ML16035A015

R.E. Ginna - Application to Revise Technical Specifications to Adopt TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.

ML16035A015
Person / Time
Site:	Ginna
Issue date:	02/04/2016
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16035A015 (20)

Text

Exelon Generation February 4, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90

Subject:

Application to Revise Technical Specifications to Adopt TSTF-490, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" In accordance with 1 O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests changes to the Technical Specifications (TS) of the R.E. Ginna Nuclear Power Plant (Ginna). The proposed amendment revises Ginna's TS "Definitions," "RCS Specific Activity," and associated Surveillance Requirements (SR). The proposed changes would replace the current TS limit for Reactor Coolant System (RCS) gross specific activity with a new limit for RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent XE-133 definition that would replace the current E-Bar average disintegration energy definition. The changes are consistent with NRG-approved Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." The request is subdivided as follows:

• Attachment 1 provides a description and evaluation of the proposed change

• Attachment 2 provides a markup of the affected TS pages

• Attachment 3 provides revisions of the affected TS Bases pages. The TS Bases pages are provided for information only and do not require NRC approval. The proposed change has been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board. EGC requests approval of the proposed license amendment by February 4, 2017. Once approved, the amendment shall be implemented within 60 days of receipt. There are no regulatory commitments contained within this letter. In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the State New York of this application for license amendment by transmitting a copy of this letter and its attachments to a designated State Official.

Document Control Desk February 4, 2016 Page 2 Should you have any questions concerning this letter, please contact Laura Lynch, at (610) 765-5729. I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 4th day of February, 2016. Respectfully, l (} I James Barstow Director -Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Change 2. Markup of Proposed Technical Specification Pages 3. Revised Technical Specification Bases Pages cc: NRC Regional Administrator, Region I NRC Senior Resident Inspector, Ginna NRC Project Manager, Ginna A. L. Peterson, NYSERDA Evaluation of Proposed Change 1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Evaluation of Proposed Change 1.0 SUMMARY DESCRIPTION In accordance with 1 O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGG) requests changes to the Technical Specifications of the R.E. Ginna Nuclear Power Plant (Ginna). The proposed changes would replace the current limits for primary coolant gross specific activity with limits for primary coolant noble gas activity. The noble gas activity will be based on DOSE EQUIVALENT XE-133 and will take into account only the noble gas activity in the primary coolant. These changes were approved in an NRG Safety Evaluation (SE) dated March 19, 2007 (Reference 1 ). Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec," was announced for availability in Reference 1 as part of the Consolidated Line Item Improvement Process (CLllP). By memorandum from the Chief, Licensing Processes Branch, to the Plant Licensing Branch Chiefs, dated March 14, 2012, the NRG staff indicated that license amendment requests (LAR) related to TSTF-490 can be accepted for review, but will be handled through the normal LAR review process, instead of the expedited six-month CLllP schedule. 2.0 DETAILED DESCRIPTION Consistent with NRG-approved TSTF-490, Revision 0, EGG proposes the following TS changes for Ginna: 1. Delete the definition of E-AVERAGE DISINTEGRATION ENERGY. 2. Add a new definition for DOSE EQUIVALENT XE-133. 3. Revise LCO 3.4.16, "RCS Specific Activity," to delete references to specific activity of the reactor coolant, and to add a reference limit for DOSE EQUIVALENT XE-133. 4. Revise LCO 3.4.16, "APPLICABILITY" to specify the LCO is applicable in MODES 1, 2, 3, and 4. 5. Modify the ACTIONS Table as follows: a. Condition B (was condition C) is added to provide a Condition and Required Action for DOSE EQUIVALENT XE-133 instead of gross specific activity. The Completion Time is changed from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A Note allowing the applicability of LCO 3.0.4.c is added, consistent with the Note to Required Action A.1. b. Condition C (was Condition B) is modified based on changes to Conditions A and B to reflect the change in the LCO Applicability. 6. SR 3.4.16.1 is revised to verify the limit for DOSE EQUIVALENT XE-133. 7. SR 3.4.16.3 is deleted. Page 1 of 5 Evaluation of Proposed Change For Ginna, minor variations exist within their TS compared to the TSTF that do not change the technical intent of the changes proposed. A list is provided below of the variations to the approved TSTF-490, Revision O:

• Reference to the NRC staff SE, dated September 27, 2006 (ML062700612) is changed to refer to the NRC staff SE in Reference 1, because the SE dated September 27, 2006, that is referred to in the model application is not publically available. The SE posted in the Federal Register on March 19, 2007 is publically available and approved for use.

• The definition of DOSE EQUIVALENT 1-131 is not being revised. The current definition of DOSE EQUIVALENT 1-131 in Ginna's TS is consistent with TSTF 490, Revision 0.

• Condition A was not modified to delete the reference to Figure 3.4.16-1. This Figure is not included in the Ginna TS, and therefore does not require deletion. Further, Condition A was not modified to define an upper limit that is applicable at all power levels. An upper limit for Dose Equivalent 1-131 is already included in Ginna's TS. These variations meet the intent of TSTF 490, Revision 0.

• SR 3.4.16.2 is revised to delete the surveillance Note, "Only required to be performed in MODE 1." Deleting this Note provides continued assessment of RCS activity for all modes of applicability since DEi will no longer be limited to MODE 1 operation. Deleting this Note is consistent with other licensees that have adopted TSTF-490.

• Figure 3.4.16-1, "Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit versus Percent of RATED THERMAL POWER," is not included in the Ginna TS, and therefore does not require deletion. This change meets the intent of TSTF 490, Revision 0.

