Text

License Amendment Request Re Full-Scope Application of NUREG-1465 Alternative Source Term Insights
	ML021350015
Person / Time
Site:	River Bend
Issue date:	04/24/2002
From:	Hinnenkamp P; Entergy Nuclear South
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NUREG-1465, RBG-45930
	Download: ML021350015 (221)
	v • d • e

Entergy Nuclear-South River Bend Station 5485 U.S. Highway 61 PO. Box 220 St. Francisville, LA 70775 vm~w nteWFax Tel 225 381 4374 225 381 4872 phinnen@entergy.com Paul D. Hinnenkamp Vice President, Operations River Bend Station RBG-45930 April 24, 2002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

River Bend Station, Unit 1 Docket No. 50-458 License Amendment Request Full-Scope Application of NUREG-1465 Alternative Source Term Insights

Dear Sir or Madam:

Pursuant to 10CFR50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for River Bend Station, Unit 1 (RBS). This amendment revises specific licensing and design bases to reflect application of alternative source term methodology.

On December 23, 1999, the NRC published new regulation 10 CFR 50.67, "Accident Source Term," in the Federal Register. This regulation provides a mechanism for licensed power reactors to replace the traditional source term used in design-basis accident 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 should apply for a license amendment under 10 CFR 50.90.

Regulatory guidance for the implementation of the alternative source term is provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000. This regulatory guide provides guidance to licensees of operating nuclear plants on acceptable applications of alternative source terms. The use of an alternative source term changes only the regulatory assumptions regarding the analytical treatment of the design basis accidents.

The alternative source term analyses for River Bend Station were performed following the guidance in accordance with Regulatory Guide 1.183 and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms." The analyses included the control rod drop, fuel handling, loss of coolant, and main steam line break accident scenarios.

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Letter RBG-45930 Page 2 of 4 The proposed changes to the current licensing basis for River Bend Station that are justified by the alternative source term analyses include:

"* Redefining Dose Equivalent 1-131.

"* Changes to Standby Gas Treatment System (SGTS) filter efficiency requirements.

"* Changes to the secondary containment post-LOCA drawdown time requirements.

"* Changes to Secondary Containment and Fuel Building Isolation Instrumentation requirements.

"* Changes to Control Room Fresh Air System (CRFA) filter efficiency requirements, unfiltered in-leakage assumptions, automatic initiation functions, and requirements during fuel movement.

"* Changes to Control Room Air Conditioning System requirements during movement of irradiated fuel.

"* Removal of Fuel Building and Fuel Building Ventilation System (HVF) requirements.

"* Changes to primary containment leakage limits.

"* Changes to drywell airlock and containment airlock seal air system leakage limits.

"* Changing containment integrity requirements during movement of irradiated fuel.

"* Changes to electrical shutdown requirements during movement of irradiated fuel.

Additional information is attached to this letter to complete the amendment request. Attachment 1 contains Entergy's Analysis of Proposed Technical Specification Change. Attachment 2 contains the marked-up pages of the Technical Specifications showing the proposed changes. consists of a marked-up copy of the Technical Specification Bases associated with this proposed change. Attachment 4 provides a commitment identification form. Attachment 5 provides a Post-LOCA Suppression Pool pH Evaluation Summary. Attachment 6 provides the results of Atmospheric Dispersion Factors (X/Q) Calculations. Attachment 7 provides a Loss of Coolant Accident (LOCA) Dose Analysis Summary. Attachment 8 provides a Fuel Handling Accident (FHA) and Light Load Drop Accident (LLA) Summary. Attachment 9 provides Control Rod Drop Accident (CRDA) Summary and Attachment 10 provides a Main Steam Line Break (MSLB) Outside of Primary Containment Summary.

The proposed change has been evaluated in accordance with 10CFR50.91 (a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.

This submittal contains commitments that are identified in Attachment 4.

The NRC has recently approved a number of TS Amendments which fully utilize the AST methodology. Entergy has reviewed the previously approved amendments and information submitted to incorporate any lessons learned.

Letter RBG-45930 Page 3 of 4 Entergy requests approval of the proposed amendment by December 31, 2002. This request date will allow time for implementation prior to RF1 1, which will occur in the Spring of 2003.

Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Greg Norris at 225-336-6391.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 24, 2002.

Sincerely, Paul D. Hinnenkamp Vice President, Operations River Bend Station, Unit 1 PDH/GPN Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Changes to TS Bases pages (Information Only)

4. List of Regulatory Commitments

5. Post-LOCA Suppression Pool pH Evaluation Summary

6. Atmospheric Dispersion Factors (x/Q) Calculations

7. Loss of Coolant Accident (LOCA) Dose Analysis Summary

8. Fuel Handling Accident (FHA) and Light Load Drop Accident (LLA) Summary

9. Control Rod Drop Accident (CRDA) Summary

10. Main Steam Line Break (MSLB) Outside of Primary Containment Summary

Letter RBG-45930 Page 4 of 4 cc: U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 U.S. Nuclear Regulatory Commission Attn: Mr. David J. Wrona MS O-7D1 Washington, DC 20555-0001 Mr. Prosanta Chowdhury Program Manager - Surveillance Division Louisiana Department of Environmental Quality Office of Radiological Emergency Plan and Response P. O. Box 82215 Baton Rouge, LA 70884-2215

Attachment 1 Letter RBG-45930 Analysis of Proposed Technical Specification Change to Letter RBG-45930 Page 1 of 17

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-47 for River Bend Station, Unit 1 (RBS). The proposed changes will revise the Operating License and design bases to reflect application of alternative source term methodology (AST).

This submittal represents a full-scope implementation of the new source term. Design basis accident analyses have been revised to define the impact of the new source term on doses to the public at the site boundary and to the operator in the control room. The analyses also bounds the Thermal Power Optimization (Appendix K power uprate) presently being planned for River Bend Station (RBS).

The accident source term is a significant aspect of the design and licensing basis of the plant.

As an input to the accident analyses that form the basis for the design and operation of the unit, a change in the source term can impact both the postulated accident consequences and the margin of safety. For this reason, the NRC has determined that any change to the design basis to use an alternative source term should be reviewed and approved by the NRC in the form of a license amendment. No specific plant modifications are required to implement alternative source term. Any modification that utilizes the benefits of alternative source term will be evaluated under 10 CFR 50.59 as allowed per Regulatory Guide 1.183.

Entergy requests approval of the proposed amendment by December 31, 2002. This request date will allow time for implementation prior to Refuel Outage 11 (RF1 1), which is scheduled for the Spring of 2003.

2.0 PROPOSED CHANGE

S AND TECHNICAL ANALYSIS Technical Specification Section: 1.1, "Definitions" Description of Change:

Re-define "Dose Equivalent 1-131" and delete reference to TID-14844 Justification:

The current definition of Dose Equivalent 1-131 refers to TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," which is the basis for the previous dose methodology used at RBS. This definition is directly applicable to the maximum allowable coolant concentrations Technical Specification (TS) Surveillance Requirement (SR) contained in TS 3.4.8, "RCS Coolant Activity." The Bases for that TS explains that the purpose of the surveillance is to ensure that the source term potentially released is bounded by that assumed in the Main Steam Line Break (MSLB) analysis. The MSLB analysis (Attachment 10) used dose conversion factors from Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989, in scaling the core concentrations to the TS maximum values. Therefore, the reference to TID-14844 is no longer appropriate and must be deleted.

to Letter RBG-45930 Page 2 of 17 Technical Specification Section: 3.3.6.2, "Secondary Containment and Fuel Building Isolation Instrumentation" Description of Change:

"* Editorial changes deleting reference to Fuel Building Isolation Instrumentation in this section.

"* Revise LCO 3.3.6.2 to indicate manual isolation.

"* Editorial change to the Applicability statement with the deletion of Table 3.3.6.2-1.

"* Deletion of Table 3.3.6.2-1.

"* Modify Condition A to delete references and requirements related to Function 2.

"* Delete Condition B.

"* Editorial Change in Condition C.

"* Editorial change in Note.

"* Delete Surveillance Requirements 1 through 4.

"* Revise Surveillance Requirement 5 to verify each Secondary Containment Isolation subsystem actuates on a manual isolation signal.

Justification:

The fuel building ventilation system is no longer credited to mitigate the consequences of any design basis accident. The fuel building was previously removed from the secondary containment envelope via RBS TS Amendment 113. Under a design basis accident - loss of coolant accident (DBA-LOCA), all potential leakage paths to the fuel building were assumed to be released directly to the environment with no credit taken for holdup, dilution, or decay by the building or for filtration by the safety related filters. RBS TS Amendment 113 also removed the requirements for the fuel building during fuel handling other than during movement of "recently irradiated fuel."

The AST Fuel Handling Accident (FHA) analysis was performed in accordance with Regulatory Guide 1.183, Appendix B. That analysis did not credit filtration by the fuel building ventilation system. A decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was assumed in the AST analysis. This decay time was chosen since it bounds the minimum plausible time that irradiated fuel could be moved (i.e.,

reactor disassembly takes > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , therefore, irradiated fuel can not be moved without at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of decay in either the containment or the fuel building). Also, Technical Requirements Manual TR 3.9.10 DECAY TIME requires that the reactor be subcritical for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to movement of irradiated fuel in the reactor pressure vessel. Thus, River Bend has both a licensing requirement and a physical limitation that would prohibit the movement of recently irradiated fuel as defined using AST assumptions (i.e. fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). The FHA dose results to Letter RBG-45930 Page 3 of 17 demonstrate that the dose consequences meet the acceptance criteria of 10CFR50.67 and Regulatory Guide 1.183.

Currently, automatic initiation of Standby Gas Treatment System (SGTS) is assumed in the LOCA analysis. This initiation could occur as a result of either a LOCA (High Drywell Pressure or Reactor Low Water Level - Level 2) or a high radiation signal in the Main Control Room Radiation Monitors (TS 3.3.7.1). The AST analysis assumes manual initiation of the SGTS, thus, the High Drywell Pressure (Function 1 in Table 3.3.6.2-1) and Reactor Low Water Level Level 2 (Function 2) signals are not required to ensure accident assumptions are met. The manual initiation signal (Function 5) will be retained since it is required to meet accident assumptions. Both the high drywell pressure and reactor water level signals will notify operators in the main control room of a potential LOCA condition. Therefore, operator notification to take the appropriate manual actions in accordance with plant procedures is assured. Review of Table 3.3.6.2-1 indicates that only SR 3.3.6.2.5 pertains to Function 5, therefore, SR 3.3.6.2.1 through SR 3.3.6.2.4 are no longer required.

Technical Specification Section: 3.3.7.1, "Control Room Fresh Air (CRFA) System Instrumentation" Description of Change:

This Technical Specification is being deleted in its entirety.

Justification:

Filtration by the CRFA system is currently credited in all of the Design Basis Accidents: LOCA, FHA, Control Rod Drop Accident (CRDA), and MSLB. These analyses credit initiation of the system via either a LOCA signal (reactor water level 2, high drywell pressure, etc.) or via a high radiation signal from the intake radiation monitors. These signals tied with automatic initiation of the CRFA filters are the basis for this TS. The Bases for this section states "The ability of the CRFA System to maintain the habitability of the MCR is explicitly assumed for certain accidents as discussed in the USAR safety analyses."

The FHA (including the Light Load Drop Accident (LLA) analysis) and MSLB dose analyses do not credit filtration by the CRFA charcoal filters. One CRDA scenario credited the CRFA system, however, it assumed manual initiation of the system 20 minutes into the event. Both air intakes for the main control room have redundant radiation monitors which annunciate in the main control room. Therefore, operator notification to take the appropriate manual actions in accordance with plant procedures is assured. Automatic initiation of the system is not required to ensure that the dose consequences of a CRDA meet the acceptance criteria from 10CFR50.67 and Regulatory Guide 1.183. The CRFA filters are also credited in the LOCA dose analysis (Attachment 7). That analysis also assumes manual initiation of the system, thus, automatic initiation of the system is not required to meet the acceptance criteria from 10CFR50.67. The ability to manually initiate the system will be demonstrated via TS SR 3.7.2.3 which is being modified appropriately by this amendment request.

The CRFA system at RBS has two redundant intakes. The system is normally aligned to the main air intake which is located on the roof of the control building. Operators may also select the remote air intake which is located in the standby cooling tower. Credit is taken for manual to Letter RBG-45930 Page 4 of 17 selection of the more favorable air intake as allowed by the Standard Review Plan (SRP),

Section 6.4. This credit is also taken in the AST LOCA and CRDA (Mechanical Vacuum Pump (MVP) case) analyses. Review of the SRP indicates that redundant safety-related radiation monitors are required in each intake for this allowance to be credited. TS 3.3.7.1 does not apply to this credit as it addresses the automatic actions from the main air intake. It is not applicable to the remote air intake. Technical Requirements Manual TLCO 3.3.7.1 provides the requirements for the remote air intake monitors. This TLCO will also be applied to the main air intake monitors since the main air intake monitors have the same licensing function as the remote air intake monitors, i.e., provide indication to operators so that they can select the more favorable intake.

Technical Specification Section: 3.6.1.2, "Primary Containment Air Locks" Description of Change:

"* Revise applicability statement to remove requirement concerning "recently irradiated fuel."

"* Revise "Condition E" to remove requirement to suspend movement of irradiated fuel assemblies in the primary containment.

"* Revise SR 3.6.1.2.1 to remove requirements for annulus bypass leakage.

"* Revise SR 3.6.1.2.4 to increase the leakage rate for the air lock seal pneumatic system from 1.28 psig/day to 1.5 psig/day.

