ML12072A205

Application to Change Dose Equivalent I-131 Spike Limit and Allowable Value for Control Room Air Intake Radiation Monitors (WBN-TS-11-08)

ML12072A205
Person / Time
Site:	Watts Bar
Issue date:	03/08/2012
From:	Shea J W Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WBN-TS-11-08
Download: ML12072A205 (34)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 March 8, 2012 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

Watts Bar Nuclear Plant Unit I -Application to Change Dose Equivalent 1-131 Spike Limit and Allowable Value for Control Room Air Intake Radiation Monitors (WBN-TS-11-08)

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA)requests a change (WBN-TS-1 1-08) to Watts Bar Nuclear Plant (WBN), Unit 1, Facility Operating License No. NPF-90.The proposed amendment will revise: 1) Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from < 9.45E-05 micro-Curie/cubic centimeter (pCi/cc)(3,308 counts per minute (cpm)) to < 1.647E-04 pCi/cc (3,308 cpm); and 2) TS 3.4.16,"RCS Specific Activity," by lowering the DOSE EQUIVALENT 1-131 spike limit from 21 micro-Curie/gram (pCi/gm) to 14 pCi/gm in Required Action A.1 and Condition C.The Enclosure to this letter provides a description, technical evaluation, regulatory evaluation, and discussion of environmental considerations of the proposed changes.Attachments 1 and 2 to the enclosure provide the existing TS and TS Bases pages marked-up to show the proposed changes. Attachments 3 and 4 to the enclosure provide the existing TS and TS Bases pages retyped to show the proposed changes.The TS Bases pages are provided to the NRC for information only.Printed on recycled paper U U.S. Nuclear Regulatory Commission Page 2 March 8, 2012 Until WBN, Unit 1 implements an NRC approved License Amendment that lowers the DOSE EQUIVALENT 1-131 spike limit from 21 pCi/gm to 14 pCi/gm, the new numerical limit for DOSE EQUIVALENT 1-131 will be controlled administratively in accordance with NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety." The WBN Plant Operations Review Committee and the TVA Nuclear Safety Review Board have reviewed these proposed changes and determined that operation of WBN Unit 1 in accordance with the proposed changes will not endanger the health and safety of the public.In accordance with 10 CFR 50.91 (b)(1), TVA is sending a copy of this letter and attachments to the Tennessee Department of Environment and Conservation.

TVA requests approval of the proposed license amendment by March 9, 2012, with implementation within 60 days of issuance.There are no commitments associated with this submittal.

If you have any questions about this change, please contact Kara Stacy at (423) 751-3489.I declare under penalty of perjury that the foregoing is true and correct. Executed on this 9th day of March 2012.Resp c ully,

Enclosure:

Evaluation of Proposed Change Icc: See Page 3 U.S. Nuclear Regulatory Commission Page 3 March 8, 2012 cc (Enclosure):

NRC Regional Administrator

-Region II NRC Resident Inspector

-Watts Bar Nuclear Plant, Unit 1 NRC Resident Inspector

-Watts Bar Nuclear Plant, Unit 2 Director, Division of Radiological Health -Tennessee State Department of Environment and Conservation ENCLOSURE TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT I EVALUATION OF PROPOSED CHANGE

Subject:

Application to Change Dose Equivalent 1-131 Spike Limit and Allowable Value for Control Room Air Intake Radiation Monitors (WBN-TS-1 1-08)1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

2.1 Proposed

Changes 2.2 Need for Proposed Changes 3. TECHNICAL EVALUATION

3.2 Safety

Analysis 3.3 Conclusions

4. REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements

4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES ATTACHMENTS

1. Proposed TS Changes (Mark-Up) for WBN, Unit 1 2. Proposed TS Bases Changes (Mark-Up) for WBN, Unit 1 3. Proposed TS Changes (Final Typed) for WBN, Unit 1 4. Proposed TS Bases Changes (Final Typed) for WBN, Unit 1 E-1 of 13 1.0

SUMMARY

DESCRIPTION The Tennessee Valley Authority (TVA) is proposing to amend Watts Bar Nuclear Plant (WBN), Unit 1, Facility Operating License No. NPF-90. The proposed amendment will revise: 1. Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from< 9.45E-05 micro-Curie/cubic centimeter (pCi/cc) (3,308 counts per minute (cpm)) to< 1.647E-04 pCi/cc (3,308 cpm); and 2. TS 3.4.16, "RCS Specific Activity," by lowering the DOSE EQUIVALENT 1-131 (DEI-131) spike limit from 21 micro-Curie/gram (pCi/gm) to 14 pCi/gm in Required Action A.1 and Condition C.Mark-ups of the affected TS and TS Bases pages are included in Attachments 1 and 2.Final typed versions of the affected TS and TS Bases pages are included in Attachments 3 and 4. The TS Bases pages are provided to the NRC for information only.2.0 DETAILED DESCRIPTION