3.0 TECHNICAL EVALUATION

EGC has reviewed References 1 and 2. EGC has applied the methodology in Reference 1 to develop the proposed TS changes. EGC has also concluded that the justifications presented in TSTF-490, Revision O and the model SE prepared by the NRC staff are applicable to Ginna and justify this amendment for the incorporation of the changes to Ginna's TS. EGC has also reviewed requests by the NRC staff for additional information from licensees that have adopted TSTF-490. To assist in the NRC staff review of this amendment request, the inputs for determining the Dose Equivalent Xenon-133 limit are summarized in Table 1. Consistent with the dose consequence analysis, the determination of DOSE EQUIVALENT XE-133 is performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12 (FGR-12) (Reference 3), as shown in column 3. To normalize each radioisotope, each FGR-12 effective dose conversion factor is divided by the FGR-12 effective dose conversion factor for Xe-133. The resultant number is each radioisotope's equivalence factor, as shown in column 4 in Table 1. The equivalence factors are then multiplied by the concentrations of noble gases based on 1 % failed fuel, as provided in column 2, Table 1. The nuclide concentrations are assumed to be the total sum of the degassed gamma activities and the gaseous gamma activities for each appropriate nuclide. The products for each radionuclide are then summed to come up with the DEX Technical Specification limits. The Dose Equivalent Xe-133 Technical Specification limit is Page 2 of 5 Evaluation of Proposed Change calculated to be 659 µCi/g; however, the DEX limit will be implemented, conseNatively, as 650 µCi/g. Table 1 -Calculation of DEX Limit for Ginna 1 2 3 4 5 UFSAR EDE DCF per Ginna Table 9.3-9 FGR 12 DEX-133 UFSAR ( Sv per Bq s m-3) Equivalence µCVg µCi I g Factor DE XE-133 Kr-85m 1.93 7.48E-15 4.79 925 Kr-85 8.21 1.19E-16 0.076 0.626 Kr-87 1.24 4.12E-14 26.4 32.7 Kr-88 3.60 1.026-13 65.4 235 Xe-131m 3.54 3.89E-16 0.249 0.882 Xe-133m 3.84 1.376-15 0.878 3.37 Xe-133 271 1.56E-15 1.00 271 Xe-135m 0.56 2.04E-14 13.1 7.31 Xe-135 9.49 1.19E-14 7.63 72.4 Xe-138 0.69 5.77E-14 37.0 25.6 DEX Limit 659

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability dated March 19, 2007 (Reference 1 ), the NRC Notice for Comment published in November 20, 2006 (Reference 2), and TSTF-490, Revision 0. 4.2 Precedent EGC is not proposing significant variations or deviations from the TS changes described in TSTF-490, Revision 0, or in the content of the NRC's model SE published in Reference 1. The NRC has previously approved similar amendment requests to the TS for Palo Verde Nuclear Generating Station Units 1, 2 and 3(ML13294A576); Braidwood Station Units 1 and 2 and Byron Station Unit Nos. 1 and 2 (ML 100690386); and Three Mile Island Nuclear Station Unit 1 (ML 100320493). Submittals by these plants to request implementation of TSTF-490 were reviewed, along with corresponding requests for additional information (RAls). The letters for issuance of amendment were also reviewed to establish the final version of the approved amendment. 4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) has reviewed the proposed no significant hazards consideration determination published in the Federal Register on March 19, 2007 (Reference 1) Page 3 of 5 Evaluation of Proposed Change as part of the CLllP. EGC has concluded that the proposed determination presented in the notices is applicable to R.E. Ginna and is providing the full evaluation. Further, the traveler and model SE discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Ginna was not licensed to the 10 CFR 50, Appendix A, GDC. The Ginna equivalent of the referenced GDC is provided in Section 3.1 of the Updated Final Safety Analysis Report (UFSAR). This Section of the Ginna UFSAR provides an analysis of plant design criteria for Ginna to the GDC criteria. Based on the analysis performed, EGC believes that the plant-specific requirements for Ginna are sufficiently similar to the Appendix A GDC and represent an adequate technical basis for adopting the proposed change. 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Reactor coolant specific activity is not an initiator for any accident previously evaluated. The Completion Time when primary coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit primary coolant noble gases to concentrations consistent with the accident analyses. The proposed change to the Completion Time has no impact on the consequences of any design basis accident since the consequences of an accident during the extended Completion Time are the same as the consequences of an accident during the Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter. The change does not create the potential for a new or different kind of accident from any previously calculated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change revises the limits on noble gas radioactivity in the primary coolant. The proposed change is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses. Page 4 of 5 Evaluation of Proposed Change Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. Based on the above, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 O CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified. 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGG has reviewed the environmental consideration included in the model SE published in the Federal Register on March 19, 2007 (Reference 1). EGG has concluded that the NRC's .findings presented therein are applicable to Ginna and the determination is hereby incorporated by reference for this application.