Justification:

Primary containment integrity is currently credited for two accident dose analyses contained in USAR Chapter 15, the DBA-LOCA and the FHA involving "recently irradiated fuel." Regulatory Guide 1.183, Appendix B, Section 5.3 states "If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time period." The AST FHA analysis (Attachment 8) does not credit containment integrity and assumes a 2-hour release in accordance with established guidance. A decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was assumed consistent with the current analyses. This decay time was chosen since it bounds the minimum plausible time that irradiated fuel could be moved (i.e., reactor disassembly takes > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , therefore, irradiated fuel can not be moved without at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of decay in either the containment or the fuel building). The FHA dose results demonstrate that the dose consequences meet the acceptance criteria of 10CFR50.67 and Regulatory Guide 1.183. Therefore, the Technical Specification Requirements are no longer applicable during fuel movement.

Annulus bypass leakage is leakage which will bypass the annulus and be released into the Auxiliary Building. All leakage paths (including the containment/fuel building personnel air lock and Inclined Fuel Transfer System [lIFTS] drain line) which can potentially bypass secondary containment are currently included in the Secondary Containment Bypass (SCB) summation (See TRM Table 3.6.1.1-1 for a list of Annulus Bypass penetrations). Since the "annulus bypass" leakage paths now lead to the Auxiliary Building, the air would be filtered by SGTS prior to release to the environment. The AST LOCA dose analysis does not assume an explicit to Letter RBG-45930 Page 5 of 17 "annulus bypass" leakage path. RBS performed a sensitivity study demonstrating that off-site doses are not sensitive to annulus bypass leakage as long as such leakage paths are filtered.

Annulus bypass leakage is only a small potion of the overall containment leakage term and both the Auxiliary Building and the Annulus are filtered by SGTS. Therefore, the impact to calculated AST doses is negligible and simplifying the model is acceptable. Thus, the annulus bypass leakage summation is no longer required and should be deleted from Technical Specifications.

The primary containment personnel air locks (PAL) are used to ingress and egress the primary containment. The PAL doors have a seal which remains pressurized via the seal air system.

Technical Specification SR 3.6.1.2.2 states that the minimum normal operation pressure for the primary containment air lock seal air flask pressure must be _>90 psig. The inflatable seals must remain above a pressure of 45 psig to maintain its integrity. The current allowable leakage rate of 1.28 psi per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on a period of 35 days ((90psig 45psig)/35days=1.286psi/day). Since dose calculations are performed for a period of 30 days a value of 1.5 psi/day is requested ((90-45)/30=1.5).

Technical Specification Section: 3.6.1.3, "Primary Containment Isolation Valves" Description of Change:

"* Revise SR 3.6.1.3.9. Specifically, increase the allowable secondary containment bypass leakage rate from "<1170,000 cc/hr when pressurized to _ŽPa" to "*<580,000 cc/hr when pressurized to _>Pa."

"* Revise SR 3.6.1.3.10 to set a single main steam line leakage limit of 50 scfh when tested at

>Pa.

"* Delete SR 3.6.1.3.12 (annulus bypass leakage rate summation).

Justification:

Secondary containment bypass leakage is leakage from the primary containment building which will bypass the annulus and auxiliary building, thus, it will potentially escape to the environment unfiltered. This leakage is independent of the overall containment rate summation (La). Prior to RBS TS Amendment 98 the Penetration Valve Leakage Control System (PVLCS) system was assumed to be manually initiated which terminated this release path early in the event. The PVLCS system was deleted via TS Amendment 98. The Inclined Fuel Transfer System (IFTS) drain line and the fuel building personnel air lock were added since the fuel building was removed from the secondary containment envelope via RBS TS Amendment 113. A review of the potential leakage paths for this summation concluded that a value of 580,000 cc/hr was appropriate for use in the AST LOCA Analysis (Attachment 7). The analysis dose consequences using this value met the criteria set forth in 10CFR50.67.

Annulus bypass leakage is leakage which will bypass the annulus and be released into the Auxiliary Building. All leakage paths (including the containment/fuel building personnel air lock and IFTS drain line) which can potentially bypass secondary containment are currently included in the Secondary Containment Bypass (SCB) summation (TS SR 3.6.1.3.9). Since the "annulus bypass" leakage paths now lead to the Auxiliary Building, the air would be filtered by SGTS prior to Letter RBG-45930 Page 6 of 17 to release to the environment. The AST LOCA dose analysis does not assume an explicit "annulus bypass" leakage path. RBS performed a sensitivity study demonstrating that off-site doses are not sensitive to annulus bypass leakage as long as such leakage paths are filtered.

Annulus bypass leakage is only a small portion of the overall containment leakage term, and both the Auxiliary Building and the Annulus are filtered by SGTS. Therefore, the impact to calculated AST doses is negligible and simplifying the model is acceptable. Thus, the annulus bypass leakage summation is no longer required and should be deleted from Technical Specifications.

The current TID dose analysis does not assume a failure of an MSIV. The AST analysis conservatively assumed that one MSIV failed (in addition to an EDG). The leakage rate of 50 scfh was assumed through the steam line containing the failed MSIV. This value corresponds to the proposed limit to be incorporated in SR 3.6.1.3.10.

Technical Specification Section: 3.6.1.10, "Primary Containment - Shutdown" Description of Change:

"* Revise applicability statement to remove requirement concerning "recently irradiated fuel."

"* Revise required action for Condition A to delete requirement to suspend movement of recently irradiated fuel.

Justification:

Regulatory Guide 1.183, Appendix B, Section 5.3 states "If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period." The AST FHA analysis (Attachment 8) does not credit containment integrity and assumes a 2-hour release in accordance with established guidance. A decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was assumed consistent with the current analyses. This decay time was chosen since it bounds the minimum plausible time that irradiated fuel could be moved (i.e.,

reactor disassembly takes > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , therefore, irradiated fuel can not be moved without at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of decay in either the containment or the fuel building). The FHA dose results demonstrate that the dose consequences meet the acceptance criteria of 10CFR50.67 and Regulatory Guide 1.183. Therefore, the Technical Specification Requirements are no longer applicable during fuel movement.

Technical Specification Section: 3.6.4.1, "Secondary Containment - Operating" Description of Change:

Revise SR 3.6.4.1.4. Specifically, revise the Auxiliary Building drawdown time from 13.5 seconds to 34.5 seconds.

to Letter RBG-45930 Page 7 of 17 Justification:

LOCA dose analyses do not credit secondary containment until an adequate vacuum is reached

(_<-0.25 in. w.g.). The AST LOCA dose analysis (Attachment 7) assumed manual initiation of SGTS and allowed an additional 10 minutes for an adequate vacuum to be established. This resulted in a total assumed Positive Pressure Period (PPP) of 30 minutes (20 minutes for Operator action and an additional 10 minutes to establish the required vacuum). Calculations used the GOTHIC computer code to demonstrate that the PPP assumed in the AST LOCA dose analysis are conservative for the annulus and auxiliary building, respectively. They determined a revised drawdown time for testing during non-accident conditions using current system parameters (flow rates, etc.). That analysis determined that an auxiliary building drawdown time of 38.5 seconds will ensure that the calculated PPP is conservative. This submittal requests a value of 34.5 seconds (-90% of the calculated value) for conservatism.

Technical Specification Section: 3.6.4.2, "Secondary Containment Isolation Dampers (SCIDs) and Fuel Building Isolation Dampers (FBIDs)"

Description of Change:

"* Editorial changes deleting reference to Fuel Building Isolation Dampers (FBIDs) in this section.

"* Revise applicability statement to remove requirement concerning "recently irradiated fuel."

"* Delete all requirements for "Condition D" Justification:

The fuel building ventilation system (including sub-components such as dampers) is no longer credited to mitigate the consequences of any design basis accident. The fuel building was previously removed from the secondary containment envelope via RBS TS Amendment 113.

Under a DBA-LOCA, all potential leakage paths to the fuel building were assumed to be released directly to the environment with no credit taken for holdup, dilution, or decay by the building or for filtration by the safety related filters. Amendment 113 also removed the requirements for the fuel building during fuel handling other than during movement of "recently irradiated fuel."

The AST Fuel Handling Accident and Light Load Accident analyses were performed in accordance with Regulatory Guide 1.183, Appendix B. That analysis did not credit filtration by the fuel building ventilation system. A decay time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was assumed consistent with the current analyses. This decay time was chosen since it bounds the minimum plausible time that irradiated fuel could be moved (i.e., reactor disassembly takes > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , therefore, irradiated fuel can not be moved without at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of decay in either the containment or the fuel building). The FHA/LLA dose results demonstrate that the dose consequences meet the acceptance criteria of 10CFR50.67 and Regulatory Guide 1.183. Therefore, the Technical Specification Requirements are no longer required for this system.

to Letter RBG-45930 Page 8 of 17 Technical Specification Section: 3.6.4.5, "Fuel Building" Description of Change:

, This TS may be deleted in its entirety.

Justification:

As discussed above, neither the fuel building nor containment integrity is credited in the FHA analysis. A 2-hour release is assumed in accordance with Regulatory Guide 1.183, Appendix B guidance. No credit is taken for filtration by the fuel building ventilation system's ESF charcoal filters. As such, this TS is not required to ensure that the dose consequences from a FHA (or LLA) meet the criteria set forth in 10CFR50.67 and Regulatory Guide 1.183. Therefore, this TS and its associated SR may be deleted.

Technical Specification Section: 3.6.4.7, "Fuel Building Ventilation System - Fuel Handling" Description of Change:

, This TS may be deleted in its entirety.

Justification:

As discussed above, neither the fuel building, nor containment integrity, is credited the FHA analysis. A 2-hour release is assumed in accordance with Regulatory Guide 1.183, Appendix B guidance. No credit is taken for filtration by the fuel building ventilation system's ESF charcoal filters. As such, this TS is not required to ensure that the dose consequences from a FHA (or LLA) meet the criteria set forth in 10CFR50.67 and Regulatory Guide 1.183. Therefore, this TS and its associated SR may be deleted.

Technical Specification Sections: 3.6.5.1, "Drywell" and 3.6.5.2, "Drywell Air Locks" Description of Change:

"* Revise SR 3.6.5.1.2 to increase the leakage rate for the air lock seal pneumatic system from 0.67 psig/day to 20.0 psig/day.

"* Revise SR 3.6.5.2.5 to increase the leakage rate for the air lock seal pneumatic system from 0.67 psig/day to 20.0 psig/day.

Justification:

Drywell integrity is credited in the containment and drywell pressure response analyses developed in support of the power uprate project approved via TS Amendment 114. The current leakage rate is based on a 30 day duration, i.e., drywell seal integrity is required for at least 30 days following a postulated LOCA. The proposed leakage rate for the drywell air lock seals is to Letter RBG-45930 Page 9 of 17 based on leakage over the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the seals could potentially fail due to the internal pressure not being adequate. A large break in the drywell would increase pressure rapidly and uncover the suppression pool vents. Containment pressure would increase steadily as a result of flow from the drywell. Following this initial blowdown period, the pressure in the drywell begins to drop due to steam condensation. Eventually the drywell pressure drops below the containment pressure and becomes a relative vacuum (see figures in RBS USAR Section 6.2). This occurs well before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Failure of the seals would tend to reduce containment pressure to less than what is analyzed since the drywell is at a lower pressure. Thus, the analyses for large breaks are conservative since the pressures they calculate bound the pressures that would be expected if the drywell air lock seals failed.

Intermediate and Small Break Accidents were also evaluated. The "short term" analyses show the drywell pressure is greater than containment pressure, however, the "long term" (>10,000 seconds) analyses show the drywell pressure drops below containment pressure well before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The reason for the differences lies with computer code limitations and decay heat assumptions. Since the drywell has a lower pressure than the containment 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the event, the calculated containment pressures are conservative and bound seal failure at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The AST LOCA dose analysis (Reference Attachment 7) treats the drywell and containment as two separate nodes. Early in the event (<20 minutes) the flow rate between the drywell and containment is based on the containment pressure response to a recirculation line break.

Regulatory Guide 1.183, Appendix A, Section 3.7 states "After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment." Since the containment and drywell are homogenized at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as required by Regulatory Guide 1.183, and seal failure would not occur until after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , there is no impact to the LOCA doses calculated using AST based assumptions.

The air lock seals remain pressurized via the seal air system. Technical Specifications SR 3.6.5.1.1 and SR 3.6.5.2.2 require that the minimum normal operating pressure for drywell seal air flask pressure must be >_75 psig. The inflatable seals must remain above a pressure of 55 psig to maintain its integrity. The current allowable leakage rate of 0.67 psi per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on a period of 30 days ((75psig - 55psig)/30days = 0.67 psi/day). The discussion above demonstrates that the seals are not required after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , therefore, a value of 20.0 psi per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is requested ((75psi - 55psi)/1day= 20.0 psi/day).

Technical Specification Section: 3.7.2, "Control Room Fresh Air (CRFA) System" Description of Change:

"* Revise applicability to remove requirements concerning movement of irradiated fuel.

"* Revise Condition C to remove requirement to suspend movement of irradiated fuel.

"* Revise Condition E to remove requirement to suspend movement of irradiated fuel.

"* Revise SR 3.7.2.3 to ensure that each CRFA subsystem actuates on a "manual initiation signal" rather than "an actual or simulated initiation signal."

to Letter RBG-45930 Page 10 of 17 Justification:

Filtration by the CRFA system is currently credited in all of the Design Basis Accidents: LOCA, FHA, CRDA, and MSLB. These analyses credit initiation of the system via either a LOCA signal (reactor water level 2, high drywell pressure, etc.) or via a high radiation signal from the intake radiation monitors. These signals tied with automatic initiation of the CRFA filters is the basis for this TS.

