines (ref.3). EDC 51930 downgrades the filters to non-safety related. R.G. 1.140 R3 will be the standard to which these filters conform to. The guide specifies the filter efficiency as 95%. Therefore using the original 90%/30% is conservative.
- 12. The filter efficiency for the ABGTS is 99% for all iodines ( ref.3 ).
- 13. It is assumed that all 24 TPBARs break (in a TPC once or twice burned assembly) or a TPBAR only accident. It is also assumed that 25% of the tritium in the spent fuel pool is released following the FHA through evaporation of the pool.
Technical Justification: All TPBARs breaking is conservative. There will not be 100% release of tritium from a TPBAR failure in a FHA because there are no high temperatures involved with the accident. Reference 26, Section 2.3, states that the release from the TPBARs will not cause the water tritium concentration to exceed 60µCi/gm. At this concentration, the total spent fuel pool inventory would be 84,490 Ci. 60µCi/gm
- 372,000 gal* 3,785.4 cc/gal* 1 gm/cc* IE-6 Ci/µCi = 84490 Ci A large fraction of the spent fuel pool will not evaporate in 8 hours ifthe spent fuel pool cooling system maintains the temperature below the boiling point. Furthermore, from page 23 of reference 38, even ifthe normal spent fuel pool cooling system is not in service, the pool will not reach 212 "F for at least 9 hours. Finally, from page 22 of reference 38, in the unlikely event that the spent fuel pool does boil, the boil off rate is 24,496.7 lb/hr, which is approximately 3000 gal/hr. Over a period of 8 hours, 3000 gal/hr is 24,000 gal, which is less than 6.5% of the pool volume. Therefore, assuming that 25% of the water evaporates in 2 hours or less is conservative.
- 14. For the RG 1.25 case, the effective volume of upper containment is taken as 1/2 the upper containment free volume.
Technical Justification: This takes into account incomplete mixing and dead end spaces and is typical for the representation of air mixing volumes.
- 15. deleted in Rl4
- 16. NUREG/CR-5009 implies that Cs-134 and Cs-137 are also in the gap. This calculation assumes these isotopes do not get released to the environs.
Technical Justification: Cs-134 decays to either Xe-134 or Ba-134, both of which are stable. Cs-137 decays to Ba-137m which in turn decays to Ba-137, which is stable. Per Regulatory Guide 1.183, particulates (Cs, Ba) have an infinite decontamination factor in the spent fuel pool/reactor vessel water. Therefore, Cs-134 and Cs-137, and their daughters, may be neglected from the calculation.
- 17. deleted in Rl4
- 18. The RG 1.25 cases utilize exponential releases. That is, the releases are governed by the mixing volume and the exhaust flow rate. This results in conservative releases compared to linear releases as more gets released in the beginning of the accident when there is less control room filtration (it takes 74 sec to isolate the control room) and also allows more to be released prior to isolation.
The RG 1.183 cases utilize linear releases. That is, all releases are constant over the 2 hour time period. This is implied in RG 1.183 by requiring all releases to be within 2 hours. (RG 1.183 section 3.3 and RG 1.183 Appendix B section 4.3). Also, this methodology is utilized by Westinghouse for SQN and other utilities.
1 I l l Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 13
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident
- 19. Only one train ofCREVS is in operation. Nonnally, each CREVS train takes suction from separate intakes with no cross communication between trains. This leads to one contaminated train, and one uncontaminated train. The only way a 2 CREVS operation could result in higher doses would be for both trains to take suction from the same vent. For this to happen, one intake path would require a failed closed intake path AND a fail open of nonnally closed passive manual damper at the beginning of the accident. An active failure of a train plus a failure of a passive component in less than 24 hours is beyond design basis.
Calculations This calculation considers several cases broken down into Regulatory Guide 1.25 and Regulatory Guide 1.183 (AST) groupings. I. Regulatory Guide 1.25 Cases: One case is for a FHA in containment with the activity released directly to the environment until containment isolation (12.7 seconds), there are no penetrations open to the AB, and the PAE filters are not credited (case series 1, found in Appendix A). Another case utilizes a containment release without isolation but with Purge Filters Credited (case series 3). Computer code STP (ref.6) is used to calculate the activity released after a FHA. Figure 1 shows the model. To insure a conservative dose, the radioisotopes are allowed only 100 hours of decay after shutdown, and are released from the containment based on PAE flow. The step source fractions of the core inventory are based on NUREG/CR-5009 and Reg.Guide 1.25. The source tenns are the 1500 EFPD maximum burnup for 18 month fuel cycle from WBNAPS3-084 (ref.14) for the standard core. These source tenns are used instead of the core average 1000 EFPD source terms because the accident involves a single fuel assembly, not the entire core (as in a LOCA). For the TPC, the source tenns for the once burned, twice burned, and 3 times burned assemblies are taken from WBNAPS3-098 (ref.29). The 24 TPBAR release apply only to the once and twice burned assemblies (the 3 times burned assembly will not have any TPBARs). The following flow chart of the STP model includes the ABGTS filters. This is an historical artifact. No Auxiliary Building releases with filtration cases are used in R15. These flows are left in the model, but the results for this flow path are not used and do not impact the results.
1 Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:14
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Figure 1 STP Model 1 Fuel 100-hr Decay 2 Containment 3 Spent Fuel Pit Air 7 ABGTS 6PAE Filter Filter 4 Containment 5 ABGTS Release Release Note: the arrow from component 2 to 4 (for crediting purge filters) and 3 to 5 does not imply a filter bypass. It indicates how STP models a filter with the "U" card, where F 2-6 = F*(efficiency), F2-4 = F*(l-efficiency) Component 1: Fuel volume=l.O (arbitrary) Component 2: Containment Air volume= 647,000 cuft (ref.30) /2 = 3.235E5 cuft (see assumption #15) Component 3: Spent Fuel Pit volume= 10,017 cuft = 39.5'x31.7'x8' (ref.31). Note: the dimensions come from ref.3 lb. The 8' dimension (air above the pool) is an arbitrary value to account for the rise of the gasses above the pool. This is reasonable and consistent with references 31 a and 31 c. Component 4: Containment Release volume =1.0 (arbitrary) Component 5: ABGTS Release volume= LO (arbitrary) Component 6: PAE Filter volume =1.0 (arbitrary) Component 7: ABGTS Filter volume =1.0 (arbitrary) Flow from containment through PAE to release (U 2 6 4)= purge rate= 14954 cfm (ref.30, note the actual value should be 14958 cfm, but this will not change the results so is not corrected)= 8.9724E5 cfh with Purge Air Exhaust filter efficiencies: 90% inorganic iodine, 30% organic (ref.3), 0% for tritium Flow from spent fuel pit through ABGTS to release (U 3 7 5) = ABGTS flow= 9900 cfm = 5.94E5 cfh (see assumption
#16) with filter efficiencies of99% for iodines. NOTE: This is not used in Rl5, but left in for historical context, and possible future revisions.
Fuel activities are as given in WBNAPS3-084 (refl4) and WBNAPS3-098 (ref.29), with the inorganic iodines equal to 99.75% of total, and organic iodines equal to 0.25% of total iodines. Peaking Factor for the highest activity fuel assembly= 1.65 (ref.4) except for tritium isotope, which is 1.0 (see assumption
#6).
ABGTS filter efficiencies: 99% (ref.3), for iodines, 0% for tritium NOTE: This is not used in Rl5, but left in for historical context, and possible future revisions. The gap activity in the damaged rods is released which consists of 10% of the inventory in the rods at the time of the accident, except for the following: Kr-85=14%, Xe-133=5%, Xe-135=2%, 1-131=12% Partition Factors: 133 for inorganic iodine.