6.0 REFERENCES

1. Federal Register Notice of Availability published on March 19, 2007, 72 FR 12838, "Notice of Availability of Model Application Concerning Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process." 2. Federal Register Notice for Comment published on November 20, 2006, 71 FR 67170, "Notice of Opportunity To Comment on Model Safety Evaluation and Model License Amendment Request on Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification; Babcock and Wilcox Pressurized Water Reactors, Westinghouse Pressurized Water Reactors, Combustion Engineering Pressurized Water Reactors Using the Consolidated Line Item Improvement Process." 3. Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil," 1993. Page 5 of 5 Markup of Proposed Technical Specification Pages R. E. Ginna Nuclear Power Plant Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION PAGES 1.1-2 3.4.16-1 3.4.16-2 CHANNEL OPERATIONAL TEST (COT) CORE AL TERA TIONS CORE OPERATING LIMITS REPORT (COLR) DOSE EQUIVALENT 1-131 E /WER/\GE aJERGY DOSE EQUIVALENT XE-133 Definitions 1.1 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. CORE ALTERATIONS shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE AL TERA TIONS shall not preclude completion of movement of a component to a safe position. The COLR is the plant specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP 30, Supplement to Part 1, pages 192-212, table entitled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity." E shall ee the mmFa§e (*.vei§hted iA fJFe190FtieA te the eeAeeAtFatieA ef ea eh FadieAuelide iA the FeaetoF eoolaAt at the tiFAe of saFAi:iliA§) of the SUFA of the a*1eFa§e 13eta a Ad §aFAFAa eAOF§ies (iA MeV) fJeF disiAE§FatieA foF ROA iodiAe isote19es, With half lives > 1 e FAiAUtes, FAakiA§ Uf) at least 96% of the total ASA iodiAe aetivity iA the eoolaAt. 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 Kr85m, 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, "External Exposure to Radionuclides in Air, Water, and Soil," 1993. R.E. Ginna Nuclear Power Plant 1.1-2 Amendment 100 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The speoifio aoti11ity of the reaotor ooolant 1Rcsi:5os-E E60TVALENTT=-! 1131 AND DOSE EQUIVALENT XE-133! specific activity shall be within limits. APPLICABILITY: MODES 1Q aREi 2, i3 and 4i MODE 3 with RCS average temperature (Tavg) > 500°F. ACTIONS A. CONDITION DOSE EQUIVALENT 1-131 specific activity not within limit. uired Action and assoc1 ompletion Time of Condi 1 ot OR Gross spec1 within limit. INSERT 2 RE. Ginna Nuclear Power Plant REQUIRED ACTION Lco 3.0.4.c is applicable. COMPLETION TIME A.1 Verify DOSE EQUIVALENT Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1-131::;; 60 µCi/gm. A.2 B.1 Restore DOSE EQUIVALENT 1-131 to within limit. Be in MODE 3 with T avg < 500°F. 3.4.16-1 7 days Amendment 88 RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS DOSE EQUIVALENT XE-133 SR 3.4.16.1 SR 3.4.16.2 Verify reactor coolant §f6SS specilc act\vity Ji4ee/E µCi/gm. 650 NOTE Only required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity 1.0 µCi/gm. -NOTE-Only required to be performed in MODE 1. R. E. Ginna Nuclear Power Plant 3.4.16-2 FREQUENCY 7 days 14 days Between 2 and 1 O hours after a THERMAL POWER change of 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 Once within 31 days after a minimum cl 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Amendment 88 INSERT 1: B. DOSE EQUIVALENT XE-133 not within limit INSERT 2: c. Required Action and associated Completion Time of Condition A or B not met. OR DOSE EQUIVALENT 1-131 specific activity > 60 µCi/gm. -NOTE-3.0.4.c is applicable Restore DOSE EQUIVALENT XE-133 to within limit. C.1 Be in MODE 3. AND C.2 Be in MODE 5. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 6 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Revised Technical Specification Bases Pages R. E. Ginna Nuclear Power Plant Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION BASES PAGES 8 3.4.16-1 B 3.4.16-2 8 3.4.16-3 8 3.4.16-4 8 3.4.16-5 RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.16 RCS Specific Activity BASES BACKGROUND APPLICABLE SAFETY ANALYSES The maximum dose that an individual can receive during an accident is specified in 10 CFR 50.67 (Ref. 1 ). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 50.67 limits 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 offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident. The specific activity limits for both DOSE EQUIVALENT 1-131 and [5oSE] !E6UiVALENTX-E=1'.fa[ gross specific activity are provided in the SRs. The allowable levels are intended to limit the dose to a small fraction of the 1 O CFR 50.67 dose limits. The limits in the LCO are standardized, based on evaluations of offsite radioactivity dose consequences for typical site locations. The evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 50.67 dose limits. The LCO limits on the specific activity of the reactor coolant ensures that the resulting doses will not exceed a small fraction of the 1 O CFR 50.67 dose limits following a SGTR accident. The SGTR dose analysis (Ref. 2) assumes the specific activity of the reactor coolant at the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gpd. The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the plant that could affect RCS specific activity, as they relate to the acceptance limits. R.E. Ginna Nuclear Power Plant 83.4.16-1 Revision 42 LCO RCS Specific Activity B 3.4.16 The analysis is for two cases of reactor coolant specific activity (Ref. 2). One case assumes specific activity at 1.0 µCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the 1-131 activity in the reactor coolant by a factor of 335 for a duration of eight hours immediately after the accident. The second case assumes the initial reactor coolant iodine activity at 60.0 µCi/gm DOSE EQUIVALENT 1-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coQlant assumes 1 % failed fuel, which closely equals the LCO limit of 1 OO!E ttGi.<§m for §Foss speoifio activity. The assumption used to calculate dose for the Control Room, Exclusion Area Boundary (EAB) and Low Population Zone are included in Reference 2. The dose analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose limits. Operation with iodine specific activity levels greater than the LCO limit is permissible provided that the activity levels do not exceed 60.0 µCi/gm. The increased permissible iodine levels are acceptable because of the low probability of a SGTR accident occurring during the es1ablished 7 day time limit. The occurrence of an SGTR accident at these permissible levels could increase dose levels, but they would still be within 1 O CFR 50.67 dose limits. RCS specific activity satisfies Criterion 2 of the NRC Policy Statement. The specific iodine to 1.0 µCi/gm DOSE EQUIVALENT is i:tGi.