ML022200382
| ML022200382 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/31/2002 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TVA-BFN-TS-405 | |
| Download: ML022200382 (173) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 July 31, 2002 TVA-BFN-TS-405 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C. 20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-259 Tennessee Valley Authority
)
50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 - LICENSE AMENDMENT - ALTERNATIVE SOURCE TERM In accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90, TVA is submitting a request for an amendment to licenses DPR-33, DPR-52 and DPR-68 that supports a full scope application of an Alternative Source Term (AST) methodology for BFN Units 1, 2, and 3. Specifically, TVA requests revision to the licensing and design basis to reflect the application of AST methodology on Units 1, 2, and 3 and approval of associated Technical Specifications (TSs) changes which are justified by the AST analyses. TVA is also proposing deletion of a completed License Condition to licenses DPR-52 and DPR-68.
On December 23, 1999, the NRC published 10 CFR 50.67, "Accident Source Term." This regulation provides a mechanism for licensed power reactors to replace the traditional source term used in design basis accident analyses with an AST. 10 CFR 50.67 requires licensees who seek to revise their current accident source term in design basis radiological consequence analyses to apply for a license amendment under 10 CFR 50.90.
Full Scope AST analyses were performed following the guidance in Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and Standard Review Plan Section 15.0.1, Printed on recycled paper
U.S. Nuclear Regulatory Commission Page 2 July 31, 2002 "Radiological Consequences Analyses using Alternative Source Terms." AST analyses were performed for the four Updated Final Safety Analysis Report (UFSAR) Chapter 14 BFN Design Basis Accidents (DBA) that could potentially result in control room and offsite doses. These include the Loss of Coolant Accident (LOCA), the Main Steam Line Break Accident, the Refueling Accident, and the Control Rod Drop Accident. The analyses demonstrated that using AST methodologies, post-accident control room and offsite doses remain within the regulatory limits.
TVA proposes implementation of this change through both revisions to the TS and UFSAR. Proposed changes in the licensing basis for BFN resulting from AST application include the following:
TS and UFSAR changes that reflect revised design requirements regarding the use of the Standby Liquid Control (SLC) System to buffer the suppression pool preventing iodine re-evolution following a postulated design basis LOCA.
TS revisions to reflect the relaxation of Secondary Containment, Standby Gas Treatment, and Control Room Emergency Ventilation System requirements. AST analyses do not take credit for secondary containment during the movement of irradiated fuel and during core alterations.
Therefore, these system TS may be made less restrictive.
" TS revisions to remove the requirements to test the charcoal filters for Standby Gas Treatment and Control Room Emergency Ventilation Systems.
AST analyses does not take credit for adsorption of elemental iodine, organic iodine, or noble gases by the charcoal. Therefore, the charcoal filters are no longer required and the associated TS may be deleted.
Additionally, the testing requirements are being revised to add limits for pressure drop without charcoal adsorbers.
TVA is also requesting deletion of Facility Operating License Condition 2.C.(4) for Units 2 and 3. The license condition required that TVA perform analyses of the design bases LOCA, confirm compliance with off-site and on-site dose limits, obtain NRC approval of the results, and make any needed modifications. These actions are complete and, therefore, this License Condition is no longer applicable.
Since the three units share a common refueling floor, the completed AST radiological dose analysis for the refueling accident is valid for all of the units.
Unit 1 is currently shutdown, defueled, and in long term layup. The AST analyses
U.S. Nuclear Regulatory Commission Page 3 July 31, 2002 for the remaining three DBAs have been performed for Units 2 and 3, but not for Unit 1. As required by existing Unit 1 License Condition 2.C.(4), TVA will verify that the required AST analyses needed for the remaining DBAs for Unit 1 are complete, and submit them for NRC review and approval prior to Unit 1 restart. Because the three units are essentially identical, TVA expects that the Unit 1 analyses will show comparable results as Units 2 and 3. Therefore, TVA is requesting this amendment and TS change be approved for Unit 1.
In support of a project to uprate the licensed thermal power of BFN Units 2 and 3, TVA determined that it was appropriate to adopt AST. This decision was communicated to the NRC staff in a meeting in Rockville, Maryland on December 5, 2001. Additional meetings were held on January 16, 2002, and July 10, 2002, between TVA and the staff to discuss the specifics of TVA's planned AST submittal, including the incorporation of Unit 1 TS changes. In those meetings, the analysis approach, submittal content, and schedule were discussed.
The current operating license allows Units 2 and 3 to operate at a maximum power level of 3458 megawatts thermal (MWt). TVA is currently engaged in an Extended Power Uprate (EPU) project to increase the maximum licensed thermal power for Units 2 and 3 to 3952 MWt. Therefore, the AST analyses which have been performed considered the core isotopic values for the current and future vendor products at EPU conditions and this license amendment is based on a bounding core isotopic inventory.
The use of AST changes the analytical treatment of the DBA radiological consequences. The use of AST has no direct impact on the probability of the evaluated DBAs. The changes in implementing AST methodology and the other changes requested by this license amendment do not increase the core damage frequency or the large early release frequency. Therefore, this TS change request is not being submitted as a "risk-informed licensing action" as defined by Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific changes to the Licensing Basis."
Several other Boiling Water Reactors (Duane Arnold, Brunswick, Grand Gulf, Hope Creek, Clinton, and Perry) have previously provided justification for the use of AST utilizing a similar approach. These applications have been approved by NRC.
TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). The BFN Plant Operations Review Committee and the Nuclear Safety Review Board have
U.S. Nuclear Regulatory Commission Page 4 July 31, 2002 reviewed this proposed change and determined that operation of BFN Units 1, 2, and 3 in accordance with the proposed change will not endanger the health and safety of the public. Additionally, in accordance with 10 CFR 50.91 (b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health. to this letter provides the description and justification of the proposed change. This includes TVA's determination that the proposed change does not involve a significant hazards consideration and is exempt from environmental review. Enclosure 2 contains copies of the appropriate marked-up TS pages from Units 1, 2, and 3 to show the proposed TS changes. UFSAR Section 3.8 is being revised to describe the new safety function of the SLC System Enclosure 3 provides marked up to indicate the proposed license change. Enclosure 4 provides the BFN Alternative Source Term Safety Assessment.
RG 1.183 recommends that changes to the UFSAR that reflect the revised analyses be submitted to the staff. Enclosure 5 provides proposed changes to UFSAR Section 14.6 that have been identified as requiring revision to reflect the AST analyses. Enclosure 5 also provides a matrix identifying other sections in the UFSAR that are currently under evaluation for change. The final UFSAR changes will be completed as required by BFN procedures following approval of this amendment request.
TVA requests the approval of the proposed license amendment for Units 1, 2, and 3 by April of 2003 and requests that the revised TS be made effective within 60 days of NRC approval. There are no new regulatory commitments associated with this submittal. If you have any questions about this change, please contact me at (256) 729-2636.
U.S. Nuclear Regulatory Commission Page 5 July 31, 2002 Pursuant to 28 U.S.C. § 1746 (1994), I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 31 st day of July, 2002.
Si cerely, T. E. Abney Manager o "nsing and I stry Affairs Encl ures cc: Se e6
U.S. Nuclear Regulatory Commission Page 6 July 31, 2002 Enclosures cc (Enclosures):
State Health Officer Alabama State Department of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 (Via NRC Electronic Distribution):
Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Mr. Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North (MS 08G9) 11555 Rockville Pike Rockville, Maryland 20852 NRC Resident Inspector Browns Ferry Nuclear Plant P. 0. Box 149 Athens, Alabama 35611
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM INDEX OF ENCLOSURES Enclosure I - DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGES
- 1.
Description of the Proposed Changes El -1 Technical Specifications Changes El -2 UFSAR Changes El -3 Administrative Change Deletion of License Condition El -3 II.
Reason For the Proposed Change El -3 III.
Background
E1-3 Control Room Habitability Discussion E1-4 IV.
Safety Evaluation E 1-6 AST Methodology E 1-6 Evaluation El-7 Results E1-7 V.
No Significant Hazards Consideration El-14 VI.
Environmental Impact Consideration El-15 - MARKED PAGES - Technical Specifications - Marked pages - UFSAR - Alternative Source Term Safety Assessment - UFSAR Chapter 14.6 Mark-Ups
ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE DESCRIPTION OF THE PROPOSED CHANGE In accordance with the provisions of 10 CFR 50.4, 10 CFR 50.67, and 10 CFR 50.90, TVA is requesting an amendment to licenses DPR-33, DPR-52, and DPR-68 to implement an Alternative Source Term (AST) for BFN Units 1, 2, and 3. This change request includes revisions to the licensing and design basis to reflect the full application of AST methodology and changes to the Technical Specifications (TS) justified by the AST analyses.
This full implementation of AST analyses will modify the licensing bases by adopting AST methodology which replaces the current accident source term with an alternative source term as prescribed in 10 CFR 50.67 and establishes the 10 CFR 50.67 total effective dose equivalent (TEDE) dose limits as a new acceptance criteria.
Since all three units share a common refueling floor the completed AST radiological dose analysis for the refueling accident is valid for all three Units.
Unit I is currently shutdown, defueled, and in long term layup. The AST analyses for the remaining Design Bases Accidents have been performed for Units 2 and 3, but not for Unit 1 As required by Unit 1 Facility License 2.C.(4),
TVA will verify that the required analyses needed for the remaining DBAs for Unit 1 are complete, and submit them for NRC review and approval prior to Unit 1 restart. Because the three units are essentially identical, it is expected that the Unit 1 analyses will show comparable results to Units 2 and 3.
Therefore, TVA is requesting this amendment and TS change be approved for Unit 1.
The current operating license allows Units 2 and 3 to operate at a maximum power level of 3458 megawatts thermal (MWt). TVA is currently engaged in an Extended Power Uprate (EPU) project to increase the maximum licensed thermal power for Units 2 and 3 to 3952 MWt. Therefore, the AST analysis which have been performed considered the core isotopic values at EPU conditions and this license amendment is based on a bounding core isotopic inventory.
This enclosure provides the description and justification of the proposed changes. This includes TVA's determination that the proposed change does not El -1
involve a significant hazards consideration and is exempt from environmental review. Enclosure 2 contains copies of the appropriate marked-up pages from Units 1, 2, and 3 TS which show the proposed changes. UFSAR Section 3.8 is being revised to describe the new safety function of the SLC System. This UFSAR section is included in Enclosure 3 marked up to indicate the proposed licensing bases changes. Enclosure 4 provides the BFN AST Safety Assessment. This enclosure provides a summary description and basis for the acceptability of the proposed changes associated with the AST methodology.
RG 1.183 recommends that changes to the UFSAR that reflect the revised analyses be submitted to the staff. Enclosure 5 provides changes to UFSAR Section 14.6 that have been identified requiring revision to reflect the AST analyses. A matrix identifying other sections in the UFSAR that are currently under evaluation for change is also provided in Enclosure 5. The final UFSAR changes will be completed as required by BFN procedures following approval of this amendment request.
The license amendment revises BFN Units 1, 2, and 3 TS and the UFSAR to implement the AST analysis. The revisions are as follows:
Technical Specification Changes "TS 3.1.7, Standby Liquid Control (SLC) System, is being changed to revise the required amount of sodium pentaborate from t3007 gallons to 04000 gallons. Additionally, a new surveillance requirement to verify that the sodium pentaborate concentration is 08.0% by weight is being added. SLC system operability will also be required in Mode 3. These changes implement AST methodology regarding the use of SLC to buffer the suppression pool following a Loss of Coolant Accident involving fuel damage.
"* TS Table 3.3.6.2-1, Secondary Containment Isolation Instrumentation, is being revised to delete the requirement for operable secondary containment instrumentation during core alterations and movement of irradiated fuel assemblies in the secondary containment. The AST analyses dose not take credit for the secondary containment function. Removal of this requirement is further justified by the AST analyses.
"* TS Table 3.3.7.1-1, Control Room Emergency Ventilation (CREV) System Instrumentation, is being revised to delete the requirement for operable CREV instrumentation during core alterations and movement of irradiated fuel assemblies in the secondary containment. The AST analyses does not take credit for automatic CREV initiation during core alterations. Removal of these requirements is further justified by the AST analyses.
"* TS 3.6.4.1, Secondary Containment, TS 3.6.4.2, Secondary Containment Isolation Valves (SCIVs), TS 3.6.4.3, Standby Gas Treatment (SGT) System and TS 3.7.3, CREV System is being revised to delete the requirement for E1-2
operability during core alterations and movement of irradiated fuel assemblies in the secondary containment. The AST analyses does not take credit for these functions. Removal of these requirements is further justified by the AST analyses.
TS 5.5.7, Ventilation Filter Testing Program (VFTP), is being revised to delete sections b and c which require testing of charcoal adsorbers in the SGT and CREV Systems, respectively. Since AST analyses takes no credit for charcoal filters, the testing requirements are being removed from the TS.
Additionally, TS 5.5.7 Section d is being revised to add limits for pressure drop testing without charcoal adsorbers (and after-filters) installed. BFN has no specific plans for physical removal of these adsorbers; however, removal would not require further license amendment.
UFSAR Changes UFSAR Section 3.8, Standby Liquid Control System, is being revised to describe the new safety function of maintaining the suppression pool water pH at or above 7.0 to prevent iodine re-evolution following a LOCA that involves fuel damage.
Administrative Change Deletion of License Condition Facility Operating License Condition 2.C.(4) for Unit 2 and Unit 3 is being deleted. This license condition required TVA to perform analyses of the design basis LOCA, confirm compliance with off-site and on-site dose limits, obtain NRC approval, and make any needed modifications. Since these requirements have been completed, this license condition is no longer applicable.
REASON FOR THE PROPOSED CHANGE Approval of this change will provide a more realistic source term for BFN that will result in a more accurate assessment of DBA radiological doses. This allows relaxation of some current licensing basis requirements. Adopting the AST may also support future evaluations and license amendments.
Ill.
BACKGROUND On December 23, 1999, the NRC published 10 CFR 50.67, "Accident Source Term." This regulation provides a mechanism for licensed power reactors to replace the current accident source term used in DBA analyses with an alternative source term. The direction provided in 10 CFR 50.67 is that licensees who seek to revise their current accident source term in design basis radiological consequences analyses apply for a license amendment under 10 CFR 50.90.
In July 2000, NRC published Regulatory Guide 1.183, "Alternative Source Terms For Evaluating Design Basis Accidents at Nuclear Power Reactors." Regulatory E1-3
Guide (RG) 1.183 provides guidance to licensees on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. Since then, several BWRs (Duane Arnold, Brunswick, Grand Gulf, Hope Creek, Clinton, and Perry) have submitted license amendments to adopt AST. These amendments have been approved by NRC.
TVA reviewed these submittals, including the associated NRC requests for additional information and Safety Evaluations for inclusion into this submittal.
AST analyses for DBAs were performed following the guidance in RG 1.183 and Standard Review Plan (SRP) 15.0.1, "Radiological Consequences Analyses Using Alternative Source Term." Acceptance criteria consistent with that required by 10 CFR 50.67 were used to replace the current design basis source term acceptance criteria. The AST analyses were performed for four BFN DBAs that could potentially result in control room and offsite doses. These include the LOCA, the Main Steam Line Break Accident, the Refueling Accident, and the Control Rod Drop Accident.
Browns Ferry is a three unit site. Units 2 and 3 are in operation, each having a licensed thermal power of 3458 MW. Unit I has a licensed thermal power of 3293 MW. Unit 1 is shutdown, defueled and in long-term layup. Activities are currently underway for restart of Unit 1 within 5 years. Each of these units is a General Electric BWR-4 boiling water reactor with a Mark I containment design.
The three units share a common refueling floor, and the three control rooms are all located in a single habitability zone. A 600 foot tall offgas stack serves all three units. Browns Ferry Units 2 and 3 have previously implemented modifications that make the main steam lines seismically rugged. This established an alternative leakage treatment leakage path using the main steam system piping and the main condenser for post accident dose mitigation for main steam isolation valve leakage.
In support of a project to uprate the licensed thermal power of BFN Units 2 and 3, TVA determined that it was appropriate to adopt AST. This decision was communicated to the NRC staff in a meeting in Rockville, Maryland on December 5, 2001. Additional meetings were held on January 16, 2002, and July 10, 2002, between TVA and the staff to discuss the specifics of TVA's planned AST submittal, including the incorporation of Unit 1 TS changes. In those meetings, the analysis approach, submittal content, and schedule were discussed.
The following provides a background discussion on control room habitability as requested by the NRC during the January 2002 meeting.
Control Room Habitability Discussion In a July 31, 1992 letter (Reference 1), TVA described corrective actions to resolve self-identified deficiencies in the design of the CREV System. These El -4
corrective actions were related to on the discovery that there was substantial unfiltered inleakage into the control room.
In October of 1997, TVA requested a license amendment to allow BFN to operate Units 2 and 3 at an uprated power level of 3458 megawatts thermal.
In reviewing the license amendment for power uprate, NRC requested additional information regarding TVA's unfiltered inleakage into the control room. In a May 7, 1998 letter (Reference 2), NRC requested that TVA include the effects of MSIV leakage to the turbine building with regard to control room dose, exclusion area boundary (EAB) dose and low population zone (LPZ) dose. In addition, NRC requested an assessment of control room dose, EAB dose, and LPZ dose due to leakage from Emergency Core Cooling Systems (ECCS) consistent with NRC SRP 15.6.5, Appendix B.
In a September 8,1998 letter (Reference 3), NRC issued a license amendment to allow operation of BFN Units 2 and 3 at 3458 megawatts thermal power. As part of the amendment, TVA concurred and NRC added a license condition that required performance of an analysis of the DBA LOCA to confirm compliance with General Design Criteria (GDC) 19 and offsite limits considering MSIV leakage and ECCS leakage and submit the results by March 31, 1999. The results of this analysis were transmitted to the NRC in a letter dated March 30, 1999 (Reference 4). This letter stated that the calculated doses were bounded by the allowable doses prescribed by 10 CFR 50 Appendix A, GDC 19 and 10 CFR 100 with the unfiltered control room inleakage.
On August 3, 1999 (Reference 5), NRC provided a Safety Evaluation (SE) acknowledging the revised dose calculation to be the analyses of record for the radiological consequence for a Design Bases Accident (DBA) LOCA.
TVA has used this NRC approved dose analysis, including the unfiltered control room inleakage, to support another license amendment. By application dated September 28, 1999 (Reference 6), supplemented February 4, 2000 (Reference 7), TVA requested a revision to the Units 2 and 3 TS to increase the allowable leakage for the main steam line isolation valves. By letter dated March 14, 2000 (Reference 8), NRC approved these TS amendments.
In the March 14, 2000, SE the staff concluded that there was reasonable assurance that the BFN control room will be habitable during a postulated DBA. This is based on (1) the relative magnitude of the infiltration currently assumed in the BFN analysis (3717 cfm of which is unfiltered), (2) the site X/Q values, (3) actions previously taken by TVA, and (4) the low probability of a design basis event occurring that could result in radioactivity releases sufficient to challenge the ability of control room personnel to protect the health and safety of the public.
E1-5
In summary, TVA conducted tests and determined the unfiltered inleakage into the control room. This inleakage has been included in the BFN licensing basis and has been accepted by NRC.
IV.
Safety Evaluation A.
Alternative Source Term BFN has performed a full scope analysis of the AST as defined in RG 1.183. A detailed description of AST analysis is provided in Enclosure 4 and the methods and results of the analysis are summarized in this section. The analysis included the following:
- 1.
Identification of the core source term based on plant specific analysis of core fission product inventory.
- 2.
Determination of the release fractions for the four BFN DBAs that could potentially result in control room and offsite doses. These are the LOCA, the main steam line break accident, the refueling accident, and the control rod drop accident.
- 3.
Calculation of fission product deposition rates and removal efficiencies.
- 4.
Calculation of offsite and control room personnel TEDE.
- 5.
Evaluation of suppression pool pH requirements to ensure that the particulate iodine deposited into the suppression pool does not re-evolve and become airborne as elemental iodine.
- 6.
Calculation of a new control room atmospheric dispersion factor (X/Q) for a main steam line break accident instantaneous ground level puff release.
- 7.
Evaluation of other related design and licensing bases such as NUREG-0737, "Clarification of TMI Action Plan Requirements."
The radiological dose analyses for AST have been performed assuming reactor operation at Extended Power Uprate conditions (3952 Mwt). This results in a conservative estimate of fission product releases for current licensed power of the units. BFN Units 2 and 3 currently have a maximum licensed thermal power of 3458 Mwt. However, TVA is actively engaged in an EPU project to increase reactor power to 3952 MWt.
AST Methodology Implementation of AST included the following:
- 1.
Development of a bounding plant-specific core fission product inventory.
El -6
- 2.
Introduction of a new X/Q for an instantaneous ground level puff release to the atmosphere for the main steam line break accident.
- 3.
No credit is taken for CREV or SGT System charcoal adsorption for any DBA.
- 4.
No credit is taken for CREV or SGT System HEPA filter particulate removal for any DBA except LOCA.
- 5.
New requirements for post-LOCA SLC System operation for suppression pool pH control along with calculation of sodium pentaborate (SPB) quantity requirements were developed.
The AST analyses were performed in accordance with RG 1.183. The results were evaluated to confirm compliance with the acceptance criteria presented in 10 CFR 50.67 and General Design Criteria 19 of 10 CFR 50, Appendix A.
Evaluation DBA accident analyses documented in Chapter 14 of the BFN UFSAR that potentially result in control room and offsite doses were addressed using methods and input assumptions consistent with the AST methodology. The following BFN DBAs were addressed:
Loss of Coolant Accident (LOCA), UFSAR Section 14.6.3 Main Steam Line Break Accident, UFSAR Section 14.6.5 Refueling Accident, UFSAR Section 14.6.4 Control Rod Drop Accident, UFSAR Section 14.6.2 The AST control room dose analyses are applicable for all three unit control rooms. The Unit 1 and 2 control rooms are shared in a common room with Unit 1 at one end and Unit 2 at the other. The Unit 3 control room, though separated from the Unit I and 2 control room, is part of the same control bay habitability zone. The refueling accident radiological consequence analysis is applicable to all three units since the refuel zone is common.
Results LOCA The radiological consequences of the DBA LOCA were analyzed. The post-accident doses are the result of the following activity considerations:
- 1.