2.1 Proposed

Changes TS 3.3.7 Currently, TS Table 3.3.7-1, "CREVS Actuation Instrumentation," defines the Allowable Value for Function 2, "Control Room Radiation Control Room Air Intakes," as"< 9.45E-05 pCi/cc (3,308 cpm)." TS Table 3.3.7-1 would be revised to denote that the Allowable Value for Function 2,"Control Room Radiation Control Room Air Intakes," as "< 1.647E-04 pCi/cc (3,308 cpm)." TS 3.4.16 Currently, Required Action A. 1 of TS 3.4.16 states: "Verify DOSE EQUIVALENT 1-131 < 21 pCi/gm." Required Action A.1 of TS 3.4.16 would be revised to state: "Verify DOSE EQUIVALENT 1-131 < 14 pCi/gm." Currently, Condition C of TS 3.4.16 states: "...DOSE EQUIVALENT 1-131 > 21 pCi/gm." Condition C of TS 3.4.16 would be revised to state: "...DOSE EQUIVALENT 1-131 > 14 pCi/gm." E-2 of 13 Bases for Technical Specifications 3.3.7 and 3.4.16 Conforming changes to the Bases for Technical Specification 3.4.16 are made to address the revision to the DEl-1 31 spike limit. No changes to the Bases for Technical Specification 3.3.7 are required to reflect the change to the Allowable Value for the main control room air intake radiation monitors.2.2 Need for Proposed Changes 2.2.1 Change to TS 3.3.7 The change to the Allowable Value for the main control room air intake radiation monitors in TS Table 3.3.7-1 is due to correcting an error in the calculation of the Allowable Value.2.2.2 Change to TS 3.4.16 TVA is proposing to lower the DEI-131 spike value due to revisions to the WBN dose calculations for the MSLB and SGTR accidents.

As discussed in Section 3.1.2, an inconsistency was discovered regarding the conversion factors used to calculate DEl-131 in the analysis of record and the definition of DE1-131 in the WBN, Unit 1 Technical Specifications.

Changes to these calculations included redevelopment of the updated DE-1 31 conversion factors based on Regulatory Guide (RG) 1.109 (Reference

1) iodine inhalation dose conversion factors and recalculating the offsite and control room doses.The revised calculations resulted in the need to further limit the iodine concentration in the reactor coolant during temporary spikes resulting from reactor power transients in order to comply with regulatory limits for doses that control room occupants could receive from postulated accidents.

TVA selected 14 gCi/gm as the appropriate TS limit, because this value permits WBN to comply with all regulatory limits and it provides sufficient operational flexibility for anticipated transient-induced iodine spikes.3.0 TECHNICAL EVALUATION

3.1 BACKGROUND

3.1.1 Change

to TS 3.3.7 The actuation instrumentation for the Control Room Emergency Ventilation System (CREVS) consists of redundant radiation monitors.

A high radiation signal from any detector will initiate its associated train of the CREVS; thus, terminating the supply of unfiltered outside air to the control room and initiating filtration and emergency pressurization of the control room. These actions are necessary to ensure the control room is kept habitable for the operators stationed there during accident recovery and post-accident operations by minimizing the radiation exposure of control room personnel.

The change to the Allowable Value for the main control room air intake radiation monitors in TS Table 3.3.7-1 is due to correcting an error in the calculation of the Allowable Value. The Allowable Value in WBN, Unit 1 TS Table 3.3.7-1 for the control E-3 of 13 room air intake radiation monitors was changed from < 9.45E-05 pCi/cc (3,308 cpm) to<5 1.647E-04 pCi/cc (3,308 cpm).Section 3.2.1 provides the derivation for this value.3.1.2 Change to TS 3.4.16 WBN, Unit 1 TS 3.4.16 uses a composite iodine concentration (DEI-131) to represent the quantity and mixture of the isotopes actually present in the primary coolant. Dose conversion factors (DCFs) are applied to the concentrations of the separate isotopes to calculate DEl-1 31. DEI-131 is then used to calculate the accident source term, which is used to calculate the offsite and control room operator doses.On April 26, 2011, Problem Evaluation Report 360041 was initiated to document that the source for the DCFs for determining DEl-1 31 and the resulting doses for these accidents in the UFSAR was inconsistent with the source prescribed by the definition of DEI-131 in WBN, Unit 1 TS 1.1, "Definitions." The UFSAR methodology utilized DCFs from industry standard ICRP-2 (Reference 2), while the definition of DEl-1 31 in WBN, Unit 1 TS 1.1 requires the use of the thyroid DCFs listed in Table E-7 of Regulatory Guide (RG) 1.109, Rev. 1. Note: a different set of DCFs from those used to calculate DEl-1 31 is used to calculate the actual dose to the thyroid organ. The source for these DCFs continues to be from ICRP-30 (Reference 3).When the DCFs from RG 1.109 are substituted for those from ICRP-2, and the current TS 3.4.16 pre-accident iodine spike limit of 21 ý.Ci/gm for DEI-131 is assumed, the calculated thyroid dose to the control room occupants exceeds the limit of 30 rem. This limit is considered to be equivalent to the limit of 10 CFR 50, Appendix A General Design Criteria (GDC)-1 9, which limits dose to control room occupants to 5 rem whole body or its equivalent to any part of the body for the duration of the accident.DEl-1 31 data for Watts Bar, Unit 1 cycle 5 (from 3/22/2002) through cycle 11 (up to 2/20/12) was reviewed to determine the maximum DEl-1 31 values. This review includes data before and after Technical Specification Revision 52, Amendment 41, 12/20/2002.