The FHA and MSLB dose analyses do not credit filtration by the CRFA charcoal filters. Since the FHA analysis assumed the system operates in normal mode for the duration, this TS is no longer applicable. The CRDA analysis (Reference Attachment 9) credited the CRFA system, however, that assumed manual initiation of the system 20 minutes into the event rather than automatic initiation. Automatic initiation of the system is not required to ensure that the dose consequences of a CRDA meet the acceptance criteria from 10CFR50.67 and Regulatory Guide 1.183. The CRFA filters are also credited in the LOCA dose analysis (Reference ). That analysis also assumes manual initiation of the system, thus, automatic initiation of the system is not required to meet the acceptance criteria from 10CFR50.67.

Technical Specification Section: 3.7.3, "Control Room AC System" Description of Change:

"* Revise applicability statement to remove requirements concerning movement of irradiated fuel.

", Revise Condition D to remove requirements concerning movement of irradiated fuel.

"* Revise Condition E to remove requirements concerning movement of irradiated fuel.

Justification:

Technical Specification Bases for Section 3.7 explains that the intent of the section is to ensure that required equipment is not affected by adverse environmental conditions, specifically, high temperatures. The applicability discussion states that "In Modes 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these modes. Therefore, maintaining the Control Room AC System OPERABLE is not required in Modes 4 and 5" except when a significant radiological release is available. The FHA no longer credits mitigation by any system including the CR AC and CRFA Systems. The dose consequences meet the criteria set forth in 10CFR50.67 and Regulatory Guide 1.183.

Therefore, this TS is no longer applicable.

to Letter RBG-45930 Page 11 of 17 Technical Specification Section: 3.8.2, "AC Sources - Shutdown" Description of Change:

"* Revise applicability statement to remove requirements concerning movement of irradiated fuel.

"* Revise Condition A to remove requirements concerning movement of irradiated fuel.

"* Revise Condition B to remove requirements concerning movement of irradiated fuel.

Justification:

The TS Bases for this section explains that this TS is applicable to ensure that "Systems needed to mitigate a fuel handling accident are available." Previously the CRFA and fuel building ventilation systems' charcoal filter trains were credited to mitigate the consequences of a FHA. The AST analyses do not credit these or any other systems to mitigate the consequences of a FHA. Based on the above discussion it can be concluded that accident assumptions will be met, therefore, this TS is no longer applicable to the movement of irradiated fuel.

Technical Specification Section: 3.8.5, "DC Sources - Shutdown" Description of Change:

"* Revise applicability statement to remove requirements concerning movement of irradiated fuel.

"* Revise Condition A to remove requirements concerning movement of irradiated fuel.

Justification:

The TS Bases for this section explains that this TS is applicable to ensure that "Systems needed to mitigate a fuel handling accident are available." Previously the CRFA and fuel building ventilation systems' charcoal filter trains were credited to mitigate the consequences of a FHA. The AST analyses do not credit these or any other systems to mitigate the consequences of a FHA. Based on the above discussion it can be concluded that accident assumptions will be met, therefore, this TS is no longer applicable to the movement of irradiated fuel.

to Letter RBG-45930 Page 12 of 17 Technical Specification Section: 3.8.8, "Inverters - Shutdown" Description of Change:

"* Revise applicability statement to remove requirements concerning movement of irradiated fuel.

"* Revise Condition A to remove requirements concerning movement of irradiated fuel.

Justification:

The TS Bases for this section explains that this TS is applicable to ensure that "Systems needed to mitigate a fuel handling accident are available." Previously the CFRA and fuel building ventilation systems' charcoal filter trains were credited to mitigate the consequences of a FHA. The AST analyses do not credit these or any other systems to mitigate the consequences of a FHA. Based on the above discussion it can be concluded that accident assumptions will be met, therefore, this TS is no longer applicable to the movement of irradiated fuel.

Technical Specification Section: 3.8.10, "Distribution Systems - Shutdown" Description of Change:

"* Revise applicability statement to remove requirements concerning movement of irradiated fuel.

"* Revise Condition A to remove requirements concerning movement of irradiated fuel.

Justification:

The TS Bases for this section explains that this TS is applicable to ensure that "Systems needed to mitigate a fuel handling accident are available." Previously the CFRA and fuel building ventilation systems' charcoal filter trains were credited to mitigate the consequences of a FHA. The AST analyses do not credit these or any other systems to mitigate the consequences of a FHA. Based on the above discussion it can be concluded that accident assumptions will be met, therefore, this TS is no longer applicable to the movement of irradiated fuel.

Technical Specification Section: 3.10, "Special Operations" Description of Change:

0 Administrative changes to reflect changes to LCO 3.3.6.2, LCO 3.6.4.2 and LCO 3.6.4.5.

to Letter RBG-45930 Page 13 of 17 Justification:

The changes to this section are administrative to reflect the changes to TS Sections 3.3.6.2, 3.6.4.2 and the deletion of TS Section 3.6.4.5.

Technical Specification Section: 5.5.7, "Ventilation Filter Testing Program (VFTP)"

Description of Change:

"* Revise 5.5.7 to delete the fuel building ventilation system (FBVS) requirements.

"* Revise the SGTS allowable penetration from 0.5% to 5.0%.

"* Revise the CRFA allowable penetration from 0.5% to 1.0%.

Justification:

The fuel building ventilation system is no longer credited to mitigate the consequences of any DBA. Section 5.5.7 ensures that the heaters preheat the air to meet humidity requirements of the ESF filter trains. Since the FBVS filter trains are no longer credited in any DBA dose analysis, the requirements may be removed from TS 5.5.7.

RBS committed to ASTM D3803-1989 via Technical Specification Amendment 115 which allows licensees to test to 50% of the margin assumed in the safety analyses. SGTS is only credited in the LOCA doses analysis which assumes 90% filter efficiency, therefore, the testing acceptance criteria is being revised to 5%. The CRFA filters are credited in the LOCA and the CRDA analyses. These analyses assume a filter efficiency of 98% which mandates a testing acceptance criteria of 1%.

Technical Specification Section: 5.5.13, "Primary Containment Leakage Rate Testing Program" Description of Change:

Revise 5.5.13 to increase the containment leakage rate from 0.26% per day to 0.325% per day.

Justification:

The containment leakage rate is a major assumption in the LOCA dose analysis. The AST LOCA analysis (Reference Attachment 7) assumed a containment leakage rate of 0.325% per day. The dose consequences of that analysis met the criteria set forth in 10CFR50.67, therefore, the increased leakage rate is acceptable.

to Letter RBG-45930 Page 14 of 17

3.0 BACKGROUND

The current dose methodology used at River Bend Station is based on NRC guidance from the 1960's and 1970's (Regulatory Guide 1.3, TID-14844, etc.). Due to the simplistic modeling techniques and limited knowledge at that time, many assumptions are grossly conservative which led to the stringent testing requirements presently contained in the RBS Technical Specifications (TS).

The Alternate Source Term (AST) project is the result of rigorous analyses performed to more accurately model the evolution of Severe Accidents. Since the methodology is more realistic, the dose consequences utilizing this methodology are often lower than the Design Basis Accidents scenarios' overly conservative assumptions. The NRC has recently approved several TS Amendments which fully utilize the AST methodology.

Regulatory Guide 1.183 was developed to provide regulatory guidance in utilizing the AST dose methodology. Analyses were developed to be consistent with the requirements provided within that Regulatory Guide. Initial application of AST to a plant must be reviewed and approved by the NRC. Once initial AST implementation has been approved by the staff and is part of the facility design basis, licensees may use 10CFR50.59 in assessing safety margins related to plant modifications and changes to procedures. Reg. Guide 1.183 allows for partial or full implementation of AST.

Note that a limited scope application of AST was previously approved by the NRC for RBS.

Specifically, AST was used to justify neglecting the Inclined Fuel Transfer System (IFTS) drain line as a source to operators stationed to manually close an isolation valve. See RBS Technical Specification Amendment 116 for more information.

4.0 TECHNICAL ANALYSIS

All of the major accidents (which could potentially lead to significant off-site doses) contained in USAR Chapter 15 were reanalyzed using the AST dose methodology. These events include the Control Rod Drop Accident (CRDA as contained in USAR Section 15.4.9), the Main Steam Line Break Outside of Containment (MSLB, USAR Section 15.6.4), the Loss of Coolant Accident (LOCA, USAR Section 15.6.5), and the Fuel Handling Accident (FHA, USAR Chapter 15.7.4). A number of supporting analyses, such as the Light Loads Analysis, were also reanalyzed using AST assumptions. Regulatory Guide 1.183 provides guidance in applying the AST insights found in NUREG-1465 to plant analyses. Attachments 7 through 10 of this submittal provide the summaries of these analyses.

5.0 REGULATORY ANALYSIS

5.1 Applicable Requlatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. The analyses used the assumptions and guidance provided by Regulatory Guide 1.183. No exceptions to Regulatory Guide 1.183 assumptions were taken for the FHA, CRDA, or MSLB analyses. The LOCA analysis had one deviation from Regulatory Guide 1.183, Appendix A assumptions. This deviation concerns the assumed liquid leakage to Letter RBG-45930 Page 15 of 17 rate from Engineered Safety Features (ESF) Emergency Core Cooling Systems (ECCS). This deviation is discussed in detail in Attachment 7 and is similar to assumptions contained in the current (TID) LOCA dose analysis.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any GDC differently than described in the SAR.

5.2 No Significant Hazards Consideration This proposed amendment to the River Bend Technical Specifications (TS) revises those specifications affected by the implementation of the alternative source term concepts in accordance with NUREG 1465. In addition, based on the alternative source term, changes are proposed to selected specifications associated with handling irradiated fuel in the primary containment or Fuel Building and CORE ALTERATIONS. The alternative source term changes affect the definitions, and the specifications for the Control Room Fresh Air System, Standby Gas Treatment System, Fuel Building Ventilation System and leakage rates for Primary Containment and the Personnel Airlocks seal air systems.

Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The alternative source term does not require modification of the facility; rather, once the occurrence of an accident has been postulated the new source term is an input to evaluate the potential consequences. The implementation of the alternative source term has been evaluated in revisions to the analyses of the limiting design basis accidents at River Bend Station. Based on the results of these analyses, it has been demonstrated that, even with the requested Technical Specification changes, the dose consequences of these limiting events are within the regulatory guidance currently approved by the NRC for use with the alternative source term. This guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms."

Because the equipment affected by the revised operational conditions is not considered an initiator to any previously analyzed accident, inoperability of the equipment cannot increase the probability of any previously evaluated accident. The proposed requirements bound the conditions of the current design basis fuel handling accident analysis which concludes that the radiological consequences are within the acceptance criteria of NUREG 0800, Section 15.7.4 and General Design Criteria 19. As noted above, with the alternative source term implementation, the acceptance criteria are also being revised. The results of the revised Fuel Handling Accident demonstrate that the dose consequences are within the NRC regulatory guidance. This guidance is presented to Letter RBG-45930 Page 16 of 17 in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms."

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes using the alternative source term dose methodology are analytical in nature and do not physically alter the facility or of any equipment within the facility. Similarly, the alternative source term does not create any new initiators or precursors of a new or different kind of accident. The proposed changes to the Technical Specifications, while they revise certain performance requirements, do not involve any physical modifications to the plant.

The proposed changes related to shutdown controls based on the alternative source term do not create the possibility of a new or different kind of accident from any previous analyzed.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The changes above are associated with the implementation of a new licensing basis for River Bend Station. Approval of the basis change from the original source term in accordance with TID-14844 to the new alternative source term of NUREG-1465 is requested by this submittal. The results of the accident analyses prepared in support of this submittal are subject to revised acceptance criteria. These analyses have been performed using conservative methodologies as outlined in the regulatory guidance and conservatively represent the requested Technical Specification changes. Safety margins and analytical conservatisms have been evaluated and are well understood. The analyzed events have been carefully selected and margin has been retained to ensure that the analyses adequately bound all postulated event scenarios. The dose consequences of these limiting events are within the acceptance criteria also found in the latest regulatory guidance. This guidance is presented in Regulatory Guide 1.183, 10CFR50.67 and Standard Review Plan Section 15.0.1, "Radiological Consequences Analyses Using Alternative Source Terms."

The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries as well as control room, are within the corresponding regulatory limits. In a similar way, the results of the existing analyses demonstrated that the dose consequences were within the applicable NRC-specified regulatory limit.

Specifically, the margin of safety for these accidents is considered to be that provided by meeting the applicable regulatory limit for Alternate Source Term methodologies, which, to Letter RBG-45930 Page 17 of 17 for most events, is conservatively set at, or below, the 10CFR50.67 limit. With respect to the control room personnel doses, the margin of safety is the difference between the 10CFR100 limits and the regulatory limit defined by 10CFR50, Appendix A, General Design Criterion (GDC) 19.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 Letter RBG-45930 Proposed Technical Specification Changes (mark-up)

Definitions 1.1 1.1 Definitions (continued)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, display, and trip functions, and channel failure trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);

and

b. Control rod movement provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in. be1.0 Po,,or~~"Iandr AotRatrSt3..~______

(conti nued)

RIVER BEND 1.0-2 Amendment No. 81

INSERT A Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1989.

Secondary Containment lIsolation 1 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment ........ Isolation&

LCO 3.3.6.2 The isolation Function hall be OPERABLE.