Primary to secondary containment leakage. This leakage is directly released into secondary containment and filtered by SGT System prior to elevated release through the plant stack with stack bypass released at E1-7
ground level. No credit is taken for SGT or CREV System charcoal adsorber action.
- 2.
ECCS leakage into the secondary containment. This leakage is directly released into the secondary containment environment and the airborne portion is filtered by SGT System prior to elevated release through the plant stack with stack bypass fraction released at ground level. No credit is taken for SFG or CREV System charcoal adsorber.
- 3.
MSIV leakage from the primary containment into the main condenser (with a fraction that bypasses the main condenser directly to the atmosphere).
Leakage passes through the alternate MSIV leakage pathway to the main condenser with credit for deposition before it is released, undiluted and unfiltered, through the turbine building vents.
- 4.
Harden Wet Well Vent leakage from primary containment. This leakage is directly released (after a eight hour delay) to an elevated release through the plant stack.
- 5.
Post-DBA LOCA radiation shine dose to personnel within the control room from activity released to the reactor building and from activity contained in Core Spray System piping.
Loss Of Coolant Accident For the AST LOCA analysis, Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and control room calculated doses remain within the regulatory limits. These results are summarized in the following table along with results for the LOCA analysis using the current source term.
E1-8
Top of Stack 5.68E-1 2.43E-1 Turbine Building 3.02E-1 1.13E-1 Roof ECCS Leakage -
1.25E-2 1.21 E-2 Base of Stack ECCS Leakage -
3.52E-1 1.12E-1 Top of Stack Shine N/A 7.62E-1 TOTAL 1.02 1.25 1.25 Regulatory Limit 25 25 5
Current Analysis 1.67E-01 (25) Gamma 4.82E-01 (25) Gamma 6.83E-01 (5) Gamma (Regulatory Limit) -
1.01E-01 (300) Beta 4.84E-01 (300) Beta 1.58E-01 (30) Beta rem 5.84 (300) Thyroid 8.6 (300) Thyroid 2.95E+01 (30) Thyroid E1-9 Base of Stack 1.08E-2 4.49E-3
Main Steam Line Break Accident For the Main Steam Line Break analysis EAB, LPZ, and control room calculated doses remain within the regulatory limits for the two cases analyzed. The control room doses were determined using the new X/Q value for an instantaneous ground level puff release. These results are summarized in the table below along with the results from the current source term analysis.
3.2 ýiCi/gm DE 1-131 1.30E-1 6.52E-2 4.09E-2 1 Current analysis are based on 32 pCi/gm DE 1-131 limit.
El-10 32 jtCi/gm DE 1-131 1.30 6.52E-1 4.09E-1 Regulatory Limit 25 25 5
Current Analysis 3.72E-01 (25) Gamma 1.86E-01 (25) Gamma 5.30E-02 (5) Gamma (Regulatory Limit) -
1.56E-01 (300) Beta 7.80E-02 (300) Beta 3.27E-02 (30) Beta remI 2.99E+01 (300) Thyroid 1.49E+01 (300) Thyroid 1.05E+01 (30) Thyroid
Refueling Accident For the AST design basis refueling accident the EAB, LPZ, and control room calculated doses are within the regulatory limits. The results are summarized in the table below along with the results of the current source term analyses.
24 Hours after shutdown 6.7E-01 3.3E-01 3.8E-01 Regulatory Limit 6.30 6.30 5
Current Analysis 3.37E-01 (25) Gamma 1.68E-01 (25) Gamma 4.94E-02 (5) Gamma (Regulatory Limit) -
5.77E-01 (300) Beta 2.89E-01 (300) Beta 4.96E-01 (30) Beta rem 3.32E+01 (300) Thyroid 1.66E+01 (300) Thyroid 1.74 (30) Thyroid El-11
Control Rod Drop Accident The radiological consequences of the design basis control rod drop accident using AST methodology were analyzed. The EAB, LPZ, and control room calculated doses remain within the regulatory limits after AST implementation.
The results are summarized in the table below along with the results of the current source term analyses.
Power Operation 1.19 6.82E-01 2.48E-01 Suppression Pool pH Control The AST LOCA analysis takes credit for minimization of re-evolution of elemental iodine from the suppression pool, which is strongly dependent on suppression pool pH. The analysis assumed that sodium pentaborate SPB was injected via SLC within several hours of the onset of a LOCA. The conservative modeling of the primary containment cabling results in the production of a large amount of hydrochloric acid. Using the assumptions of a minimum of 4000 gallons of >8% by weight injectable SPB solution, the minimum suppression pool pH at 30 days post-LOCA remains above 7.0. This pH satisfies the conditions for inhibiting the release of the chemical form of elemental iodine from the containment. This quantity of SLC is above current TS SR 3.1.7 requirements of 3007 gallons. Therefore, TS revisions are proposed which increase the quantity of SLC required to be maintained as shown in Enclosure 2.
Based on the AST analysis for suppression pool pH control, the SLC system will also be credited for limiting radiological dose following a design basis recirculation pipe break LOCAs involving fuel damage. However, the SLC system will not be re-classified as a safety system, but will retain the current classification as described in UFSAR Section 3.8.
El-12 Regulatory Limit 6.30 6.30 5
Current Analysis 1.52 (25) Gamma 8.58E-01 (25) Gamma 3.86E-02 (5) Gamma (Regulatory Limit) -
1.07 (300) Beta 6.04E-01 (300) Beta 4.32E-01 (30) Beta rem I1.58E+01 (300) Thyroid 1.58E+01 (300) Thyroid 6.3 (30) Thyroid
Main Steam Line Break Accident Puff Release Dispersion Factor In support of the AST Main Steam Line Break analysis, a new control room X/Q value for an instantaneous ground level puff release to the atmosphere was calculated for use in the radiological dose analysis. This X/Q value is shown in the table below.
I Time Period I
Control Room (sec/m 3)
I 46 secs 4.60E-4 NUREG-0737 Evaluation The revised analyses includes consideration of the impacts of AST methodology for several NUREG-0737 items. These are summarized below.
Post-Accident Vital Area Access and Sampling - The results of the revised post-accident mission dose calculations demonstrate that the current calculated doses (based on TID-14844 source terms) bound the doses that would be calculated based on AST source terms. The evaluated mission doses remain less than 5 rem TEDE (NUREG-0737, Items ll.B.2 and II.B.3).
Post-Accident Radiation Monitor - The containment high range radiation monitors used to monitor post-accident primary containment radiation levels were evaluated for the impact of AST. The monitors continue to provide their design function and envelope the projected radiation rates. (NUREG 0737, Item II.F.1).
Control Room Radiation Protection - The resultant doses to the control room for each of the four DBAs analyzed for AST have been determined.
In each case the control room dose is less than 5 rem TEDE (NUREG-0737 items III.A.1.2 and III.D.3.4).
Radioactive Sources Outside the Primary Containment - The contribution of radiological dose consequences as a result of radiation shine and ECCS leakage was determined as part of the radiological dose analysis for the LOCA and found acceptable (NUREG-0737, Item III.D.1.1).
El-13
Conclusion Radiological dose analyses were performed using AST methodology for the four BFN DBAs with a potential for control room and offsite doses. Control room and offsite doses remain within regulatory requirements.
B.
Pressure Drop Testing of ESF Ventilation System TS 5.5.7.d addresses the pressure drop test across the combined HEPA filters, prefilters, and charcoal adsorbers for the CREVS and SGT systems. As discussed earlier, AST radiological analyses do not take credit for charcoal filters in the CREVS and SGT Systems. Although BFN has no specific plans for the physical removal of these adsorbers, TS 5.5.7.d must be revised to include the case in which the charcoal adsorbers and associated after-filters may be removed. The after-filters are present to capture any charcoal fines and have not been credited for any radioactivity removal.
A plant modification to remove these filters would result in a decrease in the pressure drop through the filter trains for these systems. Accordingly, the new TS limits for the pressure drop tests have been decreased to reflect the potential removal of the charcoal adsorber and after-filter. The revised limits will ensure that appropriate testing criteria exists for the potential removal of the charcoal adsorber and resulting system modification effects.
V.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA is submitting a request for amendment to the Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 Technical Specifications (TS). The proposed amendment is a full implementation of an alternative source term (AST) for the Units 1, 2, and 3 operating licenses, adopting AST methodology by revising the current accident source term and replacing it with an accident source term as prescribed in 10 CFR 50.67.
AST analyses were performed using the guidance provided by Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000, and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms."
The four limiting design basis accidents (DBAs) considered were the Control Rod Drop Accident, the Refueling Accident, the Loss of Coolant Accident, and the Main Steam Line Break Accident.
TVA has concluded that operation of BFN Units 1, 2, and 3 in accordance with the proposed change to the TS does not involve a significant hazards 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(c).
El-14
A.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The AST and those plant systems affected by implementing AST do not initiate DBAs. The AST does not affect the design or operation of the facility; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences. The implementation of the AST has been evaluated in the analyses for the limiting DBAs at BFN.
The equipment affected by the proposed change is mitigative in nature and relied upon following an accident. The proposed changes to the TS do revise certain performance requirements. However, these changes will not involve a revision to the parameters or conditions that could contribute to the initiation of a design basis accident discussed in Chapter 14 of the BFN Updated Final Safety Analysis Report.
Plant specific radiological analyses have been performed and, based on the results of these analyses, it has been demonstrated that the dose consequences of the limiting events considered in the analyses are within the regulatory guidance provided by the NRC for use with the AST. This guidance is presented in 10 CFR 50.67, Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1. Therefore, the proposed amendment does not result in a significant increase in the consequences or a significant increase the probability of any previously evaluated accident.
B.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of AST does not alter any design basis accident initiators.
These changes do not affect the design function or mode of operations of systems, structures, or components in the facility prior to a postulated accident. Since systems, structures, and components are operated essentially no differently after the AST implementation, no new failure modes are created by this proposed change. Therefore, the proposed license amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.
C.
The proposed amendment does not involve a significant reduction in a margin of safety.
The changes proposed are associated with a revision to the licensing basis for BFN. The results of accident analyses revised in support of the proposed change are subject to the acceptance criteria in 10 CFR 50.67.
The analyzed events have been carefully selected, and the analyses supporting this submittal have been performed using approved El-15
methodologies. The dose consequences of these limiting events are within the acceptance criteria provided by the regulatory guidance as presented in 10 CFR 50.67, Regulatory Guide 1.183, and SRP 15.0.1.
Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, the changes are considered to not result in a significant reduction in a margin of safety.
VI.
ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.
El-16
References:
- 1.
TVA letter to NRC dated July 31, 1992, Browns Ferry Nuclear Plant Resolution of Control Room Emergency Ventilation System (CREV)
Issues.
- 2.
NRC letter to TVA dated May 7, 1998, Browns Ferry Nuclear Plant, Units 2 and 3: Request for Additional Information Relating To Technical Specification Change No. TS-384 - Power Uprate Operation.
- 3.
NRC Letter to NRC dated September 8, 1998, Issuance to Amendments Re: Power Uprate - Browns Ferry Plant, Units 2 and 3.
- 4.
TVA Letter to NRC dated March 30, 1999, Browns Ferry Nuclear plant (BFN) - Resolution Of Control Room Emergency Ventilation (CREV)
System Issues With Regard To License Condition Associated With Units 2 and 3 power Uprate Operating License Amendments 254 and 214.
- 5.
NRC Letter to TVA dated August 3, 1999, Safety Evaluation Supplement, Browns Ferry Nuclear Plant Units 2 and 3 - Radiological Dose Calculations Associated With Power Uprate License Amendment Nos.
254 and 214.
- 6.
TVA Letter to NRC dated September 28, 1999, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specification (TS) Change 399 Increased Main Steam Isolation Valve (MSIV) Leakage Rate Limits and Exemption From 10 CFR 50 Appendix J.
- 7.
TVA Letter to NRC dated February 4, 2000, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications Change 399 - Increased Main Steam isolation Valve (MSIV) Leakage Rate Limits and Exemption From 10 CFR 50 Appendix J.
- 8.
NRC Letter to TVA dated March 14, 2000, Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Limits on Main Steam Isolation Valve Leakage (TAC Nos. MA6405 and MA6406).
El-17
ENCLOSURE2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM MARKED PAGES - TECHNICAL SPECIFICATIONS AFFECTED PAGE LIST The following pages have been revised. On the affected pages the revised portions have been highlighted. A line has been drawn through the deleted text and a double underline for new or revised text.
Operating License Unit 2 Unit 3 Page 4 Page 4 Technical Specifications Unit 1 Unit 2 Unit 3 3.1-23 3.1-23 3.1-23 3.1-24 3.1-24 3.1-24 3.1-25 3.1-25 3.1-25 3.1-26 3.1-26 3.1-26 3.3-64 3.3-65 3.3-65 3.3-69 3.3-70 3.3-70 3.6-44 3.6-44 3.6-44 3.6-45 3.6-45 3.6-45 3.6-47 3.6-47 3.6-47 3.6-49 3.6-49 3.6-49 3.6-51 3.6-51 3.6-51 3.6-52 3.6-52 3.6-52 3.6-53 3.6-53 3.6-53 3.7-8 3.7-9 3.7-9 3.7-9 3.7-10 3.7-10 3.7-10 3.7-11 3.7-11 5.0-15 5.0-15 5.0-15 5.0-16 5.0-16 5.0-16
(3)
The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6, 1996, as supplemented May 1, August 14, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997, January 23, March 12, April 16, 20 and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment. This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
(4)
TClassro rormaan siuanorltrinin of ahe d
poien ba rate l
re olate t ccha dent tha aoffect operaorplerformaGncerwi besonducrted rionrGC to9oertng atf uprtedlmt conidiriong. Smuaitor ihngso thato alre lensiste nd wihmoergc carte coodiiong rwqire madlftenadsimltrfdlt will be valletd uida thedi acrdaelnce with AeNSeI/NS3518.Tangad the pblt omnto eilclacndtsimuaondrwinl be modified, as amnietessry ptaoincmrpodate tchantg o
idoomie oduran torsto mtenting doThithi mdfctosifnarcope.Thsamendment is effective immediately.
(5)(a Celasroentimltrtanngoeldoerurteaedcagsta (4 )
e....
affect..........
operator...........
pefrm ne.il.eco d ctdpio.o.praig.tupae condiions Simulator changes tha are.....
cosstn with poe.prt
.oniin will~~~ be......
made and. siuao fielt wil be.aldatd.n.acodane.it ANr /N 3.-95 riigad h lnMiuao wl emdfea necessary,..
to.....
inoprt chngs.dntfid.uin.sa
.u.tstn.
Thi amedmntisefectveimedatly (5)(a)....
Deleted........
2 (6).....
Deleted.....
(7)
Deleted BFN-Unit 2 4
Amendment NO.262 December 16,1999
(3)
The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's application dated September 6, 1996, as supplemented May 1, August 14, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997, January 23, March 12, April 16, 20, and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment. This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.
(4
............iiiii i H
'
- u * :r
....................... *....................*..................................l *t.:. :.................
(4)
TC lassroommand simulatrtainin of th l powern upratle recolante ccidnens toa afctopeirator perorance wi llGn~
besicondrcted rior 00 toopraing af~to lprate condeiting. Smuai torm choanges thalae consisteand wihmoergc corae cooliiong wilemlbe made.
and simulatsor fdityhill bealyidai b uited inacorac wthe R o
ANSI/ANSi3.5-nd85pproainibyanrche3 plant Fimulaorwilg be mppodifil, ans n~uieesay todiinorpioratil e hangesientife during sthartupesting.
Thigs
()
hedulense forl furlny implemernit 3and maitriingn0 efrcnt all prvison ofill Cmmistsion-apprili ove phynitc adlsecurty guardtringadqulfction, uigemrece and safe$guirdselningency plans binciludbn mnamnendmentslth raepuirsuntt provisiaions, of tnhe Misclanipeous Amendmentsmant Serc Rfequire menitsl (5) Classroom and simulatR training. onhal power, uprate relatedcaingSaesgtatd affetoperatior performaedncer wil beR conducte prire ettoed opowseratnturate codiios.ca SimurtPlator cihangevsitar ons sbistent wthruh powe uprat condtion willde mnadeorande simltor CFidlt will5 behalleidalmeted in accordance wt nec tesay toedl incorporate changesidntiiddrn.truetn.Ti amenmenUisTffetiv immediately.
22 Decmbe 1-,199 (6).................
The. license shallfull implment and,.
mintain. in..
efec all. prvsin of. the Commission-approved...............
physical securit. guardtrainin.and.qulificaton.an safeuard cotigec plan inluin amendments. made.......
pursunt.t provision of th Miclaeu Amnmet an Serc Reuiemnt revision to 10. CFR.........
73.55 (5.R287ad.72)adt.teatoiyo 10..............
CF 5.9 an.1.C R 5054p. The......
plans,..........
which.contain.Safeguard revision sumte thog Apil 16ow 1987 and "BonerySfgad Contingency.. Plan",.. wihreiiossum..dtrog
.Jn.2,98.Chne made.........in...
ac oda c with.....
10 CFR 73............5 shall.. be...
im le ened i.a coda c w ith the schedule set forth therein........................................
BFN........
UNIT 3.
- 4.
Amendment.......No....222.
December 16....1999..
SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 APPLICABILITY:
Two SLC subsystems shall be OPERABLE.
MODES 1* *2, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.
B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable, subsystem to OPERABLE status.
C. Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
02 B ftMDEG3&hAr Amendment No. 234 BFN UNIT-1 3.1-23
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution (SPB) is _Ž 07 4000 gallons.
SR 3.1.7.2 Verify continuity of explosive charge.
31 days SR 3.1.73.-A Verify the SPB concentration is _< 9.2% by weight.
OR Verify the concentration and temperature of boron in solution are within the limits of Figure 3.1.7-1.
31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to solution Once within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after discovery that SPB concentration is
> 9.2% by weight AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
.......2..
Amendment No.W 3.1-24 BFN-UNIT 1
SLC System 3.1.7 thereafter SR 3.1.74 Verify the minimum quantity of Boron-10 in the 31 days SLC solution tank and available for injection is
> 186 pounds.
(continued)
U3...Am...t Amendment No.
BFN-UNIT 1 3.1-24
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.1.7.5 Verify the SLC conditions satisfy the following equation:
(13 wt. %)(86 gpm)(1 9.8 atom%)
- where, C = sodium pentaborate solution concentration (weight percent)
Q = pump flow rate (gpm)
E =
Boron-10 enrichment (atom percent Boron-1 0)
( C
)(
o )(
E FREQUENCY 31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to the solution SR 3.1.7.a Verify each pump develops a flow rate > 39 18 months gpm at a discharge pressure _ 1275 psig.
Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel.
STAGGERED TEST BASIS SR 3.1.7.*9 Verify all piping between storage tank and 18 months pump suction is unblocked.
(continued)
Amendment No.....
BFN-UNIT 1 3.1-25
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1..
Verify sodium pentaborate enrichment is within 18 months the limits established by SR 3.1.7.$ by calculating within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verifying by AND analysis within 30 days.
After addition to SLC tank SR 3.1.7.4Q1`.
Verify each SLC subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
Amendment No.
4 BFN-UNIT I 3.1-26
Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)
Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED FUNCTION OTHER CHANNELS SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIREMENTS VALUE CONDITIONS TRIP SYSTEM
- 1.
1,2,3, 2
_ 538 inches above Low, Level 3 (a)
SR 3.3.6.2.2 vessel zero SR 3.3.6.2.3 SR 3.3.6.2.4
- 2.
Drywell Pressure - High 1,2,3 2
< 2.5 psig SR 3.3.6.2.3 SR 3.3.6.2.4
- 3.
Reactor Zone Exhaust 1,2,3, 1
< 100 mR/hr Radiation - High (a)M SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4
- 4. Refueling Floor Exhaust 1,2,3, 1
< 100 mR/hr Radiation - High (a)*
SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 (a)
During operations with a potential for draining the reactor vessel.
7 fb* ::
ring C RE A:T::: ::
S 3 du::n : m:{:;:
t fT::
1*, :ue: :c m
.lc~
i*:ndar,'
cnI
- .40 Amendment No.
3.3-64 BFN-UNIT 1
CREV System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)
Control Room Emergency Ventilation System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION A.1
1,2,3,(a) 2 B
> 538 inches Low, Level 3 SR 3.3.7.1.2 above vessel SR 3.3.7.1.5 zero SR 3.3.7.1.6
- 2.
Drywell Pressure - High 1,2,3 2
_2.5 psig SR 3.3.7.1.5 SR 3.3.7.1.6
- 3.
Reactor Zone Exhaust 1,2,3 1
<100 mR/hr Radiation - High (a)*
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 4.
Refueling Floor Exhaust 1,2,3, 1
- 100 mR/hr Radiation - High (a)*
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 5.
Control Room Air Supply Duct 1,2,3, 1
< 270 cpm Radiation - High (a)h SR 3.3.7.1.2 above SR 3.3.7.1.3 background SR 3.3.7.1.4 (a)
During operations with a potential for draining the reactor vessel.
A."!
4-uGiko 446E 6wERTIN d'~I~~~c
.T3A e n..........
Amendment No.
3.3-69 BFN-UNIT 1
Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.
APPLICABILITY:
MODES 1. 2. and 3.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, containment to or 3.
OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No.
4 BFN-UNIT 1 3.6-44
Secondary Containment 3.6.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment 1
NOTE inoperable
- rgLO.Q3int inoperable
~
~
~
~
~
.....ii
.i
~ i ~ i:
.i i:i i:*:i:*:
i:* :::i i i i
- :*...- ~~~i iiii!~i.. ' i..*
C. SusIitaen acto~ntospendo Immediately OPPDR~s.
- k A
.u p n e e dLEATOS C.~I Iniiate acton.to.sspend.Imediate.
OPDR~s.
Amendment No. 234 BFN-UNIT 1 3.6-45
SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 APPLICABILITY:
MODES 1, 2, and 3, Dur....i........
m.......
~
tdfe a~ ibie ntc o o d r Duin operations with.. a. oenilfo.rann.tereco.vse fliirin.
.~2 P
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS
NOTES -------------------------------------------------
- 1.
Penetration flow paths may be unisolated intermittently under administrative controls.
- 2.
Separate Condition entry is allowed for each penetration flow path.
- 3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one SCIV penetration flow path by inoperable, use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.
AND (continued)
Amendment No.