The maximum DEI value during this period was 7.476E-1 p.Ci/gm. Thus, the plant never exceeded the new limit of 14 ptCi/gm.Currently, the pre-accident iodine spike limit for DEl-1 31 of 14 g+/-Ci/gm is being controlled administratively in accordance with NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety." In the event DEl-1 31 is found to be > 0.265 jiCi/gm, the RCS will be sampled once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure that the DE1-131 is < 14 jiCi/gm. If DEI-131 is determined to be outside this limit, DEI-131 in the RCS would be required to be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If DEI-131 cannot be restored within this time frame, the plant would be required to be placed in MODE 3 with Tavg < 500°F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The Bases for TS 3.4.16 were also revised to explain the control at this lower value.These controls will remain in place pending resolution of the issue by the NRC approval of a license amendment request to resolve the non-conservative technical specification.

E-4 of 13 The accident dose calculations were revised to use the thyroid DCFs from RG 1.109 Rev. 1, Table E-7. When the proposed pre-accident iodine spike limit of 14 giCi/gm is used in the calculations, the resulting doses to individuals at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and control room occupants are within the applicable regulatory requirements.

3.2 SAFETY

ANALYSES 3.2.1 Change to TS 3.3.7 The change to the Allowable Value for the main control room air intake radiation monitors in TS Table 3.3.7-1 is due to correcting an error in the calculation of the Allowable Value.The Allowable Value (cpm) is derived by the following equation: Allowable Value (cpm) = log-1 {(Allowable Value (Volts) / 1.667) + 11 The range of acceptable values for Allowable Value (Volts) was determined to be 4.441 V > Allowable Value (Volts) > 3.26 V. The present Allowable Value (Volts) of 4.2 V remained within the acceptable range. As a result, the Allowable Value (cpm)presented in TS Table 3.3.7-1 remained 3,308 cpm.The Allowable Value (pCi/cc) is derived from the following equation: Allowable Value (pCi/cc) = (Allowable Value (cpm)/sensitivity))

• 1.13 The sensitivity utilized to calculate the Allowable Value in pCi/cc was corrected from 3.5E+7 cpm/g.Ci/cc to 2.27E+7 cpm/ljCi/cc.

The original sensitivity value for Xenon 133 (3.5E+7 cpm/ýCi/cc) was based on a graphical extrapolation from a vendor test report.A more recent test report from the same vendor included a higher quality graph and a vendor established value for the same point of 2.27E+7 cpm/g.Ci/cc.

The change in the sensitivity resulted in the Allowable Value (pCi/cc) for the main control room air intake radiation monitors changing from < 9.45E-05 pCi/cc to s 1.647E-04 pCi/cc. This change is acceptable, because the margin to the Analytical Limit for the resultant setpoint remains adequate.

The margin to the Analytical Limit for the previously implemented setpoint was 2.48E-6 pCi/cc considering the uncertainty from a setpoint of 2.671 volts (1.29E-5 pCi/cc equivalent at a sensitivity of 3.5E+7 cpm/pCi/cc).

The margin to the Analytical Limit for the new setpoint is 2.54E-6 pCi/cc considering the uncertainty from a setpoint of 2.671 volts (1.99E-5 pCi/cc equivalent at sensitivity of 2.27E+7 cpm/pCi/cc).

E-5 of 13

3.2.2 Change

to TS 3.4.16 3.2.2.1 Introduction to Accident Analyses The accident analyses for a MSLB and a SGTR, including the assumed parameters, are presented in Sections 15.4.2 and 15.4.3 of the WBN UFSAR, respectively.

These UFSAR Sections are unaffected by the proposed changes. The radiological consequences of these accidents are presented in Sections 15.5.4 and 15.5.5 of the WBN, Unit 1 UFSAR, respectively.

Sections 15.5.4 and 15.5.5 of the WBN, Unit 1 UFSAR will be updated to reflect the results of the revised calculations following NRC approval of this license amendment request.Section 15.6.3 (Reference

4) of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," and Appendix A of Section 15.1.5 (Reference

5) of NUREG-0800 requires license applicants to perform calculations of doses resulting from a postulated SGTR accident and a postulated MSLB outside containment.

For each accident, two cases for iodine spiking effects are required to be analyzed: 1) A reactor transient has occurred prior to the accident initiation and has raised the primary coolant iodine concentration to the maximum value permitted by the Technical Specifications (i.e., pre-accident iodine spike case), and 2) The reactor trip or the primary system depressurization associated with the postulated accident causes the iodine release rate from the fuel to increase 500 times the rate corresponding to the maximum DEl-1 31 equilibrium iodine concentration allowed by Technical Specifications (concurrent iodine spike case).During the postulated MSLB accident (Reference 6), it is assumed the operator takes action to cool down and depressurize the plant and place the residual heat removal (RHR) system into service for further RCS heat removal within eight hours after the accident.

Once the RHR system is placed into service and the RCS has been depressurized, there are no more steam releases from the intact steam generators.

