TMI-09-082, Response to Request for Additional Information Related to License Amendment Request No. 326 to Adopt TSTF-490-A, Revision 0, Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical....
| ML091870996 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 07/02/2009 |
| From: | Cowan P Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME0100, TMI-09-082 | |
| Download: ML091870996 (24) | |
Text
10 CFR 50.90 July 2,2009 TMI-09-082 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Three Mile Island Unit 1 Response to Request for Additional Information Related to License Amendment Request No. 326 to Adopt TSTF-490-A, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process"
References:
(1) AmerGen Letter 5928-08-20201, Three Mile Island, Unit 1, "License Amendment Request No. 326 to Adopt TSTF-490-A, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" dated November 6,2008.
(2) TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process, dated March 15, 2007.
(3) Federal Register Notice of Availability published on March 15, 2007 (72FR12217)
(4) Letter from P. Bamford (U. S. Nuclear Regulatory Commission) to C. Pardee (Exelon Generation Company, LLC), "Request for Additional Information Regarding License Amendment Request to Adopt Technical Specification Task Force Traveler TSTF-490-A" (TAC No. ME0100) dated June 24, 2009.
By letter dated November 6, 2008 (Reference 1), AmerGen Energy Company, LLC, now Exelon Generation Company, LLC (Exelon), requested an amendment to the Technical Specifications (TS) for Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1) consistent with U.S. Nuclear Regulatory Commission (USNRC) approved Industry Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-490-A, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process" (References 2 and 3).
U.S. Nuclear Regulatory Commission July 2,2009 Page 2 The USNRC staff has been reviewing the Reference 1 submittal and has determined that additional information is needed to complete the review. The USNRC staff formally requested additional information on June 24,2009 (Reference 4).
Exelon's responses to the USNRC questions are provided in Attachment 1 to this letter.
Revised mark-ups of the TMI Unit 1 TS Pages and TS Bases Pages are provided in Attachment 2.
Exelon has determined that the information provided in response to this request for additional information does not impact the conclusions of the No Significant Hazards Consideration as stated in Reference 1.
There are no regulatory commitments contained in this submittal.
A copy of this letter and its attachments are being provided to the designated State Official and the chief executives of the township and county in which the facility is located.
Should you have any questions concerning this letter, please contact Wendy E. Croft at (610) 765-5726.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd day of July, 2009.
{
Pamela B. Cbwan Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC . Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326
- 2. Revised Mark-ups of Technical Specification Pages and TS Bases Pages cc: Regional Administrator, USNRC Region I Project Manager, NRR, USNRC - Three Mile Island, Unit 1 Senior Resident Inspector, USNRC - Three Mile Island Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection Chairman, Board of County Commissioners of Dauphin County, PA Chairman, Board of Supervisors of Londonderry Township, Dauphin County, PA R. R. Janati, Commonwealth of Pennsylvania
ATTACHMENT 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326
Attachment 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326 NRC Question 1 The proposed Inserts 1 and 2 contained in the letter dated November 6,2008, provide definitions for DEI and DEX, respectively. These definitions indicate that DEI and DEX may be determined using several references for dose conversion factors (DCFs). However, the purpose of the limiting condition for operation (LCO) for DEI and DEX is to satisfy Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, criterion 2, which establishes an operating restriction that is an initial condition of a design-basis accident (DBA). When surveillance of the RCS radionuclides is performed, each acceptable set of DCFs will yield a different DEI and DEX. As approved by the NRC staff, the intent of Technical Specification Task Force Traveler (TSTF)-490 was to allow the licensee to select, from the acceptable list, one DCF reference for the calculation of DEI, and one DCF reference for the calculation of DEX.
Therefore, consistent with 10 CFR 50.36 and TSTF-490, the licensee should specify one DCF reference for each definition, which will be consistent with the specified LCO and DBA analysis, or justify why a list of several DCFs is consistent with the specified LCO and DBA analysis.
Therefore, please justify how the use of multiple DCFs maintains consistency with the specified LCO values and DBA analyses or provide revised definitions for DEI and DEX that specify one DCF reference for each definition.