,l I a Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 15
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident The step fractions from the fuel to the containment (or spent fuel pit) are: S = 0.1 for Kr-83m, Kr-85m, Kr-87, Kr-88, Kr-89, Xe-13lm, Xe-133m, Xe-135m, Xe-138, organic iodine (except I-131) S= 0.14 for Kr-85 S=0.05 for Xe-133 S=0.02 for Xe-135 S=0.000752 for I-132, I-133, I-134, I-135 (=0.1/133) S=0.000902 for I-131 (=0.12/133) S=0.12 for I-131 (organic iodine) All of the activity is to be released after 2 hours. To simulate this all activity remaining in the Reactor Building at the end of2 hours is put into a new "source" which is stepped to the environment. The stepping fraction is equal to what would have gotten through the filters (i.e. I-efficiency) had the isotopes been released through the filters. For the Containment case with isolation, the purge flow (F 2 4 0) is set to 0 cfm after 12.7 seconds. The activity released to the environment as calculated by STP is used as input to computer code COROD (ref.7) to determine the control room operator doses. The control room model is identical to that described in TI-RPS-198 (ref.5) except for the shine from containment which is neglected (all activity inside the containment from FHA is released). During the vertical slice review of the control room, a concern was raised that when the control room is isolated by a signal from the main control room intake radiation monitors, some amount of unfiltered activity could enter the control room before the isolation dampers close (ref.9). This could be the case for a fuel handling accident because there will be no safety injection signal to isolate the control room. The isolation dampers downstream from the radiation monitors are O-FCV-31-3 and O-FCV-31-4 (ref. IO). It is required by reference 11 that the closure time of the dampers is 14 seconds, with a signal response time of 60 seconds (ref.13), which gives a total closure time of 74 seconds. Therefore all cases will analyze the first 74 seconds without CREVS filtration. The ARCON96 X/Q values used (which supersede the Halitsky X/Q values) for the Shield Building Vent were: from ref.34: 1.09E-03 sec/m3 for 0-2 hr (since all releases are< 2 hours, X/Q values after 2 hours are unimportant. The Auxiliary Building Vent X/Q is 2.56E-3 sec/m3* Prior to isolation the intake flow is 3000 cfm (ref. I 0). (3200 cfm has been used in previous revisions and was previously shown on the drawing 1-47W866-4. The 3200 cfm will be retained in this calculation revision since this value produces conservative results.) It is assumed that the unfiltered in leakage is the same as for the isolated case (51 cfm, due to open doors, leaky valves, etc.) After isolation, the total flow rate into the control room is 711 cfm filtered plus 51 cfm unfiltered (ref.5). The circulation flow rate in the control room is the total flow- the makeup flow= 3600 - 711 = 2889 cfm (ref.5). Note: the recirculation nominal flow is 4000 cfm. The 4000 cfm- 10% is chosen as this will result in longer residence time in the Control Room. If 4000 cfm + 10% were to be used, the radioisotopes would recirculate through the filters faster, and therefore be filtered out of the Control Room faster. Cases were performed for the standard core using ARCON96 XIQ values and ICRP-30 dose conversion factors (see note on methodologies in Conclusion section). The activity released to the environment as calculated by STP is used as input to computer code FENCDOSE (ref.8) to determine the site boundary dose. The FENCDOSE model is the same as that found in reference 19. II. Regulatory Guide 1.183 (Alternate Source Term) Cases: There are two AST FHA cases. One is in the spent fuel pit/Auxiliary Building with no ABI and with unfiltered releases through the Auxiliary Building vent. A second case is an accident in the containment with unfiltered release for 12. 7 seconds through the Shield Building vent, with the remainder released unfiltered through the Auxiliary Building vent (no ABI and no filtration). The AB Vent has less favorable X/Q values than the Shield Building Vent. Computer code STP (ref.6) is used to calculate the activity released after a FHA. Figure 2 shows the model. To insure a conservative dose, the radioisotopes are allowed only 100 hours of decay after shutdown, and are released to the environment linearly. The step source fractions of the core inventory are based on Reg.Guide 1.183. The source terms are the 1500 EFPD maximum burnup for 18 month fuel cycle from WBNAPS3-084 (ref.14) for the standard core. These source terms are used instead of the core average 1000 EFPD source terms because the accident involves a single fuel assembly, not the entire core (as in a LOCA). For the TPC, the source terms for the once burned, twice burned, and 3 times burned assemblies are taken from WBNAPS3-098 (ref.29). The 24 TPBAR release apply only to the once and twice burned assemblies (the 3 times burned assembly will not have any TPBARs).
1 Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 16
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Figure 2 AST STP Model 1 Fuel 100*hr Decay 2 Release The STP model consists of the assembly inventory stepped into the Fuel component with a 1.65 peaking factor and allowed to decay for 100 hours. The remaining decayed isotopes are then stepped into the Release component based on filtration efficiency (=99% for iodines for ABI case, =0% filtered for no ABI case). Component 1: Fuel volume= 1.0 (arbitrary) Component 2: Release volume= 1.0 (arbitrary) ABGTS filter efficiencies: 99% (ref.3), for iodines, 0% for tritium The gap activity in the damaged rods is released which consists of 5% of the inventory in the rods at the time of the accident, except for the following: Kr-85=10%, 1-131=8%. H-3 = 0.25 (see assumption 13). (see Release Fraction in a table below). NOTE: Filtration is not used in Rl5, but left in for historical context, and possible future revisions. Partition Factors: 200 for all iodines (see assumption 8) (see Partition Factor in a table below). The step fractions for both cases are calculated as shown in a table below. Release Partition Release Step Fraction Class Isotopes Fraction Factor 0-74 sec 0-12.7 sec 12.7-74 sec 74-7200 sec 1 n~bl~~~-!3-~cee!.f2r_!i_S.!!3-_~-- ,..__ _Q:05 ___ 1 5.13~E-Q~ ~~~-1-~E:9_5-.J 4.257E-04 I 4.949E-02 2 Kr-85 0.1 1 1.028E-03 1.764E-04: 8.514E-04 i 9.897E-02 3 iodines excepfl:faf------~-- 0.05 200 2:569E-o64A10E:Oir2.128E-06 i 2.474E-04
- 4-or:g-a_nic iocli!!~~=~~C0j)iT.:1:fl -01)5- 200 2.569E-os 4.410E-07 i 2.128E-os 1 2.474E-04 5 H~~-- _______ _*-***--**--*---*---*--*- 0.25 1 2.569E-03 4.41 OE-04 I 2.128E-03 i 2.4 74E-01 t
6 1-131 0.08 200 4.111E-06 7.056E-07' 3.406E-06 l 3.959E-04
--i-0r!:iarlic-1~TI1 ---------*-*** 0.08 _~=-- *2oq_.__ 4.1_~1E-os 11.05sE-01 1 3.4osE-os 3.959E-o4 8 Xe-133 0.05 1 5.139E-041 8.819E-05 i 4.257E-04 4.949E-02 ,____ 9___ Xe~-135*m-- .. ****------~-~--r-* 0.05 1 5.1.39E-04) 8.B19E-05 i 4.257E-04 i 4.949E-02 Note: example 0-74 sec noble gas: 0.05/1*74/7200=5.139E-4 The activity released to the environment as calculated by STP is used as input to computer code COROD (ref.7) to determine the control room operator doses. The control room model is identical to that described in TI-RPS-198 (ref.5) except for the shine from containment which is neglected (all activity inside the containment from FHA is released). For AST, all breathing rates for all times are the same 3.47E-4 m3/sec (note: this should be 3.5E-4. Also, the TEDE as determined by COROD or FENCDOSE does not use this value. See Appendix B for conservatism and acceptability of use of FENCDOSE/COROD TEDE calculation methodology).
The ARCON96 X/Q values used for Shield Building Vent releases (which supersede the Halitsky X/Q values) are: from ref.34: l .09E-03 sec/m 3 for 0-2 hr. For Auxiliary Building Vent releases (when there is no ABI), the X/Q value is: 2.56E-3 sec/m3 for 0-2 hr. Prior to isolation the intake flow is 3000 cfm (ref.10). (3200 cfm has been used in previous revisions and was previously shown on the drawing l-47W866-4. The 3200 cfm will be retained in this calculation revision since this value produces conservative results.) It is assumed that the unfiltered inleakage is the same as for the isolated case ( 51 cfm, due to open doors, leaky valves, etc.) After isolation, the total flow rate into the control room is 711 cfm filtered plus 51 cfm unfiltered (ref.5). The circulation flow rate in the control room is the total flow- the makeup flow= 3600 - 711 = 2889 cfm (ref.5). The activity released to the environment as calculated by STP is used as input to computer code FENCDOSE (ref.8) to determine the site boundary dose. The FENCDOSE model is the same as that found in reference 19.