Therefore, MSLB accident LPZ doses are based on activity releases for the initial eight hours following the MSLB. The control room dose calculation extends beyond eight hours to 30 days, because the activity will continue to be in the control room atmosphere and additional time is required to either filter (via recirculation) or purge the activity from the control room atmosphere.

The operator actions and the thermal and hydraulic analysis previously performed to determine plant response for a postulated MSLB accident remain unaffected by this proposed change to the TSs.During the postulated SGTR accident (Reference 7), the operator is expected to readily determine that a SGTR has occurred, isolate the ruptured steam generator from the intact steam generators, isolate feedwater to the ruptured steam generator, cool down the RCS using the intact steam generators, depressurize the RCS to restore reactor coolant inventory, terminate safety injection to stop primary to secondary leakage, control charging flow, letdown, and pressurizer heaters to prevent repressurization of the RCS and reinitiation of leakage into the ruptured steam generator.

During the postulated SGTR accident, the limiting single failure is assumed to be the failure of the power operated relief valve (PORV) on the ruptured steam generator.

Failure of this valve in the open position will cause an uncontrolled depressurization of the faulted steam E-6 of 13 generator which will increase primary to secondary leakage and mass release to the atmosphere.

It is assumed the ruptured steam generator PORV fails open when the ruptured steam generator is isolated, and the valve is subsequently isolated by locally closing the associated block valve 11.0 minutes after it is assumed to fail open. The operator actions and the thermal and hydraulic analysis previously performed to determine plant response for a postulated SGTR accident remain unaffected by this proposed change to the TSs.3.2.2.2 Changes to Dose Conversion Factors and Pre-Accident Iodine Spike Value In the revised calculations, the assumed pre-accident Iodine spike value was lowered to 14 11Ci/gm from 21 gCi/gm. Reducing the value in Required Action A.1 and Condition C of TS 3.4.16 from < 21 p.Ci/gm to < 14 gCi/gm imposes a more restrictive condition on the plant. In the event DEI-131 is found to be > 0.265 gCi/gm, the RCS will be sampled once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure that the DEl-1 31 is < 14 g.Ci/gm. If DEI-131 is determined to be outside this limit, DEI-131 in the RCS would be required to be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If DEl-1 31 cannot be restored within this time frame, the plant would be required to be placed in MODE 3 with Tavg < 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.This DEI-131 value is used by computer code "STP" to calculate the released activities, which are then used as inputs to computer codes "COROD," which calculates doses to the control room occupants, and "FENCDOSE," which calculates doses to individuals at the EAB and LPZ.3.2.2.3 Changes to Atmospheric Dispersion Factors Section 2.3.4 of the WBN, Unit 1 UFSAR provides the current licensing basis regarding the derivation of the offsite atmospheric dispersion factors (X/Qs). The current offsite X/Qs are based on onsite meteorological data for the time period of 1974 through 1993 and Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," calculation methodology.

WBN, Unit 1 UFSAR Table 15A-2 provides the offsite X/Qs utilized in the MSLB and SGTR accident analyses.Table 15.5-14 of the WBN, Unit 1 UFSAR provides the current onsite X/Qs. They are based on onsite meteorological data for the time period of 1974 through 1993 and were determined utilizing the "ARCON96" computer code.New WBN, Unit 1 onsite and offsite X/Qs are utilized in the MSLB and SGTR analyses.The onsite and offsite X/Qs are calculated consistent with the current licensing basis methodology, except the meteorological data was updated to reflect a more recent 20-year time period (1991 through 2010). The onsite and offsite X/Qs utilized in the dose analyses of the MSLB and SGTR are presented in Table 1.E-7 of 13 Table 1 -Revised Offsite and Onsite Atmospheric Dispersion Factors Location XIQ (seclm 3)Low Population Zone 1.784E-4 0-2 hours 8.835E-5 2-8 hours 6.217E-5 8-24 hours 2.900E-5 1-4 days 9.811 E-6 4-30 days Exclusion Area Boundary 6.382E-4 0-2 hours Unit 1 Control Room 3.85E-3 0-2 hours 3.22E-3 2-8 hours 3.2.2.4 Results Calculations were performed to demonstrate that the resulting doses to offsite individuals and control room occupants do not exceed the regulatory criteria.Tables 2 and 3 provide the calculated doses for the accident initiated iodine spike case and the pre-accident iodine spike case for the MSLB and SGTR accidents, respectively.