TMI Unit 1 Response Exelon Generation Company, LLC (Exelon) has provided, in Attachment 2, revised mark-ups of Technical Specification (TS) pages and TS BASES pages to clarify the definitions for DOSE EQUIVALENT lodine-131 (DEI) and DOSE EQUIVALENT Xenon-133 (DEX) to specify one dose conversion factor (DCF) reference for each definition that is consistent with the applicable limiting condition for operation (LCO) and design basis accident (DBA) analysis.
Note: All of the TS pages and TS BASES pages associated with the original submittal are being re-submitted. All changes to the Inserts from the original submittal will be shown with a revision bar in the right margin.
NRC Question 2 Consistent with the safety evaluation for TSTF-490, please confirm that the site-specific limits for both DEI and DEX, and the DCFs used for the determination of DEI and DEX surveillances, are consistent with the current design bases radiological dose consequence analyses (for example, steam generator tube rupture and main steam line break). For DEX, please provide the information necessary (dose conversion factors and RCS radioisotopic concentrations) for the NRC staff to verify the proposed value in the LCO.
Page 1 of 5
Attachment 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326 TMI Unit 1 Response The current design bases radiological dose consequence analyses for the Loss of Coolant Accident and the Fuel Handling Accident are performed using Alternative Source Term with acceptance criteria in the form of total effective dose equivalent per 10 CFR 50.67. These analyses utilize dose conversion factors from EPA Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" and EPA Federal Guidance Report No.12, "External Exposure to Radionuclides in Air, Water, and Soil."
The radiological dose consequence analyses for the main steam line break and steam generator tube rupture accidents do not postulate fuel damage. These accidents are analyzed using the maximum reactor coolant system activity allowed by TS. This activity is presented in the form of DEI and noble gas activity. Since these accidents are the bounding design basis accidents that use DEI, the dose conversion factors associated with the determination of DEI used in the radiological dose consequence analyses are consistent with the site-specific limits for DEI and for the determination of DEI surveillances. These DCFs are from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." The DCFs for the determination of noble gas dose or DEX are from EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil" as described below.
DEX shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DEX shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil."
The effective dose conversion factors from Table 111.1, Dose Coefficients for Air Submersion, of EPA Federal Guidance Report No.12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993 are as follows:
Nuclide Dose Coefficient hI (Sv per So s m-3)
Kr-85m 7.48E-15 Kr-85 1.19E-16 Kr-87 4.12E-14 Kr-88 1.02E-13 Xe-131m 3.89E-16 Xe-133m 1.37E-15 Xe-133 1.56E-15 Xe-135m 2.04E-14 Xe-135 1.19E-14 Xe-138 5.77E-14 Page 2 of 5
Attachment 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326 The dose equivalence factors (DEF) are calculated by normalizing the effective dose conversion factors from Table 111.1 for each nuclide to Xe-133. The DEF terms are unitless. The significant figures for the calculated DEF terms are kept consistent with the significant figures from Table 111.1.
DEF (nuclide i) = hTnuclide i/h/e.133 DEF (Kr-85m) = 7.48E-15 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 4.79 DEF (Kr-85) = 1.19E-16 Sv per Bq s m*3/1.56E-15 Sv per Bq s m* 3 = 0.0763 DEF (Kr-87) = 4.12E-14 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 26.4 DEF (Kr-88) = 1.02E-13 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 65.4 DEF (Xe-131 m) = 3.89E-16 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 0.249 DEF (Xe-133m) = 1.37E-15 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 0.878 DEF (Xe-133) = 1.56E-15 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 1.00 DEF (Xe-135m) = 2.04E-14 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 13.1 DEF (Xe-135) = 1.19E-14 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 7.63 DEF (Xe-138) = 5.77E-14 Sv per Bq s m*3/1.56E-15 Sv per Bq s m*3 = 37.0 In summary:
Radioisotope Equivalence Factor Radioisotope Equivalence Factor Kr-85m 4.79 Xe-133m 0.878 Kr-85 0.0763 Xe-133 1.00 Kr-87 26.4 Xe-135m 13.1 Kr-88 65.4 Xe-135 7.63 Xe-131 m 0.249 Xe-138 37.0 For TMI Unit 1 the Primary Coolant Noble Gas Activity Based on 1% Fuel Defects is found in TMI-1 UFSAR Table 14.2-4 and are listed below.