<§m (*wl=lere Eis tl=le avera§e E:iisiRte§ratioR eReF!iJY of tl=le s1:1m of tl=le a*;era!iJe beta aRE:i §amma eRer§ies oftl=le eoelaRt R1:1oliaes). Tl=le limit OR DOSE I 1a1 ensures the dose to an individual during the Design Basis Accident (OBA) will be a small fraction of the allowed dose. The SGTR accident analysis (Ref. 2) shows that the dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to doses that exceed the 10 CFR 50.67 dose limits. R.E. Ginna Nuclear Power Plant B 3.4.16-2 Revision42 APPLICABILITY RCS Specific Activity B 3.4.16 In MODES 1 aA62,13. and 41 aRel iR l\A09E a witR RCS aveFa*Je teFR13eFat1:1Fe > eQQ°F, operation within the LCO limits for DOSE EQUIVALENT 1-131 and IDOSE EQUIVALENT XE-133! *JFess s13eeifie astivity are necessary to contain the potential consequences of an SGTR to within the acceptable dose values. Fer e13eratieR iR l\A09E a witl=I RCS avera*Je teFR13erat1:1Fe < eQQ2F, aRel iR ,-------, ;7' M09ES 4 aRel e, the release of raelioaeti*1ity iR the eveRt ef a SGTR is 1:1Rlikely eiRee the sat1:1ratioR 13ress1:1re of the reastor eoolaRt is Below the lift 13ress1:1re ef the maiR steaFR safety valves. ACTIONS nsert NOTE 2 A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> must be taken to demonstrate that the limits of 60 µCi/gm are not exceeded. The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is required to obtain and analyze a sample. Sampling is done to continue to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limits within 7 days if the limit violation resulted from normal iodine spiking. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the appicable MODE(S) while relying on the ACTIONS. This allowance is provided because of the significant conservatism included in the LCO limit. Also, reducing the DOSE EQUIVALENT 1-131 to within limits is accomplished through use of the Chemical and Volume Control System (CVCS) demina-alizers. This cleanup operation parallels plant restart following a reactor trip which frequently results in iodine spikes due to the large step decrease in reactor power level and RCS pressure excursion. The cleanup operation can normally be accomplished within the LCO Completion Time of 7 days. If a Reei1:1ireel l\etioR aRel the assoeiateel CeFR13letioR TiFRe of CoRelitioR /\ is Rot FRet er if tl=le 90SE I 1 a1 s13eeifie aeti'1ity is gmatm thaR 6Q ttCi}!1JFR, the reaetoF FRl:lst Be BFe1:1!1Jht ta l\A09E a witl=I RCS aveFa*Je teFR13eFat1:1Fe < 6QQPF witl=liR 8 he1:1rs. Tl=le et:laRge witl=liR 8 l=le1:1rs ta l\AOQE a aRel RCS avera!1Je teFR13erat1:1re < eQQP.F lav.*eFs the sat1:1ratiaR 13ress1:1re af the reaotor eoolaRt Below tl=le set13oiRts of the FRaiR steaFR valves aRel 13reveRts a1:1taFRatieally veRtiR!1J the SC to tl=le eRlfiFeRFReRt iR aR SGTR eveRt. Tl=le CaFR13leUaR TiFRe af 8 l>lel:lrs is reasoRaBle, Baseel OR 013eratiR§ e><13erieRee, ta reaeh l\A09E 6 Belew eQQ°F freFR full 13ewer eeRelitieRs iR aR orelerly maRRer aRel witho1:1t ehalleR!1JiR*J 13laRt systeFRs. RE. Ginna Nuclear Power Plant B 3.4.16-3 Revision 42 RCS Specific Activity B 3.4.16 C.1landC.21 SURVEILLANCE REQUIREMENTS If tRe *jFOss s13eeifie aeti'l*ity is not witRin liFAit, tRe eRan*je witRin 8 Rours to MODE a anel RCS avera§e tem13erature < 600°F lowers tRe saturation 13ressure of tRe roaster ooolant belO't'f tRe set13oints of tRe main steaFA safety valves anel 13re*.*ents automatieally ventin*j tRe SG te tRe environment in an SGf R event. TRe alloweel GoFA13letion TiFAe of 8 Rours is reasonable, easeel en e13eratin!:J ex13erienee, te reaoR MODE a below SQQP.F from full 13ewer eonelitions in an erelerly manner anel witReut ol=lallen!:Jin§ 13lant systems. SR 3.4.16.1 This SR re uires performing a gamma isotopic analysis as a measure of the §f6SS noble a specific activity of the reactor coolant at least once every 7 days. \OJl=lile easioally a 1:1uantitative FAeasure of raelionuolieles witl=l Ralf li'l'OS lon!:Jer tl=lan 16 FAinutes, mcoluelin§ ioelines, 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 §f0SS lnoble gaSJ 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 Surveillance is applicable in MODES 11. 2, 3 and 4.l &A6-2, anel in MODE a witl=l 600°F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during this time. SR3.4.16.2 This SR is 0fl-ly performed in MODE 1 to ensure iodine remains within limits during normal operation and following fast power changes when fuel failure is more likely to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 1 O hours 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 fuel failure; samples at other times would provide inaccurate results. SR a.4.16.a A raeliool=leFAieal analysis for E eleterFAination is re1:1uireel witRin a1 elays after a miniFAUFA of 2 e#eotive full 13ower elays anel 20 elays of MODE 1 013eration l=lave ela13seel sinee tl=le roaster was last sueoritioal for at least 48 l=louFS anel every 184 elays (6 montRs) tl=lereafter. Tl=lis ensures tl=lat tRe raelieaotive FAaterials are at OEtUilieriuFA se tl:ie analysis for E is re13resentative anel not sl<eweel ey a oruel burst or otl=ler similar aenorFAal event. Tl=le E eleterFAination elireotly relates te tl=le LGO anel is re1:1uireel te 13lant e13eration witl=lin tl=le s13eeifieel !:JFess aotivity LGO limit. Tl=le R.E. Ginna Nuclear Power Plant B 3.4.16-4 Revision 42 REFERENCES RCS Specific Activity B 3.4.16 aAalysis fer E is a FAeas1:1reFAeAt ef tAe a*1era§e eAer§ies 13er elisiAtO§ratieA fer isete13es witA Ralf lives leA§Or tAOA 1 e FAiA1:1tes, mEel1:1eliA§ ieeliAes. TAe Freei1:10Aey reee§Aii!es E elees Aet eAOA§O ra13idly. TAis eR is FAeelifieel by a f)Jete tAat iAelieates saFA13liA§ is eAly reei1:1irea to be 13erferFAeel iA MODE 1 s1:1eA tAat eei1:1ilibri1:1FA eeAelitieAs are 13reseRt a1:1riA§ tAe saFA13le. 1. 10 CFR 50.67. 2. Design Analysis DA-NS-2001-084, Steam Generator Tube Rupture Offsite and Control Room Doses. R.E. Ginna Nuclear Power Plant B 3.4.16-5 Revision 42 NOTE 1: In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS activity is not required. NOTE 2: With the DOSE EQUIVALENT XE-133 greaterthan 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. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODES, relying on Required Action 8.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plantremains at, or proceeds to, power operation. NOTE 3: If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is> 60.0 µCi/g, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 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. NOTE 4: 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 3.4.16.1 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.