In addition, the tables compare the resultant doses to the applicable regulatory criteria.Table 2 -MSLB Dose Calculation Result Control SRP Guidance 2-Hour 30-Day LPZ SRP Guidance Room for GDC 19 EAB (rem) (rem) for 10 CFR 100 (rem) Limits (rem) Limits (rem)Accident Initiated Iodine Spike Case (0.265 jiCi/gm steady state)Gamma 1.25E-02 5 1.04E-01 1.23E-01 2.5 Thyroid 1.73E+01 30 3.20E+00 4.59E+00 30 Pre-Accident Iodine Spike Case (14 tLCi/gm maximum peak)Gamma 7.07E-03 5 2.92E-02 1.16E-02 25 Thyroid 1.31E+01 30 2.63E+00 1.27E+00 300 E-8 of 13 Table 3 -SGTR Dose Calculation Results Control SRP Guidance 2-Hour 30-Day LPZ SRP Guidance Room (rem) for GDC 19 EAB (rem) (rem) for 10 CFR 100 Limits (rem) Limits (rem)Accident Initiated Iodine Spike Case (0.265 iXCilgm steady state)Gamma 8.11E-02 5 5.03E-01 1.47E-01 2.5 Thyroid 3.37E+00 30 6.33E+00 1.86E+00 30 Pre-Accident Iodine Spike Case (14 jiCi/gm maximum peak)Gamma 8.56E-02 5 3.50E-01 1.03E-01 25 Thyroid 2.18E+01 30 1.33E+01 3.81E+00 300 The calculated radiological consequences remain within the limits prescribed in 10 CFR 100 and GDC-1 9 and are consistent with the methodology and acceptance criteria of Section 15.6.3 of NUREG-0800 for the SGTR and Appendix A to Section 15.1.5 of NUREG-0800 for the MSLB. Following NRC approval of this license amendment request, the revised calculated dose consequences and atmospheric dispersion factors will be reflected in a future revision to the WBN, Unit 1 UFSAR, specifically Tables 15.5-14, 15.5-17, 15.5-19 and 15.A-2.

3.3 CONCLUSION

S The calculated radiological consequences remain within the limits prescribed in 10 CFR 100 and GDC-19, and are consistent with the methodology and acceptance criteria of Section 15.6.3 of NUREG-0800 for the SGTR and Appendix A to Section 15.1.5 of NUREG-0800 for the MSLB. Following NRC approval of this license amendment request, the revised calculated dose consequences and atmospheric dispersion factors will be reflected in a future revision to the WBN, Unit 1 UFSAR, specifically Tables 15.5-14, 15.5-17, 15.5-19 and 15.A-2.The change to the Allowable Value for the main control room air intake radiation monitors is acceptable, because the margin to the Analytical Limit is adequate.Therefore, the main control room radiation monitors will remain capable of performing their intended design function of isolating the main control room subsequent to an accident.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE

REGULATORY REQUIREMENTS/CRITERIA 10 CFR 50 Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 19,"Control room," limits radiological doses to control room occupants under accident conditions to 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

These limits are 5 rem whole body and 30 rem thyroid for the duration of the accident.10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," limits radiological doses to individuals from postulated accidental E-9 of 13 releases.

These limits are 25 rem to the whole body and 300 rem to the thyroid from iodine exposure for two hours at the EAB or for 30 days at the LPZ boundary.NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," prescribes the methodology and acceptance criteria for evaluating the dose consequences from postulated accidents.

For accident evaluation cases that assume an equilibrium iodine concentration for continued full power operation in combination with an assumed accident initiated iodine spike, Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure," of NUREG-0800 and Appendix A to Section 15.1.5, "Radiological Consequences of Main Steam Line Failures Outside of Containment of a PWR," of NUREG-0800 state: " For the pre-accident iodine spike case, the calculated doses should not exceed the 10 CFR 100.11 limits of 25 rem to the whole body and 300 rem to the thyroid from iodine exposure for two hours at the EAB or for 30 days at the LPZ boundary.* For the concurrent iodine spike case, the calculated doses should not exceed 10 percent of the 10 CFR 100.11 limits, or 2.5 rem to the whole body and 30 rem to the thyroid from iodine exposure for two hours at the EAB or for 30 days at the LPZ boundary.4.2 PRECEDENT TVA requested a license amendment on January 14, 2002 (Reference

8) to change WBN, Unit 1 Technical Specifications.

It requested that the value of DEl-1 31 be lowered from 60 p.Ci/gm to 21 gCi/gm DEl-1 31, and it modified the value for the Allowable Value for the main control room air intake radiation monitors.

The NRC approved the requested changes with the issuance of Amendment 41 to WBN, Unit 1 Operating License on November 18, 2002 (Reference 9).4.3 SIGNIFICANT HAZARDS CONSIDERATION TVA has concluded that the following proposed changes to the WBN, Unit 1 Technical Specifications do not involve a significant hazards consideration:

1) Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from < 9.45E-05 micro-Curie/cubic centimeter (pCi/cc) (3,308 counts per minute (cpm)) to < 1.647E-04 pCi/cc (3,308 cpm); and 2) TS 3.4.16 by lowering the DOSE EQUIVALENT 1-131 spike limit from 21 micro-Curie/gram (pCi/gm) to 14 pCi/gm in Required Action A.1 and Condition C.consideration.

TVA's conclusion is based on its evaluation in accordance with 10 CFR 50.91 (a)(1) of the three standards set forth in 10,CFR 50.92, "Issuance of amendment," as discussed below.E-10 of 13 Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated?

Response:

No.The proposed TS changes do not adversely affect any fission product barrier nor do they alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation.

They do not change any credited operator actions nor do they physically change any plant system, structure, or component.

The amount of iodine in the primary coolant and the Allowable Value for the main control room radiation monitors do not affect the initiation of any accident or transient.