, l, l i Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 17
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident III. TPBAR Only Accident The TPBAR only accident will result in 25% of the tritium inventory being released over 2 hours. Since tritium is low energy beta decay only, the Spent Fuel Pit Monitors and the Control Room Intake monitors will not respond to the tritium. Therefore, the Auxiliary Building will not be isolated (all Auxiliary Building releases will be out the Auxiliary Building Results A. Control Room Doses TPBAR Onlv Control Room Doses CR Dose (rem) Limit (rem) Core Type TPBARonlv - COROD File TS9C13E8 - Gamma O.OOE+OO 5 Beta 7.08E-02 30 Thyroid (IRCP-30) O.OOE+OO 30 TEDE 1.16E+OO 5 Revn atorv Gm"de 125 . ControIRoom Doses RG 1.25 Containment FHA Control Room Doses (rem) Core Type TPC Thrice Conventional Core TPC Once Burned TPC Twice Burned Burned COROD File TSR009R15C3A TSR009R15C3B TSR009R15C3C TSR009R15C3D Gamma 2.92E-01 3.34E-01 2.52E-01 3.29E-01 Beta 2.41E+OO 2.70E+OO 2.10E+OO 2.69E+OO Thyroid (IRCP-30) 1.07E+01 1.13E+01 8.85E+OO 1.19E+01 TEDE 6.26E-01 1.27E+OO 1.11E+OO 6.99E-01 Regulatory Guide 1.183 Alternate Source Term (AST) Control Room Doses [rem] Aux.iliatry Building FHA with no ABI (Auxiliary Building Vent Release) RG 1.183 Auxiliary Building FHA Control Room Doses (rem) Core Type TPC Thrice Conventional Core TPC Once Burned TPC Twice Burned Burned COROD File TSR009R15C4A TSR009R15C4B TSR009R15C4C TSR009R 15C4D Gamma 5.95E-01 6.81 E-01 5.13E-01 6.69E-01 Beta 4.75E+OO 5.38E+OO 4.15E+OO 5.31E+OO Thyroid (IRCP-30) 1.51 E+01 1.61 E+01 1.25E+01 1.68E+01 TEDE 1.08E+OO 2.39E+OO 2.10E+OO 1.21E+OO Contai nment FHA with 12.7 sec release, no Purge filters, remaining release through AB vent (no ABI, no filters) RG 1.183 Containment FHA Control Room Doses (rem) Core Type TPC Thrice Conventional Core TPC Once Burned TPC Twice Burned Burned COROD File TSR009R15C11A TSR009R15C11B TSR009R15C11C TSR009R 15C 11 D Gamma 5.93E-01 6.78E-01 5.11E-01 6.66E-01 Beta 4.73E+OO 5.36E+OO 4.13E+OO 5.28E+OO Thyroid (IRCP-30) 1.47E+01 1.56E+01 1.22E+01 1.63E+01 TEDE 1.06E+OO 2.33E+OO 2.05E+OO 1.19E+OO
, 1 l l Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 18
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident B. Offsite Doses TPBAR Only Offsite Doses Offsite Dose (rem) Limit (rem) Core Type 2-hr EAB 30-day LPZ - FENCDOSE File TS9F13E8 - Gamma O.OOE+OO O.OOE+OO 6.25 Beta 1.76E-02 4.92E-03 75 Thyroid (IRCP-30) O.OOE+OO O.OOE+OO 75 TEDE 2.88E-01 8.06E-02 6.25 Regulatory Guide 1.25 Offsite Doses RG 1.25 Containment FHA Offsite Doses (rem) Core Type TPC Thrice Conventional Core TPC Once Burned TPC Twice Burned Burned FENCDOSE File TSR009R 15F3A TSR009R 15F3B TSR009R 15F3C TSR009R 15F3D 2-hr EAB Gamma 4.31 E-01 4.91 E-01 3.71 E-01 4.84E-01 Beta 1.24E+OO 1.39E+OO 1.0BE+OO 1.39E+OO Thyroid (IRCP-30) 4.15E+01 4.40E+01 3.43E+01 4.61E+01 TEDE 1.85E+OO 2.27E+OO 1.83E+OO 2.06E+OO 30-day LPZ Gamma 1.21E-01 1.37E-01 1.04E-01 1.35E-01 Beta 3.47E-01 3.90E-01 3.02E-01 3.87E-01 Thyroid (IRCP-30) 1.16E+01 1.23E+01 9.60E+OO 1.29E+01 TEDE 5.17E-01 6.34E-01 5.11E-01 5.75E-01 Rem 1latorv Guide 1.183 Alternate Source Tenn (AST) Offsite Doses RG 1.183 AST FHA Offsite Doses (rem Core Type TPC Thrice Conventional Core TPC Once Burned TPC Twice Burned Burned FENCDOSE File TSR009R 15F4A TSR009R 15F4B TSR009R 15F4C TSR009R 15F4D 2-hr EAB Gamma 4.29E-01 4.89E-01 3.69E-01 4.82E-01 Beta 1.19E+OO 1.35E+OO 1.04E+OO 1.33E+OO Thyroid (IRCP-30) 5.51E+01 5.85E+01 4.56E+01 6.12E+01 TEDE 2.38E+OO 2.83E+OO 2.27E+OO 2.65E+OO 30-day LPZ Gamma 1.20E-01 1.37E-01 1.03E-01 1.35E-01 Beta 3.34E-01 3.78E-01 2.91E-01 3.73E-01 Thyroid (IRCP-30) 1.54E+01 1.63E+01 1.28E+01 1.71 E+01 TEDE 6.66E-01 7.92E-01 6.34E-01 7.41E-01 Containment FHA with 12.7 sec release, no Purge filters, remaining release through AB vent (no ABI, no filters) Same offsite doses as Auxiliary Building FHA with no ABI
, 1 I l Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page:19
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident C. Control Room Dose Margins The following Unit 1 margins were calculated from the doses in the Tables above. Where: margin = limit - dose, and percent
= (limit-dose)/limit TPBAR Only Control Room Dose Margins TPBAR Only CR Dose Margins Core Type Margin Percent (rem) (%)
Gamma 5.0 100.0 Beta 29.9 99.8 Thyroid (IRCP-30) 30.0 100.0 TEDE 3.8 76.9 Regulatory Guide 1.25 Control Room Dose Margins RG 1.25 Containment FHA Control Room Dose Margins Conventional Core TPC Once Burned TPC Twice Burned TPC Thrice Burned Core Type Margin Percent Margin Percent Margin Percent Margin Percent (rem) (%) (rem} (%) (rem) (%) (rem) (%) Gamma 4.7 94.2 4.7 93.3 4.7 95.0 4.7 93.4 Beta 27.6 92.0 27.3 91.0 27.9 93.0 27.3 91.0 Thyroid (IRCP-30) 19.3 64.4 18.7 62.2 21.2 70.5 18.1 60.4 TEDE 4.4 87.5 3.7 74.7 3.9 77.9 4.3 86.0 Regulatory Guide 1.183 Alternate Source Term (AST) Control Room Dose Margins AuxT 11ary B m'Id.mg FHA Wit
. h no ABI (AUXlTtary Bm*1d*mg V ent R e Iease)
RG 1.181 Auxiliary Buildin1:1 FHA Control Room Dose Margins Conventional Core TPC Once Burned TPC Twice Burned TPC Thrice Burned Core Type Margin Percent Margin Percent Margin Percent Margin Percent (rem) (%) (rem) (%) (rem) (%) (rem) (%) Gamma 4.4 88.1 4.3 86.4 4.5 89.7 4.3 86.6 Beta 25.3 84.2 24.6 82.1 25.9 86.2 24.7 82.3 Thyroid (IRCP-30) 14.9 49.6 14.0 46.5 17.5 58.2 13.2 44.0 TEDE 3.9 78.4 2.6 52.3 2.9 57.9 3.8 75.8 Conta inment FHA with 12.7 sec release, no Purge filters, remaining release through AB vent (no ABI, no filters) RG 1.181 Containment FHA Control Room Dose Margins Conventional Core TPC Once Burned TPC Twice Burned TPC Thrice Burned Core Type Margin Percent Margin Percent Margin Percent Margin Percent (rem) (%) (rem) (%) (rem) (%) (rem) (%) Gamma 4.4 88.1 4.3 86.4 4.5 89.8 4.3 86.7 Beta 25.3 84.2 24.6 82.1 25.9 86.2 24.7 82.4 Thyroid (IRCP-30) 15.3 51.1 14.4 48.1 17.9 59.5 13.7 45.7 TEDE 3.9 78.7 2.7 53.5 3.0 59.1 3.8 76.2