Exelon Generation February 4, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90

Subject:

Application to Revise Technical Specifications to Adopt TSTF-490, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" In accordance with 1 O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests changes to the Technical Specifications (TS) of the R.E. Ginna Nuclear Power Plant (Ginna). The proposed amendment revises Ginna's TS "Definitions," "RCS Specific Activity," and associated Surveillance Requirements (SR). The proposed changes would replace the current TS limit for Reactor Coolant System (RCS) gross specific activity with a new limit for RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent XE-133 definition that would replace the current E-Bar average disintegration energy definition. The changes are consistent with NRG-approved Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." The request is subdivided as follows:

• Attachment 1 provides a description and evaluation of the proposed change

• Attachment 2 provides a markup of the affected TS pages

• Attachment 3 provides revisions of the affected TS Bases pages. The TS Bases pages are provided for information only and do not require NRC approval. The proposed change has been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board. EGC requests approval of the proposed license amendment by February 4, 2017. Once approved, the amendment shall be implemented within 60 days of receipt. There are no regulatory commitments contained within this letter. In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the State New York of this application for license amendment by transmitting a copy of this letter and its attachments to a designated State Official.

Document Control Desk February 4, 2016 Page 2 Should you have any questions concerning this letter, please contact Laura Lynch, at (610) 765-5729. I declare under penalty of perjury that the foregoing is true and correct. This statement was executed on the 4th day of February, 2016. Respectfully, l (} I James Barstow Director -Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Change 2. Markup of Proposed Technical Specification Pages 3. Revised Technical Specification Bases Pages cc: NRC Regional Administrator, Region I NRC Senior Resident Inspector, Ginna NRC Project Manager, Ginna A. L. Peterson, NYSERDA Evaluation of Proposed Change 1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Evaluation of Proposed Change 1.0 SUMMARY DESCRIPTION In accordance with 1 O CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGG) requests changes to the Technical Specifications of the R.E. Ginna Nuclear Power Plant (Ginna). The proposed changes would replace the current limits for primary coolant gross specific activity with limits for primary coolant noble gas activity. The noble gas activity will be based on DOSE EQUIVALENT XE-133 and will take into account only the noble gas activity in the primary coolant. These changes were approved in an NRG Safety Evaluation (SE) dated March 19, 2007 (Reference 1 ). Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec," was announced for availability in Reference 1 as part of the Consolidated Line Item Improvement Process (CLllP). By memorandum from the Chief, Licensing Processes Branch, to the Plant Licensing Branch Chiefs, dated March 14, 2012, the NRG staff indicated that license amendment requests (LAR) related to TSTF-490 can be accepted for review, but will be handled through the normal LAR review process, instead of the expedited six-month CLllP schedule. 2.0 DETAILED DESCRIPTION Consistent with NRG-approved TSTF-490, Revision 0, EGG proposes the following TS changes for Ginna: 1. Delete the definition of E-AVERAGE DISINTEGRATION ENERGY. 2. Add a new definition for DOSE EQUIVALENT XE-133. 3. Revise LCO 3.4.16, "RCS Specific Activity," to delete references to specific activity of the reactor coolant, and to add a reference limit for DOSE EQUIVALENT XE-133. 4. Revise LCO 3.4.16, "APPLICABILITY" to specify the LCO is applicable in MODES 1, 2, 3, and 4. 5. Modify the ACTIONS Table as follows: a. Condition B (was condition C) is added to provide a Condition and Required Action for DOSE EQUIVALENT XE-133 instead of gross specific activity. The Completion Time is changed from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A Note allowing the applicability of LCO 3.0.4.c is added, consistent with the Note to Required Action A.1. b. Condition C (was Condition B) is modified based on changes to Conditions A and B to reflect the change in the LCO Applicability. 6. SR 3.4.16.1 is revised to verify the limit for DOSE EQUIVALENT XE-133. 7. SR 3.4.16.3 is deleted. Page 1 of 5 Evaluation of Proposed Change For Ginna, minor variations exist within their TS compared to the TSTF that do not change the technical intent of the changes proposed. A list is provided below of the variations to the approved TSTF-490, Revision O:

• Reference to the NRC staff SE, dated September 27, 2006 (ML062700612) is changed to refer to the NRC staff SE in Reference 1, because the SE dated September 27, 2006, that is referred to in the model application is not publically available. The SE posted in the Federal Register on March 19, 2007 is publically available and approved for use.

• The definition of DOSE EQUIVALENT 1-131 is not being revised. The current definition of DOSE EQUIVALENT 1-131 in Ginna's TS is consistent with TSTF 490, Revision 0.

• Condition A was not modified to delete the reference to Figure 3.4.16-1. This Figure is not included in the Ginna TS, and therefore does not require deletion. Further, Condition A was not modified to define an upper limit that is applicable at all power levels. An upper limit for Dose Equivalent 1-131 is already included in Ginna's TS. These variations meet the intent of TSTF 490, Revision 0.

• SR 3.4.16.2 is revised to delete the surveillance Note, "Only required to be performed in MODE 1." Deleting this Note provides continued assessment of RCS activity for all modes of applicability since DEi will no longer be limited to MODE 1 operation. Deleting this Note is consistent with other licensees that have adopted TSTF-490.

• Figure 3.4.16-1, "Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit versus Percent of RATED THERMAL POWER," is not included in the Ginna TS, and therefore does not require deletion. This change meets the intent of TSTF 490, Revision 0.