4 BFN-UNIT 1 3.6-47
SCIVs 3.6.4.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and 0.1 NOTE associated Completion 3.0.3
..,not Time of Condition A or B pIe4 ns÷ mb~ie
- in theSu:pnd mvemet ome:ite:
.:.:.::.;.::+
ALTERAT:**::IO:NS, or::: during OPDRVs.
n.o m*tp.:. : :........
- :iii
- ::1%
S.....
- .....:..:.:...:.,.:+. :.:.. :..... :. :.+,
D.3 1*
Initiate action to suspend Immediately OPDRVs.
Amendment No....
BFN-UNIT 1 3.6-49
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 APPLICABILITY:
Three SGT subsystems shall be OPERABLE.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No. 2.4 BFN-UNIT 1 3.6-51
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not C.1 Place two OPERABLE SGT subsystems in operation.
OR m n ANP Initiate action to suspend OPDRVs.
Immediately Immediately D. Two or three SGT D.1 Enter LCO 3.0.3.
Immediately subsystems inoperable in MODE 1, 2, or 3.
(continued)
Amendment No.
C.2v
.. -..., H,.... _ *..
BFN-UNIT 1 3.6-52
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGT El NOTE subsystems inoperable is not
- t::.*
- t*:*:t*:t*::'.*.:::**:i*::...........1.................:
during O V s.
during OPRA.
ieOPDR s duringOPDRRs.
Amendment No.2+,".'
BFN-UNIT 1 3.6-53
CREV System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Ventilation (CREV) System LCO 3.7.3 Two CREV subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2. and 3, uring operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREV subsystem A.1 Restore CREV subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
...........234 Amendment No.
BFN-UNIT 1 3.7-8
CREV System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not
' ' '*"" i*.'.*ii~ !!!*i~i~ ~ i~
i~i~........
met 64vmen
!o i"4.
.. E.E,.
- ..:T
.o d u ring..
.I.................
. +
..':* ii.-..*......... i :: i~ ~ i:.....................
.-.+... *.-.....,.*.. :.... ¥....
- ..:.........:..+........:...:...:.::.. -+ -..-..: -. k.:.
during OPDRVs.
C.1 Place OPERABLE CREV subsystem in pressurization mode.
.: :.:... :..... +......................................................
AM~
.. L..........
C.2.3 Initiate action to suspend OPDRVs.
Immediately Immediately D. Two CREV subsystems D.1 Enter LCO 3.0.3.
Immediately inoperable in MODE 1, 2, or 3.
(continued)
Amendment No. 234 3.7-9 BFN-UNIT 1
CREV System 3.7.3 ACTIONS (continued)
Amendment No.
4 BFN-UNIT I CONDITION REQUIRED ACTION COMPLETION TIME E. Two CREV subsystems inoperable riot
.:.: + :..;.:....:. :..... :.......... :. : :.
i:!:*
i:* :i*:i:*
ii*....
N:
- ii*
.*i*
A*:lTE.R.AT*!I~i~t: iSIii.: or during ir*dat"ed fu"el a:':embl'*e:
O P DRVs.
- ~
~
~ ~ ~ ~
ii
.....i~i i i i
- E.*I1 Initiate action to suspend Immediately OPDRVs.
3.7-10
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued) b.
+
E$.....................
(continued)
BFN-UNIT 1 5.0-15 Amendment No. 2 b....3.,......
-Q=
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued) 4*.
Once every 24 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below at the system flowrate specified below, + 10%:
ESF Ventilation Delta P wiih IIRitb*
A Flowrate System car charcoal (cfm) 6 (inches water)
SGT System 7
59000 CREV System 6
4 3000
- c*.
Once every 24 months demonstrate that the heaters for the SGT System dissipate _> 40 kW when tested in accordance with ANSI N510-1975.
5.5.8 Explosive Gas and Storage Tank Radioactivity Monitorinq Program This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiners, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The program shall include:
- a. The limits for concentrations of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and (continued)
BFN-UNIT 1 5.0-16 Amendment No.
SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1:a *W2, 2
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.
B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable, subsystem to OPERABLE status.
C. Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
- 2.
B e..n..............
..........EN.2me o
Amendment No. 2 BFN-UN IT 2 3.1-23
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution (SPB) is > 0O4Q gallons.
SR 3.1.7.2 Verify continuity of explosive charge.
31 days y
P.
-e h. h..6 ca.da ur Verify the SPB concentration is _< 9.2% by weight.
OR Verify the concentration and temperature of boron in solution are within the limits of Figure 3.1.7-1.
31 -days AND..
Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is..
added to solution 31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to solution Once within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after discovery that SPB concentration is
> 9.2% by weight AND I
Amendment No.
SR 317 iii BFN-UN IT 2 3.1-24
SLC System 3.1.7 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter SR 3.1.74-Verify the minimum quantity of Boron-10 in the 31 days SLC solution tank and available for injection is
Ž186 pounds.
(continued)
Amendment No.
BFN-UNIT 2 3.1-25
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.1.7.5 Verify the SLC conditions satisfy the following equation:
(13 wt. %)(86 gpm)(1 9.8 atom%)
- where, C = sodium pentaborate solution concentration (weight percent)
Q = pump flow rate (gpm)
E =
Boron-lO enrichment (atom percent Boron-I 0)
( r~
o
)~(
F FREQUENCY 31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to the solution SR 3.1.7.Z Verify each pump develops a flow rate _> 39 24 months gpm at a discharge pressure _> 1325 psig.
SR 3.1.7.18 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel.
STAGGERED TEST BASIS SR 3.1.7.59 Verify all piping between storage tank and 24 months pump suction is unblocked.
(continued)
BFN-UNIT 2 3.1-26 Amendment No. W
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.7.9 Verify sodium pentaborate enrichment is within 24 months the limits established by SR 3.1.7.$
by calculating within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verifying by AND analysis within 30 days.
After addition to SLC tank SR 3.1.7.1 Verify each SLC subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
BFN-UNIT 2 3.1-26 Amendment No. 2 Novemb..r...,.....
Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)
Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED FUNCTION OTHER CHANNELS SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIREMENTS VALUE CONDITIONS TRIP SYSTEM
1,2,3, 2
_ 528 inches above Low, Level 3 (a)
SR 3.3.6.2.2 vessel zero SR 3.3.6.2.3 SR 3.3.6.2.4
- 2.
Drywell Pressure - High 1,2,3 2
!5 2.5 psig SR 3.3.6.2.3 SR 3.3.6.2.4
- 3.
Reactor Zone Exhaust 1,2,3, 1
_< 100 mR/hr Radiation - High (a))
SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4
- 4. Refueling Floor Exhaust 1,2,3, 1
_< 100 mR/hr Radiation - High (a)*
SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 (a)
During operations with a potential for draining the reactor vessel.
(b) DLit¶ CORE ALTER/\\TIQN:rngtt tJrr
tcftU+/-t effltl lr e tdt' ontinttnt.
Amendment No. 253,2.w 3.3-65 BFN-UNIT 2
CREV System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)
Control Room Emergency Ventilation System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION A.1
- 1.
1,2,3,(a) 2 B
> 528 inches Low, Level 3 SR 3.3.7.1.2 above vessel SR 3.3.7.1.5 zero SR 3.3.7.1.6
- 2.
Drywell Pressure - High 1,2,3 2
_< 2.5 psig SR 3.3.7.1.5 SR 3.3.7.1.6
- 3. Reactor Zone Exhaust 1,2,3 1
_< 100 mR/hr Radiation - High (a)*
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 4. Refueling Floor Exhaust 1,2,3, 1
- 100 mR/hr Radiation - High (a)*
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 5.
Control Room Air Supply Duct 12,3 1
_ 270 cpm Radiation - High (a);
SR 3.3.7.1.2 above SR 3.3.7.1.3 background SR 3.3.7.1.4 (a)
During operations with a potential for draining the reactor vessel.
BFN-UNIT 2 3.3-70 Amendment No. 253, 260
Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The APPLICABILITY:
MO[
secondary containment shall be OPERABLE.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, containment to or 3.
OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No. 25 BFN-UNIT 2 3.6-44
Secondary Containment 3.6.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment r`.A NOTE...
inoperable......
.*.........i n t i:
i* i:
PDeRVs.'"
C.*1 Initiate action to suspend Immediately OPDRVs.
Amendment No. 25 BFN-UNIT 2 3.6-45
SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 APPLICABILITY:
M O D E S 1, 2, a n d 3,..................................
ON
...ýiM h :
Ug P A
.X..
"A'I...
r1t
-UMUR",
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS NOTES
- 1.
Penetration flow paths may be unisolated intermittently under administrative controls.
- 2.
Separate Condition entry is allowed for each penetration flow path.
- 3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one SCIV penetration flow path by inoperable, use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.
AND (continued)
Amendment No.
15I0 sidoa-BFN-UNIT 2 3.6-47
SCIVs 3.6.4.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.........T...
associated Completion i...n..
Time of Condition A or B
- ::i * ! ii ~
i~*;*:......:
soon~r cntinn, rr:d*ted fuel*
sse:;'
mbt'*:ie T..
- r.
during
.e...
n o t:::::::**:*
- i i :** :*; :* :* :i.........
- AAN, D. I* Initiate action to suspend Immediately OPDRVs.
Amendment No. 2 BFN-UNIT 2 3.6-49
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 APPLICABILITY:
Three SGT subsystems shall be OPERABLE.
MODES 1, 2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No. 2M BFN-UNIT 2 3.6-51
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not metduii omotl during OPDRVs.
C.1 Place two OPERABLE SGT subsystems in operation.
OR AND C.24 Initiate action to suspend OPDRVs.
Immediately Immediately D. Two or three SGT D.1 Enter LCO 3.0.3.
Immediately subsystems inoperable in MODE 1, 2, or 3.
(continued)
Amendment No.
BFN-UNIT 2 3.6-52
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGT EA
..:<2.......
subsystems inoperable tXOiLO23:i:
not S.......
during OPDRVs.
ieiidis
- ..., *.%...,'* iii~iiiii'ii~ii~i!i,...
- i...."
i:..* : !.,.....:.*.!:*!
- ...-.!I m
E.*i1 Initiate action to suspend Immediately OPDRVs.
Amendment No.
BFN-UNIT 2 3.6-53
CREV System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Ventilation (CREV) System LCO 3.7.3 APPLICABILITY:
Two CREV subsystems shall be OPERABLE.
MODES 1, 2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREV subsystem A.1 Restore CREV subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
BFN-UNIT 2 3.7-9 Amendment No. 264 p.....t.m.:.
M.`.. :$..;..:.:
CREV System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not m et -ad W h 6;.. 0U i 4 6 ent ol during OPDRVs.
C.1
- ..:..:. :..:...:..:.:.:...:+.. :.. :.:.. : ::..........
I i!iiii~~i*i*~gi~!!iiiiiiiiii Place OPERABLE CREV subsystem in pressurization mode.
OR C.21 Initiate action to suspend OPDRVs.
Immediately Immediately D. Two CREV subsystems D.1 Enter LCO 3.0.3.
Immediately inoperable in MODE 1, 2, or 3.
(continued)
BFN-UNIT 2 3.7-10 Amendment No.
CREV System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two CREV subsystems T..
`
E.
upoIniiat mctoneen tosspn Immediately OPDRVs.
BFN-UNIT 2 3.7-11 Amendment No.
4
...... t.................,.............
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testingq Program (VFTP) (continued)
~...........o echofth E F e tha mo.. lao tr tet of f th
.h r o l a s r e ~ s o s a m t y o i e e f c e
Ž90% w hen fr~ a..i.............it...............
9 (continued)
BFN-UNIT 2 5.0-15 Amendment No. 25
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)
Once every 24 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below at the system flowrate specified below, + 1 0%
ESF Ventilation System SGT System CREV System (inches water) 7 6
h.a.r.:....
.i..t.....
-Wic.
Once every 24 months demonstrate that the heaters for the SGT System dissipate _> 40 kW when tested in accordance with ANSI N510-1975.
5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiners, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The program shall include:
- a. The limits for concentrations of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and (continued)
BFN-UNIT 2 5.0-16 Amendment No. 265
.N......
m b....
r 3 Flowrate (cfm) 9000 3000
SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, i 2, ahn*d ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.
B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable, subsystem to OPERABLE status.
C. Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND Amendment No. 24 BFN-UNIT 3 3.1-23
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> solution (SPB) is _> 3OO.4.
gallons.
SR 3.1.7.2 Verify continuity of explosive charge.
31 days
- i.
i...
b y SR 3.1.7;"4 Verify the SPB concentration is _< 9.2% by weight.
OR Verify the concentration and temperature of boron in solution are within the limits of Figure 3.1.7-1.
AND Once within water or boron is added toM solution 31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to solution Once within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after discovery that SPB concentration is
> 9.2% by weight AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Amendment No.
BFN-UNIT 3 3 1-24
SLC System 3.1.7 thereafter SR 3.1.7A4`,"
Verify the minimum quantity of Boron-10 in the 31 days SLC solution tank and available for injection is
Ž186 pounds.
(continued)
Amendment No. 2 BFN-UNIT 3 3.1-24
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.1.7.4 Verify the SLC conditions satisfy the following equation:
(
C
)(
Q
)(
E
)
(13 wt. %)(86 gpm)(1 9.8 atom%)
1
- where, C = sodium pentaborate solution concentration (weight percent)
Q = pump flow rate (gpm)
E =
Boron-1 0 enrichment (atom percent Boron-1 0)
FREQUENCY 31 days AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to the solution SR 3.1.7.0Z Verify each pump develops a flow rate >_ 39 24 months gpm at a discharge pressure __ 1325 psig.
SR 3.1.7.78 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel.
STAGGERED TEST BASIS SR 3.1.7.89 Verify all piping between storage tank and 24 months pump suction is unblocked.
(continued)
BFN-UNIT 3 3.1-25 Amendment No. 2
SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3 Verify sodium pentaborate enrichment is within 24 months the limits established by SR 3.1.7.$
by calculating within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verifying by AND analysis within 30 days.
After addition to SLC tank SR 3.1.7.j Verify each SLC subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
BFN-UNIT 3 3.1-26 Amendment No. W N::::::::::::b:::::::: ::,:::::::: ::
Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)
Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED FUNCTION OTHER CHANNELS SURVEILLANCE ALLOWABLE SPECIFIED PER REQUIREMENTS VALUE CONDITIONS TRIP SYSTEM
1,2,3, 2
> 528 inches above Low, Level 3 (a)
SR 3.3.6.2.2 vessel zero SR 3.3.6.2.3 SR 3.3.6.2.4
- 2.
Drywell Pressure - High 1,2,3 2
- 2.5 psig SR 3.3.6.2.3 SR 3.3.6.2.4
- 3.
Reactor Zone Exhaust 1,2,3, 1
- 100 mR/hr Radiation - High (a)*
SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4
- 4.
Refueling Floor Exhaust 1,2,3, 1
< 100 mR/hr Radiation - High (a)M SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 (a)
During operations with a potential for draining the reactor vessel.
(b) u~hCORELTEA~l~~ ~d du~hgmo~nentof Amendment No. 212, 213, 249 3.3-65 BFN-UNIT 3
CREV System Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)
Control Room Emergency Ventilation System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION A.1
1,2,3,(a) 2 B
> 528 inches Low, Level 3 SR 3.3.7.1.2 above vessel SR 3.3.7.1.5 zero SR 3.3.7.1.6
- 2.
Drywell Pressure - High 1,2,3 2
< 2.5 psig SR 3.3.7.1.5 SR 3.3.7.1.6
- 3.
Reactor Zone Exhaust 1,2,3 1
< 100 mR/hr Radiation - High (a),
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 4.
Refueling Floor Exhaust 1,2,3, 1
< 100 mR/hr Radiation - High (a)*
SR 3.3.7.1.2 SR 3.3.7.1.5 SR 3.3.7.1.6
- 5.
Control Room Air Supply Duct 1,2,3, 1
< 270 cpm Radiation - High (a)-
SR 3.3.7.1.2 above SR 3.3.7.1.3 background SR 3.3.7.1.4 (a)
During operations with a potential for draining the reactor vessel.
uurng tr :'.LI iI-:'
j¶JrI no curing ir.cmen r rroIatc mci ecmoie n mc eseonoar; containment.
Amendment No. 2 1l-
- 2.......
i z.I 3.3-70 BFN-UNIT 3
Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LCO 3.6.4.1 The APPLICABILITY:
MOE secondary containment shall be OPERABLE.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, 2, containment to or 3.
OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No. 242 BFN-UNIT 3 3.6-44
Secondary Containment 3.6.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary containment
.i..NOT........
inoperable d iii*LC i:i li:i! i*i not!i!*}i~i D.~.
4.0p144 oemeto
- ~ i :: :: * * ;..................
i~ iil:.i:
- il~ i if ~
l l:*
i:.i:.i
- d r in g........................................................................................................
C.1i Initiate action to suspend Immediately OPDRVs.
Amendment No. 242 BFN-UNIT 3 3.6-45
SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 APPLICABILITY:
MODES 1, 2, and 3, f.......m o..m..... f.irr-......d i..........d....
f.............................. i.
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS NOTES
- 1.
Penetration flow paths may be unisolated intermittently under administrative controls.
- 2.
Separate Condition entry is allowed for each penetration flow path.
- 3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> flow paths with one SCIV penetration flow path by inoperable, use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.
AND (continued) 4 2 Amendment No.
BFN-UNIT 3 3.6-47
SCIVs 3.6.4.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and
- .*....**........`...*.......*...........*.......*.*.*......*..*.....*..*.*.....*..*.*.....*..*.....*.....
associated Completion i....
i....io..
Time of Condition A or B n
membis~ h
- 1I
- .i *..-
- i i *
,*,
- r *.,."* : :*;,:::,*
- t.* *:*J
- ':* '* : ':**::*::r:*::
- ii
- AL*TERAIONSi,.or during OPDRVs.
imp
.A D.3 Initiate action to suspend Immediately OPDRVs.
Amendment No. 2 BFN-UNIT 3 3.6-49
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, uring operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Amendment No. 242 BFN-UNIT 3 3.6-51
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not
.M a
.t........ *.....*.............................**.....*
during UPURVs.
C.1 Place two OPERABLE SGT subsystems in operation.
OR AND AM~
C.2.3 Initiate action to suspend OPDRVs.
Immediately Immediately D. Two or three SGT D.1 Enter LCO 3.0.3.
Immediately subsystems inoperable in MODE 1,2, or 3.
(continued)
Amendment No. 2 BFN-UNIT 3 3.6-52
SGT System 3.6.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGTner b
- ."-*i*::*i*...-......:.
subsystems inoperable
-ta
-,A A'*..........
d u d n g m o v e.m..*..t t p
÷........
v... -..................
during OPDRVs.
.s.
.n.s.
m..
E.*I*
Initiate action to suspend Immediately OPDRVs.
Amendment No.
3.6-53 BFN-UNIT 3
CREV System 3.7.3 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Ventilation (CREV) System LCO 3.7.3 APPLICABILITY:
Two CREV subsystems shall be OPERABLE.
MODES 1, 2, and 3,
.. r.n...............
n
°--°....
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREV subsystem A.1 Restore CREV subsystem 7 days inoperable, to OPERABLE status.
B. Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
BFN-UNIT 3 3.7-9 Amendment No. 24
.S.:...:p.t..:.....:...b..o.:r...:.....O
- ~:*ii~:.*-'.* **ii l~~.:[L':~
CREV System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not C.1 Place OPERABLE CREV subsystem in pressurization mode.
OR C.24 Initiate action to suspend OPDRVs.
Immediately Immediately D. Two CREV subsystems D.1 Enter LCO 3.0.3.
Immediately inoperable in MODE 1, 2, or 3.
(continued)
BFN-UNIT 3 3.7-10 Amendment No. 244
..... t...........
CREV System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E. Two CREV subsystems r
NOTE in o p e ra b le
- i*,i
........ ~~~~~~~~.............
iiii N i~ ii ii ii* i! ! !.................
- ::i:
- i** i~:* :* i:i:
ii]..,*.*
,* *.**,*.*-**.'x.;.*,
.:...+..:.....'..:..+....
...: + + :..:+.:......................
inoperable 009""."~3 s o ALTEi :IO~NSor:i.
during
...,td
.l
...! :..: *i.i i.i.
....i E.*i1 Initiate action to suspend Immediately OPDRVs.
BFN-UNIT 3 3.7-11 Amendment No. 244 ept.....
m b r
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued) n.............t.......
ONG0 (continued)
BFN-UNIT 3 5.0-15 Amendment No. 24.
N.....mb............
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued) d 'i-.
Once every 24 months demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below at the system flowrate specified below, + 10%:
ESF Ventilation System SGT System CREV System (inches water) 7 6
4.......
ec.
Once every 24 months demonstrate that the heaters for the SGT System dissipate > 40 kW when tested in accordance with ANSI N510-1975.
5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained downstream of the offgas recombiners, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The program shall include:
- a. The limits for concentrations of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and (continued)
BFN-UNIT 3 5.0-16 Amendment No.
..........., 1.
Flowrate (cfm) 9000 3000
ENCLOSURE3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM MARKED PAGES - UFSAR AFFECTED PAGE LIST The following pages have been revised. On the affected pages the revised portions have been highlighted. A line has been drawn through the deleted text and a double underline for new or revised text.
UFSAR Pages 3.8-1 through 3.8-10 See Attached
3.8 STANDBY LIQUID CONTROL SYSTEM 3.8.1 Safety Objective The safety objective of the Standby Liquid Control System is to provide a backup method, which is independent of the control rods, to make the reactor subcritical over its full range of operating conditions, anid t d srt tn' Making the reactor subcritical is essential to permit the nuclear system to cool to the point where corrective actions can be carried out. Th~
after~~~~
~~~..
l...f..l.t..c......itan.
3.8.2 Safety Desi..n Basis
- 1.
Backup capability for reactivity control shall be provided, independent of normal reactivity control provisions in the nuclear reactor, to shut down the reactor if the normal control is impaired so that cold shutdown (MODE 4) cannot be obtained with control rods alone.
- 2.
The backup system shall have the capacity for controlling the reactivity difference between the steady-state rated operating condition of the reactor and the cold shutdown condition (MODE 4), including shutdown margin, to assure complete shutdown from the most reactive condition at any time in the core life.
- 3.