Nuclide Concentration (uCi/gm)
Kr-85m 2.43 Kr-85 9.75 Kr-87 1.28 Kr-88 3.95 Xe-131m 2.68 Xe-133m 4.22 Xe-133 392.0 Xe-135m 0.485 Xe-135 8.37 Xe-138 0.692 Page 3 of 5
Attachment 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326 Using the dose equivalent factors from above, the DEX limit is calculated in Exelon Procedure CY-AA-130-3020, "Dose Equivalent Xenon" using the following equation.
DEX = L Nuclidei concentration @ 1% fuel defects (IJCi/gm) x Nuclidei DEF Dose Ec uivalent Xenon Calculation Radioisotope Activitv or MDA, uCi/Qm Equivalence Factor Dose Equivalent Xenon, uCi/Qm Kr-85m 2.43 4.79 11.6 Kr-85 9.75 0.0763 0.744 Kr-87 1.28 26.4 33.8 Kr-88 3.95 65.4 258 Xe-131m 2.68 0.249 0.667 Xe-133m 4.22 0.878 3.71 Xe-133 392.0 1.00 392 Xe-135m 0.485 13.1 6.35 Xe-135 8.37 7.63 63.9 Xe-138 0.692 37.0 25.6 Total Dose Equivalent Xenon-133 797 uCi/gm The DEX limit for TMI Unit 1 will be less than or equal to 7971JCi/gm.
NRC Question 3 The proposed TS Table 4.1-3, item 1(a) requires a verification that DEX is less than or equal to 797 microcuries/gram. The TS bases submitted with the application letter dated November 6, 2008 (Insert 6), state that "Due to the inherent difficulty in detecting [Krypton] Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as
[Fluorine] F-18 and 1-134, that it is acceptable to include the minimum detectable activity of Kr-85 in the [surveillance requirement] SR Table 4.1-3 calculation."
If no or little "masking" occurs, the use of the minimum detectible activity of Kr-85 would underestimate the amount of Kr-85 present in the reactor coolant sample. Therefore, please justify why use of the minimum detectable Kr-85 is an acceptable and conservative approach for performing this surveillance.
TMI Unit 1 Response The intent of the TS BASES description is to describe that there may be difficultly in detecting Kr-85 due to radioisotope masking when a radioisotope is present with similar decay energies.
In this circumstance, if Kr-85 is not detected, then the minimum detectable Kr-85 activity value will be utilized in the surveillance calculation. If a Kr-85 value is detected, then the detected value will be utilized in the surveillance calculation.
This is an acceptable and conservative approach for performing this surveillance because a Kr-85 value is assumed despite it not being detected.
Page 4 of 5
Attachment 1 Three Mile Island Unit 1 Response to Request for Additional Information Related to Technical Specification Change Request No. 326 In addition, it is important to note that this wording is identical to the NRC-approved TSTF-490, Revision 0, TS BASES wording as published in the Federal Register Notice of Availability on March 15,2007 (72FR12217). See TSTF-490, Revision 0, page "BWOG STS B 3.4.16-4."
NRC Question 4 Inserts 4 and 5, contained in the letter dated November 6,2008, refer to REFUELING as a mode in capital letters, indicating that REFUELING is defined in the TS Definition section.
However, the current TMI-1 TS Definition section contains references to REFUELING SHUTDOWN, REFUELING OPERATION, and REFUELING INTERVAL, but not REFUELING.
Please clarify the intended mode of applicability relating to inserts 4 and 5.
TMI Unit 1 Response Exelon has provided, in Attachment 2, revised mark-ups of TS pages and TS BASES pages that clarify the intended mode of applicability (REFUELING SHUTDOWN) relating to inserts 4, 5, and 6.
Note that REFUELING OPERATION and REFUELING INTERVAL, as defined in the TMI Unit 1 TS, are not modes of operation but are a specific activity descriptor and a time period definition, respectively.