APPLICABILITY: L M tolaloQ L 62r*a ACTIONS

NOT I---------------------------------------------------------

Separate Condition entry is allowed for each cI B. Ono 9F merc auteFmatic Fu nctions With iecondap' isolation capability containment o~r f I f)

Required Action and .1.1 Isolate the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 3 ,."

associated Completion penetration flow path(s).

Time of Condition Apri.)

not met. OR (continued)

RIVER BEND ý3.3-58 Amendment No. 84,113

Secondary Containment Isolation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME (continued) Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolation dampers inoperable.

AND Place the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ventilation subsystem in operation.

OR Declare associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ventilation subsystem inoperable.

SURVEILLANCE REQUIREMENTS

........................ NO T E - ---------------------------------------------------------.

Surveillances, entry into associated Conditions aand Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains secondary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 C P-AG G-12 hour S R 3.3.6.2.2 1ýdr WXELFWJTGA--~

RIVER BEND 3.3-59 Amendment No. 84-, 113

Secondary Containment

Isolation I 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.2.3 -G-lIb.-hr ent.cp>e trp un1t.

SR 3.3.6.2.4 4.8e~

SR 3.3.6.2.5 f-O "P18 months svv4erv*e- soLc 0Ai Su~~~5~~~$~~' c atwc\~n 4 v&loei lv l~ov RIVER BEND 3.3-60 Amendment No. 84-,113

Secondary Containment Isolationn 1Z ý3.3.6.2

ý S Xt3 %A'oAct\\j Lef+/- P,ŽYk\.

RIVER BEND 3.3-61 Amendment No. 81-, 113

CRF Cyseni r~Lrmnati~on 3.3 INSTRUMENTATION 3.3.7.1 P emFeh --- iU1%.11t I 'De-L 3.3.7.1 System instrumentation for each Function in Table 3.3.7.1-1 shall be OPERABLE.

APPLICABIL Y: According to Table 3.3.7.1-1.

ACTIONS

-- - - - -- i - . - . - .*- T.------- NOTE ----------- ---............ . ..........

Separate Condition en y is allowed for each chan 1.

CONDITION REQU DACTION COMPLETION TIME

\ A. One or more channels required inoperable. A.1 Enter the Condition

referenced in Immediately

/th~* Tble 3.3.71-1 for B. As required by B.1 Declare ass ciated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action .1 CRFA subsyste discovery of and referenced in inoperable. loss of CRFA Table 3.3.7. 1. binitiation I *

both trip in capability AND B.2 Place channel in 24 ho rs trip.

.7(contin

-I RIVER BEND 3.3-68 Amendment No. 81

C.1 Declare associated CRFA subsystem inoperable.

Table 3.3

,e'apability in both trip systems AND C.2 Place channel 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months trip. /

D. As required by 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months from Required Action A.1 discovery of and referenced in loss of CRFA Table 3.3.7.1-1. initiation capability i n both trip systems 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months trip.

E. E.I Place the associated CRFA subsystem in I emergency mode.

C, or OR E.2 Declare associated CRFA subsystem inoperable.

TKýý' a 0 RIVER BEND 3.3-69 Amendment No. K- 95

zlA'System Isrmnto SURVEILLANCE REQUIREMENTS

.NOTES -- - -- - - -- - --..

1. Refer to Table 3.3.7.1-1 to determine which SRs apply fo each Function.

2. When a channel is placed in an inoperable status sol y for performance of required Surveillances, entry into associated Condi ions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provide the associated Function maintains CRFA initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CH . 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.7.1.2 Perform CHr ELFUNCTIONAL TEST. 92 days SR 3.3.7.1.3 /Calratethe trip units. 92 days SR 3.3.7.1. Perform CHANNEL CALIBRATION. 18 months SR 3..7.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months

-7J F:2 9 5Sage ý +eyt (0I'o~jfL~I RIVER BEND 3-.3-70 Amendment No. 81

APPLICABLE CONDITIONS MCODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIEC. PER TRIP REOtJI SURVEILL.ANCE ALLOWABLE FUNCTION CONDITIONS SYSTEMt ACTI A.1 REQUIREMENTS VALUE

1. Reactor vessel water 1.2.3 2 B SR 3.3.7.1.1 z -47 inches Level -Low Low.Level 2 SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5 2, Drywell Pressure -High 1.2.3 2 C SR 3.3.7.1.1 1.88 Dsid SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

3. Control Room Local 1.2.3 1 0 SR 3.3.7.1.1 Intake ventilation s 0 97 x 1O-5 (a).(b) SR 3.3.7.1.2 £JXifcc Radiation monitor SR 3.3.7.1.4 SR 3.3.7 1.5 (a) Dun! operations withi a potential for draining th~e reactor vessel.

(b) ring movement of irradiated fuel assemiblies inthe primary contair went or fuel building.

t-lh Ispotýe-RIVER BEND 3.3-71 Amendment No. 8f-g5- 119

Primary Containment Air Locks 3.6.1.2 3.6 CONTAINMENT SYSTEMS 3.6.1.2 Primary Containment Air Locks LCO 3.6.1.2 Two primary containment air locks shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

NOTES--- ----------------------

1. Entry and exit is permissible to perform repairs of the affected air lock components.

2. Separate Condition entry is allowed for each air lock.

3. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more primary ------------ NOTES--------

containment air locks 1. Required Actions A.1, with one primary A.2, and A.3 are not containment air lock applicable if both doors door inoperable, in the same air lock are inoperable and Condition C is entered.

2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

(continued)

RIVER BEND 3.6-3 Amendment No. 8- 85

Primary Containment Air Locks 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore air lock to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, or C not met in MODE 1, 2, or 3. D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and '.1 Suspend movcmcnt f immediatcly associated Completion irradiated fucl Time of Condition A, a.s...bi ~~90 psig.~~

~~SR 3.6.1.2.3 ----------------- NOTE--------------~~

~~Only required to be performed upon entry or exit through the primary containment air lock.~~

~~Verify only one door in the primary 184 days containment air lock can be opened at a time.~~

~~(continued)~~

~~RIVER BEND 3.6-7 Amendment No. Wi,84~~

~~Primary Containment Air3.6.1.2 Locks SURVEILLANCE REQUIREMENTS (continued)~~

FREQUENCY SURVEILLANCE Verify, from an initial pressure of 18 months SR 3.6.1.2.4 90 psig, the primary containment air lock seal pneumatic system pressure does nC decay at a rate equivalent to

~~>1.28 sig for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .~~

~~3.6-8 Amendment No. 81 RIVER BEND~~

~~PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)~~

~~SURVEILLANCE FREQUENCY SR 3.6.1.3.8 Verify in-leakage rate of~~

~~340 scfh for 18 months each of the following valve groups when tested at 11.5 psid for MS-PLCS valves.~~
~~a. Division I MS-PLCS valves~~
~~b. Division II MS-PLCS valves SR 3.6.1.3.9 ---------------- NOTE---------------~~
Only required to be met in MODES 1, 2, and 3.
Verify the combined leakage rate for all In accordance secondary contaii ent bypass leakage with the paths is * ,, cc/hr when pressurized Primary to Ž Pa, 5 **Containment Leakage Rate Testing Program (continued)
RIVER BEND 3.6-18 Amendment No. 84984 98
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.10 ------------------ NOTE ---------------------
Only required to be met in MODES 1, 2, and 3.
~~.-iy leakage rate through the valves In accordance served by each division of MS-PLCS is with the Primary~~
_ 150 scfh per division when tested at > Pa. Containment Leakage Rate Testing Program SR 3.6.1.3.11 ------------------- NOTE--------------
Only required to be met in MODES 1, 2, and 3.
Verify combined leakage rate through In accordance with hydrostatically tested lines that penetrate the the Primary primary containment is within limits. Containment Leakage Rate Testing Program (continued)
J.e- (20- Ck In 6z Ve e CL 3 e-j 0
RIVER BEND 3.6-19 Amendment No. 84, 84
PCIVs 3.6.1.3 SR RIVER BEND 3.6-20 Amendment No. 8+,84
Primary Containment-Shutdown 3.6.1.10 3.6 CONTAINMENT SYSTEMS 3.6.1.10 Primary Containment-Shutdown LCO 3.6.1.10 Primary containment shall be OPERABLE.
I APPLICABILITY: Durng
~
o':mAt of recently irradiatod fuel Assomblia, in the
. npta i ..ment.
r.._ry A I
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.4 Suspend ...D ,-a- 1. #fe*4--*e.e I inoperable. recently irradited-4fuel ass~enlies inthe primary A./ Initiate action to suspend Immediately
" OPDRVs.
RIVER BEND 3.6-31 Amendment No. 8+- 119
Secondary Conta nment -Operati ng 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.4.1.2 Verify all secondary containment equipment 31 days hatches are closed and sealed and loop seals filled.
SR 3.6.4.1.3 Verify each secondary containment access door 31 days is closed, except when the access opening is being used for entry and exit.
SR 3 6.4 1.4 Verify each standby gas treatment (SGT) 18 months on subsystem will draw down the shield building a STAGGERED annulus and auxiliary building to z 0.5 and TEST BASIS
, 0.25 inch of vacuum r----
water gauge in s 18.5 and ! seconds.
respectively.
SR 3.6.4 1.5 Deleted Not Applicable SR 3.6.4.1.6 Verify each SGT subsystem can maintain ; 0.5 18 months on a and a 0.25 inch of vacuum water gauge in the STAGGERED TEST shield building annulus and auxiliary BASIS building. respectively. for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
SR 3.6.4.1.7 Deleted Not Applicable RIVER BEND 3.6-47 Amendment No. 8&-. 95-. 113
Ds tj SC HI 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs) an- Fu. Build!ng 1selti E LCO 3.6. 4.2 Each SCID s shall be OPERABLE.
APPLICABILITY: MODES 1. 2. and 3 for secondary containment isolation.
building for- f6@l buildin zlain 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 a licable Conditions and Required Actions for systems made inoperable by SCIDs Cy ..~~
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more penetration A.1 Isolate the affected 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months flow paths with one SCID penetration flow path 4 i inoperable. by use of at least one closed and de-activated automatic damper.
closed manual damper, or blind flange.
AND (continued)
RIVER BEND 3.6-48 Amendment No. 8+-.113
SC!"Ds i i CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------NOTE --------
Isolation devices in high radiation areas may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path is isolated.
B. One or more penetration B.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months flow qpa t o penetration flow path SCIDs e yt)oBIgs by use of at least one inopera le.- - closed and de-activated automatic damper.
closed manual damper.
or blind flange.
C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B AND not met for SCIDs in MODE 1. 2. or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Re.ui tion and- ------------- NOTE--------
associated C LCO 3.0.31 is not appli e.
Time of Condition A or not met for FBIDs during movement of recently Suspend mov f Immediately irradiated fuel recently irradiated assemblie he fuel fuel assemblies in the building. Sfuel RIVER BEND 3.6-49 Amendment No. 8-1-. 113
3.6 CONTAINMENT SYSTEMS 3645 (j bx-During movement of recently irradiated fuel ies in the fuel building. /
ACTTONS

~~3 ------------~~

ot -- ------ NOTE LCO 3,0 3 is not ap :)1 clýbe.:
.k C.JNDITION A Fue rbulding Sus end movement of Immediately rnc~eraoi e rýýently irradiated fuel assemblies in the fuel building.
'y fuel building vacuum is a 0.
of vacuum water gauge.
Verify all fuel building equipment hatch covers are installed.
Verify each fuel building access door is closed, except when the access opening is being used for entry and exit.
RIVER BEND 3.6-55 Amendment No. 8--. 95-. 113
3.6 CONTAINMENT SYSTEMS 3.6.4.7 F"^l ,,id Vit÷nt, FzHndn LC 3.6.4. Two fuel building ventilation charcoal filtration subsystems shal be OPERABLE and one shall be operating in emergency mode.
APPLICABILI1 DuringbuilIdi ng. of recently irradiated fuel assemblies movement h fuel the ACT IONS