l ,, I
- Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:20
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident D. Offsite Dose Margins TPBAR Only Offsite Dose Margins TPBAR Only Offsite Dose Margins 2-hr EAB 30-day LPZ Core Type Margin Percent Margin Percent (rem) (%) (rem) (%) Gamma 6.3 100.0 6.3 100.0 Beta 75.0 100.0 75.0 100.0 Thyroid (IRCP-30) 75.0 100.0 75.0 100.0 TEDE 6.0 95.4 6.2 98.7 Regulatorv Guide 1.25 Offsite Dose Margins RG 1.25 Containment FHA Offsite Dose Mal' ins Conventional Core TPC Once Burned TPC Twice Burned TPC Thrice Burned Core Type Margin Percent Margin Percent Margin Percent Margin Percent (rem) (%) (rem) (%) (rem) (%) (rem) (%) 2-hr EAB Gamma 5.8 93.1 5.8 92.1 5.9 94.1 5.8 92.3 Beta 73.8 98.3 73.6 98.1 73.9 98.6 73.6 98.2 Thyroid (IRCP-30) 33.5 44.7 31.0 41.3 40.7 54.2 28.9 38.6 TEDE 4.4 70.4 4.0 63.7 4.4 70.8 4.2 67.1 30-da" LPZ Gamma 6.1 98.1 6.1 97.8 6.1 98.3 6.1 97.8 Beta 74.7 99.5 74.6 99.5 74.7 99.6 74.6 99.5 Thyroid (IRCP-30) 63.4 84.5 62.7 83.6 65.4 87.2 62.1 82.8 TEDE 5.7 91.7 5.6 89.9 5.7 91.8 5.7 90.8 Regulatory Guide l.183 Alternate Source Term (AST) Offsite Dose Margins RG 1.183 AST FHA Offsite Dose Margins Conventional Core TPC Once Burned TPC Twice Burned TPC Thrice Burned Core Type Margin Percent Margin Percent Margin Percent Margin Percent (rem) (%) (rem) (%) (rem) (%) (rem) (%) 2-hr EAB Gamma 5.8 93.1 5.8 92.2 5.9 94.1 5.8 92.3 Beta 73.8 98.4 73.6 98.2 74.0 98.6 73.7 98.2 Thyroid (IRCP-30) 19.9 26.5 16.6 22.1 29.4 39.2 13.8 18.4 TEDE 3.9 61.9 3.4 54.7 4.0 63.7 3.6 57.6 30-day LPZ Gamma 6.1 98.1 6.1 97.8 6.1 98.4 6.1 97.8 Beta 74.7 99.6 74.6 99.5 74.7 99.6 74.6 99.5 Thyroid (IRCP-30) 59.6 79.5 58.7 78.2 62.3 83.0 57.9 77.2 TEDE 5.6 89.3 5.5 87.3 5.6 89.9 5.5 88.2 Containment FHA with 12.7 sec release, no Purge filters, remaining release through AB vent (no ABI, no filters) Same offsite dose margins as Auxiliary Building FHA with no ABI
l! Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 21
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Conclusions The control room operator doses resulting from a Fuel Handling Accident are less than the 10CFR50, Appendix A. GDC 19 limits of5 rem gamma, 30 rem beta, 30 rem thyroid, and less than the 10CFR50.67 limit of 5 rem TEDE. The 2 hour Site Boundary (SB)/Exclusion Area Boundary and 30 day Low Population Zone (LPZ) doses from a FHA are less than 25% of the 10CFRlOO limits of25 rem gamma, 300 rem beta, and 300 rem thyroid (=6.25 rem gamma, 75 rem beta, 75 rem thyroid, 6.25 rem TEDE). IOCFR50.67 provides the TEDE equivalence to the gamma limits. The Auxiliary Building/Spent Fuel Pit releases with no ABI or filtration are through the Auxiliary Building vent which have worse X/Q values than the containment FHA with Shield Building releases. Therefore the Auxiliary Building AST FHA will bound a containment AST FHA. TPBAR only accident doses are all less than the l OCFRl 00/ 10CFR50.67 offsite control room limits and 10CFR50 App.A GDC 19 limits. Note on methodologies used: This calculation determined the doses using different methodologies. The gamma, beta and Thyroid (JCRP-30) doses are all based on TID-14844 methodologies utilizing the ICRP-30 iodjne dose conversion factors. The other methodology used is the TEDE (Total Effective Dose Equivalent). The TEDE presents an overall weighted dose and is more representative of the impact of all isotopes on the body as a whole. The TEDE dose is required for AST, however is not required for RG 1.25 methodology. It is important to note that tritium does not impact the thyroid doses utilizing the TID-14844 methodology, because only iodine is applied to the thyroid dose. However, in fact tritium does contribute to the thyroid dose, as well as other organs of the body. This is why the TEDE is a more representative dose when discussing the impact of tritium. The effect of the increased time delay had been evaluated in this calculation, and it has been shown that adequate margin remains between the calculated doses and regulatory limits to accommodate instrument errors and uncertainties. Refer ences I. Discrepancy Report No. 274, RO, 10114/88
- 2. l-RE 130 RI 8, "Demonstrated Accuracy Calculation for Containment Building Purge Air Exhaust Monitors"
- 3. Regulatory Guide 1.52, "Design, Testing And Maintenance Criteria For Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration And Adsorption Units Of Light-Water-Cooled Nuclear Power Plants,"
Revision 2, March 1978.
- 4. Safety Guide 25, "Assumptions Used For Evaluating The Potential Radiological Consequences Of A Fuel Handling Accident In The Fuel Handling And Storage Facility For Boiling And Light Water Reactors," 3/23/72
- 5. TI-RPS-198 R24 "Dose to Control Room Personnel Due to a Regulatory Guide L.4 Loss of Coolant Accident,"
- 6. Computer Code STP R7, Code I.D. 262165
- 7. Computer Code COROD R7, Code l.D. 262347
- 8. Computer Code FENCDOSE R5, Code I.D. 262358
- 9. Discrepancy Report No.209, I0/07/88
- 10. WBN CCD drawing l-47W866-4 R43.