The time required for actuation and effectiveness of the backup reactivity control shall be consistent with the nuclear reactivity rate of change predicted between rated operating and cold shutdown conditions (MODE 4). A scram of the reactor or operational control of fast reactivity transients is not specified to be accomplished by this system.
- 4.
Means shall be provided by which the functional performance capability of the ba,*::k-G.o:*,.*tr:*:*
oti~
system components can be verified periodically under conditions approaching actual use requirements. Demineralized water, rather than the actual neutron absorber solution, is injected into the reactor to test the operation of all components of the redundant control system.
- 5.
The neutron absorber shall be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage, dilution, or imperfect mixing.
- 6.
The system shall be reliable to a degree consistent with its role as a special safety system.
3.8-1
- 7.
The possibility of unintentional or accidental shutdown of the reactor by this system shall be minimized.
3.8.3 Description (Fiqures 3.8-1, 3.8-2. 3.8-3. 3.8-5. and 3.8-6)
The Standby Liquid Control System is manually initiated from the Main Control Room to pump a boron neutron absorber solution into the reactor ift*
I.
the operator determines the reactor cannot be shut down or kept shut down with the control rods,,
f.l.........r.......
The Standby Liquid Control System is required o to shut down the reactor at a steady rate within the capacity of the shutdown cooling systems and to keep the reactor from going critical again as it cools.
The Standby Liquid Control System is needed dply in the improbable event that not enough control rods can be inserted in the reactor core to accomplish subcriticality in the normal manner.
The system consists of a boron solution tank, a test water tank, two positive-displacement pumps, two explosive-actuated valves, and associated local valves and controls. They are mounted in the Reactor Building outside the primary containment. The liquid is piped into the reactor vessel via the differential pressure and liquid control line and discharged near the bottom of the core lower support plate through a standpipe so it mixes with the cooling water rising through the core (see Sections 4.2, "Reactor Vessel and Appurtenances Mechanical Design," and 3.3, "Reactor Vessel Internals Mechanical Design").
The Boron-1 0 isotope absorbs thermal neutrons and thereby terminates the nuclear 3.8-2
fission chain reaction in the uranium fuel.
The specified neutron absorber solution is enriched sodium pentaborate (Na 2B, 0 016-10H20). It consists of a mixture of borax, enriched boric acid, and demineralized water prepared in accordance with approved plant procedures to ensure the proper volume and enriched sodium pentaborate concentration is present in the standby liquid control tank. A sparger is provided in the tank for mixing, using air. To prevent system plugging, the tank outlet is raised above the bottom of the tank and is fitted with a strainer.
At all times when it is possible to make the reactor critical, the configuration of the Standby Liquid Control System shall satisfy the following equation:
(C
)(Q)(
E)
Ž1.0 (13 WT%) (86GPM)
(19.8ATOM%)
C = sodium pentaborate solution weight percent concentration Q = SLCS pump flow rate in gpm E = Boron-1 0 atom percent enrichment in the sodium pentaborate solution The soluvtion con edntrton isontormll limited oto pa mainuo t
.e weight pfercent toA prineclude snwntdu preiptarteiontof the seodium pentboae. 1isThem sturan~ttion th
- o.
th d mixed yEso fl.
o w isr40.hic.
.h providsa 1 the.....r..
mu arogit gtbeow the lowesto temeatur~ e predicted for tnhectLoS equoipmen area. TOan haing detomponednts prtofvtde backu assuanc thatthe odium prentaboratei solutioni tmpeintained winl thevaer tal k aso 0°h soefldium thpehnicaloratesolution cncntato Th ouincnetainis nrallyolmied to abmxiumof9.2 weight percent poiethcnetrinadtmeatueof h
preludinae uwathnthed lmt preciitation of the soiutentcaboael h
s atufiratos ighon o
temperature, ofithe 9.oprclwlqi eent r
solto s hihe petrovidses an 1alathrma margi cnto beowm Thek loestvemperdiatureas predicted for the Sconro equimetaa. Tn etn Ecmpontsiprovidespbackmnump assuraned thaijet the sodium iopentborathe solutiorn50t 15mntemprtr willnevroxfalltelow 50 The) sopedium pentheaborateo solution conctentration isallowed rator bess elaight operetin provided
.The onenraio and sytempderatreofth 125emiuteis (50pi.Tetorlifvle r
e tapproximately 50gm1dpedn4nth2mutofslto pin thetak exceed the reactor operating pressure by a sufficient margin to avoid valve leakage.
3.8-3
To prevent bypass flow from one pump in case of relief valve failure in the line from the other pump, a check valve is installed downstream of each relief valve line in the pump discharge pipe.
A bladder-type pneumatic-hydraulic accumulator is installed on the piping near each relief valve to dampen pulsations from the pumps to protect the system.
The two explosive-actuated injection valves provide high assurance of opening when needed and ensure that the boron solution will not leak into the reactor even when the pumps are being tested. The valves have a demonstrated firing reliability in excess of 99.99 percent. Each explosive valve is closed by a plug in the inlet chamber. The plug is circumscribed with a deep groove so the end will readily shear off when pushed with the valve plunger. This opens the inlet hole through the plug. The sheared end is pushed out of the way in the chamber and is shaped so it will not block the ports after release.
The shearing plunger is actuated by an explosive charge with dual ignition primers inserted in the side chamber of the valve. Ignition circuit continuity is monitored by a trickle current, and an alarm occurs in the control room if either circuit opens. Indicator lights show which primer circuit is opened. To service a valve after firing, a 6-inch length of pipe (spool piece) must be removed immediately upstream of the valve to gain access to the shear plug.
The Standby Liquid Control System is actuated by a five-position spring return to "normal" keylock switch located on the control room console. The keylock feature ensures that switching from the "stop" position is a deliberate act (safety design basis 7). Momentarily placing the switch to either "start A" or "start B" position starts the respective injection pump, opens both explosive valves, and closes the Reactor Water Cleanup System isolation valves to prevent loss or dilution of the boron solution.
A green light in the control room indicates that power is available to the pump motor contactor, but that the contactor is open (pump not running). A red light indicates the contactor is closed (pump running). A white light indicates that the motor has tripped or the local handswitch is in the test position.
A red light beside the switch turns on when liquid is flowing through an orifice flow switch downstream of the explosive valves. If the flow light or pump lights indicate that the liquid may not be flowing, the operator can immediately turn the switch to the other side, which actuates the alternate pump. Crosspiping and check valves assure a flow path through either pump and either explosive valve. The chosen pump will start even though its local switch at the pump is in the "stop" position for test or maintenance.
Pump discharge pressure indication is also provided in the control room.
Equipment drains and tank overflow are piped not to the waste system but to separate containers (such as 55-gallon drums) that can be removed and disposed of 3.8-4
independently to prevent any trace of the boron solution from inadvertently reaching the reactor.
Instrumentation is provided locally at the standby liquid control tank consisting of solution temperature indication and control, tank level, and heater status.
Instrumentation and control logic is presented in Figures 3.8-4 and 3.8-7, Mechanical Logic Diagram.
3.8.4 Safety Evaluation The Standby Liquid Control System is a special safety system not required for normal plant operation, and is never expected to be needed for reactor shutdown because of the large number of control rods available to shut down the reactor.
TO assure iithe availabilty' othSadb;Lqid CotoSyttost fth The system is designed to make the reactor subcritical from rated power to a cold shutdown (MODE 4) at any time in core life. The reactivity compensation provided will reduce reactor power from rated to the after-heat level and allow cooling the nuclear system to normal temperature with the control rods remaining withdrawn in the rated power pattern. It includes the reactivity gains due to complete decay of the rated power xenon inventory. It also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced Doppler effect in uranium, reduction of neutron leakage from the boiling to cold condition, and decreasing control rod worth as the moderator cools. A licensing analysis is performed each cycle to verify adequate SLCS shutdown capacity. The analysis assumes the specified minimum final concentration of boron in the reactor core and allows for calculational uncertainties. The SLCS shutdown capacity is reported in Appendix N.
The specified minimum average concentration of natural boron in the reactor to provide the specified shutdown margin, after operation of the Standby Liquid Control System, is 660 ppm (parts per million). The minimum quantity of sodium pentaborate to be injected into the reactor is calculated based on the required 660 ppm average concentration in the reactor coolant, Boron-1 0 enrichment, the quantity of reactor coolant in the reactor vessel, recirculation loops, and the entire RHR System in the shutdown cooling mode, at 70OF and reactor normal water level. The result is increased by 25 percent to allow for imperfect mixing, leakage, and volume in other piping connected to the reactor. This minimum concentration is achieved by preparing the solution as defined in paragraph 3.8.3 and maintaining it above saturation temperature. This satisfies safety design basis 5.
3.8-5
Cooldown of the nuclear system will take several hours, at a minimum, to remove the thermal energy stored in the reactor, cooling water, and associated equipment, and to remove most of the radioactive decay heat. The controlled limit for the reactor coolant temperature cooldown is 1000 F per hour. Normal operating temperature is about 5500 F. Usually, shutting down the plant with the main condenser and various shutdown cooling systems will take 10 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the reactor vessel is opened, and much longer to reach room temperature (70 0F). Room temperature is the condition of maximum reactivity and, therefore, the condition which requires the maximum boron concentration. Thus safety design basis 2 is met.
The specified boron injection rate is limited to the range of 7 to 40 ppm per minute change of boron concentration in the reactor pressure vessel and recirculation loop piping water volumes. The lower rate ensures that the boron is injected into the reactor in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, which is considerably faster than the cooldown rate. The upper limit injection rate insures that there is sufficient mixing such that the boron does not recirculate through the core in uneven concentrations which could possibly cause asymmetric power oscillations in the core. This satisfies safety design basis 3.
3.84.2 Sup-r~es'sion Foot oH Cnrd fobr p.stýLQQA.events that nolefldmaeto maintain the ýsuorsinol Ha or above 7.0. The rediolonicafl. dsanlssfrteDALAassu mes doncenftratons Qf i dih. so.is 6
nitentwt a suoeso pg Ha r abveT (ie.. r e.-
evo t i dn of iodn t t
.e
.o tim n o o r is
.o
- e.
3.8-6
.3 8.. 3 C.m m on Ia t
i......
The Standby Liquid Control System is designed as a Class I system for withstanding the specified earthquake loadings (see Appendix C). Nonprocess equipment such as the test tank is designed as Class I1. The system piping and equipment are designed, installed, and tested in accordance with USAS B31.1.0,Section I.
The Standby Liquid Control System is not required to be designed to meet the single failure criterion because it serves as a backup to the control rods. System reliability is enhanced by providing redundancy of pumps and valves. Hence, redundancy is not required for the tank heater or the heating cable.
The Standby Liquid Control System is required to be operable in the event of a station power failure so the pumps, valves, and controls are powered from the standby AC power supply in the absence of normal power. The pumps and valves are powered and controlled from separate buses and circuits so that a single failure will not prevent system operation. The essential instruments and lights are powered from the 120-V AC instrument power supply.
The Standby Liquid Control System and pumps have sufficient pressure margin, up to the system relief valve nominal setting of 1425 psig, to assure solution injection into the reactor at a pressure of at least three percent above the lowest setpoint of the main steam relief valves (1140 psig pre-uprated; 1174 psig uprated). The nuclear system is protected from overpressurization during operation of the Standby Liquid Control System positive displacements pumps by the nuclear system main steam relief valves.
Only one of the two standby liquid control pumps is needed for proper system 3.8-7
operation. If one pump is inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. The system pumps are powered by a diesel backed source and are not load shed. The period during which one redundant component upstream of the explosive valves may be out of operation will be consistent with the very small probability of failure of both the control rod shutdown capability and the alternate component in the Standby Liquid Control System, together with the fact that nuclear system cooldown takes 10 or more hours while liquid control solution injection takes about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Fo heSanb iid
~.
availablefortystinsg andrestorng the Standby Liquid Control System ti not o
trible aItlcanbe fsting an the Standby Liquid Control System stf to oerb 3.8.5 Inspection and Testinpq Operational testing of the Standby Liquid Control System is performed in at least two parts to avoid injecting boron into the reactor inadvertently. By opening two closed valves to the solution tank, the boron solution may be recirculated by turning on either pump with its local switch. With the valves to and from the solution tank closed and the three valves opened to and from the test tank, the demineralized water in the test tank can be recirculated by turning on either pump locally. After pumping boron solution, demineralized water is pumped to flush out the pumps and pipes. Functional testing of the injection portion of the system is accomplished by closing the open valve from the solution tank, opening the closed valve from the test tank, and actuating the switch in the control room to either the A or B circuit. This starts one pump and ignites one of the explosive actuated injection valves to open. The lights and alarms in the control room indicate that the system is functioning. This satisfies safety design basis 4.
After the functional test, the affected injection valve and explosive charge must be replaced and all the valves returned to their normal positions as indicated in Figures 3.8-1, 3.8-2, 3.8-3, 3.8-5, and 3.8-6.
By closing a local normally open valve to the reactor in the containment, leakage through the injection valves can be detected at a test connection in the line between the containment isolation check valves. (A position indicator light in the control room indicates when the local valve is full open and ready for operation.) Leakage from the reactor through the first check valve can be detected by opening the same test connection whenever the reactor is pressurized.
3.8-8
The test tank contains sufficient demineralized water for testing pump operation.
Demineralized water from the makeup or condensate storage system is available at 30 gpm for refilling or flushing the system.
Should the boron solution ever be injected into the reactor, either intentionally or inadvertently, then after making certain that the normal reactivity controls will keep the reactor subcritical, the boron is removed from the reactor coolant system by flushing for gross dilution followed by operation of the reactor cleanup system. There is practically no effect on reactor operations when the boron concentration has been reduced below about 50 ppm.
The sodium pentaborate solution weight percent in the SLCS storage tank is periodically determined by titration or equivalent chemical analysis. The Boron-1 0 isotopic atom percent concentration of the solution is also determined periodically, utilizing mass spectrometry or equivalent technology.
The gas pressure in the two accumulators is measured periodically to detect leakage.
A pressure gauge and portable nitrogen supply are required to test and recharge the accumulators.
3.8-9
ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM SAFETY ASSESSMENT
BFN Alternate Source Term Safety Assessment TABLE OF CONTENTS
- 1. INTRODUCTION.................................................................................................
1 1.1 Evaluation Overview and Objective..........................................................
1 1.2 Major Aspects of AST Analyses......................................................................
1 1.3 S u m m a ry...................................................................................................
.. 2
- 2. EVALUATION.....................................................................................................
3 2.1 Scope......................................................................................................
3 2.1.1 Accident Radiological Consequence Analyses.............................
3 2.1.2 Suppression Pool PH Control......................................................
3 2.1.3 Main Steam Line Break Accident Puff Release Dispersion F a cto r......................................................................................
... 4 2.1.4 NUREG-0737 Evaluation.............................................................
4 2.1.5 Environmental Qualification........................................................
4 2.2 Method of Evaluation................................................................................
4 2.2.1 Accident Radiological Consequence Analyses.............................
4 2.2.2 Suppression Pool PH Control Calculations..................................
7 2.2.3 Main Steam Line Break Accident Instantaneous Ground Level Puff Release Dispersion Factor.........................................
7 2.2.4 NUREG-0737 Evaluation............................................................
8 2.3 Inputs and Assumptions.............................................................................
8 2.3.1 Accident Radiological Consequence Analyses.............................
8 2.3.1.1 LOCA Inputs and Assumptions....................................
10 2.3.1.2 Main Steam Line Break Accident Inputs and Assumptions...................................................................
13 2.3.1.3 Refueling Accident Inputs and Assumptions................. 14 2.3.1.4 Control Rod Drop Accident Inputs and Assumptions........ 15 2.3.2 Suppression Pool pH Control......................................................
15 2.3.3 Main Steam Line Break Accident Puff Release Dispersion Factor. 17 2.3.4 NUREG-0737 Evaluation.............................................................
17 3. R E S U L T S..................................................................................................................
3 5 3.1 Evaluation Results...................................................................................
35 3.1.1 Accident Radiological Consequence Analyses...........................
35 3.1.1.1 LOCA.............................................................................
35 3.1.1.2 Main Steam Line Break Accident.................................. 36 3.1.1.3 Refueling Accident........................................................
36 3.1.1.4 Control Rod Drop Accident............................................
36 3.1.2 Suppression Pool PH Control......................................................
36 3.1.3 Main Steam Line Break Accident Instantaneous Ground Level Puff Release Dispersion Factor.......................................
37 3.1.4 NUREG-0737 Evaluation..........................................................
37 3.2 Summary..................................................................................................
37 ii
BFN Alternate Source Term Safety Assessment 4. R E F E R E N C E S..........................................................................................................
4 2 iii
BFN Alternate Source Term Safety Assessment LIST OF TABLES Table Table 2-1 Table 2-2 Table 2-3 Table 2-4 Table 2-5 Table 2-6 Table 2-7 Table 2-8 Table 2-9 Table 2-10 Table 2-11 Table 2-12 Table 2-13 Table 2-14 Table 2-15 Table 2-16 Table 2-17 Table 3-1 Table 3-2 Table 3-3 Table 3-4 Table 3-5 Title Computer Codes Used in AST Fission Product Inventory X/Q Values for Radiological Dose Calculations Top of Stack Releases X/Q Values for Radiological Dose Calculations Base of Stack Releases X/Q Values for Radiological Dose Calculations Refueling Vent Releases X/Q Values for Radiological Dose Calculations Turbine Building Exhaust Releases X/Q Values for Radiological Dose Calculations Turbine Building Roof Ventilator Releases Fuel Data LOCA Release Fractions as Release Rates Over the Duration CREV/SGT Functions Modeled in Dose Analyses Accident Radiological Consequence Analyses Inputs LOCA Inputs Main Steam Line Break Accident Inputs Refueling Accident Inputs Control Rod Drop Accident Inputs Suppression Pool pH Control Inputs Main Steam Line Break Accident Puff Release X/Q Inputs LOCA Radiological Consequence Analysis Main Steam Line Break Accident Radiological Consequence Analysis Refueling Accident Radiological Consequence Analysis Control Rod Drop Accident Radiological Consequence Analysis Main Steam Line Break Accident Instantaneous Ground Level Puff Release X/Q Value iv BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment LIST OF FIGURES Figure Figure 2-1 Figure 2-2 Figure 3-1 Title LOCA Transport Model Control Rod Drop Accident Transport Model Suppression Pool pH Response V
BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS pCi/gm micro-curies per gram
" HG inches of mercury AST Alternative Source Term BFN Browns Ferry Nuclear Plant BLEU Blended Low Enriched Uranium BWR Boiling Water Reactor CAD Containment Air Dilution cfm cubic feet per minute CREV Control Room Emergency Ventilation Csl Cesium Iodine DBA Design Basis Accident DE Dose Equivalent EAB Exclusion Area Boundary ECCS Emergency Core Cooling System EPU Extended Power Uprate ft feet GDC General Design Criterion GE General Electric gpm gallons per minute H+
Hydrogen Ion HEPA High Efficiency Particulate Air HNO 3 Nitric Acid hrs hours HWWV Hardened Wetwell Vent in inch Ibm pounds-mass LOCA Loss of Coolant Accident LPZ Low Population Zone m/s meters per second vi BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment ACRONYMS AND ABBREVIATIONS MSIV Main Steam Isolation Valve MSL Main Steam Line MWt Megawatt thermal OH-Hydroxyl Ions PCIS Primary Containment Isolation Signal pH Hydrogen Ion Concentration psid pounds per square inch differential rem roentgen equivalent man RG Regulatory Guide scfh standard cubic feet per hour SGT Standby Gas Treatment System secs seconds SER Safety Evaluation Report SLC Standby Liquid Control System SPB Sodium Pentaborate SR Surveillance Requirement SRP Standard Review Plan TEDE Total Effective Dose Equivalent TS Technical Specification TVA Tennessee Valley Authority UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink X/Q Atmospheric Dispersion Factor vii BFN Alternate Source Term Safety Assessment
- 1.
INTRODUCTION 1.1 Evaluation Overview and Objective The objective of this safety assessment is to document BFN implementation of the Alternative Source Term (AST). The implementation of AST is governed by 10 CFR 50.67, the guidelines of the Standard Review Plan (SRP) Section 15.0.1 (Reference 1), and Regulatory Guide (RG) 1.183 (Reference 2).
BFN has elected to perform a full scope implementation of the AST as defined in RG 1.183. The implementation consists of the following:
- 1.
Identification of the core source term based on plant specific analysis of core fission product inventory.
- 2.
Determination of the release fractions for the four Updated Final Safety Analysis Report (UFSAR) Boiling Water Reactor (BWR) Design Basis Accidents (DBAs) that could potentially result in control room and offsite doses. These are the loss of coolant accident (LOCA), the main steam line break accident, the refueling accident, and the control rod drop accident.
- 3.
Calculation of fission product deposition rates and removal efficiencies.
- 4.
Calculation of offsite and control room personnel Total Effective Dose Equivalent (TEDE).
- 5.
Evaluation of suppression pool pH to ensure that the particulate iodine deposited into the suppression pool during a DBA LOCA does not re-evolve and become airborne as elemental iodine.
- 6.
Calculation of a new control room atmospheric dispersion factor (X/Q) for a main steam line break accident instantaneous ground level puff release.
- 7.
Evaluation of other related design and licensing bases such as NUREG-0737 (Reference 3).
The radiological dose analyses have been performed assuming reactor operation at a thermal power of 4031 MWt (102% of 3952 MWt). This results in a conservative estimate of fission product releases for operation at current licensed power of 3458 MWt.
1.2 Major Aspects of AST Analyses Implementation of AST includes changes to the methodology presently used at BFN. These include:
- 1.
Development of a bounding plant-specific core fission product inventory.
1 BFN Alternate Source Term Safety Assessment
- 2.
Analysis of a new X/Q for an instantaneous ground level puff release to the atmosphere for the main steam line break accident.
- 3.
No credit is taken for Control Room Emergency Ventilation (CREV)
System or standby gas treatment (SGT) System charcoal adsorption for any DBA.
- 4.
No credit is taken for CREV System or SGT System HEPA filter particulate removal for any DBA except LOCA.
- 5.
New requirements were developed for post-LOCA standby liquid control (SLC) System operation for suppression pool pH control along with calculation of minimum sodium penteborate quantity requirements.
1.3 Summary Implementation of the AST as the plant radiological consequence analyses licensing basis requires a license amendment per the requirements of 10 CFR 50.67. The enclosed AST analyses demonstrate the offsite and control room post-accident radiological doses remain within regulatory limits.