3.0 TECHNICAL EVALUATION

EGC has reviewed References 1 and 2. EGC has applied the methodology in Reference 1 to develop the proposed TS changes. EGC has also concluded that the justifications presented in TSTF-490, Revision O and the model SE prepared by the NRC staff are applicable to Ginna and justify this amendment for the incorporation of the changes to Ginna's TS. EGC has also reviewed requests by the NRC staff for additional information from licensees that have adopted TSTF-490. To assist in the NRC staff review of this amendment request, the inputs for determining the Dose Equivalent Xenon-133 limit are summarized in Table 1. Consistent with the dose consequence analysis, the determination of DOSE EQUIVALENT XE-133 is performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12 (FGR-12) (Reference 3), as shown in column 3. To normalize each radioisotope, each FGR-12 effective dose conversion factor is divided by the FGR-12 effective dose conversion factor for Xe-133. The resultant number is each radioisotope's equivalence factor, as shown in column 4 in Table 1. The equivalence factors are then multiplied by the concentrations of noble gases based on 1 % failed fuel, as provided in column 2, Table 1. The nuclide concentrations are assumed to be the total sum of the degassed gamma activities and the gaseous gamma activities for each appropriate nuclide. The products for each radionuclide are then summed to come up with the DEX Technical Specification limits. The Dose Equivalent Xe-133 Technical Specification limit is Page 2 of 5 Evaluation of Proposed Change calculated to be 659 µCi/g; however, the DEX limit will be implemented, conseNatively, as 650 µCi/g. Table 1 -Calculation of DEX Limit for Ginna 1 2 3 4 5 UFSAR EDE DCF per Ginna Table 9.3-9 FGR 12 DEX-133 UFSAR ( Sv per Bq s m-3) Equivalence µCVg µCi I g Factor DE XE-133 Kr-85m 1.93 7.48E-15 4.79 925 Kr-85 8.21 1.19E-16 0.076 0.626 Kr-87 1.24 4.12E-14 26.4 32.7 Kr-88 3.60 1.026-13 65.4 235 Xe-131m 3.54 3.89E-16 0.249 0.882 Xe-133m 3.84 1.376-15 0.878 3.37 Xe-133 271 1.56E-15 1.00 271 Xe-135m 0.56 2.04E-14 13.1 7.31 Xe-135 9.49 1.19E-14 7.63 72.4 Xe-138 0.69 5.77E-14 37.0 25.6 DEX Limit 659

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability dated March 19, 2007 (Reference 1 ), the NRC Notice for Comment published in November 20, 2006 (Reference 2), and TSTF-490, Revision 0. 4.2 Precedent EGC is not proposing significant variations or deviations from the TS changes described in TSTF-490, Revision 0, or in the content of the NRC's model SE published in Reference 1. The NRC has previously approved similar amendment requests to the TS for Palo Verde Nuclear Generating Station Units 1, 2 and 3(ML13294A576); Braidwood Station Units 1 and 2 and Byron Station Unit Nos. 1 and 2 (ML 100690386); and Three Mile Island Nuclear Station Unit 1 (ML 100320493). Submittals by these plants to request implementation of TSTF-490 were reviewed, along with corresponding requests for additional information (RAls). The letters for issuance of amendment were also reviewed to establish the final version of the approved amendment. 4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) has reviewed the proposed no significant hazards consideration determination published in the Federal Register on March 19, 2007 (Reference 1) Page 3 of 5 Evaluation of Proposed Change as part of the CLllP. EGC has concluded that the proposed determination presented in the notices is applicable to R.E. Ginna and is providing the full evaluation. Further, the traveler and model SE discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Ginna was not licensed to the 10 CFR 50, Appendix A, GDC. The Ginna equivalent of the referenced GDC is provided in Section 3.1 of the Updated Final Safety Analysis Report (UFSAR). This Section of the Ginna UFSAR provides an analysis of plant design criteria for Ginna to the GDC criteria. Based on the analysis performed, EGC believes that the plant-specific requirements for Ginna are sufficiently similar to the Appendix A GDC and represent an adequate technical basis for adopting the proposed change. 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Reactor coolant specific activity is not an initiator for any accident previously evaluated. The Completion Time when primary coolant gross activity is not within limit is not an initiator for any accident previously evaluated. The current variable limit on primary coolant iodine concentration is not an initiator to any accident previously evaluated. As a result, the proposed change does not significantly increase the probability of an accident. The proposed change will limit primary coolant noble gases to concentrations consistent with the accident analyses. The proposed change to the Completion Time has no impact on the consequences of any design basis accident since the consequences of an accident during the extended Completion Time are the same as the consequences of an accident during the Completion Time. As a result, the consequences of any accident previously evaluated are not significantly increased. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change in specific activity limits does not alter any physical part of the plant nor does it affect any plant operating parameter. The change does not create the potential for a new or different kind of accident from any previously calculated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change revises the limits on noble gas radioactivity in the primary coolant. The proposed change is consistent with the assumptions in the safety analyses and will ensure the monitored values protect the initial assumptions in the safety analyses. Page 4 of 5 Evaluation of Proposed Change Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. Based on the above, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 O CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified. 4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGG has reviewed the environmental consideration included in the model SE published in the Federal Register on March 19, 2007 (Reference 1). EGG has concluded that the NRC's .findings presented therein are applicable to Ginna and the determination is hereby incorporated by reference for this application.