Therefore, the proposed amendment does not result in a significant increase in the probability of an accident previously evaluated.

The changes do not adversely affect the protective and mitigative capabilities of the plant. The SSCs will continue to perform their intended safety functions.

The proposed reduction in the amount of DOSE EQUIVALENT 1-131 (DEI-131) in the reactor coolant following a load transient has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.The calculated radiological doses remain within the limits prescribed in 10 CFR 100 and GDC-19 and are consistent with the methodology and acceptance criteria of Section 15.6.3 of NUREG-0800 and Appendix A of Section 15.1.5 of NUREG-0800.

The change to the Allowable Value for the main control room radiation monitors continues to ensure that the monitors are capable of performing their intended design function of isolating the main control room subsequent to an accident.Therefore, the proposed amendment does not involve a significant increase in the 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 TS changes do not alter the configuration of the plant nor do they directly affect plant operation.

The proposed TS changes do not result in the installation of any new equipment or system or the modification of any existing equipment or systems. No new operation procedures, conditions, or modes are created. As a result, the proposed TS changes do not introduce any new failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing basis. There will be no adverse effects or challenges imposed on any safety-related system as a result of these changes. Therefore, the proposed amendment does not create the possibility of a new or different kind of accident.from any previously evaluated.

E-11 of 13

3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:

No.The calculated radiological doses remain within the limits prescribed in 10 CFR 100 and GDC-19, and are consistent with the methodology and acceptance criteria of Section 15.6.3 of NUREG-0800 and Appendix A of Section 15.1.5 of NUREG-0800.

The Allowable Value for the main control room radiation monitor continues to ensure that the monitors are capable of performing their intended design function of isolating the main control room subsequent to an accident.Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.Based on the above, TVA concludes that the proposed changes to the WBN Unit 1 TSs do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 CONCLUSION

S 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 TVA has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the proposed amendment.

6.0 REFERENCES

1. NRC Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Rev. 1, 1977.2. ICRP-2, "Permissible dose for internal radiation," 1959.3. ICRP-30, "Limits for Intakes of Radionuclides by Workers," 1972.E-12 of 13

4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.6.3, Radiological Consequences of Steam Generator Tube Failure.5. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.1.5, Appendix A, Radiological Consequences of Main Steam Line Failure Outside Containment of a PWR.6. WBN Updated Final Safety Analysis Report, Section 15.4.2, "Major Rupture of a Main Steam Line." 7. WBN Updated Final Safety Analysis Report, Section 15.4.3, "Steam Generator Tube Rupture." 8. TVA letter to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 -Technical Specification (TS) Change No. WBN-TS-01-12

-Reactor Coolant System (RCS)Specific Activity, dated January, 14, 2002.9. NRC letter to TVA, Watts Bar Nuclear Plant, Unit 1 -Issuance of Amendment Regarding Reactor Coolant System Specific Activity (TAC No. MB3831).E-13 of 13 ATTACHMENT I PROPOSED TS CHANGES (MARK-UPS)

FOR WBN, UNIT 1 CREVS Actuation Instrumentation

3.3.7 Table

3.3.7-1 (page 1 of 1)CREVS Actuation Instrumentation FUNCTION REQUIRED SURVEILLANCE ALLOWABLE VALUE CHANNELS REQUIREMENTS

1. Manual Initiation 2 trains SR 3.3.7.3 NA 2. Control Room Radiation 2 SR 3.3.7.1 <9A545-l.647E-04 AC/cc Control Room Air Intakes SR 3.3.7.2 (3,308 cpm)SR 3.3.7.4 3. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

Watts Bar-Unit 1 3.3-60 Amendment 41 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:

The specific activity of the reactor coolant shall be within limits.MODES I and 2, MODE 3 with RCS average temperature (Tavg) -> 500 0 F.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131 ------------------

NOTE -----------

> 0.265 pCi/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 1-131 <:2kCi/gm AND \-A.2 Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 1-131 to within limit.B. Gross specific activity of the B.1 Perform SR 3.4.16.2.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reactor coolant not within limit.AND B.2 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tavg < 500 0 F.(continued)

Wafts Bar-Unit 1 3.4-39 Amendment 41,55 RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time Tavg < 500 0 F.of Condition A not met.OR DOSE EQUIVALENT 1-131 >24- p.Ci/gm.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific 7 days activity < 100/ E jiCi/gm.SR 3.4.16.2 ----------------------

NOTE ------------------

Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 specific 14 days activity < 0.265 jiCi/gm.AND 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 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 (continued)

Watts Bar-Unit 1 3 3.4-40 Amendment 41 ATTACHMENT 2 PROPOSED TS BASES CHANGES (MARK-UPS)

FOR WBN, UNIT I RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

The analysis for the SGTR and MSLB 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 analyses are for two cases of reactor coolant specific activity.