2 BFN Alternate Source Term Safety Assessment
- 2.
EVALUATION 2.1 Scope 2.1.1 Accident Radiological Consequence Analyses The DBA accident analyses documented in Chapter 14 of the BFN UFSAR (Reference 4) that could potentially result in control room and offsite doses were addressed using methods and input assumptions consistent with the AST. The following DBAs were addressed:
LOCA, UFSAR Section 14.6.3 Main Steam Line Break Accident, UFSAR Section 14.6.5 Refueling Accident, UFSAR Section 14.6.4 Control Rod Drop Accident, UFSAR Section 14.6.2 The analysis was performed per RG 1.183. The results were evaluated to confirm compliance with the acceptance criteria presented in 10 CFR 50.67 and GDC 19 of 10 CFR 50, Appendix A. Computer codes used in the DBA analyses results are listed in Table 2-1.
The AST control room dose analyses are applicable for all three unit control rooms. The Unit 1 and 2 control rooms are shared in a common room with Unit 1 at one end and Unit 2 at the other. The Unit 3 control room, though separated from the Units 1 and 2 control room, is part of the same control bay habitability zone.
The refuel zone is a common three-unit zone consequently; the refueling accident radiological consequence analysis is the only analyses applicable to all three units. Since the Unit 1 is in an extended shutdown, the remaining three DBA radiological consequence analyses have not been performed for Unit 1. However, TVA expects that the results will be similar to Units 2 and 3.
2.1.2 Suppression Pool PH Control A calculation was performed to evaluate the suppression pool pH in the event of a DBA LOCA. The objective of the analysis was to demonstrate that the suppression pool pH remains at or above 7.0, thus ensuring that the particulate iodine (cesium iodide - Csl) deposited into the suppression pool during this event does not re-evolve and become airborne as elemental iodine. The analysis credits the pH buffering effect of sodium pentaborate introduced into the suppression pool post-LOCA by SLC operation to maintain the pH above 7.0.
3 BFN Alternate Source Term Safety Assessment
2.1.3 Main Steam Line Break Accident Puff Release Dispersion Factor A new control room X/Q was determined for use in the main steam line break accident analysis. This X/Q reflects an instantaneous ground level puff release to the atmosphere in accordance with Regulatory Guide 1.183 Appendix D.
2.1.4 NUREG-0737 Evaluation An evaluation was performed to identify potential impacts of applying AST methodologies on compliance with NUREG-0737 requirements. This evaluation included the following:
Revision of the current radiological dose analyses for post-accident vital area access and post-accident sampling (NUREG-0737, Item ll.B.2 and Item ll.B.3),
Revision of the current radiological dose analyses for the post accident containment high range radiation monitors (NUREG-0737, Item II.F.1),
Revision of control room post-accident radiological dose analyses for emergency support facility upgrades and control room habitability (NUREG-0737, Items III.A.1.2 and III.D.3.4), and Consideration of post-accident sources of radiation and radioactivity outside the primary containment in terms of impact on dose analysis related to integrity of systems outside containment likely to contain radioactive material (NUREG-0737, Item III.D.1.1).
2.1.5 Environmental Qualification The radiation doses used for the environmental qualification analyses at the original licensed thermal power conditions were calculated using source terms determined by TID-14844 (Reference 5) methodology. The radiation doses used for the environmental qualification analyses at both current licensed thermal power and Extended Power Uprate (EPU) conditions are adjusted upward from the original values based on the determined source terms of the ORIGEN computer code for the respective power level.
2.2 Method of Evaluation 2.2.1 Accident Radiological Consequence Analyses Analyses were prepared for the simulation of the radionuclide release, transport, removal, and doses estimated for the postulated accidents listed in Section 2.1.1.
4 BFN Alternate Source Term Safety Assessment
The ORIGEN code (Reference 6) was used to calculate plant-specific fission product inventories which bound the effect of two-year fuel cycles, power operation at EPU conditions (4031 MWt (102% of 3952 MWt)), and using current and anticipated fuel designs. The fission product inventory for General Electric (GE)-14, Framatome Atrium-10 fuel, and Framatome Blended Low Enriched Uranium (BLEU) fuel designs were evaluated.
Bounding values of fission product activity were determined for each radionuclide in the DBA radiological analyses. Fission product activities were calculated for immediately after shutdown and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following shutdown. The values are shown in Table 2-2.
The RADTRAD computer code Version 3.02(a) (Reference 7) was used for the DBA dose calculations. The computer code STARDOSE (Reference 8) was used to check the RADTRAD results. The RADTRAD and STARDOSE programs are radiological consequence analysis codes used to determine post-accident doses at offsite and control room locations. The STARDOSE code is the proprietary property of Polestar Applied Technology, Inc., and the NRC has previously reviewed results obtained from the application of this code.
The existing UFSAR X/Q values were developed prior to and used in support of the license amendment request (References 9 and 10) for increased main steam isolation valve (MSIV) leakage rate limits. Control room X/Q values for the base of the stack releases were calculated using the computer code ARCON96 (Reference 11). For sites such as BFN with control room ventilation intakes that are close to the base of tall stacks, ARCON96 underpredicts the X/Q values for top of stack releases; therefore, top of stack releases to the control room intakes were evaluated using the methods of Regulatory Guides 1.111 (Reference 12) and 1.145 (Reference 13). The X/Q values associated with top of stack, base of stack, and turbine building releases were reviewed by the NRC in the Safety Evaluation for Amendments 263 and 223 for BFN Units 2 and 3, respectively (Reference 14). The existing X/Q values applicable to the time periods, distances, and geometric relationships are shown in Tables 2-3 through 2-7. Existing values for X/Q were used for AST radiological dose analyses except for the establishment of a new control room X/Q value associated with an instantaneous ground level puff release for the case of a main steam line break accident (see Section 2.2.3).
The post-LOCA shine dose to personnel in the control room includes the radiation shine from the secondary containment airborne activity and gamma dose from Core Spray System piping, which is in close proximity to the control building. An evaluation was performed of the existing TID-14844 analysis to determine application values for AST. For radiation from the Core Spray System piping, a comparison of gamma radiation plots from the 5
BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term suppression pool water was performed for high energy photons to determine similarity of shapes for the TID-14844 source term and the AST source term.
For the secondary containment airborne shine dose, a shine dose multiplier for AST airborne radioiodines was developed to enable direct comparison of the TID-14844 and the AST shine dose. To support this comparison, the activity for TID-1 4844 was increased to account for the increase in power level. The resulting comparison of several key nuclides found that the AST 1-131 and 1-133 activities in the reactor building are approximately a factor of 3 lower at 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and, a factor of 30 lower at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> compared to the TID-14844 levels at the same times. Considering the highest multiplier for the AST radionuclides (used to account for the activity other than iodine, especially for cesium) for I to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the effective iodine activity airborne in the reactor building for AST would be about the same before 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and about a factor of 10 lower at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> compared to TID-14844. For noble gases, the AST activities are about a factor of two lower than the TID-1 4844 source term at two hours, and by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, they are about the same.
The evaluation established that the integrated gamma dose from Core Spray System piping is slightly higher than previous over the 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> duration of the accident for the AST. However, only about 25 percent of the total 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> control room shine dose is due to the Core Spray System piping contribution. The control room shine dose from airborne activity in the secondary containment will be substantially reduced for the AST as compared to the TID-14844 source term. Therefore, the existing integrated control room shine dose, even if increased by the power ratio of EPU, is acceptable for a combination of EPU operation and AST application. This evaluation was checked using the MicroShield code, Version 5.03 (Reference 15). MicroShield is a point kernel integration code used for general purpose gamma shielding analysis. MicroShield has been used in safety-related applications by many nuclear plants in the United States. In this BFN application, it has been used as a means for design verification as an independent analysis.
For the main steam line break accident, radiation shine from the turbine building was conservatively handled assuming all released inventory is in the turbine building for two hours. Radiation shine from the airborne activity having escaped the turbine building is handled explicitly by the TVA computer code COROD. The calculation incorporates the control building dimensions and concrete roof (2.25 ft thick) in conjunction with the main steam line break accident released radioisotopes in a cloud above the control building.
6 Safety Assessment
2.2.2 Suppression Pool PH Control Calculations The calculation methodology for suppression pool pH control was based on the approach outlined in NUREG-1465 (Reference 16) and NUREG/CR-5950, (Reference 17). Specifically, credit was taken for sodium pentaborate (SPB) addition to the suppression pool water as a result of SLC operation. The pH of the suppression pool water was then calculated using the STARpH code (Reference 18). This same methodology and code for calculation of transient suppression pool pH (including the formation of acids by radiation effects on drywell components) was applied to the Hope Creek AST application (Reference 19).
Calculations were performed to verify sufficient SPB solution is available to maintain the suppression pool pH at or above 7.0 for 30 days post accident.
The design inputs were conservatively established to maximize the post LOCA production of acids and to minimize the post-LOCA production and/or addition of bases. Other design input values such as initial suppression pool volume and pH were selected to minimize the calculated pH. It was determined that the calculated required quantity of SPB was in excess of the current TS limit. Therefore, a change to TS 3.1.7, Standby Liquid Control System (SLC), is being proposed increase the required amount of SPB.
2.2.3 Main Steam Line Break Accident Instantaneous Ground Level Puff Release Dispersion Factor To meet RG 1.183 requirements for a main steam line break a new X/Q for a puff release was calculated. An instantaneous ground level puff release is assumed. The calculation of the main steam line break accident ground level puff release dispersion factor uses plant parameters for the main steam line break accident (e.g., mass of liquid-steam mixture released, timing of release, temperature of the liquid-steam mixture) to obtain the initial conditions of the released steam puff. The steam puff is treated as a "bubble" with a given transit time up to and across the control room intake.
Once introduced into the atmosphere, the steam bubble rises at a rate corresponding to the buoyancy force (resulting from the density difference between ambient air and hot steam) equaling the drag force resulting from the friction between the bubble mass and the surrounding air. Mixing of the steam with surrounding air reduces the bubble's buoyancy, but also increases dilution. Different bubble shapes and degrees of air entrainment are considered, and the worst case is used (i.e., minimum dilution). No credit is taken for concentration gradients within the rising bubble. In particular, no credit is taken for a vertical concentration gradient; (i.e., the concentration at the elevation of the control room air intake is assumed to be the same as that of the leading edge of the rising bubble).
The bubble is assumed to be released from the turbine building at a distance from the nearest control room intake that is exceeded by 90% of the potential 7
BFN Alternate Source Term Safety Assessment
release locations. No credit is taken for wind direction; (i.e., it is assumed that the centerline of the bubble trajectory always passes over one control room intake). The diameter of the bubble (even with substantial initial air entrainment) is less than the distance between the two air intakes The minimum dilution effect was quantified as a dilution factor of 0.25; (i.e., a factor of four decrease in the initial concentration of activity in the release).
This is an average value during the time of passage over the control room air intake.
2.2.4 NUREG-0737 Evaluation Post Accident Vital Area Access and Sampling - Post-accident personnel missions resulting in mission doses (including post-accident sampling) were identified. Plant calculations used in support of plant post-accident vital area access (prepared in accordance with the requirements of NUREG-0737, Item ll.B.2 and ll.B.3) were revised for impact by AST. The revisions considered the comparative radiation levels from AST and the existing TID-14844 methodology source terms (such as airborne activity in the reactor building and turbine building, and also as activity in the suppression pool water).
Post-Accident Radiation Monitor - Post-accident containment high range radiation monitoring calculations were revised for impact by AST (NUREG-0737, Item II.F.1).
Control Room Radiation Protection - The control room radiological dose impact of AST has been specifically calculated for each of the four DBAs analyzed for AST implementation (NUREG-0737, Item Ill. D.3.4).
Radioactive Sources Outside the Primary Containment - The DBA LOCA control room dose analysis, as well as that for offsite doses, includes the effects of coolant leakage outside the primary containment and (for the control room dose analyses only) the shine contribution from Core Spray System piping (NUREG -0737, Item 111.D.1.1).
2.3 Inputs and Assumptions 2.3.1 Accident Radiological Consequence Analyses For AST accident radiological consequences, analyses were performed for the four DBAs that could potentially result in control room and offsite doses.
These are the LOCA, main steam line break accident, refueling accident, and control rod drop accident.
Plant-specific fuel design parameters were used in the fission product and transuranic nuclide inventories for the accident analyses. Table 2-8 summarizes key fuel cycle parameters.
8 BFN Alternate Source Term Safety Assessment
The reactor core inventory of activity for the AST dose analyses is based on an average burn-up of 35 to 37 GWd/MT depending on the fuel type. For the control rod drop accident and refueling accident analyses, a RG 1.183 minimum core radial peaking factor of 1.5 was used. For the refueling accident analysis, the core isotopic inventory after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of decay was used.
The release source term is developed using the radionuclide isotopes listed in Table 2-2 and the release fractions from Table 1 of RG 1.183. The radionuclides that are included are those identified as being potentially important contributors to TEDE in NUREG/CR-6604 (Reference 7). Release fractions for LOCA as release rates are shown in Table 2-9.
No credit is taken for the adsorption of elemental iodine, organic iodine, or noble gases by charcoal in the SGT or CREV systems for any of the four DBAs. SGT and CREV system HEPA filters are credited for removal of 90 percent of the particulate activity in the LOCA analysis. HEPA filter removal of particulates activity was not credited for the remaining three DBA analyses. A comparison of CREV/SGT functions modeled in the AST radiological dose analyses is presented in Table 2-10.
The CREV System is automatically initiated by a Group 6 primary containment isolation signal (PCIS), by high radiation at the control bay air intakes, or it can be manually initiated by the control room operators. The PClS Group 6 trip signal is initiated by reactor vessel low water level, drywell high pressure, or reactor building ventilation high radiation. For a LOCA, the PCIS Group 6 initiation of CREV will occur significantly prior to the control room experiencing conditions which would result in excessive doses to the control room operators and, hence, significantly prior to an initiation on control bay air intake high radiation. The adequacy of the control bay radiation monitoring setpoint was reviewed as part of the AST NUREG-0737 evaluation.
In accordance with Standard Review Plan Section 6.4 (Reference 25), the doses due to airborne activity released from the turbine building may be divided by a factor of two because the CREV intakes are on opposite sides of the building and the makeup flow is equal from each intake.
The BFN Emergency Core Cooling Systems (ECCS) are designed, maintained, and tested to minimize the radiological consequences following a postulated DBA. The AST analyses inputs and assumptions are consistent with the design and licensing for these systems.
An assumed unfiltered inleakage rate of 3717 cubic feet per minute into the control room habitability zone was used. This inleakage rate was acknowledged by NRC in the Safety Evaluation for Amendments 263 and 223 for BFN Units 2 and 3, respectively (Reference 14).
9 BFN Alternate Source Term Safety Assessment
The standard breathing rates specified in RG 1.183 have been used. The key accident radiological consequence analyses inputs are summarized in Table 2-11.
2.3.1.1 LOCA Inputs and Assumptions The key inputs used in this analysis are included in Table 2-12. These inputs and assumptions fall into three categories: Radionuclide Release Inputs, Radionuclide Transport Inputs, and Radionuclide Removal Inputs.
LOCA Release Inputs The BFN TSs specify a maximum allowable primary containment leakage rate of two percent primary containment air weight per day. This leakage rate was assumed in the AST analyses. ECCS leakage was considered in accordance with the guidance from SRP 15.6.5, Appendix B (Reference 20). A five gallon per minute (gpm) leak rate into the reactor building was used starting at the onset of the event. This leakage is more than the operational ECCS leakage for BFN (approximately I gpm). The plant TSs total limit allowable MSIV leakage of 150 scfh (maximum of 100 scfh in any one line) was assumed by the analyses. The analyses conservatively assume no reduction in these leak rates over the 30 day duration of the dose calculation.
Primary Containment leakage via the hardened wetwell vent (HWWV) bypasses secondary containment and is released unfiltered to the atmosphere via the top of the stack. A leakage of 10 scfh is conservatively assumed to begin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the beginning of the event.
This delay is based upon the leakage rate and the large volume of the HWWV piping between the drywell and the stack.
LOCA Transport Inputs All three trains of SGT are conservatively assumed to be in operation at the beginning of the accident. This maximizes the release from secondary containment. If only two of the SGT trains are in operation, a short time period exists at the start of the accident during which the secondary containment can become pressurized relative to the outside environment. However, negative pressure would be re-established in secondary containment prior to the gap release at two minutes specified by RG 1.183. Accordingly, three train operation of SGT is the conservative case. The reactor building pressure is negative throughout the RG 1.183 release phases and the primary containment leakage (with the exception of the MSIV leakage and the leakage through the HWWVV after eight hours) is assumed to be collected by SGT and directed to the stack. A portion of the stack flow (10 scfm) is assumed to leak through the stack backdraft isolation dampers and released as a ground level 10 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment release at the base of the stack. This amount of leakage is within the bounds of procedural controls.
Since the main steam lines and the main condenser are seismically rugged, and are assumed to remain intact, the MSIV leakage eventually collects in the main condenser (except for a small portion that is assured to bypass the main condenser). The LOCA analysis also assumes that one of the four inboard MSIVs fails to close (this postulated single failure results in the worst case dose consequences). Therefore, three of the steam lines have a closed space between the inboard and outboard MSIVs. The piping volume between the outboard MSIVs and the assorted valving downstream (i.e., main turbine stop valves, main turbine bypass valves, reactor feed pump high pressure steam stop valves, etc.) also comprises a large, closed space. In each of the three steam lines that are fully isolated, a well mixed control volume is defined in the space between the closed MSIVs as well as in the space downstream of the outboard MSIVs.
Only the control volumes in the horizontal portions of this main steam piping are credited in the analyses for activity disposition. The space down stream of the MSIV in the faulted steam line (the one with only the outboard MSIV closed) is credited with an isolated control volume only in the space from the outboard MSIV to the point where the drain line pathway to the main condenser connects to the steam line. This volume is consistent with others in that it is made up of horizontal piping also.
For conservatism, a maximum MSIV leakage per line of 100 scfh is assumed to exist in the faulted line. One of the fully isolated lines is assumed to leak at 50 scfh, while the other two are assumed to be leak tight. This set of assumptions minimizes credit for retention in the steam lines.
The pressure in the space between the closed MSIVs is assumed to be that of the containment, but the temperature is assumed to be the normal operating conditions of the steam line. In the steam line outboard of the MSIVs, the pressure is assumed to be atmospheric, the temperature is also assumed to be the normal operating. The condenser is assumed to be at standard conditions. MSIV leakage at the test pressure is converted into volumetric flow rates based upon post-LOCA drywell temperature and pressure.
The MSIV leakage from the main condenser is assumed to be released directly to the environment as a turbine building release with no credit for turbine building hold-up.
The control room would automatically isolate and the CREV is automatically initiated at the onset of the accident due to high drywell 11 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment pressure or low reactor level trip signals. However, for conservatism, a 10 minute delay in CREV initiation is assumed in the dose analysis. A schematic of the transport model is provided in Figure 2-1.
LOCA Removal Inputs LOCA activity release is partially removed by natural deposition in the drywell, natural deposition in the main steam lines and the condenser, and by removal of particulates by the SGT and CREV HEPA filters.
In the Drywell Natural removal of activity is credited in the drywell using the 10th percentile values from the models of RADTRAD Table 2.2.2.1-3. The elemental iodine is assumed to have the same removal rate as the particulate (noting that the total surface area of the particulate is substantially greater than that of the drywell structures and that the Standard Review Plan Section 6.5.2 elemental iodine wall deposition rate (X4) is greater in any case). No credit is taken for organic iodine and noble gas removal.
In the Steam Lines The AEB-98-03 (Reference 22) model is used to obtain the deposition velocity for particulate. The AEB-98-03 model assumes a well mixed control volume. Since the flow in the steam line is expected to be plug flow (because of the values of the assumed MSIV leakage), it is justified (according to AEB-98-03) to use the median value for the deposition velocity found in Appendix A of that document. The horizontal cross-section of the steam line is used as the surface area for deposition.
The RADTRAD Bixler model is used for deposition of elemental iodine. As with particulate deposition, no credit is taken for cooling of the steam lines with time and the associated increase in residence time.
For the steam lines in which both MSIVs are closed, there are two steam line control volumes in series. In the second (outboard) control volume, it would be expected that the particulate concentration and the representative deposition velocity would be lower. Therefore, the distribution of deposition velocity for particulates in the second control volume has been adjusted to reflect the faster settling particles that have already been removed in the first control volume. The median deposition velocity in the first control volume is 1.1 7E-3 m/s (the AEB 98-03 median value), but it is calculated to be 2.7E-4 m/s in the second control volume.
12 BFN Alternate Source Term Safety Assessment
Removal In the Condenser Particulate deposition in the main condenser is treated using the same approach as that for the steam lines. The effective volume of the main condenser (for hold-up) is based on crediting 90% of the nominal condenser volume and none of the volume of the low pressure turbine.
Since the efficiency of the condenser in removing both particulate and elemental iodine is determined by the relative removal and leakage lambdas (and since the main condenser volume is in the denominator for both), the only things determining the condenser removal efficiency are: (1) deposition velocity, (2) deposition area, and (3) volumetric flow out of the main condenser. For particulate, the sedimentation velocity in the main condenser is assumed to be the flow-weighted average of the median values exiting the two steam lines with leakage, and that flow-weighted average is 3.47E-4 m/s. The sedimentation area is the assumed effective volume of the main condenser divided by the sedimentation height of the main condenser.
The elemental iodine removal rate in the main condenser could be appropriately calculated using the 4.9 meter/hour (1.36E-3 m/s) deposition velocity from the SRP 6.5.2 for Xw, but instead, it is assumed to be the same as particulate. This is especially conservative, because not only is the SRP 6.5.2 elemental iodine deposition velocity nearly four times greater than that for sedimentation; but also, elemental iodine deposition occurs on vertical and overhead surfaces as well as on horizontal surfaces facing upward.
Combined Efficiencies for Steam Lines and Main Condenser The steam line and main condenser removal efficiencies for particulate and elemental iodine may be combined by weighting the steam line removal according to flow and then placing these removal efficiencies in series with that of the main condenser. These efficiencies included a condenser bypass term of 0.5 percent of the total MSIV leakage.