6.0 REFERENCES

1. Federal Register Notice of Availability published on March 19, 2007, 72 FR 12838, "Notice of Availability of Model Application Concerning Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification Using the Consolidated Line Item Improvement Process." 2. Federal Register Notice for Comment published on November 20, 2006, 71 FR 67170, "Notice of Opportunity To Comment on Model Safety Evaluation and Model License Amendment Request on Technical Specification Improvement Regarding Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification; Babcock and Wilcox Pressurized Water Reactors, Westinghouse Pressurized Water Reactors, Combustion Engineering Pressurized Water Reactors Using the Consolidated Line Item Improvement Process." 3. Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil," 1993. Page 5 of 5 Markup of Proposed Technical Specification Pages R. E. Ginna Nuclear Power Plant Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION PAGES 1.1-2 3.4.16-1 3.4.16-2 CHANNEL OPERATIONAL TEST (COT) CORE AL TERA TIONS CORE OPERATING LIMITS REPORT (COLR) DOSE EQUIVALENT 1-131 E /WER/\GE aJERGY DOSE EQUIVALENT XE-133 Definitions 1.1 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. CORE ALTERATIONS shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE AL TERA TIONS shall not preclude completion of movement of a component to a safe position. The COLR is the plant specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications. DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP 30, Supplement to Part 1, pages 192-212, table entitled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity." E shall ee the mmFa§e (*.vei§hted iA fJFe190FtieA te the eeAeeAtFatieA ef ea eh FadieAuelide iA the FeaetoF eoolaAt at the tiFAe of saFAi:iliA§) of the SUFA of the a*1eFa§e 13eta a Ad §aFAFAa eAOF§ies (iA MeV) fJeF disiAE§FatieA foF ROA iodiAe isote19es, With half lives > 1 e FAiAUtes, FAakiA§ Uf) at least 96% of the total ASA iodiAe aetivity iA the eoolaAt. 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 Kr85m, 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, "External Exposure to Radionuclides in Air, Water, and Soil," 1993. R.E. Ginna Nuclear Power Plant 1.1-2 Amendment 100 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The speoifio aoti11ity of the reaotor ooolant 1Rcsi:5os-E E60TVALENTT=-! 1131 AND DOSE EQUIVALENT XE-133! specific activity shall be within limits. APPLICABILITY: MODES 1Q aREi 2, i3 and 4i MODE 3 with RCS average temperature (Tavg) > 500°F. ACTIONS A. CONDITION DOSE EQUIVALENT 1-131 specific activity not within limit. uired Action and assoc1 ompletion Time of Condi 1 ot OR Gross spec1 within limit. INSERT 2 RE. Ginna Nuclear Power Plant REQUIRED ACTION Lco 3.0.4.c is applicable. COMPLETION TIME A.1 Verify DOSE EQUIVALENT Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1-131::;; 60 µCi/gm. A.2 B.1 Restore DOSE EQUIVALENT 1-131 to within limit. Be in MODE 3 with T avg < 500°F. 3.4.16-1 7 days Amendment 88 RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS DOSE EQUIVALENT XE-133 SR 3.4.16.1 SR 3.4.16.2 Verify reactor coolant §f6SS specilc act\vity Ji4ee/E µCi/gm. 650 NOTE Only required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity 1.0 µCi/gm. -NOTE-Only required to be performed in MODE 1. R. E. Ginna Nuclear Power Plant 3.4.16-2 FREQUENCY 7 days 14 days Between 2 and 1 O hours after a THERMAL POWER change of 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 Once within 31 days after a minimum cl 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Amendment 88 INSERT 1: B. DOSE EQUIVALENT XE-133 not within limit INSERT 2: c. Required Action and associated Completion Time of Condition A or B not met. OR DOSE EQUIVALENT 1-131 specific activity > 60 µCi/gm. -NOTE-3.0.4.c is applicable Restore DOSE EQUIVALENT XE-133 to within limit. C.1 Be in MODE 3. AND C.2 Be in MODE 5. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 6 hours 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Revised Technical Specification Bases Pages R. E. Ginna Nuclear Power Plant Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION BASES PAGES 8 3.4.16-1 B 3.4.16-2 8 3.4.16-3 8 3.4.16-4 8 3.4.16-5 RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.16 RCS Specific Activity BASES BACKGROUND APPLICABLE SAFETY ANALYSES The maximum dose that an individual can receive during an accident is specified in 10 CFR 50.67 (Ref. 1 ). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 50.67 limits 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 offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident. The specific activity limits for both DOSE EQUIVALENT 1-131 and [5oSE] !E6UiVALENTX-E=1'.fa[ gross specific activity are provided in the SRs. The allowable levels are intended to limit the dose to a small fraction of the 1 O CFR 50.67 dose limits. The limits in the LCO are standardized, based on evaluations of offsite radioactivity dose consequences for typical site locations. The evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 50.67 dose limits. The LCO limits on the specific activity of the reactor coolant ensures that the resulting doses will not exceed a small fraction of the 1 O CFR 50.67 dose limits following a SGTR accident. The SGTR dose analysis (Ref. 2) assumes the specific activity of the reactor coolant at the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gpd. The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the plant that could affect RCS specific activity, as they relate to the acceptance limits. R.E. Ginna Nuclear Power Plant 83.4.16-1 Revision 42 LCO RCS Specific Activity B 3.4.16 The analysis is for two cases of reactor coolant specific activity (Ref. 2). One case assumes specific activity at 1.0 µCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the 1-131 activity in the reactor coolant by a factor of 335 for a duration of eight hours immediately after the accident. The second case assumes the initial reactor coolant iodine activity at 60.0 µCi/gm DOSE EQUIVALENT 1-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coQlant assumes 1 % failed fuel, which closely equals the LCO limit of 1 OO!E ttGi.<§m for §Foss speoifio activity. The assumption used to calculate dose for the Control Room, Exclusion Area Boundary (EAB) and Low Population Zone are included in Reference 2. The dose analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose limits. Operation with iodine specific activity levels greater than the LCO limit is permissible provided that the activity levels do not exceed 60.0 µCi/gm. The increased permissible iodine levels are acceptable because of the low probability of a SGTR accident occurring during the es1ablished 7 day time limit. The occurrence of an SGTR accident at these permissible levels could increase dose levels, but they would still be within 1 O CFR 50.67 dose limits. RCS specific activity satisfies Criterion 2 of the NRC Policy Statement. The specific iodine to 1.0 µCi/gm DOSE EQUIVALENT is i:tGi.