Page 5 of 5
ATTACHMENT 2 Three Mile Island Unit 1 Revised Mark-Ups of Technical Specification Pages and TS BASES Pages
LIST OF FIGURES FIGURE TITLE 2.1-1 Core Protection Safety Limit TMI-I 2-4a 2.1-2 DELETED 2.1-3 Core Protection Safety Bases TMI-l 2-4c 2.3-1 TMI-l Protection System Maximum Allowable Setpoints 2-11 2.3-2 DELETED 3.1-1 Reactor Coolant System Heatup/Cooldown Limitations 3-5a (Applicable thru 29 EFPY) 3.1-2 Reactor Coolant Inservice Leak and Hydrostatic Test 3-5b (Applicable thru 29 EFPY) I
=:;=~~
3.l-2a ~
3.1-3 DELETED 3.3-1 Makeup Tank Pressure vs Level Limits 3-24a 3.5-2A DELETED thru 3.5-2M 3.5-1 Incore Instrumentation Specification Axial Imbalance Indication 3-39a 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication 3-39b 3.5-3 Incore Instrumentation Specification 3-39c 3.11-1 Transfer Path to and from Cask Loading Pit 3-56b 4.17-1 Snubber Functional Test - Sample Plan 2 4-67 5-1 Extended Plot Plan TMI N/A 5-2 Site Topography 5 Mile Radius N/A 5-3 Gaseous Effluent Release Points and Liquid Effluent Outfall Locations N/A 5-4 Minimum Burnup Requirements for Fuel in Region II ofthe Pool A 5-7a Storage Racks 5-5 Minimum Burnup Requirements for Fuel in the Pool "B" Storage Racks 5-7b vu Amendment Nos. 11, 17,29. 39, 45, 50, 59. 71, 106, 109, 120, 126, 134, 142, 150, 1M, 167, 168, 184,211,22~ ~
1.9 DELETED 1.10 DELETED
- 1. 11 DELETED 1.12 DOSE EQUIVALENT 1-131
~
! 'f'fIe-tlOSL-~UIVALENT 1-131 shall be that concentration of 1-131 - _----
- !~ert.1. (microcurie~amt-*.lch alone would produce the sa. thYOlid-~he quantity and isotopic 1ilfnUf'e-~3I. 1-132 . . .~r:l34 and 1-135 actually present. The thyroid dose 10n factors used for this calculation shall be th~ts in Table Ul-14844, *Calculation of Dhtance FI~-POWer and Test Reactor Sites-. [or-tn--Table E-7 of NRC
~+It~Uide 1.109, Revision I, October 1 9 7 7 . ] - - - - - - - -
1.13 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
- 1. 14 DELETED 1.15 OfFSITE DOSE CALCULATION MANUAL (ODeN)
The OFFSITE DOSE CALCULATION MANUAL (ODeM) shall contain the methodology and parlllltters used in the calculation of offsite doses resulting f~
radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid effluent MOnitoring Alanl/Tri, Setpoints, and in the conduct of the Radiological Environ..ntal Monitoring Progr... The OOCM shall also contain .
(1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Progr.s required by Section 6.8.4 and (2) descriptions of the info".ation that should be included in the Annual RadiologicaJEnviroftllental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.3 and 6.9.4.
1.16 PROCESS CONTROL PROGRAM (PCP)
The PROCESS CONTROL PROGRAM (PCP) shall contain the current fOrlalas, sa.,11ng, analyses, test, and dete,..inations to be ..de to ensure that proc~sin9 and packaging of solid radioactive wastes based on delOnstrated proclSsing of actual or si.lated wet solid wastes will be accOIIPHshed in such a way as to assure cDIP]iance with 10 eFR Parts 20, 61, and 11~ State regulations, burial ground requir...nts, and other requir. .nts governing the disposal of solid radioactive waste.
- 1. 17 GASEOUS RADWASTE TREATMENT The GASEOUS RADWASTE TREATMENT SYSTEM is the syst.. designed and installed to reduce radioactive gaseous effluent by collecting priMary coolant syst.. off gases from the primary syst** and prOViding for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
1-6 Amendment'No. lZ, J~l, 11J.~
Insert 1:
DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same thyroid dose when inhaled as the combined activities of iodine isotopes 1-131,1-132,1-133,1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."