2.3.1.2 Main Steam Line Break Accident Inputs and Assumptions The main steam line break accident assumes a double ended break of one main steam line outside the secondary containment with displacement of the pipe ends that permits maximum blowdown rates.
The analysis also assumes isolation of the control room habitability zone and the initiation of the CREV System by the control room normal ventilation intake radiation monitors on high radiation.
13 BFN Alternate Source Term Safety Assessment
The radiological consequences of the design basis main steam line break accident were analyzed using TVA's STP, COROD, and FENCDOSE codes. The evaluation of fuel performance for the main steam line break accident determined that no fuel rod failures are postulated for this event.
Two cases were evaluated that corresponded to the iodine concentration in the primary coolant:
An assumed pre-accident spike of 32 pCi/gm Dose Equivalent (DE) 1-131 (conservative value based on the TS maximum allowed value of 26 jtCi/gm DE 1-131).
"° A value of 3.2 ltCi/gm DE 1-131 corresponding to the maximum TS value allowed for continued operation.
The break mass released includes the line inventory plus the system mass released through the break prior to isolation. Break isolation was assumed in 5.5 seconds. This assumption is consistent with the isolation time used in evaluation of the pressure, temperature, pipe whip, and jet impingement effects for main steam line breaks outside of the drywell.
This is the maximum isolation time for an MSIV given the expected 3 to 5 second isolation and includes isolation instrumentation response time.
This results in the maximum radiological release for analysis. The analysis assumes an instantaneous ground level puff release.
RG 1.183, Section 4.4 of Appendix D, indicates that the iodine species released from the main steam line should be assumed to be 95 percent Csl as an aerosol, 4.85 percent elemental, and 0.15 percent organic. The main steam line break accident analysis assumes all iodine to be elemental. This difference is inconsequential for the BFN AST analysis since no credit is taken for filtration or other removal mechanisms of iodine, such as plateout, sedimentation, condensation, or decay. The key inputs used in this analysis are included in Table 2-13.
2.3.1.3 Refueling Accident Inputs and Assumptions This postulated refueling accident involves the drop of a fuel assembly on top of the reactor core during refueling operations. The drop over the reactor core is more limiting than the drop over the spent fuel pool since the kinetic energy for the drop over the reactor core area (greater than 23 feet) produces a larger number of damaged fuel pins on impact than the shorter drops that could occur over the fuel pool.
All the refueling accident activity is assumed released to the environment from the refuel building ventilation system with no credit for reactor building holdup or dilution. Not crediting any dilution, holdup, or cleanup by SGT of the activity released from the pool represents a more conservative basis than that used in the existing licensing basis analysis.
14 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term All current fuel types are bounded by this analyses. The key inputs used in this analysis are included in Table 2-14.
2.3.1.4 Control Rod Drop Accident Inputs and Assumptions The BFN analysis for the control rod drop accident considers the worst-case radiological exposure release path. The analysis assumes the condenser is evacuated for 30 days after the rod drop either by the steam jet air ejectors (steady state operation) or the Mechanical Vacuum Pump (MVP) operation (at startup) (Reference 23). While plant interlocks and procedures essentially prevent power operation with the MVPs in service, exhaust via the MVPs provide the greatest amount of activity released, and therefore this pathway is used for the analysis.
The activity released from the core is instantaneously released to the main condenser. From the condenser the activity flows to the stack where fumigation conditions are considered from 0 to 30 minutes. It is assumed ten scfm of leakage enters the stack room where mixing occurs prior to release to the environment at the base of the stack. Releases from the damaged fuel, and deposition in the main condenser are per Appendix C of RG 1.183. A schematic of the transport model is provided in Figure 2-2. The key inputs used in this analysis are included in Table 2
- 15.
2.3.2 Suppression Pool pH Control NUREG-1465 notes that SRP 6.5.5 (Reference 24), allows credit for fission product scrubbing in the suppression pool. Although fission product removal by suppression pool scrubbing is not credited in the BFN analyses, natural removal by sedimentation is credited; and this will lead to a large fraction of activity being deposited in the pool water. The pool water will also retain soluble gaseous and soluble fission products such as iodides and cesium, but not noble gases. Once deposited the iodine will remain in solution as long as the suppression pool pH is maintained at or above 7.0.
It is expected that the initial effects on post-accident suppression pool pH will come from rapid fission product transport and the formation of cesium compounds, which would result in increasing the suppression pool pH.
However, cesium compounds are not credited in the long-term pH analyses and the determination of the final (30 day) pH value. As radiolytic production of nitric acid and hydrochloric acid proceeds, and these acids are transported to the pool over the first days of the event, the pH would become more acidic.
Upon detection of high drywell radiation associated with the postulated activity release, plant procedures will be revised to require manual initiation of SLC injection. The buffering effect of SLC injection within several hours is 15 Safety Assessment
sufficient to offset the effects of these acids that are transported to the pool and maintain suppression pool pH at or above 7.0.
The current design function of the SLC System is to provide a backup method, independent of control rods, to make the reactor subcritical over a full range of operating conditions. The system actuation requirements for reactivity control are explicitly addressed in the BFN Emergency Operating Instructions (EOIs). The SLC system is designed as a seismic Class I system for withstanding specified earthquake loadings. Additionally, the SLC System pumps, valves, and controls are powered from the diesel generator in absence of normal power. The current TS requires the system be maintained in an operable status whenever the reactor is in modes 1 or 2.
The SLC System is currently classified as a special safety system as defined in UFSAR Section 1.2. The SLC System will also be credited for limiting radiological dose following LOCAs involving fuel damage in accordance with the AST analyses for suppression pool pH control; however, the system will remain classified as a special safety system instead of being classified as a safety system.
A core damage event large enough to release substantial quantities of fission products into the drywell will result in high drywell radiation alarms. The operational response procedures will be revised to include instructions to manually actuate the SLC System injection. The AST analysis provides for SLC System actuation at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of accident initiation and completion of injection of an adequate volume and content of SPB within several hours, which will ensure the suppression pool pH remains at or above 7.0 for 30 days.
Initiation of the SLC System following fuel damage to control suppression pool pH is a new operator action during a DBA LOCA response. High radiation indicative of fuel failure would be sensed by two radiation monitors in the drywell and two radiation monitors in the pressure suppression chamber. Upon reaching a high radiation level, the "Drywell/Suppr Chamber Radiation High" annunciator on Panel 9-7 in the main control room would alert the operator to the fuel damage. The Alarm Response Procedure (ARP) will direct the operator to initiate SLC System injection based on the high radiation level.
Initiation of the SLC System will be accomplished from the main control room with a simple keylock switch manipulation. This switch is located on control room panel 9-5 and actuation of this switch is the only action necessary to initiate injection of the sodium pentaborate into the reactor vessel. The new SLC System function to control suppression pool pH does not involve any change to the actions needed to be performed to initiate SLC system injection. Indication of proper SLC System operation is provided in the control room as described in UFSAR Section 3.8.
16 BFN Alternate Source Term Safety Assessment
During this postulated event, plant operators will be responding to the event as directed by the plant Emergency Operating Instructions (EOI). Adequate time is available for SLC System initiation during these events. Immediate initiation of the SLC System is not vital since the analysis allows for two hours before initiation. Operators are familiar with operation of the SLC System due to previous training for Anticipated Transients Without Scram (ATWS) events. Training on this new operator action will also be provided to the operators.
With certain post LOCA conditions, existing BFN procedures direct the operations of systems to accomplish a total floodup of the primary containment. This floodup uses the Ultimate Heat Sink (Tennessee River) as the preferential source of makeup water since it is the only safety related makeup water source. A review of a previous ten years of data reflect that the minimum river pH has been above a pH of 7.0, with the exception of one data point, over this time period. Although the condensate storage tank (CST) could be used as a makeup source, it is not safety related and does not have sufficient volume to flood containment without repeated refilling or the use of additional CSTs. Consequently, the addition of a large amount of water from the UHS to the suppression pool and containment inventory will not result in a pH below 7.0.
2.3.3 Main Steam Line Break Accident Puff Release Dispersion Factor A new control room X/Q value for an instantaneous ground level puff release to the atmosphere was determined for use in the main steam line break accident radiological dose analysis. The inputs used in the determination of the X/Q value are provided in Table 2-17.
2.3.4 NUREG-0737 Evaluation The inputs and assumptions utilized in the NUREG-0737 evaluation include the AST plant-specific fission products inventories and other applicable inputs as described in Section 2.3.1.
17 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Computer Code Version or Revision Comments Used to generate the existing ARCON96 1996 See Note1 X/Q's NUREG/CR - 6331, Rev. 1, May 1997 NRC Sponsored Used to determine control room COROD R5 TVA Code See Note' operator doses for main steam line break accident Used to determine the offsite FENCDOSE R4 TVA Code See Note' doses for main steam line break accident Used to model complex Source Transport R6 TVA Code See Note' systems that take into account Program (STP) radioactive decay and production of daughter isotopes.
Output can be in activity levels or gamma spectra.
Point Kernel Integration code MicroShield 5.03 Code used in nuclear used for general purpose radiological analyses.
gamma shielding analysis.
Grove Engineering.
Used in safety-related applications by many nuclear plants in the U.S.
Used to calculate fission ORIGEN ORIGEN2 (GE)
The codes are either product inventories SAS2H/ORIGEN-S referenced by RG 1.183 or consistent with NRC recommendation.
ORNLITM-7175 NUREG/CR-0200R6 Used to develop photon QADISOTP R1 TVA Code used in spectrum for Main Steam Line nuclear radiological Break Accident analyses 1 Results were reviewed and approved by NRC in Safety Evaluation for Amendment Nos. 263 and 223 for BFN Units 2 and 3.
18 Task BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Computer Code Version or Revision Comments Used to determine the direct QAD-P5Z R6 TVA Point Kernel gamma shine dose due to the Code used in nuclear released isotopes in the turbine radiological analyses.
building for the Main Steam Line Break Accident Used for the LOCA and Control RADTRAD 3.02a Referenced by Rod Drop Accident Dose RG 1.183 Calculations NUREG/CR-6604 USNRC April 1998 Used to perform independent STARDOSE 03/01/1997 Polestar Applied check of LOCA and Control Technology code Rod Drop Accident.
NUREG/CR-5106 Used to evaluate Suppression STARpH 1.04 Utilized in other AST Pool Water pH as a function of Submittals &
time Developed by Polestar.
NRC reviewed and approved for use of STARpH for Hope Creek. (Reference 19)
Used to perform a independent QADMODE Version 5.03 Point Kernel Gamma check of MicroShield.
Ray Shielding Code with Geometric Progression Building Factors 19 Task BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Ci/MWt t=O
+
i Co58 Co60 Kr83M Kr85 Kr85M Rb86 Kr87 Kr88 Kr89 Sr89 Sr90 Y90 Sr91 Y91 Sr92 Y92 Y93 Zr95 Nb95 Zr97 Mo99 Tc99M Rul03 Rul05 RH105 Ru106 Sb127 Te127 Te127M Sb129 Te129 Tel29M Te131M 1131 Ci/MWt t =24 hr 1.430E+02 1.425E+02 3.432E+03 3.601 E+02 7.329E+03 6.372E+01 1.446E+04 2.009E+04 2.521E+04 2.786E+04 3.165E+03 3.283E+03 3.487E+04 3.583E+04 3.677E+04 3.696E+04 4.147E+04 4.880E+04 4.897E+04 4.953E+04 5.088E+04 4.454E+04 4.094E+04 2.710E+04 2.559E+04 1.488E+04 2.796E+03 2.773E+03 3.721 E+02 8.457E+03 8.326E+03 1.615E+03 5.155E+03 2.669E+04 Isotope 1.416E+02 1.424E+02 1.387E+01 3.601E+02 1.811E+02 6.141E+01 3.051 E-02 5.743E+01 0.OOOE+00 2.748E+04 3.165E+03 3.273E+03 6.103E+03 3.564E+04 7.922E+01 1.168E+03 8.084E+03 4.822E+04 4.897E+04 1.851E+04 3.956E+04 3.772E+04 4.018E+04 6.615E+02 1.840E+04 1.486E+04 2.369E+03 2.580E+03 3.719E+02 1.952E+02 1.236E+03 1.590E+03 2.976E+03 2.481 E+04 Ci/MWt t=O i
+
Xel31M Tel 32 1132 1133 Xe133 Xel33M 1134 Cs134 1135 Xel 35 Xe135M Cs136 Xe137 Cs137 Ba137M Xe138 Ba139 Bal40 Lal40 Lal4l Cel4l La142 Ce143 Pr143 Ce144 Nd147 Np239 Pu238 Pu239 Pu240 Pu241 Am241 Cm242 Cm244 20 Isotope 3.544E+02 3.829E+04 3.885E+04 5.534E+04 5.504E+04 1.734E+03 6.141 E+04 5.703E+03 5.250E+04 1.971E+04 1.135E+04 1.941E+03 5.023E+04 4.037E+03 3.829E+03 4.757E+04 4.930E+04 4.909E+04 5.231 E+04 4.498E+04 4.535E+04 4.397E+04 4.245E+04 4.113E+04 3.810E+04 1.806E+04 5.201 E+05 2.805E+02 1.234E+01 1.730E+01 4.450E+03 5.449E+00 1.234E+03 5.697E+01 Ci/MWt t=24 hr 3.487E+02 3.089E+04 3.184E+04 2.559E+04 5.303E+04 1.562E+03 1.450E-03 5.697E+03 4.189E+03 1.429E+04 6.823E+02 1.841E+03 0.000E+00 4.037E+03 3.810E+03 1.172E-26 4.170E-01 4.644E+04 5.079E+04 7.085E+02 4.463E+04 1.035E+00 2.597E+04 4.075E+04 3.810E+04 1.698E+04 3.902E+05 2.805E+02 1.238E+01 1.730E+01 4.448E+03 5.470E+00 1.234E+03 5.697E+01 I
20
BFN Alternate Source Term Safety Assessment Time Period Control Room (sec/m3)
Unit 1 Intake Unit 3 Intake EAB(2)
(sec/m3)
LPZ (sec/m3)
Fumigation 3.40E-5 3.02E-5 2.35E-51 1.26E-5 0-2 hrs 9.08E-13 1.41 E-7 1.19E-6 1 1.13E-6 2-8 hrs 3.41 E-13 4.50E-8 5.75E-7 8-24 hrs 2.09E-13 2.54E-8 4.10E-7 1-4 days 7.21 E-14 7.36E-9 1.97E-7 4-30 days 1.57E-14 1.24E-9 6.88E-8 These values were incorrectly listed in Reference 14; however, the correct values were used as the basis of Reference 14.
2 Maximum EAB TEDE for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.
Time Period Control Room (sec/m3)
Unit 1 Intake Unit 3 Intake EAB12)
(sec/m 3)
LPZ (sec/m3) 0-2 hrs 2.OOE-4 8.60E-5 2.62E-4 1.31 E-4 2-8 hrs 1.28E-4 6.46E-5 6.61 E-5 8-24 hrs 5.72E-5 2.80E-5 4.69E-5 1-4 days 4.05E-5 2.00E-5 2.23E-5 1 4-30 days 3.09E-5 1.53E-5 7.96E-6 21 I Typo in Reference 14; same as value for turbine building release.
2 Maximum EAB TEDE for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period BFN Alternate Source Term Safety Assessment
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Safety Assessment FuelType GE14 A10 BLEU Initial Bundle Mass of 182.0 177.7 177.7 Uranium (kg)
Initial Core Average 4.6 4.5 4.95 Enrichment (U-235 wt%)
Core Average Bundle 5.28 5.28 5.28 Power (MWt/bundle)
End of Cycle Core 35.0 37.0 37.0 Average Exposure (GWd/MT) 0 - 120 No Release 120 - 1920 Gases Xe, Kr - 0.1/hr (0.05 total)
Elemental I - 4.9E-3/hr (2.4E-3 total)
Organic I - 1.5E-4/hr (7.5E-5 total)
Aerosols I, Br - 0.095/hr (0.0475 total)
Cs, Rb - 0.1/hr (0.05 total) 1920 - 7320 Gases Xe, Kr - 0.63/hr (0.95 total)
Elemental I - 8.1 E-3/hr (1.2E-2 total)
Organic I - 2.5E-4/hr (3.8E-4 total)
Aerosols I, Br - 0.158/hr (0.2375 total)
Cs, Rb - 0.133/hr (0.2 total)
Te Group - 0.033/hr (0.05 total)
Ba, Sr - 0.013/hr (0.02 total)
Noble Metals - 1.7E-3/hr (2.5E-3 total)
La Group -.. 3E-4/hr (2E-4 total)
Ce Group - 3.3E-4/hr (5E-4 total) 23 BFN Alternate Source Term
BFN Alternate Source Term Safety Assessment CREV Pressurization Mode HEPA Particulate Removal Charcoal Adsorber Flow/
Secondary Containment SGT HEPA Particulate Removal LOCA Y
Y N
Y Y
N Main Steam Y
N1 N
N2 N2 N
Line Break Accident Refueling N
N1 N
N3 N3 N
Accident Control Rod Y
N N
N2 N2 N
Drop Accident 1
No particulates are released to the atmosphere; therefore no particulate filtering is necessary in analysis.
2 No release to secondary containment.
3 No credit taken for holdup or filtering in secondary containment.
24 DBA Dose Analysis Charcoal Adsorber BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment CREV Intake Flow Rate 6717 scfm 25 CREV Makeup Filtered Flow Rate 3000 scfm CREV Unfiltered Inleakage Rate 3717 scfm CREV HEPA Filter Efficiency 90% Particulate CREV Charcoal Adsorption Efficiency No credit taken Control Room volume 210,000 ft3 SGT Flow Rate 24,750 scfm SGT HEPA Filter Efficiency 90% Particulate SGT Charcoal Adsorption Efficiency No credit taken Environment Breathing Rate 0-8 hours: 3.5E-04 m3/sec 8-24 hours: 1.8E-04 m3/sec 1-30 days: 2.3E-04 m3/sec Control Room Breathing Rate 3.5E-04 m3/sec Control Room Occupancy Factors 0-1 day: 1.0 1-4 days: 0.6 4-30 days: 0.4 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Fission Products Release Fractions Regulatory Guide 1.183 Table 1 BWR Core Inventory Fraction Released Into Containment Group Noble Gases 0.05 Halogens 0.05 Alkali Metals 0.05 Tellurium Metals 0.00 Ba, Sr 0.00 Noble Metals 0.00 Cerium Group 0.00 Lanthanides 0.00 Gap Early Release In-vessel Phase Phase Total 0.95 0.25 0.20 0.05 0.02 0.0025 0.005 0.005 1.0 0.3 0.25 0.05 0.02 0.0025 0.0005 0.0002 Fission Product Release Timing Regulatory Guide 1.183 Table 4 LOCA Release Phases BWR Phase Onset Duration Gap release 2 min 0.5 hr Early In-Vessel 0.5 hr 1.5 hr Fission Product Iodine Chemical Form Particulate 95%
Elemental 4.85%
Organic 0.15%
Control Room Isolation/CREV Initiation 10 minutes ECCS Leakage Release Fractions Ten percent of the radioiodine in the leaked coolant is assumed to become airborne in the reactor building (secondary containment). Of this activity, 97% is assumed to be elemental iodine and 3% is assumed to be organic iodine.
Primary Containment Leak Rate (30 days) 2 % containment air weight/day Secondary Containment Bypass Leak HWWV = 10 scfh beginning at t>8 hours Rate (30 Days)
Assumed ECCS Leak Rate (30 days) 5 gpm ECCS Leakage Temperature
<212°F 26 Safety Assessment
BFN Alternate Source Term Safety Assessment MSIV Leak Rate at test pressure of 25 psig 150 scfh total 100 scfh maximum for one line Leakage at base of stack (stack bypass) 10 scfm MSIV Leakage that Bypasses Main 0.5%
Condenser (percentage of total MSIV leakage)
CAD vent rate 139 scfm for 24 hrs
@ 10 days, 20 days, 29 days Drywell Airspace 159,000 ft3 (Min value used for dose calculation)
Torus Airspace 119,400 ft3 (Minimum)
Suppression Pool 121,500 ft3 (Minimum)
Reactor Building Free Volume 1,931,502 ft3 (50% of this value used due to incomplete mixing)
Stack Room 69,120 ft3 (50% of this value used due to incomplete mixing)
High Pressure Turbine 568.6 ft3 (No credit taken)
Low Pressure Turbine 51,000 ft3 (No credit taken)
Drywell Natural Deposition Particulate: Power's Model, 1 0 th percentile values(conservative compared to SRP 6.5.2 X*.
Elemental: Same as particulate.