- 11. WB-DC-40-36.1 RI 0 "The Classification of Heating, Ventilating and Air Conditioning System"
- 12. WBN PER 01-000080-000
- 13. WBNTSR-028 RIO. "Main Control Room Emergency and Normal Air Intake Monitors Required Range, Safety Limits, Response Time and Accuracy
- 14. WBNAPS3-084 RO "Source Terms For 1500 EFPD Bumup"
- 15. Code ofFederal Regulations, Title 10, Chapters 50 and 100
- 16. System Description N3-79-4001 Rl8 "Fuel Handling and Storage System"
- 17. System Description N3-30AB-4002 R4 "Auxiliary Building - Heating, Ventilation and Air Conditioning System"
- 18. Watts Bar Technical Specifications Bases Section 3.9.4
- 19. TI-RPS-197 R22 "Offsite Doses Due to a Regulatory Guide 1.4 Loss of Coolant Accident"
- 20. WBPER930129 RO
- 21. EPM-RA V-081193 R2 "Isolation Damper Leakage Rate for the Aux Bldg Supply and Exhaust Fans"
l' ! Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page: 22
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident
- 22. DCN M-29141-D
- 23. WBPER960798
- 24. WBNAPS3-095 R3 "Offsite and Control Room Dose Due to a FHA with 15 Minutes Unfiltered Release and Various Flow Rates"
- 25. NUREG/CR-5009 "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors" Feb. 1988
- 26. TTQP-1-091 RI 0 "Tritium Technology Program - Unclassified TPBAR Tritium Releases, f ncluding Tritium" - Att. 4
- 27. N3-78-4001 Rl 7 "Spent Fuel Pool Cooling and Cleaning System" 28a. Memorandum from James S. Chardos to Cheryl K. Thornhill, TVATP-00-068, October 23, 2000 "Verification of Design Inputs for Calculations of Breached TPBAR Leaching in the Spent Fuel Pool" Att. 5 28b. Memorandum from Cheryl K. Thornhill to James S. Chardos, TTQP-00-175, September 19, 2000 "Verification of Design Inputs for Calculations of Breached TPBAR Leaching in the Spent Fuel Pool" - Att. 6
- 29. WBNAPS3-098 R2 "Source Tenns for WBN Tritium Production Core"
- 30. TT-535 RS "Max. Expected Airborne Concentration in Primary Containment, Turbine Building, Auxiliary Building, and Instrument Room During Nonnal Operation" 3 la. WBNTSR-023 R7 "Response Time, Range, and Accuracy for the Spent Fuel Pool Radiation Monitors (TSFPRM)" - not used as design input, only for comparative purposes 31b. DCN W-23167-A 3 lc. WBNTSR-020 R8 "Safety Limit For the Spent Fuel Pool Radiation Monitors" - not used as design input, only for comparative purposes
- 32. CCD drawing 1-47W866-IO R35
- 33. Lederer and Shirley "Table ofisotopes" 7th edition
- 34. WBNAPS3-104 R4 " WBN Control Room X/Q"
- 35. TTD-14844 "Calculation of Distance Factors For Power and Test Reactor Sites"
- 36. Regulatory Guide 1.183 RO " Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Plants"
- 37. Regulatory Guide 1.140 R3, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-water-cooled Nuclear Power Plants"
- 38. WBNOSG4-I 69 R2, "Spent Fuel Pool temperature Analysis during 10CFR50 Appendix R Conditions"
- 39. PER 252012
- 40. WBN Drawing 1-47£866-20 R4
- 41. PER 428401
- 42. PER 775553
- 43. PER l 14028
Il, l, Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:23
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Appendix A: RG 1.25 Containment FHA with 12. 7 sec unfiltered release, containment closed to Auxiliary Building This appendix supplies an additional case performed in previous revisions, with the following criteria: RG 1.25 Containment FHA with 12.7 seconds unfiltered release (no purge filters) and the containment is closed to the Auxiliary Building. Control Room Doses freml: Containment FHA with 12. 7 sec containment isolation, containment closed to AB, no Purge Filters Run: TS9Cl3%#* Conventional TPC Once TPC Twice TPC Thrice limit o/o=A,B,CorD; #=I Core Burned Burned Burned Gamma l.037E-02 l.l84E-02 8.930E-03 l.165E-02 5 Beta 8.553E-02 9.591E-02 7.441E-02 9.532E-02 30 Thyroid (ICRP-30) 4.762E+-OO 5.054E+OO 3.943E+OO 5.291E+OO 30 TEDE l.603E-OI 1.913E-Ol l.534E-Ol 1.782E-OI 5 Offsite Doses frem Run: TS9Fl3%# Conventional Core TPCOnce Burned TPC Twice Burned TPC Thrice Burned o/o=A,B,CorD; #=l 2-HrEAB 30-DavLPZ 2-HrEAB 30-DavLPZ 2-HrEAB 30-DayLPZ 2-HrEAB 30-DayLPZ limit Gamma 4.545E-03 l.271E-03 5.148E-03 l.439E-03 3.894E-03 l.089E-03 5.095E-03 l.424E-03 6.25 Beta l.227E-02 3.430E-03 1.376E-02 3.846E-03 l.067E-02 2.981E-03 l.368E-02 3.823E-03 75 Thyroid (ICRP-30) l.615E+OO 4.513E-Ol l.713E+OO 4.790E-Ol l.337E+-00 3.737E-Ol 1.794E+-OO 5.0l4E-Ol 75 TEDE 6.605E-02 l.846E-02 7.309E-02 2.043E-02 5.756E-02 l.609E-02 7.339E-02 2.052E-02 6.25 controJRoom M aTll ms: Conv. Core TPC Once Burned TPCTwice Burned TPC Thrice Burned Mari;~in Percent Man!in Percent Mar2in Percent Man!in Percent Gamma 4.99 99.79 4.99 99.76 4.99 99.82 4.99 99.77 Beta 29.91 99.71 29.90 99.68 29.93 99.75 29.90 99.68 Thyroid (ICRP-30) 25.24 84.13 24.95 83.15 26.06 86.86 24.71 82.36 TEDE 4.84 96.79 4.81 96.17 4.85 96.93 4.82 96.44 Offi. site Dose M ar1nns: Conv. Core TPCOnce Burned TPC Twice Burned TPC Thrice Burned 30-DayLPZ 2-hrEAB Man!in % 30-DayLPZ Man!in % 2-hrEAB Mar1dn % 30-DayLPZ Man!ln .,,. 2-hrEAB Marein % 30-DayLPZ Marldn % 2-hrEAB Marldn */o Marldn o;. Gamma 6.25 99.93 6.25 99.98 6.24 99.92 6.25 99.98 6.25 99.94 6.25 99.98 6.24 99.92 6.25 99.98 Beta 74.99 99.98 75.00 100.00 74.99 99.98 75.00 99.99 74.99 99.99 75.00 100.00 74.99 99.98 75.00 99.99 Thyroid (ICRP-30) 73.39 97.85 74.55 99.40 73.29 97.72 74.52 99.36 73.66 98.22 74.63 99.50 73.21 97.61 74.50 99.33 TEDE 6.18 98.94 6.23 99.70 6.18 98.83 6.23 99.67 6.19 99.08 6.23 99.74 6.18 98.83 6.23 99.67 Discussion and
Conclusion:
This RG 1.25 case with the containment FHA and an unfiltered 12.7 sec release, has all limits less than the IOCFRlOO offsite dose limits and 10CFRSO App.A GDC 19 dose limits.
11 ! I Calculation sheet Document: WBNTSR009 j Rev.: 016 j Plant: WBN I Units 1,2 I Page: 24
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Appendix B TEDE Calculation per Regulatory Guide 1.183 Purpose The purpose of this Appendix is to calculate the TEDE value as outlined in Regulatory Guide 1.183. COROD and FENCDOSE calculate TEDE differently and so a comparison is needed (PER 114028 Action 10 and PER 428401 ). Based on the results in the body of the calculation, the TPC once burned case is bounding for the control room dose and the thrice burned TPC case is bounding for the offsite dose. Therefore these cases will be utilized to compare to the RG l.183 methodology. Introduction R.G. 1.183 specifies calculating the TEDE by summing the Committed Effective Dose Eq11ivalent (CEDE) with the Deep Dose Equivalent (DDE). Position 4.1.2 states that exposure to CEDE factors in Table 2.1 of Federal Guidance Report (FGR) 11 are acceptable to calculate the CEDE. When calculating the CEDE value, a breathing rate of 3.5E-4 m3/sec should be used for the first 8 hours, I .8E-4 m3/sec from 8-24 hrs, and 2.3E-4 m3/sec for times greater than 24 hrs (position 4.1.3). Position 4. 1.4 states that DDE DCFs from Table ITl. l of Federal Guidance report 12 are acceptable to use when calculating the DDE. The following table provides the DCFs from these two tables. T abl e I RGI 183 acceptable DCFs. DDE CEDE 3 Sv/(Bq s m*3) rem/(Ci hr m" ) Sv/Ba rem/Ci KRM 83 1.SOE-18 2.00E-02 O.OOE+OO O.OOE+OO KRM 85 7.48E-15 9.96E+Ol O.OOE+OO O.OOE+OO KR 85 1.19E-16 l.59E+OO O.OOE+OO O.OOE+OO KR 87 4.12E-14 5.49E+02 O.OOE+OO 0.00E+OO KR 88 1.02E-13 l.36E+03 O.OOB+OO O.OOE+OO KR 89 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 13 1 3.89E-16 5.18E+OO O.OOE+OO O.OOE+OO XEM 133 l.37E-15 l.82E+OI O.OOE+OO O.OOE+OO XE 133 J .56E-15 2.08E+OI O.OOE+OO O.OOE+OO XEM 135 2.04E-14 2.72E+02 0.00E+OO 0.00E+OO XE 135 1.19E-14 J.59E+02 O.OOE+OO O.OOE+OO XE 138 5.77E-14 7.69E+02 O.OOE+OO O.OOE+oO r 131 l.82E-14 2.42E+02 8.89E-09 3.29E+04 I 132 1.12E-13 l.49E+03 l.03E-10 3.81E+o2 r 133 2.94E-14 3.92E+02 l.58E-09 5.85E+03 I 134 l.30E-13 l.73E+03 3.55E-l l l.31E+02 r 135 7.98E-14 l.06E+o3 3.32E-10 l.23E+o3 H3 3.31E-19 4.41E-03 l.73E-11 6.40E+Ol The DDE is converted to rem/(Ci hr m"3) by multiplying the values from Table HI.I by 3.7El2 (rem/Ci)/(Sv/Bq)
- 3600 sec/hr.