1.24 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is a TMI-l specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be detennined for each reload cycle in accordance with SpeciflCalion 6.9.5. Plant operation within these operating limits is addressed in individual specifications.
1.25 FREQUENCY NOTATION The FREQUENCY NOTATION specified for die perfonnance of Surveillance Requirements shall correspond co die intervals dermed in Table 1.2. All Surveillance Requirements shall be perfonned within the specified time interval with a maximum allowable extension not co exceed 25 II of the surveillance interval. The 2S ~ extension applies to an frequency intervals with the exception of "F. No extension is allowed for intervals designated "F."
II TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY S Sbiftly (once per 12 houn)
D . Daily (once per 24 houn)
W Weekly (once per 7 days)
M Monthly (once per 31 days)
Q Quanedy (once per 92 days)
. S/A Semi-Annually (once per 184 days)
R Refueling Interval (once per 24 months)
PSIU Prior to each reactor startup, if not done during the pnMous 7 days PS/A W'1thiD six (6) moDtbs prior to ada reactor startup P Completed prior to each release N/A (NA) Not applicable Ii Once per 18 months F Not to exceed 24 months 1-8 Amendment No. 12, 1J1, US, 111, r7J ,~
Insert 2:
1.26 DOSE EQUIVALENT Xe-133 Dose Equivalent Xe-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT Xe-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12.
3.1.4 &£ACTOR COOLANT SY~TIYlTY 3.1.4.1 LIMITING CONDITION FOR OPERATION
~
The s,eElifie aeti'lity Mthe primary eeelant shall be limitecj to. .Jl.-
... b,ss than 9r "tual to Q.3S mieroeu"e/gmm DOSE EQUIVALENT 1-131, Ilftd .JL-
- b. Less than or equal to 190ffi microeurieslgram' .J2--
3.1.4.2 . N'PLll::"'BIUTY. at all blJ1e..xcept refueling JZ..-'~
3.1.4.3 ACTION:
MOMS. Pow.. OpelM"" Stan*Up, Hot Stand~ ~
- a. With the specific activity oftbe-pt imar, coolant greater than 9.35 microcurie/gram DOSE £...1 EQUIVALENT I- 131 fer more dlen 48 hottl'S" dttrinl ene eonttnttotts time inten'al or -e..
O~Q. .cJiRg .elimit line sho'NfI on Figttre 3.1-2a, be in at least nOT SHUTDOWN widlin, e..
(; hOUFS. PO'>>er 9perati9A may eOAtiAu, whIR DOSE EQUIVALENT I I J I is below --e.
9JS mierocuriefgl I:J"~ e..
- b. ')lith die sl"eifie eeti...ity of the p"mary soolant greater than I QQIJ; mieFOSU"estgFam ee 12..
ill at least HOT SHUTDOWN within (; hours. Power operation may eontintle when ,e...
primary seelant 86th/ity is less than looiB miergeurieslgram. e..
MODES: .etd all times 1*6ept refueling. e.
'"C. Wid. die specific acti"it)l of the primaryeoolant grearerthan O.3S mieroellfiel-grem DOSE,p .
EQUIVALENT la J3 J or greater than 19&1£ microc:uriesI-gJ am perfcrm the sampling and .t2-ullysis NftuiNments of Taele 4.1 3 until the speElifie 86tivity of the primary eeelen. is I?-
NstoNd to withift its limits. .e-
- mi~"3r.ily of tile primary _
e in
....... 1MI1lle .....I,;"g a """' .........
. . . houftdary will be well within the Pitt 100 Ii.it fiell Y' 8' steam generator ttlbe [Upton: accident or
...e
-e I ste&ft'l line break Heident with pestulated Heicjen. indtteed steam generato, tube leakage in eonjtfnetien 0 with 8ft assumecJ SleecJ)' state primary te seeondary steam generater leakage rate of 1.0 GPM. The ...allle?
for the limits Oft specific acti"t'ity represent limits based on a parametrie eYaluation hy the NRC ohYPiea~
tite ~oeations. These "Slues I:J e conservative, in dlat the specific site pararneten of TMJ-I, such as site boundar}, location and meteorological conditions, were not considered in this e"f'IlJtlation. .e...