~~. NOTE--- -- -- -- --- -- -- -- --- -- -- -- -~~

LCO 3-0.3 is not applic le.
REQUIRED ACTION COMPLETION TIME A. One fuel building A. Restore f 1 building 7 days ventilation charcoal ventila on charcoal filtration subsystem filtr ion subsystem to inoperable OPE BLE status.
B. Required Action and B.1 Suspen ovement of Immediately associated Completion recentlyraite Time of Condition A not fuel assemb ies in the met. fuel buiIdin Two fuel building ventilation char al filtration sub stems i noperabl e.
One el building yen ilation charcoal f tration subsystem not n operation.
RIVER BEND 3.6-58 Amendment No. 8-+-. 113
Fuel Building Ventilation bystem-+-uel Hana*ng 3 4..7 SUOREILLANCE REQUIREMENTS SURVEILLANCE FEQUENCY SR 3.6.4.7. Verify one fuel building ventilation 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months charcoal filtration subsystem in SR 3.6.4.7.2 Opera each fuel building ven lation 31 days charcoa filtration subsyste for 2o10 con "nuous hours wit /eaters SR 3.6.4.7.3 Perform fuel buil ng ventilation In accordance charcoal filtra on ilter testing in with the VFTP ehf accordance Testing Pro l 3wi am thYe (VFTP) . tilation Filter 3.6.4.7.4 Veri each fuel building ventilation 18months SR
. Verify each fuel building ventilation \ mnh SSR 3.6.4 m
filter cooling bypass charcoal filtration damper can be opened and the fan started.
pole, -TY41`ýýtoý\OMý Aloty))ý 3.6-59 Amendment No. 81 RIVER BEND
Drywel 1 3.6.5.1 3.6 CONTAINMENT SYSTEMS 3.6.5.1 Drywell LCO 3.6.5.1 The drywell shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell inoperable. A.1 Restore drywell to I hour OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1.1 Verify personnel door inflatable seal air 7 days flask pressure z 75 psig.
SR 3.6.5.1.2 Verify from an initial pressure of 75 18 months psig, the personnel door inflatable seal pneumatic system pressure d "_snot decay at a rate equivalent to .4ý.-psig for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
(continued)
RIVER BEND 3.6-60 Amendment No. 81
Drywell Air Lock 3.6.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.2.1 Deleted SR 3.6.5.2.2 Verify drywell air lock seal air flask 7 days pressure is > 75 psig.
SR 3.6.5.2.3 ---------------- NOTE---------------
Only required to be performed upon entry into drywell.
Verify only one door in the drywell air lock can be opened at a time. 24 months SR 3.6.5.2.4 Deleted SR 3.6.5.2.5 Verify, from an initial pressure of 18 months 75 psig, the drywell air lock seal pneumatic system pressure does not decay at a rate equivalent to >S psig for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
I RIVER BEND 3.6-66 Amendment No. 8K, 87
CRFA System 3.7.2 3.7 PLANT SYSTEM 3.7.2 Control Roam Fresh Air (CRFA) System LCO 3.7.2 Two CRFA subsystems shall be OPERABLE.
APPLICABILITY: MODES 1. 2. and 3, I
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME.
A. One CRFA subsystem A.1 Restore CRFA subsystem to 7 days inoperable. OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)
RIVER BEND 3.7-5 Amendment No. 8&- 119
CRFA System 3.7.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and -------------NOTE---------
associated Completion LCO 3.0.3 is not applicable.
Time of C tion A not met iu~g I ofir**rwd c C. 1 Place OPERABLE CRFA Immediately
-fuel a.semblies in the subsystem in emergency mode.
during OPDRVs.. OR I'a e ,
Am_,,
C.I*nitiate action to Immediately suspend OPDRVs.
D. Two CRFA subsystems 0.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1.
~~2. or 3.~~
(cont i nued)
R IVER BEND 3.7-6 Amendment No. &+ 119
CRFA System 3.7.2 I
I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Operate each CRFA subsystem for ? 10 continuous 31 days hours with the heaters operating.
SR 3.7.2.2 Perform required CRFA filter testing in In accordance accordance with the Ventilation Filter Testing with the VFTP Program (VFTP).
SR 3.7.2.3 Verify each CRfA subsystem actuates ona - 18 months
ýý-initiation signal.
(continued)
RIVER BEND 3.7-7 Amendment No. 8+- 119
Control Room AC System 3.7.3 3.7 PLANT SYTEMS 3.7.3 Control Room Air Conditioning (AC) System LCO 3.7.3 Two control room AC subsystems shall be OPERABLE.
APPLICABILITY: MODES 1. 2. and 3.
IL During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control room AC A.1 Restore control room AC 30 days subsystem inoperable, subsystem to OPERABLE status.
B. Two control room AC B.1 Verify control room area Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months subsystems inoperable. temperature s 104 0 F.
AND B.2 Restore one control room AC 7 days subsystem to OPERABLE status.
C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Associated Completion Time of Condition A or B AND not met in MODE 1. 2.
or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)
RIVER SEND, 3.7-9 Amendment No. 8+- 119
Control Room AC System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and ------------ NOTE ------------
associated Completion LCO 3.0.3 is not applicable.
Time of Condition A-----------not--------
t
~~.1 Place OPERABLE control room Immediately a~lwnblias in the AC subsystem in operation.~~
II D.J S Iiate action suspend OPDRVs. to Innediately (continued)
RIVER BEND 3.7-10 Amendment No. 8-- 119
Control Room AC System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME COMPLETION TIME E. Required Action and Li uspeiid-m~vementL of Immediately associated Completion irradiated fNei assem1icz Time of Condition B not in the Wrmar-Y Gentaipment met. dufnn V-A and fulel-building.-
i.o~l~ ind-ae ftle r Srm - mýn
_.;t E/ Initiate action to suspend Immediately OPORVs.
1 I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify each control room AC subsystem has the 18 months capability to remove the assumed heat load.
RIVER BEND 3.7-11 3A*endment No. 8+- 119
AC Sources-Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources-Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:
~~a. One qualified circuit between the offsite transmission network and the onsite Class I E AC electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems-Shutdown";~~
~~b. One diesel generator (DG) capable of supplying one division of the Division I or II onsite Class 1 E AC electrical power distribution subsystem(s) required by LCO 3.8.10; and~~
c. One qualified circuit, other than the circuit in LCO 3.8.2.a, between the offsite transmission and the Division III onsite Class I E electrical power distribution subsystem, or the Division III DG capable of supplying the Division III onsite Class 1E AC electrical power distribution subsystem, when the Division III onsite Class 1 E electrical power distribution subsystem is required by LCO 3.8.10.
APPLICABILITY: MODES 4an5 oemrfu building.
RIVER BEND 3.8-17 Amendment No. 84 95
AC Sources-Shutdown 3.8.2 ACTIONS
~~, I I"*"r Ir'-~~
I r----------------------------------------------------------
LCO 3.0.3 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. LCO Item a not met. ---- --------------NOTE-------
Enter applicable Condition and Required Actions of LCO 3.8.10, when any required division is de-energized as a result of Condition A.
A. 1 Declare affected Immediately required feature(s) with no offsite power available from a required circuit inoperable.
OR A.2.1 Suspend CORE Immediately ALTERATIONS.
AND ImeditI a,nd fuel buid A.2./ Initiate action to Immediately "2 suspend operations with a potential for draining the reactor vessel (OPDRVs).
AND (continued) I RIVER BEND 3.8-18 Amendment No. 81
AC Sources-Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2./' Initiate action to restore Immediately required offsite power circuit to OPERABLE status.
B. LCO Item b not met. B.1 Suspend CORE Immediately ALTERATIONS.
AND
. Initiate action to suspend Immediately 3.;' Initiate action to restore Immediately 3 required DG to OPERABLE status.
C. LCO Item c not met. C.1 Declare High Pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Core Spray System and Standby Service Water System pump 2C inoperable.
RIVER BEND 3.8-19 Amendment No. 81
DC Sources-Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 The following shall be OPERABLE:
a. One Class 1E DC electrical power subsystem capable of supplying one division of the Division I or II onsite Class 1 E DC electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems-Shutdown";
b. One Class 1 E battery or battery charger, other than the DC electrical power subsystem in LCO 3.8.5.a, capable of supplying the remaining Division I or II onsite Class I E DC electrical power distribution subsystem(s) when required by LCO 3.8.10; and
c. The Division III DC electrical power subsystem capable of supplying the Division III onsite Class 1E DC electrical power distribution subsystem, when the Division III onsite Class 1 E DC electrical power distribution subsystem is required by LCO 3.8.10.
APPLICABILITY: MODES 4 and 5/
eentainmcne.t orF fuel building.
RIVER BEND 3.8-28 Amendment No. 81
DC Sources-Shutdown 3.8.5 ACTIONS
-NOTE--------------------------------------------------------------
LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.1 Declare affected Immediately electrical power required feature(s) subsystems inoperable. inoperable.
OR A.2.1 Suspend CORE Immediately ALTERATIONS.
AND
.22 Stispemd m,,yeff,,nt ef ssecmblies !R*the ab R-Rcd f- 'P1 hLifldlog.
Immediately A.2.X Initiate action to
'2 suspend operations with a potential for draining the reactor vessel.
AND Immediately
<* Initiate action to restore required DC electrical power subsystems to OPERABLE status.
_______________________ L _______________
RIVER BEND 3.8-29 Amendment No. 81
Inverters-Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Inverters-Shutdown LCO 3.8.8 One Divisional inverter shall be OPERABLE capable of supplying one division of the Division I or II onsite Class 1 E uninterruptible AC vital bus electrical power distribution subsystem(s) required by LCO 3.8.10, "Distribution Systems-Shutdown".
APPLICABILITY: MODES 4 and 5A ACTIONS
-NOTE -------------------------------------------------------------
LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION [_COMPLETION TIME A. One or more required A.1 Declare affected Immediately inverters inoperable. required feature(s) inoperable.
OR A.2.1 Suspend CORE Immediately ALTERATIONS.
asNDmblies the AND (continued)
RIVER BEND 3.8-36 Amendment No. 81
Inverters-Shutdown 3.8.8 ACTIONS CONDITIONS REQUIRED ACTION COMPLETION TIME A. (continued) A.2/ Initiate action to Immediately "Z suspend operations with a potential for draining the reactor vessel.
AND A.2. Initiate action to restore Immediately required inverters to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage, frequency, and 7 days alignments to required AC vital buses.
RIVER BEND 3.8-37 Amendment No. 81
Distribution Systems-Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems-Shutdown LCO 3.8.10 The necessary portions of the Division I, Division II, and Division III AC, DC, and Division I and II AC vital bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.
APPLICABILITY: MODES 4 and 5 ACTIONS
-NOTE-LCO 3.0.3 is not applicable CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC, DC, or AC vital bus supported required electrical power feature(s) inoperable.
distribution subsystems inoperable. OR A.2.1 Suspend CORE Immediately ALTERATIONS.
.* 0'fecd eat ed f'- e !,
, ssscmblie- in ths primnry eenta',nmznt and AND (continued)
RIVER BEND 3.8-41 Amendment No. 81
Distribution Systems-Shutdown 3.8.10 ACTIONS CONDITIONS REQUIRED ACTIONS COMPLETION TIME A. (continued) A.2 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.
AND A.2.ý Initiate actions to Immediately restore required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.
AND A.2.4 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to 7 days required AC, DC, and AC vital bus electrical power distribution subsystems.
RIVER BEND 3.8-42 Amendment No. 81
Inserv-1ce Leak and Hydrostatic Testing Operat'ion 3.10.1 2 S*ZAL
1 OPERATIONS 3 2* 1 rservice Leak and Hydrostatic Testing Operation t2* 3. !2 .The average reactor coolant temperature specified in Table 1.1-1 for MODE 4 may be changed to "NA." and operation considered not to be in MODE 3. and the requirements of LCO 3.4.10, "Residual Heat Removal (RHR) Shutdown Cooling System -Cold Shutdown." may be suspended. to allow performance of an inservice leak or hydrostatic test provided the following MODE 3 LCOs are met