The CEDE values are converted by multiplying the values from Table 2.1 by 3.7E 12 (rem/Ci)/(Sv/Bq)
! Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:25
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Calculation The following describes how the DDE and CEDE were calculated using methods as recommended in RG 1.183 for offsite and control room dose. Deep Dose Equivalent The DDE is calculated as follows: Concentration (Ci*hr/m3) x DCFooE (rem/Ci*hr*m-3) Committed Effective Dose Equivalent The CEDE is calculated as follows: 3 Concentration (Ci*hr/m3) x breathing rate (m /hr) x DCFcEDE(rem/Ci) 3 Since the FHA lasts only 2 hours the breathing rate will be 3.5E-4 m /sec = 1.26 m3/hr Concentration COROD calculates and prints out the concentration in Ci hr/m3 for each time period. Therefore this can be easily obtained from the output. The following is the concentration as output in COROD for the once burned case (TSR009R15C4B): Table 2 - Concentrations (Ci*hr/m3) from COROD Case B4 lsotpe Concentration in Control Room (Ci hr/m 3 ) Isotope 0-74 sec 74 sec-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days KRM 83 2.32E-23 3.93E-20 5.62E-20 2.15E-21 3.21E-25 O.OOE+OO KRM 85 1.51E-13 2.89E-10 6.53E-10 1.02E-10 5.03E-13 2.02E-23 KR 85 3.33E-08 7.03E-05 2.37E-04 1.17E-04 7.23E-06 2.00E-11 KR 87 3.28E-30 5.04E-27 5.38E-27 7.13E-29 6.76E-34 O.OOE+OO KR 88 3.87E-17 7.03E-14 1.29E-13 1.09E-14 1.21E-17 5.91 E-31 KR 89 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 131 2.53E-08 5.32E-05 1.78E-04 8.64E-05 5.12E-06 1.19E-11 XEM 133 6.14E-08 1.29E-04 4.19E-04 1.87E-04 9.27E-06 1.02E-11 XE 133 3.59E-06 7.56E-03 2.51E-02 1.19E-02 6.68E-04 1.25E-09 XEM 135 2.29E-11 1.57E-08 3.22E-09 1.32E-16 O.OOE+OO O.OOE+OO XE 135 6.86E-09 1.38E-05 3.80E-05 1.04E-05 1.84E-07 2.17E-15 XE 138 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 131 1.37E-08 3.91E-06 5.73E-06 2.02E-06 1.17E-07 2.50E-13 I 132 1.11 E-21 2.65E-19 2.04E-19 7.76E-21 3.45E-24 O.OOE+OO I 133 9.23E-10 2.59E-07 3.53E-07 9.70E-08 3.41E-09 8.55E-16 I 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 135 6.76E-13 1.82E-10 2.08E-10 3.26E-11 3.56E-13 5.17E-22 I* 131 3.43E-11 9.80E-09 1.44E-08 5.07E-09 2.93E-10 6.26E-16 I* 132 2.79E-24 6.64E-22 5.11E-22 1.95E-23 8.66E-27 O.OOE+OO I* 133 2.32E-12 6.50E-10 8.84E-10 2.43E-10 8.55E-12 2.15E-18 I* 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I* 135 1.70E-15 4.55E-13 5.22E-13 8.17E-14 8.92E-16 1.30E-24 H 3 1.20E-06 2.53E-03 8.51E-03 4.21E-03 2.60E-04 7.18E-10
ll Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:26
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident 3 FENCDOSE does not explicitly print out concentration in terms ofCi*br/m . Therefore it must be back calculated. FENCDOSE calculates the TEDE by multiplying the DCFs found in Table 5-1 ofEPA-400-R-92-001 by the concentration. T he TEDE values for each isotope in each time period are printed out. Therefore to obtain the concentration for offsite dose, the TEDE values must be divided by the DCF values. The following are the TEDE values from FENCDOSE TPC thrice burned case (TSR009R l5F4D), the DCFs used in FENCDOSE, and the back calculated concentration. TEDE (rem) DCF (Table 5.1) Concentration (Ci hr/m3) Isotope 0-2 HRS 2-HR EAB rem/(uCi cm*3 hr) 0-2 HRS 2-HR EAB KRM 83 O.OOE+OO O.OOE+OO 1.00E+20 O.OOE+OO O.OOE+OO KRM 85 9.07E-09 3.24E-08 9.30E+01 9.75E-11 3.49E-10 KR 85 7.26E-05 2.60E-04 1.30E+OO 5.59E-05 2.00E-04 KR 87 1.05E-24 3.74E-24 5.10E+02 2.05E-27 7.34E-27 KR 88 3.12E-11 1.12E-10 1.30E+03 2.40E-14 8.58E-14 KR 89 O.OOE+OO O.OOE+OO 1.20E+03 O.OOE+OO O.OOE+OO XEM 131 1.23E-04 4.40E-04 4.90E+OO 2.51E-05 8.98E-05 XEM 133 9.16E-04 3.28E-03 1.70E+01 5.39E-05 1.93E-04 XE 133 6.17E-02 2.21E-01 2.00E+01 3.08E-03 1.10E-02 XEM 135 5.00E-06 1.79E-05 2.50E+02 2.00E-08 7.15E-08 XE 135 8.08E-04 2.89E-03 1.40E+02 5.77E-06 2.06E-05 XE 138 O.OOE+OO O.OOE+OO 7.20E+02 O.OOE+OO O.OOE+OO I 131 6.63E-01 2.37E+OO 5.30E+04 1.25E-05 4.48E-05 I 132 4.92E-15 1.76E-14 4.90E+03 1.00E-18 3.60E-18 I 133 1.19E-02 4.26E-02 1.50E+04 7.93E-07 2.84E-06 I 134 O.OOE+OO O.OOE+OO 3.10E+03 O.OOE+OO O.OOE+OO I 135 131 4.70E-06 1.66E-03 1.68E-05 5.95E-03 8.10E+03 5.30E+04 5.80E-10 3.14E-08 2.08E-09 1.12E-07 I" 132 133 1.24E-17 2.98E-05 4.42E-17 1.07E-04 4.90E+03 1.50E+04 2.52E-21 1.99E-09 9.01 E-21 7.11E-09 I* 134 O.OOE+OO O.OOE+OO 3.10E+03 O.OOE+OO O.OOE+OO I* 135 1.18E-08 4.22E-08 8.10E+03 1.46E-12 5.20E-12 H 3 O.OOE+OO O.OOE+OO 7.70E+01 O.OOE+OO O.OOE+OO
- Note that conversion from µCi to Ci and cm3 to m3, cancel each other out. Therefore the correct units are obtained without multiplying by any conversion factors.
- Note that there is no DCF in Table 5-1 ofEPA-400-R-92-001 for Kr-83m. The value listed in the above Table is suspect, but does not change the results and conclusions of the calculation.
I ,
! Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:27
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Offsite - TPC Thrice Burned - Case D4 Offsite DOE (rem) Offsite CEDE (rem) Isotope 0-2 HRS 2-HR EAB 0-2 HRS 2-HR EAB KRM 83 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KRM 85 9.71E-09 3.47E-08 O.OOE+OO O.OOE+OO KR 85 8.88E-05 3.18E-04 O.OOE+OO O.OOE+OO KR 87 1.13E-24 4.03E-24 O.OOE+OO O.OOE+OO KR 88 3.26E-11 1.17E-10 O.OOE+OO O.OOE+OO KR 89 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 131 1.30E-04 4.65E-04 O.OOE+OO O.OOE+OO XEM 133 9.81E-04 3.51E-03 O.OOE+OO O.OOE+OO XE 133 6.42E-02 2.30E-01 O.OOE+OO O.OOE+OO XEM 135 5.44E-06 1.95E-05 O.OOE+OO O.OOE+OO XE 135 9.17E-04 3.28E-03 O.OOE+OO O.OOE+OO XE 138 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 131 3.03E-03 1.08E-02 5.19E-01 1.86E+OO I 132 1.50E-15 5.36E-15 4.82E-16 1.73E-15 I 133 3.11E-04 1.11E-03 5.85E-03 2.09E-02 I 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 135 6.15E-07 2.20E-06 8.99E-07 3.22E-06 I* 131 7.59E-06 2.72E-05 1.30E-03 4.65E-03 I* 132 3.76E-18 1.34E-17 1.21E-18 4.33E-18 I* 133 7.80E-07 2.79E-06 1.47E-05 5.24E-05 I* 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I* 135 1.54E-09 5.52E-09 2.26E-09 8.0?E-09 H 3 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Total 6.96E-02 2.49E-01 5.26E-01 1.88E+OO 2 hr EAB TEDE = 2.49E-Ol + I.88EOO = 2.13EOO rem 30 day LPZ TEDE = 6.96E-2 + 5.26E-l = 5.96E-Ol rem
l Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page:28
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Control Room - TPC Once Burned - Case B4 Based on position 4.2.6 ofRG l.183, the occupancy factor for personnel in the control is 100% for the first 24 hrs, 60% between 1 and 4 days and 40% between 4 days and 30 days. This is taken into account by multiplying the sum of all isotopes of each time step by the appropriate occupancy factors. Position 4.2.7 allows correcting the DDE DCFs for the difference between a finite cloud geometry and a semi-infinite cloud geometry. The correction factor is given as v0*338/l l 73, where Vis the control room volume in ft3
- The control room volume found in the COROD runs is 257198 ft3
- Therefore the correction factor is 5.75E-2. This is multiplied to the DDE totaled over all time periods with occupancy factors included.