- is shall be the avela~ (weighted ~n proponlon to the cotlcenbatioll oreach radionueli~ in th~
reactor coolant at the time ofsamphng) of the sum of the average beta and gamma energies per*' lL disifttegra.ioft (in MeV) for isotopes, other than iodines, with halfliyes greater-than IS minutes, making up at least 95'ft ofthe total non-iodine acti~ ity in the eeelent.
- The time period begifts ~m me time die S8R1ple is taken. _.e-3-8 Amendment No. lJ)ft, .J..Io1 , ~
Insert 3:
RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT Xe-133 specific activity shall be limited to:
- a. Less than or equal to 0.35 microcuries/gram DOSE EQUIVALENT 1-131, and
- b. Less than or equal to 797 microcuries/gram DOSE EQUIVALENT Xe-133.
Insert 4:
APPLICABILITY: At all times except REFUELING SHUTDOWN and COLD SHUTDOWN.
Insert 5:
MODES: At all times except REFUELING SHUTDOWN and COLD SHUTDOWN a.1 With DOSE EQUIVALENT 1-131 not within limit, perform the sampling and analysis requirements of Table 4.1.3 until the RCS DOSE EQUIVALENT 1-131 is restored to within limit, AND a.2 Verify that DOSE EQUIVALENT 1-131 is less than or equal to 60 microcuries/gram, AND a.3 Restore DOSE EQUIVALENT 1-131 to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
a.4 If the requirements of a.1, a.2 or a.3 cannot be met, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
b.1 With DOSE EQUIVALENT Xe-133 not within limit, restore DOSE EQUIVALENT Xe-133 to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
b.2 If the requirements of b.1 cannot be met, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Insert 6:
The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 100.11 (Ref. 1) or 10 CFR 50.67 for accidents analyzed using AST. Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.
The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.
The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT Xe-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2).
The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite and control room doses meet the appropriate 10CFR100.11 (Ref. 1) and 10CFR50 Appendix A GDC19 (Ref. 5) acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 IlCi/gm DOSE EQUIVALENT 1-131 from LCO 3.13, "Secondary Specific Activity."
The analyses for the SLB and SGTR accidents establish the acceptance limits for RCS specific activity. Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.
The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 IlCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60.0 IlCi/gm DOSE EQUIVALENT 1-131 due to an iodine spike caused by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 797IJCi/gm DOSE EQUIVALENT Xe-133.
The SGTR analysis assumes a rise in pressure in the ruptured SG causes radioactively contaminated steam to discharge to the atmosphere through the atmospheric dump valves or the main steam safety valves. The atmospheric discharge stops when the turbine bypass to the condenser removes the excess energy to rapidly reduce the RCS pressure and close the valves. The unaffected SG removes core decay heat by venting steam until the cooldown ends and the Decay Heat Removal (DHR) system is placed in service.
Insert 6 Continued:
The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SG removes core decay heat by venting steam to the atmosphere until the cooldown ends and the DHR system is placed in service.
Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 60.0 IJCi/gm for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.
RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The iodine specific activity in the reactor coolant is limited to 0.35 IlCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 7971JCi/gm DOSE EQUIVALENT Xe-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate 10CFR1 00.11 (Ref. 1) and 10CFR50 Appendix A GDC19 (Ref. 5) acceptance criteria.
The SLB and SGTR accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the 10CFR1 00.11 (Ref. 1) and 10CFR50 Appendix A GDC19 (Ref. 5) acceptance criteria.
APPLICABILITY In all MODES other than REFUELING SHUTDOWN and COLD SHUTDOWN, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT Xe-133 is necessary to limit the potential consequences of a SLB or SGTR to within the 10CFR100.11 acceptance criteria (Ref. 1) and 10CFR50 Appendix A GDC 19 acceptance criteria (Ref. 5).
In the REFUELING SHUTDOWN and COLD SHUTDOWN MODES, the steam generators are transitioning to decay heat removal and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.