One case assumes specific activity at 0.265 jiCi/gm DOSE EQUIVALENT 1-131 with an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant by a factor of 500 times the iodine production rate necessary 14 to maintain a steady state iodine concentration of 0.265 jiCi/gm DOSE_QUIVALENT 1-131. The second case assumes the initial reactor coolant iodine act-ivi Ci/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 coolant equals the LCO limit of 100/ E pCi/gm for gross specific activity.Rgarding tne Gpecitic actIvity valuecF forGE U 1U11IVAI NIT 1121 de"1 ned above, the F=unctional Evalu _ E lation for Problem Eivaluationc Repeot 3600411 (Ref. 3)the corroc~t limit Or, 14 p~i/gmr. Theo 41 P~iigA limit ic boinRg adminictrativoly Gt-reG"d through Step 5 Ad 65 Of Sec;tion 3.0, "Operator Action+," of A^l 28,"High in Reac_"tor Coolan.t," and th, nGo@ prec.ding Step 5 until NRC's djppF-eva ul a LIUlU:lul[~LRequesut updating I ucnnicai DP8G1+1GAH0R 3.4.16 GaR 138 9-MaRORA-d.

The analysis also assumes a loss of offsite power at the same time as the SGTR and MSLB event. The SGTR causes a reduction in reactor coolant inventory.

The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature AT signal. The MSLB results in a reactor trip due to low steam pressure.The coincident loss of offsite power causes the steam dump valves to close to protect the condenser.

The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.The safety analysis shows the radiological consequences of a SGTR and MSLB accident are within the appropriate 10 CFR 100 and 10 CFR 50, Appendix A, GDC 19 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed[tCi/gm DOSE EQUIVALENT 1-131, in the applicable specification, for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The safety analysis has concurrent and pre-accident iodine spiking levels up t iCi/gm DOSE EQUIVALENT 1-131.M.14 Regard the Gpecific activity va'-e for DOSE EQUI ALENT I d. A 4 2E)2A 1 deuR,.Rt-that tho pro accidont iodino spike of 214 ,,t!i/g@m i inorroct and that the corrct lim;it i; 14 Pi;gm. The 414 pi/gm limit is being dm;gnirtvy G3014~el4018 hFeh .tep 5 and 6 of Sec-tion 3.,prtrAtOn"of AGl 28,"High At Viity in Rcactor Ceolant," aRd tho nets prcodi~ngStep 5 until NRC's approval of a LicbtRo Amondmont Roquest updating Technical Specifleation 3.1.16 Can bor heobtainod.(continued)

Watts Bar-Unit 1 B 3.4-94 Revision 52, 111 Amendment 41 RCS Specific Activity B 3.4.16 BASES (continued)

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at 14 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 limit of 2 gm is not exceeded.

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 done to continue to provide a trend.R9ga.ding th. pcific actiVitY value for DOrE EQUI 'VAL'ENT 11-31 defined abovo, the Func-tionlEvAluio for Problemn Evalu ation Repeot 3600114 (Ref. 3)do...ment.

that the pro a... do. ' odno cpk of 21 i.. Onrrect and that the GGrroct limit is 14 p~iigm. The 14 p~ilgm limit is boeing adminicitrativeY8I ee4t-eFIed through Stop 5 and 6 of Section 3.0, "Operator A.tienc," of 28,"High ,Aetivuty

'in CoolaRt," and the notc prceoding Step 5 until NRG', approval of a Liconco. Amcndmcnt Rc.ucct -1'datinc Technical Spocification The DOSE EQUIVALENT 1-131 must be restored to within limits 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 required, 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 applicable MODE(S) while relying on the ACTIONS. 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 plant remains at, or proceeds to power operation.

B.1 and B.2 With the gross specific activity in excess of the allowed limit, an analysis must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to determine DOSE EQUIVALENT 1-131. 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.The change within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3 and RCS average temperature

< 500OF lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.(continued)

Watts Bar-Unit 1 B 3.4-96 Revision 52, 68, 111 Amendment 41, 55 RCS Specific Activity B 3.4.16 BASES ACTIONS (continued) cA1 If a Required Action and the associated Completion Tim f Condition A is not met or if the DOSE EQUIVALENT 1-131 is greater than 24 pCi/gm, the reactor must be brought to MODE 3 with RCS average temperature

< 500OF within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.Rogarding the Gpeeifie aetivty- 'akie for DOSE EQUI.ALE,,NT 11 31 dofino above, the Func-tfion.al Ev.alua'tion for Problom Evaluation Report 360041 (Ref. 3)docuwmonts that the prc accidont iodina spoke of 21 t*i~i~gm it inorrect and that the correc limti 14 p~ilgm. The 14 p~i/gm limit i6 bofing adminictrativoly GGR4Felled through Stop 5 and 6 Of SectionR 3.0, "Operator Actiont,," of A0l 28,"High i

Colant," and the noto preceding Step 5 until N.RC'c approval e a 'of a Lic Amendmen-t Request updating TOchnical Sp-,ecificion SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once every 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, 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 gross 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 1 and 2, and in MODE 3 with Tavg at least 500 0 F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following rapid power changes when fuel failure is more apt 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 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 fuel failure; samples at other times would provide inaccurate results.(continued)

Watts Bar-Unit 1 B 3.4-97 Revision 52, 111 Amendment 41 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.16.3 A radiochemical analysis for E determination is required every 184 days (6 months) with the plant operating in MODE 1 equilibrium conditions.