Drywell Accident Conditions (maximum)
P = 48.3 psig, T = 294.9 Degrees F Surface Area for Elemental Iodine 3409 m2 Deposition in Drywell 27 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Condenser Volume 90 Percent of 136,000 ft3 or 122,400 ft3 28 Steam Line Conditions Saturated Conditions at 1050 psia Steam Line Volume: Inboard to Outboard 53.7 ft3 MSIV Steam Line Volume: Outboard MSIV to 173.1 ft3 drain line Sedimentation Height 27.2 ft Removal Efficiency Removal Efficiency for for Aerosol Particles Elemental Iodine Steam Line Leakage 99.87%
99.01%
(Drywell to Main Condenser)
(These removal efficiencies applied to a leakage entering the main condenser volume include removal in the condenser downstream)
Main Condenser Bypass 89.33%
16.37%
(Drywell to Environment)
I Safety Assessment
BFN Alternate Source Term Safety Assessment Mass Release 11,975 Ibm steam 42,215 Ibm water (saturated @ 898psia)
MSIV Isolation Time 5.5 seconds DE 1-131 Equilibrium Value 3.2 ýtCi/gm DE-1-131 Pre-Accident Spike 32 ptCi/gm (Conservative to TS value of 26,Ci/gm)
Iodine Species Release Fraction All Assumed Elemental Number of Failed Rods 111 Radial Peaking Factor 1.5 Fuel Decay Period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Pool Water Iodine Decontamination Factor Elemental 500 Organic 1
Release Period Instantaneous Reactor Building Ground Release Reactor Building Refueling Zone Vent Location (No credit for holdup or SGT operation)
Release Fractions Noble Gases excluding Kr-85 5 percent Kr-85 10 percent 1-131 8 percent Iodines except 1-131 5 percent 29 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Number of Failed Rods 850 Percent Fuel Melt for Failed Rods 0.77 %
Radial Peaking Factor 1.50 Release Period 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Main Condenser and Low Pressure 187,000 ft3 Turbine Free Volume Stack Room Volume 69,120 ft3 (50% of this value used due to incomplete mixing)
Assumed Base of Stack Leakage 10 cfm Mechanical Vacuum Pump Flowrate 1850 scfm @ 7" Hg Gap Release Fractions Noble Gas 10%
Iodine 10%
Br 5%
Cs, Rb 12%
Te Group 0%
Ba, Sr 0%
Noble Mtls 0%
Ce Group 0%
La Group 0%
Core Melt Release Fractions Noble Gas 100%
Iodine 50%
Br 30%
Cs, Rb 25%
Te Group 5%
Ba,Sr 2%
Noble Mtls 0.25%
Ce Group 0.05%
La Group 0.02%
Activity that reaches the condenser Noble Gas 100%
Iodine 10%
Br 1%
Cs, Rb 1%
Te Group 1%
Ba,Sr 1%
Noble Mtls 1%
Ce Group 1%
La Group 1%
30 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Activity released from the condenser Noble Gas 100%
Iodine 10%
Br 1%
Cs, Rb 1%
Te Group 1%
Ba, Sr 1%
Noble MtIs 1%
Ce Group 1%
La Group 1%
Maximum Suppression Pool Volume 131,400 ft3 Containment Free Volume 278,400 ft3 Reactor Coolant System Inventory 1..226E-06 Ibm Sodium Pentaborate Injectable Volume 4000 gal SLC (Na 20*5B203*10H20) injected 8 weight percent Sodium Pentaborate Enrichment 62.9 mole% BI0 Initial Suppression Pool pH 5.3 Average suppression pool temperature 132°F Mass of Polyvinyl Chloride Jacket in the Drywell 2865 Ibm Mass of Hypalon Jacket in the Drywell 868 Ibm Average Cable Outside Diameter 0.89 inches Average Cable Jacket Thickness 72 mils Percent of Drywell Cable in Conduit 30%
Conduit Material Aluminum Conduit wall thickness 0.1 inch Conduit air gap 0.25 inch 31 BFN Alternate Source Term Safety Assessment
BEN Alternate Source Term Safety Assessment Mass Release 11,975 Ibm steam 42,215 Ibm water (saturated @ 898psia)
Assumed instantaneous release Bubble Geometry Spherical & Hemispherical Cases Considered Turbine Building Perimeter Dimension
-1500 ft 32 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Figure 2-1: LOCA Transport Model 1
Drywell/Torus Mixing (After Release) 2/3 Primary containment leakage 4
ECCS Leakage 5
SGT Flow 6
CAD venting 7
Base of stack release (stack bypass) 8 Stack release 9
HWWV Leakage 10 MSIV Leakage 11 MSIV Leakage - Condenser bypass 12 Condenser leakage 13 No credit taken for holdup in the Turbine Building 14 CREV filtered/unfiltered intake 15 CREV exhaust 33 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term Safety Assessment Figure 2-2: Control Rod Drop Accident Transport Model daps 34
- 3.
RESULTS 3.1 Evaluation Results 3.1.1 Accident Radioloqical Consequence Analyses The postulated accident radiological consequence analyses were updated for AST implementation impact. Comparison of updated AST doses to existing licensing basis doses considers impact from the assumed operation at EPU conditions (4031 MWt (102% of 3952 MWt)) as well as the change in analysis methodology.
3.1.1.1 LOCA The radiological consequences of the DBA LOCA were analyzed using the RADTRAD code and the inputs/assumptions defined in Section 2.3.1.1 of this report. The post accident doses are the result of the following activity considerations:
- 1.
Primary to secondary containment leakage. This leakage is directly released into secondary containment and filtered by SGT System prior to elevated release through the plant stack with stack bypass released at ground level. No credit is taken for charcoal adsorber action.
- 2.
ECCS leakage into the secondary containment. This leakage is directly released into the secondary containment environment and the airborne portion is filtered by SGT System prior to elevated release through the plant stack with stack bypass released at ground level. No credit is taken for charcoal adsorber.
- 3.
MSIV leakage from the primary containment into the main condenser (with a fraction that bypasses the main condenser directly to the atmosphere). Leakage passes through the alternate MSIV leakage pathway to the main condenser with credit for deposition before it is released, undiluted and unfiltered, through the turbine building vents.
- 4.
HWVVV leakage from primary containment. This leakage is directly released (after a eight hour delay) to an elevated release through the plant stack.
- 5.
Post-DBA LOCA radiation shine dose to personnel within the control room from activity released to the reactor building and from activity contained in Core Spray System piping.
35 Safety Assessment
The EAB, LPZ, and control room calculated doses are within the regulatory limits. Table 3-1 presents the results of the LOCA radiological consequence analysis.
3.1.1.2 Main Steam Line Break Accident The EAB, LPZ and control room calculated doses are within the regulatory limits for the cases analyzed. The control room doses were determined using the new X/Q value for the instantaneous puff release.
Table 3-2 presents the results of the main steam line break accident radiological consequence analysis.
3.1.1.3 Refueling Accident The radiological consequences of the design basis refueling accident were analyzed using a simplified configuration of one unique release pathway using the turbine building exhaust release X/Q for the EAB and LPZ, and the refueling X/Q for the control room along with the inputs/assumptions defined in Section 2.3.1.3 of this report. The EAB, LPZ, and control room calculated doses are within the regulatory limits.
Table 3-3 presents the results of the refueling accident radiological consequence analysis.
3.1.1.4 Control Rod Drop Accident The radiological consequences of the design basis control rod drop accident were analyzed using the RADTRAD code and the inputs/assumptions defined in Section 2.3.1.4 of this report. The EAB, LPZ, and control room calculated doses are within the regulatory limits.
Table 3-4 presents the results of the control rod drop accident analysis.
3.1.2 Suppression Pool PH Control The re-evolution of elemental iodine from the suppression pool is strongly dependent on suppression pool pH. The analysis assumed that SPB was injected via SLC within several hours of the onset of a LOCA. The conservative modeling of the primary containment cabling results in the production of a large amount of hydrochloric acid. The minimum suppression pool pH at 30 days post-LOCA remains above 7.0, which satisfies the conditions for inhibiting the release of the chemical form of elemental iodine in the elemental form from the suppression pool water. The suppression pool pH response over time is shown in Figure 3-1.
The quantity of SLC calculated as necessary to meet AST requirements is above the current TS requirements; therefore, TS revisions are proposed which increase the quantity of SLC required. Based on these TS changes, AST analysis for suppression pool pH control, the SLC system will be credited for limiting radiological dose following LOCAs involving fuel damage.
36 BFN Alternate Source Term Safety Assessment
BFN Alternate Source Term 3.1.3 Main Steam Line Break Accident Instantaneous Ground Level Puff Release Dispersion Factor The new control room X/Q value for an instantaneous ground level puff release to the atmosphere was calculated for use in the main steam line break accident radiological dose analysis. The X/Q value is shown in Table 3-5.
3.1.4 NUREG-0737 Evaluation The results of the NUREG-0737 evaluation are summarized below.
Post-Accident Vital Area Access and Sampling - The results of the revision of post-accident mission doses demonstrate that the current calculated doses (based on TID-14844 source terms) bound the doses that would be calculated based on AST source terms. The evaluated mission doses for BFN remain less than 5 rem TEDE.
Post-Accident Radiation Monitor - The containment high range radiation monitors used to monitor post-accident primary containment radiation levels were evaluated for the impact of AST. The monitors continue to provide their design function and envelop the projected radiation exposure rates.
"* Control Room Radiation Protection - The resultant doses to the control room for each of the four DBAs analyzed for AST have been determined.
The results of these analyses are presented in Section 3.1.1.
Radioactive Sources Outside the Primary Containment - The contribution of radiological dose consequences as a result of core spray piping shine and ECCS leakage was determined as part of the radiological dose analysis for the LOCA. The results of this analysis are presented in Section 3.1.1.1.
3.2 Summary Implementation of the AST as the plant radiological consequence analyses licensing basis requires a license amendment pursuant to the requirements of 10 CFR 50.67. Radiological dose analyses were performed for the four DBAs with a potential for offsite/control room dose. Doses calculated with the AST for accidents involving damaged fuel reflect delayed and/or reduced activity releases (relative to those of TID-14844 and RG 1.3) to the primary containment, reactor building, and or/or steam lines, as applicable. Offsite and control room doses remain within regulatory requirements.
37 Safety Assessment
BFN Alternate Source Term Base of Stack 1.08E-2 4.49E-3 Top of Stack 5.68E-1 2.43E-1 Turbine Building 3.02E-1 1.13E-1 Roof ECCS Leakage -
1.25E-2 1.21E-2 Base of Stack ECCS Leakage -
3.52E-1 1.12E-1 Top of Stack Shine N/A 7.62E-1 TOTAL 1.02 1.25 1.25 Regulatory Limit 25 25 5
Current Analysis 1.67E-01 (25) Gamma 4.82E-01 (25) Gamma 6.83E-01 (5) Gamma (Regulatory Limit) -
1.01E-01 (300) Beta 4.84E-01 (300) Beta 1.58E-01 (30) Beta 5.84 (300) Thyroid 8.6 (300) Thyroid 2.95E+01 (30) Thyroid rem 38 Safety Assessment
CD CD cn CD CL 0
- 0) m~
(D 0
(D C:
~CD (D
C")
Cmrri 005 3Z~~
rQ
- 0) wcoo C)- 0 C)
G).nr
-- Ica CF) wA 0
C,,
0C:
CD I1
- 0) m' 0
--I 0
m 0
CA) 00 m
6 2CD 3 cD CD 0
(D 0 >.
mmm
-C)0 Ca (OO)
- 2.
mmrn
+66 0.~
CD 0
N)
(71 N)
C,,
c,,
m CA) 0
- 0)
N) 0 Co m
m CA) 0 N) m NJ m
- 0) w)
0 w)
(0
BFN Alternate Source Term Power Operation 1.19 6.82E-01 2.48E-01 40 Regulatory Limit 6.30 6.30 5
Current Analysis 1.52 (25) Gamma 8.58E-01 (25) Gamma 3.86E-02 (5) Gamma (Regulatory Limit) -
1.07 (300) Beta 6.04E-01 (300) Beta 4.32E-01 (30) Beta rem I 1.58E+01 (300) Thyroid I 1.58E+01 (300) Thyroid 6.3 (30) Thyroid Time Period Control Room (sec/ma) 46 secs 4.60E-4 Safety Assessment
BFN Alternate Source Term Safety Assessment Figure 3-1: Suppression Pool pH Response Figure 1 Plot of pH vs. Time 8.6 8.5 8.4 8.3 8.2 8.1
- " 8 7.9 7.8 7.7 7.6 7.5 0
200 400 600 800 Hours after shutdown 41 BFN Alternate Source Term Safety Assessment
- 4.
REFERENCES
- 1.
NRC Standard Review Plan 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms," Revision 0, Dated July 2000.
- 2.
NRC Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents At Nuclear Power Reactors," Dated July 2000.
- 3.
NRC NUREG-0737, "Clarification of TMI Action Plan Requirements," Dated November 1980.
- 4.
TVA Browns Ferry Nuclear Plant, "Updated Final Safety Analysis Report,"
Amendment 19.
- 5.
Technical Information Document (TID) - 14844, "Calculation of Distance Factors for Power And Test Reactor Sites," U.S. Atomic Energy Commission, Dated March 23, 1962.
- 6.
ORIGEN Computer Code, Oak Ridge National Laboratory.
- 7.
NUREG/CR-6604, RADTRAD Computer Code,: "A simplified model for Radionuclide Transport and Removal And Dose Estimation," Dated April 1998 and Supplement 1, Dated June 8, 1999.
- 8.
STARDOSE Model report, Polestar Applied Technology, Inc., Dated January 31, 1997.
- 9.
TVA Letter to NRC Dated September 8, 1999, Browns Ferry Nuclear Plant (BFN)
- Units 2 and 3 - Technical Specification (TS) Change 399 - Increased Main Steam Isolation Valve (MSIV) Leakage Rate Limits And Exemption From 10 CFR Appendix J.
- 10.
TVA Letter Dated February 4, 2000, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Response to Request for Additional Information Regarding Technical Specifications Change 399 - Increased Main Steam Isolation Valve (MSIV)
Leakage Rate Limits and Exemption From 10 CFR 50 Appendix J - Revised TS Pages for Increased MSIV leakage Limits.
- 11.
NRC NUREG - 6331, "Atmospheric Dispersion Relative Concentrations in Building Wakes," Revision 1, May 1997, ARCON 96, RSICC Computer Code Collection No. CCC-664.
- 12.
NRC Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Reactors," Dated March 1, 1996.
- 13.
NRC Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments of Nuclear Power Plants," Revision 2.
42 BFN Alternate Source Term Safety Assessment
- 14.
NRC Letter to TVA dated March 14, 2000, Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Limits on Main Steam Isolation Valve Leakage (TAC Nos. MA6405 and MA6406).
- 15.
MicroShield, Version 5.0.3, Grove Engineering
- 16.
NRC NUREG 1465, "Accident Source Terms for Light Water Reactors for Light Water Nuclear Power Plants," Dated February 1995.
- 17.
NRC NUREG/CR 5950, "Iodine Evolution and pH Control," Published December, 1992.
- 18.
STARpH, "A Code for Evaluating Containment Water Pool pH during Accidents,"
R4, February 2000, Polestar Applied Technology, Inc.
- 19.
NRC Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 134 to Facility Operating License NO. NPF-57, PSEG Nuclear, LLC, Atlantic City Electric Company, Hope Creek Generating Station, Docket No. 50-354, Dated October 3, 2001.
- 20.
NRC Standard Review Plan 15.6.5, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage From Engineered Safety Feature Components Outside Containment," Dated July 1981.
- 21.
NRC Standard Review Plan 6.5.2, "Containment Spray As a Fission Product Cleanup System," Revision 1, Dated July 1981.
- 22.
AEB-98-03, "Assessment of Radiological Consequences For Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Term," Dated December 9, 1998.
- 23.
NRC Standard Review Plan 15.4.9, "Spectrum of Rod Drop Accidents (BWR)",
Revision 2, Dated July 1981.
- 24.
NRC Standard Review Plan 6.5.5, "Suppression Pool as a Fission Product Cleanup System," Dated December 1998.
- 25.
NRC Standard Review Plan 6.4, "Control Room Habitability Systems," Dated July 1981.
43 Safety Assessment
ENCLOSURE5 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 PROPOSED LICENSE AMENDMENT ALTERNATIVE SOURCE TERM UPDATED FINAL SAFETY ANALYSIS REPORT CHAPTER 14.6 MARKUPS AFFECTED PAGE LIST TVA anticipates that the following Chapter 14 pages will require revision by AST.
The revised text has been highlighted with a line drawn through the deleted text and new or revised text indicated with a double underline.
A matrix identifying other sections in the UFSAR that are currently under evaluation for change is also provided in this enclosure. The final UFSAR changes will be completed as required by BFN procedures following approval of this change.
Marked Pages See Attached
UFSAR Review Matrix Chapter/
Affected Sections*
Comments Appendix 1
1.2,1.4,1.6,1.8 2
None 3
3.8 See Enclosure 3 4
4.5, 4.7, 4.8, 4.11 5
5.2, 5.3 6
6.1 7
7.3, 7.12 8
None 9
None 10 10.21 11 None 12 None 13 None 14 14.3, 14.6 Section 14.6 attached A
None B
See marked-up TS pages Technical Specifications C
None D
Yes E
None F
None G
None H
N/A Deleted I
None J
None K
None L
None M
None N
None
- These sections have been currently identified as requiring changes to support AST.
Fission Product Release From Fuel The following assumptions were used in the initial calculation of fission product activity release from the fuel.
- a.
Eight hundred fifty fuel rods fail, per General Electric (GE) Licensing Topical Report, NEDO-31400A.
- b.
The reactor has been operatir failed atre tiau to haved operate 0rut apowe The rods that failed are assumed to have operated at a power peaking factor of 1.*51:.
- c.
Of the rods that fail, 0.077% of the fuel melts, per NEDO-31400A. The following percentages of radioactive material are released to the reactor coolant from the failed fuel rods":
m...
t...
d...
Oathrhalan Nohbfle Me1b~e n Cerum k
ou
- Noble,
!i:
10M%
2%
0.....5%.....
8 Regulatory Guide 1.183 and NUREG-0800, Section 15.4.9.
BFN AST UFSAR Changes Page 1 14.6.2.4
14.6.2.5 Fission Product Transport The following assumptions were used in calculating the amounts of fission product activity transported from the reactor vessel to the main condenser (ntaer) a.....
The r.iclto flo rate........
is percen.of. rte..nd.th..tea.flowto.th condonser~ ~~~~.
isfiepec.tofrae. The.2..
prcent.rcircu..ton.flowand.fiv percent
~
~
~ ~..
steam. f..o.are.he.aximm
.f.'A.
ate. expct.dwhenth..eactr..
bein taken to....
powe
.n the main condenser is..
st......
being evauaedbyth mechani..al..vacuum.pump..(M....
The..e.irc..ati.n.low.rat..is.used.i determining the...o..me.of..oolant.in.which.the..activity.rele.s.d.from.the. fue..
is deposited. The five. peren sta flwrt-getrta htwihwudb acdn. This........
hsu pto is cosrvtvebcus t eulsitetrnpoto reach",
thýea turbine antcnener 8
- o. W~&
caryovr inthe ain teamline is ssumd tobe, -0+1preto4hoa mass................
of.st.m.tansfrred.o.th..on en M e.u..t of...
th-ta separation~~
effected..................
b.tesm..e f eaaosue n hsratrvse show ~ ~
~
~~010 tha waeFaoesls ta ecn vnaitraestmflwTh camMvW fratio perit rmptaio ofth.a.gn.civt
.crie.o.h.mi co de se in..the..wat..r.entrained....in.the...t.....
~
~
~
ag 2
14.6.2.6 Fission Product Release to Environs The following assumptions and initial conditions were used in the calculation of fission product activity released to the environs (initial core):
- a*
The accident is assumed to occur while condenser vacuum is being maintained with the mechanical vacuum pump During normal operation, vacuum is maintained with the steam-jet-air ejector, the discharge, from which, is through a holdup (time delay) and filter system. The assumed operation of the mechanical vacuum pump results in the discharge of the condenser activity directly to the environment via the elevated release..point but without the benefits of holdup (decay) or filtration beyond th condense.
ci All of the noble gas activity transferred to the condenser is assumed to be airborne in the condenser. The halogen and orticla activity transferred to the condenser experiences the removal effects of the condens~ate areail The rate at which the condenser activity is discharged to the environment is dependent upon the free volume of the turbine and condenser and the discharge rate of the mechanical vacuum pump. The numerical values appropriate to these parameters are 187,000 ft3 (lwpesr ubn oueplus condenser free volume) and 1,850 cfm mechanical vacuum pump discharge rate.
h A continuous ground level release of 10 cfm occurs at the base of the stack.
The 10 cfm leakage mixes within the rooms at the base of the stack (34,560 ft3, 50% of 69,120 ft3 because of incomplete mixing).
nAtmospheric dispersion coefficients, X/Q, for elevated releases under fumigation conditions, elevated releases under normal atmospheric conditions and ground level releases at the base of the stack are used. X/Q values applicable to the time periods, distances, and geometric relationships (offsite and control room) are shown in Table 14.6-8. Control room X/Q values for the base of the stack releases are calculated using the computer code ARCON96. For sites, such as BFN, with control room ventilation intakes that are close to the base of tall stacks, ARCON96 underpredicts the X/Q values for top of stack releases; BFN AST UFSAR Changes Page 3
therefore, top of stack releases to the control room intakes are evaluated using the methods of Regulatory Guides 1.145 and 1.111.
Based upon these conditions, the fission product release rate to the environment is shown in Table 14.6-1.
14.6.2.7 Radiological Effects The BFN analysis for the CRDA consists of two potential release paths; condenser leakage at 1 % per day into the turbine building or through SJAE and offgas system as analyzed by the NEDO-31400A, and the MVP discharge as analyzed in accordance with RýKe"`"Ul-`atovGie16.SP-49 h
onr.6im~os isd ed by-2 beau~o te iltinefec o te ua ir intalii onlfiguaino h oto bay,,
The "worst-case" radiological exposure resulting from the activity discharged from a CRDA and a Rubl"e*ta dG*ui*e 1.183 4&4 source term would be from the MVP release path. The resulting control room dose is less than the g*mm*30 mb:t,:1q*O R:.*e thy.o:d. The G EAB and LPZ doses from
- ...:.....~~i:i*, i::
the MVP are well below the :R.
-1ov Gutide 1.1S3.$,@-4i,-4--g.reference values of 75 REMthyrod and66. REM TE'D:E:woe-o BFN AST UFSAR Changes Page 4
Primary Containment Response BFN Units 2 and 3 use the Mark I primary containment design. The main function of the Mark I containment design is to accommodate pressure and temperature conditions within the drywell resulting from a LOCA or a reactor blowdown through the MSRV discharge piping and, thereby, to limit the release of fission products to values which will ensure off-site dose rates below the C.-:FR-1-O GFRI16b,*76 limits. In the event of a pipe break in the drywell, water and/or steam from the reactor pressure vessel (RPV) are discharged into the drywell. The resulting increase in the drywell pressure forces the water and steam, along with non-condensable gases initially existing in the drywell, through the vents which connect the drywell to the suppression pool. During a reactor blowdown through the SRVs, the steam is directly discharged into the suppression pool. The reactor blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel dome pressure and the mass and energy of the fluid inventory in the RPV.
The long-term heatup of the suppression pool following a LOCA is governed by the capability of the Residual Heat Removal (RHR) System to remove decay heat which is transferred from the RPV to the suppression pool.
The Primary Containment System requirements are:
Design Pressure Design Temperature 56 psig 281°F Minimum containment overpressure following a LOCA and its affect on NPSH for Core Spray and RHR pumps is discussed in Chapter 6.5.5.
BFN AST UFSAR Changes Page 5 14.6.3.3
Fission Products Released to Primary Containment The following assumptions and initial conditions were used in calculating the amounts of fission products released from the nuclear system to the drywell:
- a.
Source terms based on the ORIGEN computer code with a 1.02 multiplier per Regulatory Guide
- .i-i41.183.