ACTIONS With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that the specific activity is ~ 60.0 IJCi/gm. The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is required to obtain and analyze a sample. Sampling is continued every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide a trend.
Insert 6 Continued:
The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.
With the DOSE EQUIVALENT Xe-133 greater than the LCO limit, DOSE EQUIVALENT Xe-133 must be restored to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.
If the Required Actions of 3.1.4.3.a and 3.1.4.3.b are not met, or if the DOSE EQUIVALENT 1-131 is> 60.0 IJCi/gm, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS Table 4.1-3 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.
Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The 7-day Frequency considers the low probability of a gross fuel failure during this time.
Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR Table 4.1-3 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT Xe-133 is not detected, it should be assumed to be present at the minimum detectable activity.
The SR allows entry into and operation in HOT SHUTDOWN, HOT STANDBY, and STARTUP prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering POWER OPERATION.
The Table 4.1-3 surveillance for isotopic analysis for DOSE EQUIVALENT 1-131 concentration is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14-day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power change> 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.
Insert 6 Continued:
The SR allows entry into and operation in HOT SHUTDOWN, HOT STANDBY, and STARTUP prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering POWER OPERATION.
REFERENCES
- 1. 10CFR 100.11.
- 3. FSAR, Section 14.1.2.9.
- 4. FSAR, Section 14.1.2.10.
cnON statement pennitting POWER OPERAnON to continue for limited time periods w* the prim lant's specific activity greater than 0.35 microcurie/gram DOSE EQUIVALENT I, but within the a able limit shown on Figure 3.1-21, accommodates possible iodine spiki enomenon which may occur lowing changes in THERMAL POWER.
Proceeding to HOT S prevents the release of activity should earn generator tube rupture since the saturation pressure of *mary coolant is below the lift ssure of the atmospheric steam relief valves.
The surveillance requirements provide adequate ce that excessive specific activity levels in the primary coolant will be detected in sufficient ti 0 corrective action. Infonnation obtained on iodine spiking will be used to assess the p eters associa with spiking phenomena. A reduction in frequency of isotopic analyses followi power changes may be issible ifjustified by the data obtained.
The NRC statfhas perf1 ed a generic analysis of airborne radiation released v e Reactor Building Purge Isolation Va s. The dose contribution due to the radiation contained in the al d steam released through the p isolation valves prior to closure was found to be acceptable provided th requirem s of Specifications 3.1.4.1,3.1.4.2 and 3.1.4.3 are met.
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Dose e .valent 1-131 Primary Coolant Specific Activity Limit Versus Percent L POWER (with the Primary Coolant Specific Activity >0.35 ~Cilgram
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I THE Eq valent 1-131). \ I I ~ \
V Amendment No. ~. J,67. 204 3-9b \\1I
TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Item Check Frequen0' a.~~~u..Q
- 1. Reactor Coolant At least once each 7 days during POWER l~ (r15e.v-t. "1- OPERAnON/HOT STANDBY. START UP, ILl aM HOT SHUTDOWN.'
- b. Isotopic Analysis for DOSE EQUIVALENT i) I per 14 days during power operations.
1-131 Concentration ii) One Sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a TIfERMAL POWER change exceeding 1S% of the RATED TIfERMAL POWER within a one hour period during power operatioD,Allllft up &Ad lKK .aD~
iii) # Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds O.35j.Ci/gram DOSE EQUIVALENT 1-131 or-'l-
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- d. Chemistry (Cl, F and 02) 5 times/week when Tavg IS GREATER TIIAN 200°F.
- e. Boron concentration 2 times/week
- f. Tritiwn Radioactivity MonthJy
- 2. Borated Water Boron concentration Weekly and after each makeup when reactor coolant Storage Tank system pressure is gn:ater than 300 psig or Tavg is greater Water Sample than 2000F.
- 3. Core Flooding Tank Boron concentration MonthJy and after each makeup when RCS pressure is Water Sample greater than 700 psig.
Insert 7:
Verify reactor coolant DOSE EQUIVALENT Xe-133 specific activity is less than or equal to 797 microcuries/gram.