The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Frequency of 184 days recognizes E does not change rapidly.This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES

1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance," 1973.2. Watts Bar FSAR, Section 15.4, "Condition IV -Limiting Faults.".Functional Evwluation for Problom Ewauation Repo.t 369001 (T-35 110501 809)Revision 111 Watts Bar-Unit 1 B 3.4-98 ATTACHMENT 3 PROPOSED TS CHANGES (FINAL TYPED) FOR WBN, UNIT 1 CREVS Actuation Instrumentation

3.3.7 Table

3.3.7-1 (page 1 of 1)CREVS Actuation Instrumentation FUNCTION REQUIRED SURVEILLANCE ALLOWABLE VALUE CHANNELS REQUIREMENTS

1. Manual Initiation 2 trains SR 3.3.7.3 NA 2. Control Room Radiation 2 SR 3.3.7.1 <1.647E-04 pC/cc Control Room Air Intakes SR 3.3.7.2 (3,308 cpm)SR 3.3.7.4 3. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

Watts Bar-Unit 1 3.3-60 Amendment 41, RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:

The specific activity of the reactor coolant shall be within limits.MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) > 500-F.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131 -----------------

NOTE -----------

> 0.265 pICi/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT 1-131 Once per 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s< 14 [tCi/gm.AND A.2 Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 1-131 to within limit.B. Gross specific activity of the B.1 Perform SR 3.4.16.2.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reactor coolant not within limit.AND B.2 Be in MODE 3 with Tavg < 500 0 F. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)

Wafts Bar-Unit 1 3.4-1 Amendment 41, 55, RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time Tavg < 500 0 F.of Condition A not met.OR DOSE EQUIVALENT 1-131 >14 ptCi/gm.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific 7 days activity < 100/ _E l.Ci/gm.SR 3.4.16.2 ----------------------

NOTE ------------------

Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 specific 14 days activity < 0.265 pCi/gm.AND 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 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 (continued)

Watts Bar-Unit 1 3.4-2 Amendment 41, ATTACHMENT 4 PROPOSED TS BASES CHANGES (FINAL TYPED) FOR WBN, UNIT 1 RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

The analysis for the SGTR and MSLB 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 analyses are for two cases of reactor coolant specific activity.

One case assumes specific activity at 0.265 p.Ci/gm DOSE EQUIVALENT 1-131 with an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant by a factor of 500 times the iodine production rate necessary to maintain a steady state iodine concentration of 0.265 pCi/gm DOSE EQUIVALENT 1-131. The second case assumes the initial reactor coolant iodine activity at 14 pCi/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 coolant equals the LCO limit of 100/ E pICi/gm for gross specific activity.The analysis also assumes a loss of offsite power at the same time as the SGTR and MSLB event. The SGTR causes a reduction in reactor coolant inventory.

The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature AT signal. The MSLB results in a reactor trip due to low steam pressure.The coincident loss of offsite power causes the steam dump valves to close to protect the condenser.

The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.The safety analysis shows the radiological consequences of a SGTR and MSLB accident are within the appropriate 10 CFR 100 and 10 CFR 50, Appendix A, GDC 19 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 14 pCi/gm DOSE EQUIVALENT 1-131, in the applicable specification, for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The safety analysis has concurrent and pre-accident iodine spiking levels up to 14 pCi/gm DOSE EQUIVALENT 1-131.(continued)

Watts Bar-Unit 1 B 3.4-94 Revision 52, 111 Amendment 41, RCS Specific Activity B 3.4.16 BASES (continued)

ACTIONS A.1 and A.2 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 limit of 14 jICi/gm is not exceeded.

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 done to continue to provide a trend.The DOSE EQUIVALENT 1-131 must be restored to within limits 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 required, 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 applicable MODE(S) while relying on the ACTIONS. 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 plant remains at, or proceeds to power operation.

B.1 and B.2 With the gross specific activity in excess of the allowed limit, an analysis must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to determine DOSE EQUIVALENT 1-131. 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.The change within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3 and RCS average temperature

< 500OF lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.(continued)

Watts Bar-Unit 1 B 3.4-96 Revision 52, 68, 111 Amendment 41, 55, RCS Specific Activity B 3.4.16 BASES ACTIONS C.1 (continued)

If a Required Action and the associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is greater than 14 1 iCi/gm, the reactor must be brought to MODE 3 with RCS average temperature

< 500°F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once every 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, 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 gross 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 1 and 2, and in MODE 3 with Tavg at least 500 0 F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following rapid power changes when fuel failure is more apt 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 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 fuel failure; samples at other times would provide inaccurate results.(continued)

Watts Bar-Unit 1 B 3.4-97 Revision 52, 111 Amendment 41, RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.16.3 A radiochemical analysis for E determination is required every 184 days (6 months) with the plant operating in MODE 1 equilibrium conditions.

The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for t is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Frequency of 184 days recognizes E does not change rapidly.This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES

1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance," 1973.2. Watts Bar FSAR, Section 15.4, "Condition IV -Limiting Faults." Watts Bar-Unit 1 B 3.4-98 Revision 111 Amendment