~~a. LCO 3.3.6.2. "Secondary Containment &~~
is ltin AI. -2 apd- 2 .f b LCO 3 6 4.1. "Secondary Containment -Operating":
LCO 3 6.4.2. "Secondary Containment Isolation Dampers d LCD 3.6 4.3. "Standby Gas Treatment (SGT) System", ' t j e :
LCO 3 .6.4.4. "Annulus Mixing System" : I--
LCC 36 4. "-uel Building":
A,-' - - - '- MODE 4 with average reactor coolant temperature > 200'F.
RIRBENC 3.10-1 Amendment No. 8+-. 113
Programs and Manuals 5.5 5.5 Programs and Manuals ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days At least once per 31 days Monthly Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days above
~~b. The provisions of SR 3.( ).2 are applicable to the required frequencies for performing inservice testing activities;~~
~~c. The provisions of SR 3. 0.3 are applicable to inservice testing activities; and~~
~~d. Nothing in the ASME Boiler and pressure Vessel Code shall be construed to supersede the requirements of any TS.~~
5.5.7 Ventilation Filter Testing Program (VFTP)
A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2.
a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1989 at the system flowrate specified below +/- 10%:
ESF Ventilation System Flowrate cd otiu4,000 (continued) 5.0-11 Amendment No. 81 RIVER BEND
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)
b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2. and ANSI N510-1989 at the system flowrate specified below +/- 10%:
ESF Ventilation System Flowrate TSG 12,500 c~fm
ýFAS 4,000 cfm
c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C and the relative humidity specified below:
~~ESF Ventilation System Penetration RH~~
d. 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 when tested in accordance with Regulatory Guide 1.52.
__ Revision 2. and ANSI N510-1989 at the system flowrate
.*specified below +/- 10%:
ESF Ventilation System Delta P Flowrate 8" WG 42,000 cfm (continued)
RIVER- BEND 5.0-12 Amendment No. " 115
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)
~~e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI N510-1989:~~
ESF Ventilation System Wattage The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the main condenser offgas treatment system and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The program shall include:
a. The limits for concentrations of hydrogen in the main condenser offgas treatment system 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
b. A surveillance program to ensure that the quantity of radioactive material contained in any unprotected outdoor tank is limited to r 10 curies, excluding tritium and dissolved or entrained noble gases.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.
5.5.9 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The (continued)
RIVER BEND 5.0-13 Amendment No. 81
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Technical Specifications (TS) Bases Control Program (continued)
~~c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.~~
d. Proposed changes that do not meet the criteria of either Specification 5.5.11.b.1 or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
5.5.12 DELETED 5.5.13 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pas is 7.6 psig.
The maxi Ilowable primary containment leakage rate, La, at Pa, shall be--~oJ\of--primary containment air weight per day.
0,32- 7o The Primar; ontainment leakage rate acceptance criterion is
~~1.0 L. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are~~
~~0.60 La for the Type B and Type C tests and : 0.75 La for Type A tests.~~
The provisions of SR 3.0.2 do not apply to test frequencies specified in the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
RIVER BEND 5.0-16 Amendment No. 84 84 95
Attachment 3 Letter RBG-45930 Changes to Technical Specification Bases Pages Information Only
Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 2. Primary Containment and Drywell Isolation SAFETY ANALYSES, LCO, and 2.a. Reactor Vessel Water Level--Low Low, Level 2 APPLICABILITY (continued) Low RPV water level indicates the capability to c the fuel may be threatened. The valves whose pene tions communicate with the primary containment arr solated to primary-containment T isolation of the limit the release of onfission Level products.
2 supports ctions to ensure that offsite dose limits of 10 CFR are not exceeded.
The Reactor Vessel Water Level--Lo Low, Level 2 Function associated with isolation is implicitly assumed in the USAR analysis as these leakage paths are assumed to be isolated post LOCA. In addition, Function 2.a provides an isolation signal to certain drywell isolation valves. The isolation of the drywell isolation valves, in combination with other accident mitigation systems, functions to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the the primary containment.
Reactor Vessel Water Level--Low Low, Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level--Low Low, Level 2 Function are available and are required to be OPERABLE to ensure no single instrument failure can preclude the isolation function.
The Reactor Vessel Water Level--Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level--Low Low, Level 2 Allowable Value (LCO 3.3.5.1),
since isolation of these valves is not critical to orderly plant shutdown.
This Function isolates the Group 1, 7, 8, 9, 15, and 16 act ,ua t i o n The valves. isolation of valve Grou 9 also includes t
,of f* ;*
a tuationo-Zof the conta ýnmeit hydrogen ana yzers.
2.b. Drywell Pressure-Hith The High drywell pressure can indicata.,t break in the RCPB. on high isolation of some of the automatic isolation valves (continued)
B 3.3-143 Revision No. 0 RIVER BEND
Primary Containment and Drywell Isolation Instrumentation SB 3.3.6.1 BAS ES APPLICABLE 2.b. Drywell Pressure-High (continued)
SAFETY ANALYSES, LCO, and drywell pressure supports actions to ensure that offsite APPLICABILITY dose limits of 10 CFR 100 are not exceeded. The Drywell Pressure-High Function associated with isolation of the primary containment is implicitly assumed in the USAR accident analysis as these leakage paths are assumed to be isolated post LOCA. In addition, Function 2.b provides an isolation signal to certain drywell isolation valves. The isolation of the drywell isolation valves, in combination with other accident mitigation systems, functions to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the the primary containment.
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure- High per Function are available and are required to be OPERABLE to ensure that no*
single instrument failure can preclude the isolation function.
The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this may be indicative of a LOCA inside primary containment.
This Function isolates the Group 1, 3, and 8 valves. The isolation of valve Grou 8also includes the actuation of con ainment hydrogen analyzers.
2.c. Containment Purge Isolation Radiation-High The Containment Purge Isolation Radiation - High isolation instrumentation is provided to contain the radioactivity released to the Primary Containment after either an MSIV closure event or the Design Basis LOCA event. The allowable setpoint for the Containment Purge Isolation Radiation High instrumentation is set at a value determined by the radioactivity released by an MSIV closure event. Since the radioactivity levels for the MSIV closure-.event are less than for the LOCA event, the allowable setpoint ensures isolation capability for both events. When high purge exhaust radiation is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fisston products.
(conti nued)
I RIVER BEND B 3.3-144 Revision No. 2-6
Secondary Containment nd Pieo uli Isolation t o*
B 3.3.6.2 B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary ContainmentQ Isolation BASES BACKGROUND The secondary containment isolation '2 initiates closure of appropriate secondary containment isolation dampers (SCIDs) and starts appropriate ventilation subsystems.
systems. in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1), such that offsite radiation exposures are maintained within the requirements of 10 CFR4 that are part of the NRC staff approved licensing basis. Secondary containment isolation and establishment of vacuum within the assumed time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment or during certain operations when primary containment is not r uired to be OPERABLE ar , maintainedithin aplicable limits.
switchos that are nieessary to o- i-nitiation of secondary containment isolation. Most hanncl inludc electrnicc ntn (e.g., trip units) that comparcz measured input signl wt pre-esta-blis-hed-I* ,
s-etpoints.
,*.,..*.--w, *- .. . . ..-,
When the setpoint; is cxcccdcd. the
, v * , I channel outnilt relay actulatec, which then outputs an .isolationl signal to the isolat-ioin logic. Functional di.riyis provided axhaustinitiation of-pn thea logicpu sis nl from~e4 ~~~eah (conti nued)
RIVER BEND B 3.3-170 Revision No. 6-5
Secondary Containment - Isolation BASES APPLICABLE Thsi Is generated by ondary containment, SAFETY ANALYSES. isolation in umentation are imp' ly assumed in the LCO. and safety an ses of References nd 2 to initiate clos e of appropria o ventilation subsyste limit to fuel APPLICABILITY of ' and damper oe start he1oI igas generated. the _
ldngsli~ini uetainesr h the ven~til]ati~on "
stmipr r igne filtration li "offsi dose.
Refer to LCO 3.6 ndarY ntainment Isolation Dampers (CIr~ns) A *d
-@I,-lei Tenliatinnp Damnpo-e (*T~ne 11and LCO 3.6. .3. andýGas Treatment ( ystem." Applicable Safety Analyses Bases for more detail of the safety analyses.
The secondary containment isolation instrumentation G
______________ satisfies Criterion of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
'The OPERABILITY/e the isolation 'ins mentation de de nteupon the OPERABILI of the individual ' strumentation ch el Functions. ach Function must ve the required n er of OPERABLE annels with their etpointseset withi the specified Allowa e Values, as shown n Table 3.3.6.2-1. he actual setp*nt is calibrated sistent with appli ble setpoint me 1odology assumptio . A channel is ino rable if its ctual ip setpoint is no within its require llowable Val e. Each channel must als espond within its sumed respon time, where appropriate.
Allowable lues are specified r each Func on specified in the Table. oinal trip setpoint are specifi in setpoint calcul ions. The nominal tpoints ar elected to ensur hat the etpoints do not exc the Allow le Values between HANNEL /
RATIONS. Operati with a tr setpoint less co ervative an the nominal tr setpoint. t within its Allo ble Value, is acceptable.
Trip setpoint are those edetermined value of output at whi h an actin d take ce. The setpoin are compared t he actualp ess para er (e.g.. reactor essel water lev and when measured tput value of th rocess parameter exceeds the etpoint. th associated devic e.g.. trip unit) changes s e. The a ytic limits are rived from the ' iting values the proc s parameters obt ed from the safe analysis. The "Allowablie alues are derived rom the analyti imits. corrected for cali ration. process. d some of the i trument errors. The
. ..... (cntinued)
RIVER BEND B 3.3-171 Revision No. 6-5
Secondary Conta i nment, Isolation Sk_
B 3.3.6.2 BASES APPLICABLE trip setpoints e then deternn accounting for th remainin SAFETY ANALYSES, instrument e rs (e.g., drif . The trip setpo s erived in LCO. and this mann provide adequa protection becau instrumentati APPLICABILITY uncert ties. process ects. calibratlo olerances. ins ument (continued) fui o inhrh vrnensa eýdy 10 cFR 5.9) dri and severe env onnent errors (fo channels that are st In general. the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIDs.{-FI.Qnand the SGT System are required.
The specific Applicable Safety Analsecablity discussions are listed belowF '.-G -*--:XA-. '
Reco esl trLvel-Low Lw. v L'ow reactor pr sure vessel (RPV) wate level indica~tes ýhat the Scapability t ocool the fuel may be *Jreatened. Should rPV water *
//level decr se too far, fuel da_ col e sult_._ A Isolation of the sec dary containment and a uation of the SGT stem are initi in order to minimi the potential of a offsite dose rel se. The Reactor Vess Water Level-Low Lo . Level 2 Functio
\ one of the Functions sumed to be OPERAB and capable of roviding isolation an initiation sgnals. The isolation and initiation o on Reactor Ve*el W er Level-Low Low.
Level 2 support ac ons to ensure that y offsite releases ar within the limit calculated in the s ety analysis.
Reactor Vess Water Level-Low L . Level 2 signals are itiated from level ransmitters that s e the difference betw n the pressure ue to a constant co mn of water (referenc leg) and the pressur due to the actual ater level (variable 1 ) inthe vesse I Four channels of eactor Vessel Water L el-Low Low.
Lev 2 Function are a ilable and are requir to be OPERABLE to ure that no sing1 instrument failure ca reclude the function (latoon TeReactor ye el Water Level -Low L . Level 2 Allowable alue wschosen t bethe same as the Hi Pressure Core Spr Lvel-Lo ow. Level 2 Allowabl Value (LCO 3.3.5. , "Emergency Core C ing System (ECCS) ins rumentation.' and CO 3.3.5.2.
"Reac rCore Isolation Cooli g (RCIC) System strumentation"),
sin this could indicate the capability to ool the fuel is being t ea ene .... ...
.. (continued)
RIVER BEND B 3.3-172 Revision No. 6-5
Secondary Containment Isol ation
~~B 3.3i.6.2 BASES BASES APPLICABLE SAFETY ANALYSES.~~
LCO. and APPLICABILITY r ired to b/0PERABLE in DmES 1.2. ahOY where co siderable aeg h br/aks exsJi y of pipe resulting/in signific ecXrCoatSr1e RS:*shrnt releases o7sprobabil'1 Tehe aturelitios W as.
Ze steam and/ tof hesev In MODES/ DES: andthus5. this2 Functio is iad theXprobability radioacti conseq f h vnt r ue to the/ RCS pres sureind not equired.
2p Dywl resssu 'B) Ani;T hh' High dr eli pressure an indicat a break in t reactor coolan press e boundary I PB). An i lation of th secondary cont inment and ac uation of e SGT System re initiated in or r to minimiz the poten a*l of an offs e dose release. The i olation of hi h drywell ressure suppo s actions to e ure hat any offs' e release are within t limits calcul ed in the safety anal is. Howe r, the Drywel Pressure-High unction associat with isola ion is not a umed in any US accident or transie analysis. It is retain for the seco ary containment isola *on instru tation as r ired by the N approved High drywell essure sign s are initia from pressure transmitters that sense e pressure i the drywell. Fo r channels o Drywell Pr sure-High Fun ion are availa e and are requiredo be OPERAB to ensure t t no single in rument fail ure an preclud the isolatio function.
The lowable Va ue was chose o be the same s the ECCS Dr 11 Pressu -High Functi Allowable Va e (LCO 3.3.5 ince this i indicative o0a loss of cool rt accident.
The Drywe in Pressure-Hi Function is equired to OPERAB& MODES 1. .and 3 where onsiderable nergy exists n the RCS: us. there is probabilit of pipe brea resulting i significant eleases of dioactive (continued)
RIVER BEND B 3.3-173 Revision No. 6-5
Secondary Containment enl Isolation B 3.3.6.2 BASES APPLICABLE. Dr(contr e-)/
eu h SAFETY ANALYSES. (continued)
LCO. and APPLICABILITY steam and as. This Fun ion is not r uired in MO S 4 and 5 because e probability and conseque es of these ents are low due to he RCS pressure and tempera ure limitati s of these n 4. F 1 il in Vn " i n xh Ra i ion-Hi i High fuel ilding exhaus radiation is n indication possible gross fai ure of the fu cladding. T e release may ave origina from the pr' ary containm t due to a br ak in the RCPB or the uel building ue to a fuel andling accid nt. When Exhau t Radiation-HgIh is detecte fuel buildi isolation and act tion f the sociated vent ation filtra ion system are in" ated to lim the release f fission pr ucts as assumed n t e USAR s fety analyses (Re. 1).
The Exhaust adiation-Hig signals are 'itiated from ra ation detectors at are loca on the vent ation exhaust ct coming from the uel buildin entilation. he signal from ach detector is inpu to an indiv ual monitor ose trip output are assigned to an 'solation ch nel.