- b.
The reactor has been operating at design power (345*
- MWt) for a 24 month fuel cycle. The totalaveras;e.fuel burnup is e,100 ct fl
.... r dainventorys 5 of 37c isotope nr*:io:ir to the acc*iident.:!**:::*:i:*
cis on hndu re ret ofr the pqilibrim raryontaine nt ob e 1s hAeondary containmen a
are oftinen peneration sleake the TechnicliSec lea ag li it Pri ar nen at o ph r is.............................
via.ma.n.steam.isolation conele ider t nolude 0."h s i drenssure As.bine anthe conEesr Pri try t dontainent atmospheren ofis euiiru relasddietl oatectianidby Gasnratentor Systemoe duriutnfguc operation ocrl~d f thisContaincent AmspeicDlui1 CD pyrtent Pim arym cortinmntc atodesp here isreleasedtor thelto oate stackins viia.leakag ofhemihardenem werel veone isolatien vanves.uThe Emeguen Core Tablen 14.6-tgivs theCS leudak core**;
invntor ofe seachiotoendteiary containment.
h olwn assmpia ons werkae.ue 14.6lultig.heam5ns F
ission Prdcpeesromut Primeared Crmteprmrontainment:
Fiso rdcsaerlasedfrmThe primary containment to re thee secondary{ii i *i~i conTainenvi primary t
eoda containmentpetrio leakrage wat theecnicaltw Specifcantio contaimen atopherea (is reesddietytfteSady)a.rametSse inclcuatinth amut of fisio products relase frmteprmrtonamet TEQ~ft3. ThMe. dQ-3e1 volmis I59O0 f, ndR-t-rs as66cvoumWi BFN AST UFSAR Changes Page 6 14.6.3.4
- c.
The four main steam lines are assumed to leak a total of 4-68 150 scfh which Q..s the Technical Specification limit.
- d.
CAD System flow rate is 139 cfm for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 days, 20 days, and 29 days.
- e.
The hardened wetwell vent isolation valves leak a total of 10 scfh to the top of the offgas stack. This leakaae is assumied tobeWin at eiaht hours
- f.
Five gpm ECCS leakage into secondary containment in accordance with NUREG-0800, Section 15.6.5, Appendix B.
W".
No credit is f
tae stwvixremovalj nin the contai*n'Mnt 14.6.3.6 Fission Product Release to Environs Secondary Containment Releases The fission product activity in the secondary containment at any time (t) is a function of the leakage rate from the primary containment, the volumetric discharge rate from the secondary containment and radioactive decay. During normal power operation, the secondary containment ventilation rate is 75 air changes per day; however, the normal BFN AST UFSAR Changes Page 7
ventilation system is turned off and the Standby Gas Treatment System (SGTS) is initiated as a result of low reactor water level, high drywell pressure, or high radiation in the Reactor Building. Any fission product removal effects in the secondary containment such as plateout are neglected. The fission product activity released to the environs is dependent upon the fission product inventory airborne in the secondary containment, the volumetric flow from the secondary containment, and the efficiency of the various components of the SGTS.
The following assumptions were used to calculate the fission product activity released to the environment from the secondary containment:
- a.
The primary containment atmosphere leakage to secondary containment mixes instantaneously and uniformly within the secondary containment.
- b.
The effective mixing volume of the secondary containment is 1,931,502 ft3 (50%
- c.
The SGTS removes fission products from secondary containment. If only two of the SGTS trains are in operation (i.e., SGTS flow of 16,200 cfm), a short period exists at the start of the accident during which the secondary containment becomes pressurized relative to the outside environment..ur.ng.this sh....t.me oriort ision oroduct release t;6i imes soeif#ied by RG 1.1iiiii*!*ii83.
Once the secondary containment pressure is reduced below atmospheric pressure, all releases from secondary containment to the environment are through the SGTS filters via the plant stack. If all three trains of SGTS are in operation (i.e., SGTS flow of 24,750 cfm), all releases to the environment from secondary containment are through the SGTS filters via the plant stack. The case with three trains in operation is the limiting condition.
- d.
The Containment Atmospheric Dilution (CAD) System operates for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a flow rate of 139 cfm at 10 days, 20 days, and 29 days post accident.
This flow is filtered via the SGTS filters.
- e.
The EGGS systems leak reactor coolant directly. to the secondary{ containment.
The maximum water temperature is 4*7-l
,.ta2120i*F. The
- G~ volume available for mixing" is 444i*2: i.41il~~i~E5 ft3. Ten percent of the iodine in the EGGS waiter eka is assumed to become airborne.
- f.
Filter efficiency for the SGTS was taken as 90 percent for organic and *%
inorganic (elemental) iodine.
- g.
Release to the environment from the plant stack is composed of three flow paths. A continuous ground level release of 10 cfm occurs at the base of the stack. This flow results from SGTS leakage through the backdraft dampers in the base of the stack. Subsection 5.3.3, "Secondary Containment System BEN AST UFSAR Changes Page 8
Description" describes the backdraft dampers. The 10 cfm leakage mixes uniformly within the rooms at the base of the stack (4 i,
ethe*:iiQ*:m Volue, f 9 2 ft3). The remaining SGTS flow exits the stack at a height of 183 meters above ground elevation. The hardened wetwell vent isolation valves leak a total of 10 scfh to the top of the offgas stack with a delay of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the leakage to reach the stack. The hardened wetwell vent isolation valve leakage enters the stack above the divider deck and exits the top of the stack.
- h.
Fumigation conditions exist for tfirst 30 minutes when tih oistaccident:
Atmospheric dispersion coefficients, X/Q, for elevated releases under fumigation conditions, elevated releases under normal atmospheric conditions and ground level releases at the base of the stack are used. X/Q values applicable to the time periods, distances, and geometric relationships (offsite and control room) are shown in Table 14.6-8. Control room X/Q values for the base of the stack releases are calculated using the computer code ARCON96. For sites, such as BFN, with control room ventilation intakes that are close to the base of tall stacks, ARCON96 underpredicts the X/Q values for top of stack releases; therefore, top of stack releases to the control room intakes are evaluated using the methods of Regulatory Guides 1.145 and 1.111.
"MSeac-ma do.io.
Nte thatte eec s fato to e
Teeatilatgone~does from brimtyhothnen top Mands botismo transferred 1tco reteain Toine mhighprnessure ansdlow p vacthr f1.r am line an2 to the ICRP
.. dnservion
..ater
.te
.o tr.eaten durTn tlo path form to 30 s.tiamle frinhelamae i
filrlm dthie tubin an13d5ermgae.t h ubiedc n
Main~~ctv Steam Islaio Vav Laa elae Thbequently ros exandusetedtosethe atmosphrelviaste oturbinsem bioldtinroof velts (MIV):*i*
The leakag]e fromprimaycontainmetviatheMfS is.` transferre 1)toh*
0e man th turbie (highupessr nolwpesuefi the four steam lines and
- 2) cnene to hodu nltheoufisn products in the MSIV leakage effluent. The following assumptions were used to calculate the fission product activity released to the environment from the turbine building:
- a.
The four main steam lines are assumed to leak a total of 4":'150 scfh which beurid j~ the Technical Specification limit. The direct leakage path to the turbines processes only 0.5% of the total leakage. The remainder goes to the condenser via the ALT flow path. The main steam piping from the outermost isolation valve up to the turbine stop valve, the bypass/drain piping to the main condenser and the main condenser will retain their structural integrity during and following a safe-shutdown earthquake (SSE).
- b.
Thd'e freout on tre lweprssur tureite i 51,00..
dande h the fr Rol R.
N rdtiaem ora holdu Dricnthe urin te btuamlding.
adTh anc~dn h.Grsn bevenlcuatmopedruicaA dispe.son co~
tewlmxdefficientsc/,
oelae mrode th d
turbin te buildingroo velnts. Rppmovable in sthea ime conriols volumnes, seand o tgevolmetri relationshipsr(offsit andcountrol from are shownino Tabe 14.6-8.~r vCocntro dtrioo ton valuesmaencalculatned reusing the opteamrnsi cAlCuOated u1i.6. the Baixorlodqicl offectsC-64 Eeeta oie eoali h mhOAn pronides ter iost cservtvere rassumeditolbelae the same psifory oandcl sconThearee voluaiment ofd theuorssurvesa turies isu51i000dftiganbtheifreecvolumeio d therm condensertisctikent a
0 offstn thentrotl ond ensoner volueso. 3100f 3 Ofie Grounsevestopei iprincefcins (Q o eessfo h
Ofsturbines builingeroof vensutn applicabe atovithe tielperiods distaeniones t
and a
goneometrc rtelatonshis of (offaite andcodntro aroom)e are s-hown inTal168
- f.
ICRP 30
- iodine dose conversion factors are appl:~i ied.*:*~
i teeclso srecoundary cotanEnts and,46 mthus),
sevsasd the bourspnding 30desg asi acidienit ine areapltinzoe(LZ boundary (EAB,(146 meters),.n h
orsodn 3-a EEdssa h low.......
poultinzoe.LZ)bondr.(.20meer)
BFN AST UFSAR Changes Page 10 Thp nff--zifin rin.qim.q nrim rnlrtJlnf.r PH i vcinn
The largest calculated total offsite dose is well within the 10 CFR
`80 f0.67 lim't Control Room accounts for the atmospheric dispersion to the dual control room intakes by use of appropriate X/Qs and models the control bay habitability zone filtered pressurization flow (3000 cfm), unfiltered inleakage (3717 cfm), Control Room Emergency Ventilation stem ( RE.VS) fter absorp on 90%
n................
.................. )
Y (
)
P
(
occupancy times, breathing rates in accordance with Regulatory Guide I ~183! and calculates the *:m*:arbet:::
7-::nd:::h i:: :::iE:doses Atmospheric dispersion coefficients are based on release point, geometric relationship of the release point,.and receptor and atmospheric conditions based on site specific meteorological data. The:
The direct gamma-dose contribution from the piping inside secondary containment~ iad t s o r c a e an..............
t...
are included. One section of core spray piping in each unit is routed just outside the common Control Building/Reactor Building wall. This piping will be carrying prs~esuppression eh*mer water in the event of a LOCA.
All of these exposure mechanisms (filtered pressurization flow, unfiltered inleakage, eeesand direct dose) are combined to produce a total control room dose for the
- :*.* *.i.*
- .* J*.*#.*..*.*;.
J
[:
rlaconse fnd the atmseherith thrpeersiofn tn operdatcon wthral roominakes byeune ofile and via the plant sack are negligible Since cREvS has dual aIr Intakes placed on opposite sides of the control building and can function with a single active failure in the inlet isolation system, in accordance with NUREG-0800, the control room dose is divided by a factor of 2 to account for dilution effects. The 30 day integrated post accident doses in the control room are within the limits of 5 REM receptooa, a0mospheri, condi0iots thbased thon as specified in 10 CER 5067-ila BEN AST UFSAR Changes Page 11
14.6.4 Refueling Accident The current safety evaluation for the Refueling Accident is contained in the licensing topical report for nuclear fuel, "General Electric Standard Application For Reactor Fuel," NEDE-2401 1-P-A, and subsequent revisions thereto. Accidents that result in the release of radioactive materials directly to the secondary containment are events that can occur when the primary containment is open. A survey of the various plant conditions that could exist when the primary containment is open reveals that the greatest potential for the release of radioactive material exists when the primary containment head and reactor vessel head have been removed. With the primary containment open and the reactor vessel head off, radioactive material released as a result of fuel failure is available for transport directly to the reactor building.
Various mechanisms for fuel failure under this condition have been investigated.
Refueling Interlocks will prevent any condition which could lead to inadvertent criticality due to control rod withdrawal error during refueling operations when the mode switch is in the Refuel position. The Reactor Protection System is capable of initiating a reactor scram in time to prevent fuel damage for errors or malfunctions occurring during deliberate criticality tests with the reactor vessel head off. The possibility of mechanically damaging the fuel has been investigated.
The design basis accident for this case is one in which one fuel assembly is assumed to fall onto the top of the reactor core.
The discussion in Subsections 14.6.4.1 and 14.6.4.2 provides the analyses for the dropping of a 7 x 7 assembly and a 8 x 8 assembly. The analyses for all current General Electric product line fuel bundle designs are contained in supplements to NEDE-2401 1-P-A. The NEDE evaluates each new fuel design against the 7x7 fuel design for the original core load. The 7x7 fuel handling accident resulted in 111 failed fuel rods. Fo 8 fuel dei.
h atvt eesddetau~adigacd will bei~ h~n 80% of the ~i6acU@ait! el*e byth oigina* The hitoicl aoind currten
=Page 14
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16ivt 'ilb ostedby.% the o
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- f. other u...l ha b e..p.rformed as a forA ecoftseuel;4 tvees fti isb
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aatdin ofih1BFNiU ASTt! UFAUhne Pagoe 13k'j1
Fission Product Release From Fuel The folloing radiological consequences arc baseid on axa fuel. Fission product release estimates for the accident are based on the following assumptions:
- a.
i oraco ue a M*n aveAr~rageMM`5.raito nin or i~ucy tcsg er uto24 hor rirt teacdet hi su ptio rsu in a qiiru time allows time for the reactor to be shut down, the nuclear system depressurized, the reactor vessel head removed, and the reactor vessel upper internals removed. It is not expected that these evolutions could be accomplished in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b GuThe activityinthefuelbundlesdetý erineds the w
a yoRhorGENscod.
a co oowerof 431 MI moifie wit a pwer eakig fator f.1..andReguator Guie -4~113 owe fcto o 102 it a ecy f 2 hur........
dc.
One hundred eleven fuel rods are assumed to fail. This was the conclusion of the analysis of mechanical damage to the fuel based on the GE 7x7 fuel design.
14.6.4.4 Fission Product Release to Secondary Containment The following assumptions were used to calculate the fission product release to the
- ... *.. *. `.
secondary containment (oer*
jReaulatorvGuMide 118:
- a.
Fraction of Fuel Rod Inventory Released (ininite decamination ornulides othe6r"tha ioin n.l
- ~e)
Noble Gases (Except Kr 85)
Kr 85 Iodines exce b11131
.......1.....,..
3@IQ
- ..percent percent 0& -
ercent BFN AST UFSAR Changes Page 14 14.6.4.3
- b.
Iodine Decontamination Factor in Reactor Cavity Pool Water 14.6.4.5 Fission Product Release to Environs The following assumptions and initial conditions are used in calculating the dose existing at the exclusion area boundary and, at the low population zone6-,i~ihedjf c6bntrolroom.oieratof due to fission product release.
I:
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- b. Nocredi is tken fr isoationof th contol.rom.n.rfor....f..e..n.b..th BFN AST UFSAR Changes Page 15
==============
=================
CI n:lRoo F........m.
000 W, The design basis fuel handling accident assumes that during the refueling period a fuel bundle is dropped into the reactor cavity pool. The dropped fuel bundle strikes additional bundles in the reactor core fracturing 111 fuel pins (assuming GE 7x7 fuel u. '.
- t'. *.'
t.'.............
- :* ii.......
design). Ta ecntfth h1oe iQtps netoypus eFcent of 611i noble eass ivetor (ecet K 85whch s 0 prcet f tis nvntoy~The frventory.
dsopctibedabove will be released from the fractured fuel rods. An Vera!l eIentrive decontamination factoro Comiior eentlaiodine I
of foFonrdhe are applicable for iodine released at depth under water. The radioactive releases to the air space above the pool are released t-h rough the p-refue ling zoneiventilation and the StandbyGa~
Treatmet System nastantaneoul i~tIe 'tWsher wt n ol assumptions used to evaluate the fuel handling design basis accident event are defined in Nuclear Regulatory Commissions Regulatory Guide' 1.--2&% 1*.1.83. Further guidance is contained in the standard review plans in NUREG-BOD, Section RDA.0.
.i.i t..*.ken r.....
the
..r.......
o e th.
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p.lled t.ru the vi at..i....
.yetc The total activity released is greater for a fuel handling accident in the reactor cavity pool than for an accident in the fuel storage pool. Normally, the number of fuel rods fractured in a drop into the reactor vessel pool is slightly larger than the number of rods fractured in a drop into the storage pool. This provides a bigger source for the vessel event.
le.
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tf:the BFN AST UFSAR Changes Page 16
the dG.se d level I.. I.................................................
-h6'fel6b'§e of thdIbA I..................
I..........
14.6.4.6 Radiological Effects The radiological exposures following the refueling accident have been evaluated in the control room, at the site boundary, and at the LPZ boundary. The calculated dose assumes that all of the byp activity is exhausted instantneou ly through a roof configurationaok ofe thooto avniain iNN*::*:*:i:*:NN*f:.N*:*~i!*.i:*:.*:N*NN*:*i~i*N i::*:i~iii*::.i*:*
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As pa~of theBFN Po er Up tc Ii~sing.........he.........P...r..r.t.o..th radio og~
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14.6.5.2 Radioactive Material Release 14.6.5.2.1 Assumptions The following assumptions are used in the calculation of the quantity and types of radioactive material released from the nuclear system process barrier outside the secondary containment:
- a.
The amounts of steam and liquid discharged are as calculated from the analysis of the nuclear system transient.
- b.
The concentrations of biologically significant radionuclides contained in the coolant discharged as liquid (which subsequently flashes to steam) and the coolant discharged as steam are based on the ANSI/ANS-1 8.1-1984, "Radioactive Source Term for Normal Operation of Light Water Reactors" methodology. The halogens considered are 1-131, 1-132, 1-133, 1-134, and 1-135.
The values obtained by the ANSI/ANS-18.1 evaluation are then scaled to represent a dose equivalent 1-131 concentration of 32 ýtCi/W"J which is aeater than the 26 u.ifam max*ýl im.i,ý'10*'iium i."b.".*--"i..Tec.'hn'icalI Speecificat-bn limit and 10 times the equilibrium value for continued full power operation allowed by Technical Specifications. $io hsvlei 0times the~o quirumxaufocotnued ful poer
~ortio alowe byTehnical Specifications ýands serat orders or:
- c.
The concentration of noble gases leaving the reactor vessel at the time of the accident are based on the ANSI/ANS-1 8.1 concentrations with an appropriate scaling based on N EDO-i10871, "Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Terms".
- d.
It is assumed that the main steam isolation valves are fully closed at 5.5 seconds after the pipe break occurs. This allows 500 milliseconds for the generation of the automatic isolation signal and 5 seconds for the valves to close. The valves and valve control circuitry are designed to provide main steam line isolation in no more than 5.5 seconds. The actual closure time setting for the isolation valves is less than 5 seconds.
- e.
Due to the short half-life of nitrogen-16 the radiological effects from this isotope are of no major concern and are not considered in the analysis.
- f. Atmospheric dispersion coefficients, X/Q, for elevated releases under fumigation conditions, elevated releases under normal atmospheric conditions and ground level releases at thsebase of the stack are used. Aniccntainsith a aropriat1evil
- d. Itisassued that thle mair sthea isntolatroom walvs areflly loshed ~ acco.rdecnce ai theip burea Ge o11cur.
Thise aowis D. X/Q values applicable to the time periods, distances and gneometric relationshiPs.(offsiteand, control. rom) are sahown ivn Tarbl e 14.6-8.
arntroli rm pXrov les mri scteal l ine its BEN AST UFSAR Changes Page 19
p terod..*
14.6.5.2.2 Fission Product Release From Break Using the above assumptions, the following amounts of radioactive materials are released from the nuclear system process barrier:
Noble gases Iodine 131 Iodine 132 Iodine 133 Iodine 134 Iodine 135
ý413-T, x 102 C i
- 5. 42 83.5x 101 Ci 4327-x 102 Ci 6
x 49X 102 Ci X 10 O ci The above releases take into account the total amount of liquid released as well as the liquid converted to steam during the accident.
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..........iiii~~~~i i ii*i iiiiiiii~ ~ ~ii iiiiii~i i ! i whereni=L 14.6.5.3 Radiologqical Effects The control room dose is divided by 2 because of tl" intake rmnfim irnfinn nf thp.nntrnl hnv vxzntilfinn Since all of the activitly is: relasd o h rwonetlin the fo4m.fi a puff, the doses indcatd
.e maximmvlesrgdlsofhtdoepidisengelue.
It is concluded that no danger to the health and safety of the public results as a consequence of this accident.
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Table 14.6-8 (Sheet 1)
VALUES FOR X/Q FOR ACCIDENT DOSE CALCULATIONS Time Period Top of Stack Releases (LOCA.& CRDA.*
0-0.5 hrs*
0.5-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days Base of Stack Releases (LOCA, & CRDA, 0-2 hrs 2-8 hrs 8-24 hrs 1-4 days 4-30 days Ul Intake 3.40E-5 9.08E-1 3 3.41 E-13 2.09E-13 7.21 E-1 4 1.57E-1 4 2.OOE-4 1.28E-4 5.72E-5 4.05E-5 3.09E-5 Control Room (sec/m3)
Unit 3 Intake 3.02E-5 1.41 E-7 4.50E-8 2.54E-8 7.36E-9 1.24E-9 8.60E-5 6.46E-5 2.80E-5 2.OOE-5 1.53E-5 Site Boundary (sec/m 3) 2.35E-5 1.19E-6 2.62E-4 LPZ Boundary (sec/m 3) 1.26E-5 1.13E-6 5.75E-7 4.1 0E-7 1.97E-7 6.88E-8 1.31 E-4 6.61 E-5 4.69E-5 2.23E-5 7.96E-6 BFN AST UFSAR Changes Page 26
Table 14.6-8 (Sheet 2)
VALUES FOR X/Q FOR ACCIDENT DOSE CALCULATIONS Time Period Refuelinq Vent Releases (FHA Only) 0-2 hrs Turbine Building Exhaust Release (MSLB Only) 0-2 hrs
- Bounded by the Unit 3 Intake Ul Intake 4.60E-4 Control Room (sec/m3)
Unit 3 Intake Site Boundary (sec/m3)
LPZ Boundary (sec/m3) 2.62E-4 1.31 E-4 2.~E **6.41E.4 S;(*,
741 ý4 q
4*9 2.17E-4 1.64E-4 7.89E-5 4.33E-5 3.35E-5 2.62E-4 1.31 E-4 6.61 E-5 4.69E-5 2.23E-5 7,96E-6 Note: Current UFSAR value reflects change to correct typo since issuance of Amendment 19.