Control Room DDE (rem) Isotope 0-74 sec 74 sec-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days KRM 83 4.65E-25 7.85E-22 1.12E-21 4.29E-23 6.42E-27 O.OOE+OO KRM 85 1.50E-11 2.88E-08 6.51E-08 1.02E-08 5.01E-11 2.01 E-21 KR 85 5.29E-08 1.12E-04 3.77E-04 1.86E-04 1.15E-05 3.18E-11 KR 87 1.80E-27 2.77E-24 2.95E-24 3.91E-26 3.71E-31 O.OOE+OO KR 88 5.27E-14 9.56E-11 1.76E-10 1.48E-11 1.64E-14 8.03E-28 KR 89 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 131 1.31 E-07 2.76E-04 9.23E-04 4.47E-04 2.65E-05 6.15E-11 XEM 133 1.12E-06 2.34E-03 7.62E-03 3.39E-03 1.69E-04 1.85E-10 XE 133 7.47E-05 1.57E-01 5.22E-01 2.47E-01 1.39E-02 2.59E-08 XEM 135 6.23E-09 4.27E-06 8.77E-07 3.59E-14 O.OOE+OO O.OOE+OO XE 135 1.09E-06 2.20E-03 6.04E-03 1.66E-03 2.92E-05 3.46E-13 XE 138 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 131 3.31E-06 9.46E-04 1.39E-03 4.89E-04 2.83E-05 6.05E-11 I 132 1.66E-18 3.95E-16 3.04E-16 1.16E-17 5.14E-21 O.OOE+OO I 133 3.62E-07 1.02E-04 1.38E-04 3.80E-05 1.34E-06 3.35E-13 I 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 135 7.17E-10 1.92E-07 2.21E-07 3.45E-08 3.77E-10 5.48E-19 I* 131 8.29E-09 2.37E-06 3.48E-06 1.23E-06 7.10E-08 1.52E-13 I* 132 4.15E-21 9.89E-19 7.61E-19 2.90E-20 1.29E-23 O.OOE+OO I* 133 9.07E-10 2.55E-07 3.47E-07 9.53E-08 3.35E-09 8.41E-16 I* 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I* 135 1.80E-12 4.82E-10 5.54E-10 8.66E-11 9.45E-13 1.37E-21 H 3 5.27E-09 1.11 E-05 3.75E-05 1.86E-05 1.15E-06 3.17E-12 total 8.08E-05 1.63E-01 5.38E-01 2.53E-01 1.42E-02 2.63E-08 w/Occupancy 8.08E-05 1.63E-01 5.38E-01 2.53E-01 8.50E-03 1.05E-08 DDE Total 9.63E-01 Corrected 5.54E-02
I l Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:29
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Control Room CEDE (rem) Isotope 0-74 sec 74 sec-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days KRM 83 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KRM 85 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KR 85 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KR 87 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KR 88 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO KR 89 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 131 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 133 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XE 133 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XEM 135 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XE 135 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO XE 138 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 131 5.67E-04 1.62E-01 2.38E-01 8.38E-02 4.85E-03 1.04E-08 I 132 5.34E-19 1.27E-16 9.78E-17 3.72E-18 1.66E-21 O.OOE+OO I 133 6.81E-06 1.91 E-03 2.60E-03 7.15E-04 2.51E-05 6.31 E-12 I 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I 135 1.05E-09 2.81E-07 3.23E-07 5.05E-08 5.51E-10 8.02E-19 I* 131 1.42E-06 4.06E-04 5.96E-04 2.10E-04 1.22E-05 2.60E-11 I* 132 1.34E-21 3.19E-19 2.45E-19 9.34E-21 4.16E-24 O.OOE+OO I* 133 1.71E-08 4.79E-06 6.52E-06 1.79E-06 6.30E-08 1.58E-14 I* 134 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I* 135 2.63E-12 7.05E-10 8.09E-10 1.27E-10 1.38E-12 2.01E-21 H 3 9.64E-05 2.04E-01 6.87E-01 3.39E-01 2.10E-02 5.79E-08 total 6.71E-04 3.68E-01 9.27E-01 4.24E-01 2.58E-02 6.83E-08 w/Occupancy 6.71E-04 3.68E-01 9.27E-01 4.24E-01 1.55E-02 2.73E-08 CEDE Total 1.736E+OO TEDE = 5.54£-02 + 1.736 = 1.79 rem. Conclusions Based on the results of this appendix it can be seen that the method COROD and FENCDOSE utilizes to calculate TEDE is conservative as it produces higher results. RG 1.183 Dose FENCDOSE/COROD Location Difference (rem) (rem) 2 hr EAB 2.13E+OO 2.65E+OO 24% 30 day LPZ 5.96E-01 7.41E-01 24% Control Room 1.79E+OO 2.39E+OO 33%
l l Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page:30
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Attachment 1* - TRITIUM TECHNOLOGY PROGRAM UNCLASSIFIED TPBAR RELEASES, INCLUDING TRITIUM TTQP-1-091 Revision 10 Effective Date:
,1 I a Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page:31
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident I TrQP-1--091 TRITIUM TECHNOLOGY PROGRAM UNCLASSIFIED TPBAR RELEASES, INCLUDING TRITIUM Revision 10 Prepared By:
~~;;~
D.D.L :rl:f" Reviewed By: 1/u.-/oe, tR.eviewer Date Conearreace:
~+*i<'i>
~:~~ :::! .. :::;;;; "J.h 7fat' T.M. Brewer, Quality Eqbleer Date
~~ 3/2~/C14:>.
.B.D. Reid, Desilll Tuk Manager I Ba&
App:nmll: Cl?. . *~~ C.K. Thonahill, TTP Project Mauger 3/z."f /off Date
, 1 l l Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:32
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Tritium Technology Program Unclassified TPBAR Releases, Including Tritium TIQP-1-091 Revision 10 Page 1 of8
1.0 INTRODUCTION
This document provides a complete listing of all unclassified tritium release values that should be assumed for unclassified analysis. Much of the information is brought forth from the related documents listed in Section 4.0 to provide a single-source listing of unclassified release values. Some information, however, is new or updated based on current design analysis and available experimental data. This document provides unclassified information for a larger number of release scenarios than previously analyzed. This information is summarized in Tables 1, 2, and 3. In addition, a section is included to address lithium and aluminum release in the event of a 24-TPBAR breach in the spent fuel pool. 2.0
SUMMARY
OF UNCLASSIFIED RELEASES, INCLUDING TRITIUM All tritium-producing burnable absorber rod (TPBAR) analysis assumes a maximum of 1.2 grams of tritium per TPBAR will be generated during an 18-month operating cycle. 2.1 Intact TPBAR In-reactor Tritium Permeation The in-reactor tritium permeation rate deduced from RCS tritium activity for the group of240 TPBARs in Watts Bar Nuclear Cycle, 6 averaged over a year extending to end-of-cycle, was 2.4 +/- 1.8 Ci!IPBAR/year (95% confidence interval) (Lanning and Pagh, 2005). The 95% upper bound of 2.4 +1.8 = 4.2 Ci!IPBAR/year is recommended as the basis for assessing the tritium release from intact TPBARs. 2.2 In-reactor Tritium Release from a Failed TPBAR The first scenario involves a TPBAR that may have a fabrication defect or may be damaged prior to insertion into the reactor for irradiation. In this case, 100 percent of the tritium generated in the TPBAR is assumed to be released to the reactor coolant as it is generated. 2.3 TPBAR Releases from Spent Fuel Pool Accidents 2.3 .1 Spent Fuel Pool Tritium Concentration Limit It has been determined that following the simultaneous breach of24 TPBARs, the Tennessee Valley Authority take-action limit for tritium concentration in the spent fuel pool water will not be exceeded. The concentration limit is 60 microcuries per milliliter. The best estimate of total tritium release in this event is less than 25% of the TPBAR inventory. r. ~s llsst ssti111ats tRli11m Fe/ease is less #laR ai~ ef#ls +PSAR iR'J8R#efy-. The release will not be instantaneous, but will occur at a steady rate over a time period substantially greater than 8 hours. The rate will thus be less than 3% (of initial inventory) per hour. l--091RovlOdnft.docRl9W~QP I 991.*ee
l, l I Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:33
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Tritium Technology Program Unclassified TPBAR Releases, Including Tritium TTQP-1-091 Revision 10 Pagel ors 2.3.2 Instantaneous Tritium Release per TPBAR In particular, the instantaneous release of tritium from breached TPBARs in the spent fuel pool (as gas within the released gas from the TPBARs) will not exceed 0.001 CiffPBAR 2.3.3 Lithium and Aluminum Release In the event of a 24-TPBAR breach in the spent fuel, the following concentration limits for lithium and aluminum will not be exceeded:
- 400 ppb lithium
- SO ppb aluminum.