<§m (*wl=lere Eis tl=le avera§e E:iisiRte§ratioR eReF!iJY of tl=le s1:1m of tl=le a*;era!iJe beta aRE:i §amma eRer§ies oftl=le eoelaRt R1:1oliaes). Tl=le limit OR DOSE I 1a1 ensures the dose to an individual during the Design Basis Accident (OBA) will be a small fraction of the allowed dose. The SGTR accident analysis (Ref. 2) shows that the dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to doses that exceed the 10 CFR 50.67 dose limits. R.E. Ginna Nuclear Power Plant B 3.4.16-2 Revision42 APPLICABILITY RCS Specific Activity B 3.4.16 In MODES 1 aA62,13. and 41 aRel iR l\A09E a witR RCS aveFa*Je teFR13eFat1:1Fe > eQQ°F, operation within the LCO limits for DOSE EQUIVALENT 1-131 and IDOSE EQUIVALENT XE-133! *JFess s13eeifie astivity are necessary to contain the potential consequences of an SGTR to within the acceptable dose values. Fer e13eratieR iR l\A09E a witl=I RCS avera*Je teFR13erat1:1Fe < eQQ2F, aRel iR ,-------, ;7' M09ES 4 aRel e, the release of raelioaeti*1ity iR the eveRt ef a SGTR is 1:1Rlikely eiRee the sat1:1ratioR 13ress1:1re of the reastor eoolaRt is Below the lift 13ress1:1re ef the maiR steaFR safety valves. ACTIONS nsert NOTE 2 A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> must be taken to demonstrate that the limits of 60 µCi/gm are not exceeded. The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is required to obtain and analyze a sample. Sampling is done to continue to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limits within 7 days if the limit violation resulted from normal iodine spiking. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the appicable MODE(S) while relying on the ACTIONS. This allowance is provided because of the significant conservatism included in the LCO limit. Also, reducing the DOSE EQUIVALENT 1-131 to within limits is accomplished through use of the Chemical and Volume Control System (CVCS) demina-alizers. This cleanup operation parallels plant restart following a reactor trip which frequently results in iodine spikes due to the large step decrease in reactor power level and RCS pressure excursion. The cleanup operation can normally be accomplished within the LCO Completion Time of 7 days. If a Reei1:1ireel l\etioR aRel the assoeiateel CeFR13letioR TiFRe of CoRelitioR /\ is Rot FRet er if tl=le 90SE I 1 a1 s13eeifie aeti'1ity is gmatm thaR 6Q ttCi}!1JFR, the reaetoF FRl:lst Be BFe1:1!1Jht ta l\A09E a witl=I RCS aveFa*Je teFR13eFat1:1Fe < 6QQPF witl=liR 8 he1:1rs. Tl=le et:laRge witl=liR 8 l=le1:1rs ta l\AOQE a aRel RCS avera!1Je teFR13erat1:1re < eQQP.F lav.*eFs the sat1:1ratiaR 13ress1:1re af the reaotor eoolaRt Below tl=le set13oiRts of the FRaiR steaFR valves aRel 13reveRts a1:1taFRatieally veRtiR!1J the SC to tl=le eRlfiFeRFReRt iR aR SGTR eveRt. Tl=le CaFR13leUaR TiFRe af 8 l>lel:lrs is reasoRaBle, Baseel OR 013eratiR§ e><13erieRee, ta reaeh l\A09E 6 Belew eQQ°F freFR full 13ewer eeRelitieRs iR aR orelerly maRRer aRel witho1:1t ehalleR!1JiR*J 13laRt systeFRs. RE. Ginna Nuclear Power Plant B 3.4.16-3 Revision 42 RCS Specific Activity B 3.4.16 C.1landC.21 SURVEILLANCE REQUIREMENTS If tRe *jFOss s13eeifie aeti'l*ity is not witRin liFAit, tRe eRan*je witRin 8 Rours to MODE a anel RCS avera§e tem13erature < 600°F lowers tRe saturation 13ressure of tRe roaster ooolant belO't'f tRe set13oints of tRe main steaFA safety valves anel 13re*.*ents automatieally ventin*j tRe SG te tRe environment in an SGf R event. TRe alloweel GoFA13letion TiFAe of 8 Rours is reasonable, easeel en e13eratin!:J ex13erienee, te reaoR MODE a below SQQP.F from full 13ewer eonelitions in an erelerly manner anel witReut ol=lallen!:Jin§ 13lant systems. SR 3.4.16.1 This SR re uires performing a gamma isotopic analysis as a measure of the §f6SS noble a specific activity of the reactor coolant at least once every 7 days. \OJl=lile easioally a 1:1uantitative FAeasure of raelionuolieles witl=l Ralf li'l'OS lon!:Jer tl=lan 16 FAinutes, mcoluelin§ ioelines, 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 §f0SS lnoble gaSJ 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 Surveillance is applicable in MODES 11. 2, 3 and 4.l &A6-2, anel in MODE a witl=l 600°F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during this time. SR3.4.16.2 This SR is 0fl-ly performed in MODE 1 to ensure iodine remains within limits during normal operation and following fast power changes when fuel failure is more likely to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 1 O hours 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 fuel failure; samples at other times would provide inaccurate results. SR a.4.16.a A raeliool=leFAieal analysis for E eleterFAination is re1:1uireel witRin a1 elays after a miniFAUFA of 2 e#eotive full 13ower elays anel 20 elays of MODE 1 013eration l=lave ela13seel sinee tl=le roaster was last sueoritioal for at least 48 l=louFS anel every 184 elays (6 montRs) tl=lereafter. Tl=lis ensures tl=lat tRe raelieaotive FAaterials are at OEtUilieriuFA se tl:ie analysis for E is re13resentative anel not sl<eweel ey a oruel burst or otl=ler similar aenorFAal event. Tl=le E eleterFAination elireotly relates te tl=le LGO anel is re1:1uireel te 13lant e13eration witl=lin tl=le s13eeifieel !:JFess aotivity LGO limit. Tl=le R.E. Ginna Nuclear Power Plant B 3.4.16-4 Revision 42 REFERENCES RCS Specific Activity B 3.4.16 aAalysis fer E is a FAeas1:1reFAeAt ef tAe a*1era§e eAer§ies 13er elisiAtO§ratieA fer isete13es witA Ralf lives leA§Or tAOA 1 e FAiA1:1tes, mEel1:1eliA§ ieeliAes. TAe Freei1:10Aey reee§Aii!es E elees Aet eAOA§O ra13idly. TAis eR is FAeelifieel by a f)Jete tAat iAelieates saFA13liA§ is eAly reei1:1irea to be 13erferFAeel iA MODE 1 s1:1eA tAat eei1:1ilibri1:1FA eeAelitieAs are 13reseRt a1:1riA§ tAe saFA13le. 1. 10 CFR 50.67. 2. Design Analysis DA-NS-2001-084, Steam Generator Tube Rupture Offsite and Control Room Doses. R.E. Ginna Nuclear Power Plant B 3.4.16-5 Revision 42 NOTE 1: In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS activity is not required. NOTE 2: With the DOSE EQUIVALENT XE-133 greaterthan 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. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODES, relying on Required Action 8.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plantremains at, or proceeds to, power operation. NOTE 3: If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is> 60.0 µCi/g, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 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. NOTE 4: 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 3.4.16.1 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.