The Allowable V ues are chos to promptly d ect gross fail e of the fuel cl dding.
he Exhaust adiation-Hi Function is r uired to be OP BLE during mo nt of rece ly irradiated uel assemblies n the fuel building ause the c pability of d ecting radiati releases due to uel failures due to fuel u covery or drop;d fuel ass les) must b provided to e ure that offs* e dose limits are* exceeded."Recently irridiated fuel" fuel that( has (x*iedl part of/a critical r ctor core wilt n the previous o
(continued)
RIVER BEND 8 3.3-174 Revision No. 6-5
Secondary Containment - Isolation B-3.3.6.2 BASES APPLICABLE SAFETY ANALYSES.
LCO. and The Manual Initiation push button channels introduce signals APPLICABILITY into the secondary containment W fu 419-iAdi \
(conti nued) **- '- an.
e ý--= .... ........
.. . manual noied _
isollcation ....
capability. I -*
hc-*e-rme "There are four push buttons for the logic, two manual initiation push buttons per trip system. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.
Four channels of the Manual Initiation Function are available and are required to be OPERABLE in MODES 1. 2. and 3. since these are the MODES and other specified conditions in which the Secondary Containment Isolation . Functions are required to be OPERABLE.
Moving reccntly ir-radi4ated fuel assemblicz in the fuel build~ing rcqurcz nly that por-tion of the Manual Initiaticn Functien
-associatcd with the fuel building te be OPERABLE.
ACTIONS A Note has been provided to modify the ACTIONS related to isolation instrumentation channels. Section 1.3. Completion Times. specifies that once a Condition has been entered.
subsequent divisions, subsystems. components. or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
However. the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable isolation instrumentation channel.
(continued)
RIVER BEND B 3.3-175 Revision No. 6-5
Secondary Containment a Isolation
--
~~B3. 3. 6.2 BASES ACTIONS A.1 (continued)~~
Because of the diversity of sensors available to provide isolation signals and the redundanc of the isolation design. an allowable utof service time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
has been shown to be acceptable (Refs. 3 and 4) to permit restoration of any inoperable channel to OPERABLE status.
L,,Uu, r, ,ouil- uurT servIlce'lme conbi*Leni witn'sel
,*a ,*, ,* hisout of service time is only acceptable provided the associated FunctiQn i.S*sti maintaining isolation capability
_ If the inoperable channel cannot be res-ored to OPERABLErstatus within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability. restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Conditions must be entered and its Required Actions taken.
Required Actio B.1 is intended t ensure that appropriate act ons are taken lf multiple, inoper e. untripped channels wi.W the same F ion result in a plete loss of automaticioclation cap lity for the asso ated penetration flow pat s) or a c plete loss of aut atic initiation capabili for the various ventilation subs ems. A Function is cons' ered to be maintaining i ation capability when -s icient channels are OPERABLE o n trip, such that one t 'p system will generate trip si 1 from the given Funct' on avalid signal. T ensu that one of the two S in the asso ated ptration flow path and e venti ion subsyst an be initiated on an isolIt n signal from the given unction. For the "Functions with two a-out-of-two logic tri ystems (Functions I.
and 2). this we require one trip ss to have two channel each OPERABL r in trip. The Condiiosn does not i ncl ude Manual In' ation Function (Func n 5), since it is no assumed inany cident or transientolysis. Thus. a tot oss of man initiation capabiliý for 24 hours (as al ed by Required A ion A.1) is allowed.
...(continued)
RIVER BEND B 3.3-176 Revision No. 6-5
Secondary Containment Isolation -. 36.
B 3.3.6.2 BASES ACTIONS /.1.1.2.,[ 2.1. and f.2.2 (continued)
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
If any Required Action and associated Completion Time of Condition A -are not met. thelate the associated secondary containment 6 and start the associated ventilation subsystems cannot be ensured. Therefore.
further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated penetration flow path(s) and starting the associated ventilation subsystems (Required Actions,.1.1 andS.2.1) performs the intended function eand allows operations to continue.
Alternatively, declaring the associated SCIDs or associated ventilation subsystem inoperable (Required ActionsV.1.2 andk'.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2. LCO 3.6.4.3.
6il LCO 3.6.4.4 . )provide appropriate actions for the inoperable components. I One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without challenging plant systems.
REQUIREMENTS inhtrunmnt ut e locatd in the SRs column of The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains secondary containment isolation capability. Upon completion of the Surveillance, or
.expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Action(s) taken.
(continued)
RIVER BEND B 3.3-177 Revision No. 6-5
Secondary Containment 6Iso1ation BASES SURVEILLANCE This Note is based on the reliability analysis (Refs. 3 REQUIREMENTS and 4) assumption of the average time required to perform (continued) channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the SCIDs will isolate the associated penetration flow paths and the associated ventilation subsystems will initiate when necessary.
Performanc of the CHANNEL CHEC once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross ailure of instrumen ion has not occurred. A CHANNEL CHECK s normally a compar on of the indicated p-ameter for one ins ument channel to a imilar parameter on ot r channels. It i ased on the assu ion that instrument c nels monitoring he same parameter hould read approximatel the same value.
Significant devi ions between the instr ent channels could be an indication excessive instrument ift in one of the channels or thing even more ser us. A CHANNEL CHECK will detect gr s channel failure: th . it is key to verify*_g the instru tation continues to o rate properly betwee each CHAN CALIBRATION.
Agreement criteria are termined by the plant taff. based on a combination of the c nnel instrument uncert nties. including indication and rea bility. If a channel
~~outside the criteria, it ma e an indication that e instrument has drifted outside its li t.~~
The Freque y is based on operati g experience that d nstrates channel f lure is rare. The NNEL CHECK supple m s less fm but more frequ acks of channels dur g normal oper nal use of the di ays associated wit he channels re ird by the LCO.
A CHANNEL FU TIONAL TEST is perf on each r 'ired channel to ensure at the entire chann will perform e intended function.
(continued)
RIVER BEND 8 3.3-178 Revision No. 6-5
Secondary Containment ol t Isolation c BASES SURVEILLANCE REQUIREMENTS Any setpoint a 'ustment shall be left se consistent with the
-assumptions the current plant specifc setpoint methodology.
The Fre ncy of 92 days is based u on the reliability analysis of Refences 3 and 4'./
Calibration of trip uni provides a check of e actual trip setpoints. The chann must be declared *o* erable if the trip setting is discover to be less conserva *ve than the Allowable Value specified i Table 3.3.6.2-1. If he trip setting is discovered to b less conservative th accounted for in the appropriate s oint methodology, b is not beyond the Allowable Value. perfo rance is still withi the requirements of theylant.
safety ana sis. Under these ditions. the setpoint 5dst be readjust to be equal to or ore conservative than a ounted for in the ppropriate setpoin methodology.
Th Frequency of 92 da is based on the relia iity analysis of R ferences 3 and 4.
CHANNEL CA BRATION is a complet check of the instrume loop and the nsor. This test ver ies the channel r;spo Is to the measur parameter within h necessary range and curacy.
CHAN L CALIBRATION leaves he channel adjusted account for in rurnent drifts betwee successive calibrati s consistent with e plant specifi set int methodology.
The Frequency is sed upon the assumpti of the magnitude of equipment drift in the setpoint analysi (conti nued)
RIVER BEND B 3.3-179 Revision No. 6-5
Secondary Containment Rlsolat'on(
BASES SURVEILLANCE SR 3.3.6.2.5 REQUIREMENTS (continued) Thhe LOCGC SYSTEM FUNCTIONAL TEST demeztrs tho sys unctional testing, performed on SCIDs and the associatedf-- *r,*
ventilation subsystems i LCO 3.6.4.2, LCO 3.6.4.3, LCO 3.6.4.4;LCO 3.6.4.6, . respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
REFERENCES 1. USAR, Section 6.3.
~~2. USAR, Chapter 15.~~
~~3. NEDC-31677-P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," July 1990.~~
~~4. NEDC-30851-P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentations Common to RPS and ECCS Instrumentation," March 1989.~~
RIVER BEND B 3.3-180 Revision No. 6-5
GCRFA-Systm-f ntryffefntration B 3.3 INSTRUMENTATION t0% -,
B 3. 3. 7. 1 0 1- 0 i-IF r
'%n Pole- PDeAY O BA~
BAKGROUND The CRFA Sys is designed to prove,
/m rai loglY
/nvironment to ensure t ¢habitability of the. /
S~controlled of c rol room operators under4, control oom for the safety c
/ al p1 Pnt conditions. Two ind e'ndent CRFA subsystems are "
( ~~~eac,hcapabl e of ful fiIlli ng t*>e stated safety function.J The System Si strumentation and contro, for the CRFA or press ize the automatically initiate a'ion to isolate main control room (MCRto minimize the consequ ces of radioactive material *n the control room en vi nment.
In the event of eactor Vessel Water Lev -Low Low, Level 2, Drywel Pressure- High, or Cant 01 Room Local Intake Ven til ion Radiation Monitor ignal, the CRFA MCR Systemn aly started in the eme ency mode. The air Is automati is then r irculated through the arcoal filter, and Suffici t outside air is drawn n through the normal intake to ke the MCR slightly pres rized with respect to ad* ent areas.
e CRFA System instrum tation has two trip system .one trip system initiates ne CRFA subsystem, while t second trip system initiat the other CRFA subsystem ef. 1).
Each trip system r ceives input from the Func ons listed above. The Funct ons are arranged as follo for each trip system. The R ector Vessel Water Level - w Low, Level 2 and Drywell Pressure -High are arrang ed ogether in a one-out-of-t'wo taken twice logic. Th Control Room Local Intake V Ktilation Radiation Monito are arranged in a one-o u of-one logic. The chann sinclude electronic equi ent (e.g.,, trip units) t compares measured inpu si als with pre-established etpoints. When the setp *nt
. exceeded, the channel o put relay actuates, whic then outputs a CRFA System miniat signal to the ini ation logic.
APPLICABL The ability of th CRFA System to maintain e habitability SAFETY A LYSES, of the MCR is e licitly assumed for cert n accidents as LCOO a discussed in e.USAR safety analyses ( fs. 2 and 3).
APPLI ABILITY CRFA System peration ensures that th radiation exposure of control ro personnel, through the uration of any one of (continued)
B 3.3-198 Revision No. 3-4 RIVER BEND
SAET NAYES, GDC I Yof 10 CFR 50, Appendix A. J*
LCO, and /
IAPPLICABILITY C*A System instrumentation sa **sfiesCriterion 3 of the NRC
/(continued) i olicy Statement. .. /
The OPERABILITY of the C A System instrumentationis dependent upon the OPE BILITY of the individual instrumentation chan 1 Functions specified in S
Table 3.3.7.1-1. ch Function must have a rquired number of OPERABLE chan is, with their setpoints ithin the specified Allo le Values, where appropr ate. A channel i inoperable i its actual trip setpoint not within its required Al owable Value. The actual etpoint is calibrated consiste with applicable setpoint ethodology assumptions.
Allo le Values are specified each CRFA System Functio sp ified in the Table. Nomin trip setpoints are s cified in the setpoint ca ulations. These nominal etpoints are selected to sure that the setpoints d not exceed the Allowable Val between successive CHANN.2a CALIBRATIONS. Operatio with a trip setpoint tha is less conservative than the ominal trip setpoint, b within it Allowable Value, is cceptable.
Trip setpoints e those predetermined v ues of output at which an actio should take place. Th setpoints are compared to e actual process p~arame er (e.g., reactor vessel wat level), and when the m asured output value of the proc s parameter exceeds the etpoint, the associated device e*g., trip unit) change state. The analytic limit are rived from the limiting alues of the process par eters obtained from th safety analysis. The Allo ble Vues are derived from t analytic limits, correcte for alibratlon, process, a some of the instrument er ors.
The trip setpoints ar then determined, accounti for the remaining instrumen errors (e.g., drift). Th trip setpoints derived n this manner provide ade ate protection because instrum tation uncertainties, pro ss effects, calibration t erances, instrument drift and severe environment rrors (for channels that st function in harsh environme s as defined by 10 CFR 50 9) are accounted for.
The sp cific Applicable Safety A yses, LCO, and Appli ability discussions are Vs ted below on a Function by Fu tion basis.
(continued)
RIVER BEND B 3.3-199 Revision No. 0
CR-F_"FA-ý SytmIntent~aioný
-SAFETY ANALYSES, /
LCO, and Low rea or pressure vessel (RPV) w er level indicates that APPLICABILITY the c ability to coal the fuel m be threatened. A 1 (continued) rea or vessel water level coul indicate a LOCA, and ill a omatically initiate the CRF System, since this co ld be precursor to a potential r diation release and su sequent radiation exposure to cont ol room personnel.
Reactor Vessel Water L el-Low Low, Level 2 s' nals are initiated from four 1 vel transmitters that nse the difference between e pressure due to a co stant column of water (reference g) and the pressure du to the actual water level (va able leg) in the vesse . Four channels of Reactor Vessel ater Level-Low Low, vel 2 Function are available (t a channels per trip sys em). and are required to be OPERABL to ensure that no sing instrument failure can preclude RFA System initiation, M he Allowable Value for the a r Vessel Water Level ow Low, Level 2 is chosen to be he same as the Seconda Containment Isolation Reac or Vessel Water Level- w Low, Level 2 Allowable alue he Reactor Vessel Water Level-Low Low, Level 2 F nction is required to be OPERABI in MODES 1, 2, and 3 toensure that the control room per nnel are protected. In DES 4 and 5, the probability of vessel draindown event r of a LOCA, is minimal. Therefo this Function is not r uired. In addition, the trol Room Ventilation diation Monitor Function provi es adequate protection.
S~~2. Dr ywe Pressure--High // "
High essure in the drywell uld indicate a break in the reac or coolant pressure bo ary (RCPB). A high rywell pr sure signal could mdi te a LOCA and will omatically i tiate the CRFA System, since this could be precursor to potential radiation r ease and subsequent adiation exposure to control ro personnel.
(continued)
RIVER BEND 8 3.3-200 Revision No. 0
kPPLICABLE 2. Dr ell Pressure-Hiqhn (cont' ed))
S'AFETY ANALYSES.
LCO. and Dr ell Pressure-High signa are initiated from four APPLICABILITY ressure transmitters tha sense drywell pressure. F r channels of Drywell Pressure-Hig Function are available (tw channels per trip system) and are quired to be OPERABLE to e ure that no single instrument f lure can preclude CRFA Sys m initiation.
The Drywell Pre ure-High Allowable Value w chosen to be the same as the Secondary Containment Isolati n Drywell Pressure-High Allowable Vue (LCO 3.3.6.2).
The Or ell Pressure-High Function s required to be OPERA in MODES 2. and 3 to ensure that ontrol room personnel e pro ected during a LOCA. In ES 4 and 5. the Drywel P ssure-High Function is no required since there nsufficient energy in the eactor to pressurize e drywell t the Drywell Pressure-Hi setpoint.