2.4 Tritium Releases from TPBARs within Storage Canisters (<200"F) The upper-bounding tritium partial pressure within storage canisters containing lead test assembly (LT A) TPBARs and sections is not expected to exceed 20 torr under nominal storage conditions (-86"F). The quoted bounding pressure for maximum temperatures (<200°F) is estimated by increasing this figure by the ratio of Kelvin temperatures, to 25 torr. Tritium release from extracted TPBARs in storage will not exceed l % of the declared post-extraction residual tritium (Clemmer et al. 1984; and Johnson et al. 1976). In both cases, the form of the released tritium will be tritiated water vapor or condensate (HTO). 2.S TPBAR Transportation Cask Event Releases 2.S. l Intact TPBARs 2.S.1.1 For TPBAR temperatures ranging from ambient to less than 200"F, and for casks containing 1,200 or less TPBARs, the tritium release from the entire cask loading would be less than 0.19 mCi per hour, based on extrapolation from an in-reactor upper bound observed permeation rate of 4.2 CiffPBAR/year. The tritium would be released from the TPBARs in the form of molecular tritium gas (i.e., T2 or HT). 1--091Rovl0draft.doolU9\' 1 ~QP I 1191 du
! Calculation sheet Document: WBNTSR009 IRev.: 016 I Plant: WBN I Units 1,2 I Page:34
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident Tritium Technology Program Unclassified TPBAR Releases, Including Tritium TTQP-1-091 Revision 10 Pagel of8 2.5.1.2 For TPBAR temperatures ranging from 200"F to 650"F, the average tritium release would be less than 0.48 mCi per TPBAR per hour based on the upper-bound in-reactor release rate of 4.2 Ci!I'PBAR/year. The tritium would be released from the TPBARs in the form of tritium gas. 2.5.1.3 For TPBAR temperatures ranging from 650"F up to 1050"F (565°C), the tritium release should be considered to be an instantaneous release ofless than 0.5 Ci per TPBAR per hour. Again, the tritium would be released from the TPBARs in the form of tritium gas. The potential for TPBAR rupture was assessed at 1050"F because this is one of the temperature break-points in the Modal Study matrix cited earlier (Laity 1998). It was determined that the TPBARs are unlikely to rupture at temperatures less than 1050"F, but may rupture at higher temperatures. 2.5.1.4 Helium release from intact TPBARs is negligible. 2.5.2 Event-failed TPBARs 2.5.2.1 For TPBAR temperatures ranging from ambient to 200°F, the tritium release from a TPBAR whose cladding fails mechanically (e.g., due to impact forces) after cask loading should be considered to be less than 0.1 Ci per TPBAR per hour, not to exceed 1% of the tritium inventory in the lithium aluminate pellets. The release should be considered to be in the form of tritiated water and a very small fraction of methane. 2.5.2.2 For TPBAR temperatures ranging from 200"F to 650°F, the tritium release from a TPBAR whose cladding fails mechanically (e.g., due to impact forces) after cask loading should be considered to be less than 55 curies total due to desorption release. The release should be considered to be in the form oftritiated water and a very small fraction of methane. 2.5.2.3 For TPBAR temperatures ranging from 6500F to 1050"F, the tritium release should be considered to be up to 100% of the TPBAR tritium inventory, in the form oftritiated water and methane. l-091bvl0draft.dool\191.1~ I 99Uoe
l Calculation sheet Document: WBNTSR009 I Rev.: 016 I Plant: WBN I Units 1,2 I Page:35
Subject:
Control Room Operator and Offsite Doses From a Fuel Handling Accident October 23, 2000 TVATP-00-068 Ma. Cheryl K. Thornhill
'l"l'QP Project Manager Paoif ic Northwest National Laboratory E'. o. Box 999 Richland, WA 99352 SUSJIC'l'I VElU:FICATION or DESIGN INPUTS roa CALCULATIONS or Bl\ZACHID TPBAJ. LBACKIN~ lN tKE SPENT rtl'EL POOL REF: C. K. Thornhill to J. S. Chardoa latter dated S*p~'lllll:>er 19, 2000, smne subject Dear Cheryl; TVA has reviewed the desiqn assumptions in the referenced letter and finds them tD be correct except for assumption nwnber 2.
The value ror tritium should be 60 uc/ml not 60 me/ml. If you have any questions, plaaaa call,
.7SC/LDR co: F. A. JCoontz, EQB lA-WBN D. M. Larever, OPS 2B-SON
.7. A. Flaniqan, BR 3r-c EDMS WT 38*1C
l l
! Calculation sheet Document: WBNTSR009 I Rev... 016 l Plant: WBN I Units. 1,2 . l Page:36
Subject:
Control Room Operator an d Offsite Doses From a Fuel Handling Accident 1>0'c:I 'l:llOl Padfic Nol1flwest National laboratory o-....i by . . . . . . . . LJ.s. ~°""av SiplQaber lP, 2~ llfr.1-. s. a.nsa., PrqjM~ ADM-lV*WSN TDlellcc Vllley ~ bco: DD Lumlng
', Wm1BarNuc11.rp11m.
BDRoid S)lllna Ctly, DI 37381 .Ge Soraili:u. Dau- M:. Clwdoe: llcai:dl Tl.llS.llPllllLJl.00-175 VBRmCA.noN JN OP DISION lNPtJTs FOJ. CALctlLA.110Ns OJI BlW.CHBD 1'PBAR. LEACHING 1Jm Sl'IHT llJ!L'POOL
'l1aia ... --ID dociutlim1hl plut 'Plllilo dmlp ~ piapond lbr-in c:abaladnc the lritlmn~ mm llraicbcd 'IP.BA.Rs ill t1i1.,. fall JICIOl n. ~ o1111e bdiina calculatiom Is to dc4ne bow quiclrJy ftlllldill ld:lam, ftay, Ill~ lO llddt"91bt ~ cOmcbed
'lP&U.t. T - Vlllsy 4mllar.b:y(TVA) ii Aqllltfld ID coiaazmt!ie ~ofilll fDlloMDa
-........,.._ .fbt- ~ *'bvdt Wims :aar llld SeqaoyalaNucJllrPllms.
The PIQpClud daila . . .Olll Dlade:
- 1. *~~of24'I'PB.Ak '
- 2. Tlli: adiaQJimlt btdtiam CIOlll:llllZatf hi tbe spai filll Pool"* .ii ~
J. 'Dlupmdlllpooh1111tcr'VO!amll1J72,000pllau,
- 4. Tiit.,.. w poo1-. diiplll ls 39 -
1" ~ dlllp llllllbpdalah1bcapa&e1 .......... aler.-d1ID be CCIDler'fllive, A.~11i.,.. poo1"°1ame Wllllld btlll&ao-'7afl. .
"* ll71111-...,, quoadans, ,... ~llniae Raid lll.50N'12-4J3J, lbjk cc; 0. W. TaYlor, DOI-HQ. DP~1
- 1. IC. Tanmr, DOB-RI, 902 &ttall. s .... 1*....i
- P.O. .Boo: 999
- 21.w....r, 'IV.\ 99.a&a}}