3 Coto Roo LaInkeVtiton adiation Monitors MCR. A hig radiation level may se a threat to MCR personnel; tua d ector indicating thi condition automatically signals initiati n of the CRFA Systtem The ntrol Room Local In e Ventilation Radiati Monitors Fu tion consists of tw independent monitors. wo channels of C trol Room Local In ke Ventilation Radiati n Monitors are vailable and are r uired to be OPERABLE ensure that no single instrument ailure can preclude C ASystem initiation.
The Allowable V ue was selected to en re protection of the control roomp sonnel.
The Contro, Room Local Intake ye ilation Radiation Monits Function *srequired to be OP LE in MODES 1.' 2.' and . and during erations with a pot ial for draining the ractor vesse (OPDRVs) and moverne of irradiated fuel in he primary ccont inment or fuel buil *ng to ensure that contr 1 room pe onnel are protect during a LOCA. fuel ha ling event-. or a RIVER BEND B 3.3-201 Revision No. 6-13
B SES (continued)
ACTIONS A Note as been.provided to mod y the ACTIONS related to CRFA ystem instrumentation c nnels. Section 1.3, Co letion Times, specifies hat once a Condition has en tered, subsequent divis ns, subsystems, componens, or variables expressed in e Condition discovered t e inoperable or not wit n limits will not result n separate entry into the Condi on. Section 1.3 also s cifies that Required Actions o the Condition continue apply for eac additional failu , with Completion Times ased on initial entry into the ondition. However, the quired Actions fo inoperable C A System instrumentation hannels provide appropriat compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separat Condition entry for each noperable CRFA System instr entation channel.
.1 Required Action A.1 di cts entry into the ppropriate Condition referenced *n Table 3.3.7.1-1. he applicable Condition specified n the Table is Fun tion dependent.
Each time an inop able channel is di covered, Condition A is entered for at channel and pro des for transfer to th appropriate su sequent Condition.
B.1 and B Becaus of the diversity sensors available to ovide init tion signals and e redundancy of the CRF System des gn, an allowable t of service time o0f 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months has b n shown to be acc table (Refs. 4 and 5) o permit estoration of any noperable channel to RABLE status.
However, this ou of service time is onl acceptable provided the a ociated Function is st 1 maintaining CRFA System initia on capability. A Fun ion is considered t be maintain' g CRFA System initiat n capability when sufficien channels are OPERABLE r in trip, such tha one trip sys emwill generate an i tiation signal from he given Fnction on a valid si aI. This would re re one trip.
trip isystem to have two ch nels, each OPERABLE r in in situation (loss o CRFA System initia on c ability), the 24 hou allowance of Requs d Action B.2 is t appropriate. If e Function is not intaining CRFA ystein initiation ca ability, both CRFA ubsystems must be declared inoperabl within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of scovery of loss of (continued RIVER BEND 6 3.3-202 RevisiosA No. 0
CRFA System initiati capability in both trip ystems. If the inoperable ch nel cannot be restored to ERABLE status within the allo le out of service time, e channel must be placed in e tripped condition per R uired Action B.2.
Placing the inoperable channel in trip ould conservatively compensat for the inoperability, re ore capability to acco"nno t a single failure, and low operation to co ntin e.Alternately, if it is ot desired to place the chan Ilin trip (e.g., as in t case where placing the initiatin) in erable channel in trip w ld result in an Action taken.
ndition E must be entere and its Required Because of the d* ersity of sensors avail e to provide initiation sig ls and the redundancy of the CRFA System design, an al owable out of service ti e of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months has been shown b be acceptable (Refs. 4 and 6) to permit restorati of any inoperable chan el to OPERABLE status.
However this out of service tim is only acceptable CRF,ý provi d the associated Functi is still maintaining Syst m initiation capability A Function is considere to be aintaining CRFA System initiation capability whe at one.
fficient channels are 0RABLE or in trip, such rip system will genera an initiation signal fr m the given Function on a v id signal. This would rquire one trip system to have wo channels, each OPERA .Eor in trip.
In this situation loss of CRFA System ini ation capability), th 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance of Re ired Action CRFA C.2 is not appropria .If the Function is n maintaining System initi tion capability,_both C A 'Subsystems must be declared i perable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months discovery of loss of CRFA Sys initiation capabilit in both trip systems. If the ino erable channel cannot restored to OPERABLE sta us wwiithi the allowable out of rvice time, the channel myst be aced in the tri pped c dition, per Required Actiatively C2 PI ing the inoperabl e c nnel in trip would conser mpensate for the nop rability, restore capabi I y to cconinodate a single ilure, and allow operati n to continue. Alternat y, if it is not de~sired b0placethethe channel in trip (e g., as in the case wher placing inoperable chann in trip would result i an initiation),
Condition E mus be entered and its Re red Actions taken.
(continued)
Revision No. 0 RIVER BEND 8 3.3-203
(cniud Bec ase of the diversity of sen 'rs available to provide i itiation signals and the re ndancy of the CRFA Sy em /
out ofJt'onervice time of 6 hour is of any inoperabl ychannel to \
/ / ~~pro~vided anto allowable design, permit restor OPERABLE status. Howev , this out of servic time is only acceptable provided Ke associated Function s*still /
maintaining CRFA S tem initiation capabil iy. A Function is considered to e maintaining CRFA Sy em initiation capability wh sufficient channels ar OPERABLE or in trip.\
such that o trip system will gener e an initiation signal \
from the yen Function on a vali signal. This would requiree ne trip system, to have wo channels, each OPER E or in ip. In this situatio loss of CRFA System mi ation capability), the hour allowance of Re ired A ion 0.2 is not app opri e. If the Function i not aintaining CRFA System nitiation capability, oth CRFA subsystems must be decyared inoperable within hour of discovery of loss of RFA System initiation capability in
/T el cannot be
/
both trip systems. If the inoperable ch restored to OPER LE status within the lowable out of service time, e channel must be p1 ed in the tripped condition, p Required Action 0.2 Placing the inoperable channel in ip performs the mt ded function of the asociate , if RFA subsystem in the to place channel arts theAltent
( mode). it is not desire theolati is ch nnel in trip (e.g. as in the case where is not desi d to start the s system), Condition E m t be entered an its Required Acti ns taken.
he 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Compl ion Time is based on te consideration that this Fun ion provides the primar signal to start the CRFA System hus ensuring that the esign basis of the CRFA W h any Required Action and associated Compi ion Time not et, the associated CR subsystem must be aced in the emergency mode of op ation (Required Acti n 0.1) to ensure that control room rsonnel will be pro ected in the event B 3.3-204 Revision No. 3-4 RIVER BEND
~RAysem LnsuevnTatiuii SBA ES
~~ACTIONS .1adE2(cniud CRFA subsystem in op ation must provide for aut atically reinitiating the s system upon restoration of ower following a loss f power to the CRFA subsys em(s).~~
Alternately,if it is not desired to st the subsystem, the CRFA s system associated with ino erable, untripped channels ust be declared inoperable ithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
(
The 1 our Completion Time is in nded to allow the operato tim to place the CRFA subsyst in operation. The I hour allowing time for reoaino it ripping pletion C i)hle Time is acceptabl ecause minimizes risk Is,/
of chan or for placing the associ ed CRFA subsystem in opera on.
SURVEILLANC/ As noted at the beg* ning of the SRs, the SRs each CRFA "
REQUIREME WS System Instrum~ent ion Function are located i the SRs column of Table .3.7.1-1.
The Surveill nces are also modified by Note to indicate
( that when for perf channel is placed in an J operable status solely ance of required Survei ances, entry into associ ed Conditions and Requir for mai tains CRFA System initi Actions may be delayed to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided th associated Function on capability. Upon c pletion of the Surveill ce, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months flowance, the channel t be returned to OPERABLE ctus or the applicable Conf ion entered and Req d ions taken. This Note is ased on the reliab ity i lysis (Refs. 4, 5, and 6 assumption of the averag ime require to perform chann surveillance. That anal is demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does no significantly reduce the p ability that the CRFA S em will initiate P formarice of the CHANNEL ECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sures at a gross failure of 'strumentation has not occ red. A CHANNEL CHECK is norma y a comparison of the indj ated parameter for one in rument channel to a simil parameter on. other channels. t is based on the assump on that instrument channe s monitoring the same par eter should read approximat y the same value. Signi cant deviations (continued)
RIVER BEND B 3.3-205 Revision No. 0
BAS SURVEILLANCE SR 3.3.7//nine)
REQUIREMENTS/ betw n the instrument channe could be an indicatiIn of \
ex ssive instrument drift one of the channels r mething even more seri s. A CHANNEL CHECK wi1I detect gross channel failure; hus, it-is key to ver' ying the instrumentation cont ues to operate proper between each CHANNEL CALIBRATI .
Agreement cri na are determined by e plant staff based on a combin ion of the channel in ument uncertainties, including ndication and readabili y. If a channel is outsi de he criteria, it may b an indication that the instr ent has drifted outsi its limit.
T Frequency is based u n operating experience at emonstrates channel f lure is rare. The CHA L CHECK supplements less fo 1, but more frequent, c ecks of channel stat~us duri g normal operational us of the displays associated with annels required by the 0.
A CHAN FUNCTIONAL TEST is p formed on each req i ed chann to ensure that th e ire channel will p form the int ded function. Any se oint adjustment sh be c sistent with the ass tions of the curre plant pecific setpoint met dology.
The Frequency of 9 days is based on e reliability analyses of Refe nces 4, 5, and 6.
~~./ // . ~SR. 3.3.7. .3/ . '. .. _.~~
The cal ration of trip its provides a check of t actual trip etpoints. Any se point adjustment shall be onsisten wit the assumptions the current plant speci c setpoi m hodology. The c nnel must be declared in erable if th rip setting is d covered to be less conse ative than the Allowable Value f the trip setting is scovered to be less conserva ve than accounted for in he appropriate setpoint met odology, but is not beyo the Allowable Valu the channe performance is still wi in the requirements o the plan safety analysis. Unde hese conditions, the (continued RIVER BEND 8 3.3-206 Revision No. 0
em ru in must be to be equal to or m Bsetpoint aed te conservative than a ounted for in the appropri te setpoint methodology.
The Frequency f 92 days is based on the eliability analyses of eferences 4, 5, and 6.
A C NNEL CALIBRATION is a co lete check of the ins ument 1 p and the sensor. This st verifies the chann esponds to the measured rameter within the ne, ssary range and accuracy. CHINEL CALIBRATION leave the channel adjusted to account f instrument drifts bepeen successi calibrations consist nt with the plant spe fic setpoint methodology..
The Frequency based on the assump 'on of the magnitude f equipment
/drt in the setpoint an ysis.
The OGIC SYSTEM FUNCTIO TEST demonstrates the 0 BILITY of the req ed initiation logic for *specific annel. The system nctional testing perform in LCO 3.7.3, "Control oom Fresh Air (CRFA) Sys in," overlaps this Surveillance o provide complete testi of the assumed safety function The 18 mont Frequency is based on t need to perform this Surveilla e under the conditions at apply during a plant outage; the potential for an planned transient if the
~~2. USAR, Sect* n 6.4.~~
~~3. USAR, apter 15.~~
RD7ontno.d0 RIVER BEND B .3.3-2Z07 Revision No. 0
CRFA~~~~~
"Luietton
,sz BA S REFERENCE (cont' ued) 4. GENE-770- -1, "Bases for Chang to Surveillance est Interv s and Allowed Out-of- rvice Times for Sele ed Instrumentation T nical Specificatj ns,"
Fe uary 1991.
~~5. EDC-31677P-A, "Tech cal Specification mprovement Analysis for BWR I ation Actuation I trumentation,"~~
~~July 1990.~~
~~6. NEDC-30851P-A Supplement 2, "Tec ical Specific ion Improvement nalysis for BWR Is ation Instrum tation Common to PS and ECCS Instru ntation," Marc 1989.~~
L,,-D4 81,ýc RIVER BEND B. 3.3- 208 Revision No. 0
RCS Specific Activity B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.8 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity.
The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.
Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure, in the event of a release of any radioactive material to the environment.
during a DBA, radiation doses are maintained within the limits of 10 CFR f4b"-Ref. 1).
This LCO conta1 yr-etod-i specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed in terms of a DOSE EQUIVALENT 1-131. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months radiation dose to an individual at the site boundary to a small fraction of the 10 CFRi1lJnit.
APPLICABLE Analytical methods and assumptions involving radioactive SAFETY ANALYSES material in the primary coolant are presented in the USAR (Ref. 2). The specific activity in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely.
This MSLB release forms the basis for determining offsite doses (Ref. 2). The limits on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses at the site boundary, resulting from an MSLB outside containment during steady state operton, will not exceed 10% of the dose guidelines of 10 CFR* S (
on t in~ue~d)
RIVER BEND B 3.4-39 Revision No. 0
RCS Specific Activity B 3.4.8 BASES APPLICABLE The limits on specific activity are values from a parametric SAFETY ANALYSES evaluation of typical site locations. These limits are (continued) conservative because the evaluation considered more restrictive parameters than for a specific site, such as the location of the site boundary and the meteorological conditions of the site.
RCS specific activity satisfies Criterion 2 of the NRC Policy Statement.
LCO The specific iodine activity is limited to : 0.2 juCi/gm DOSE EQUIVALENT 1-131. This limit ensures the source term assumed in the safety analysis for the MSLB is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR iits.
APPLICABILITY In MODE 1, and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are applicable since there is an escape path for release of radioactive material from the primary coolant to the environment in the event of an MSLB outside of primary containment.
In MODES 2 and 3 with the main steam lines isolated, such limits do not apply since an escape path does not exist. In MODES 4 and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced.
ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LCO DOSE EQUIVALENT 1-131 limit, but is r 4.0 pCi/gm, samples must be analyzed for DOSE EQUIVALENT 1-131 at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In addition, the specific activity must be restored to the LCO limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on the time needed to take and analyze a sample. The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time to restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems.
(continued)
RIVER BEND B. 3.4-40 Revision No. 0
RCS Specific Activity B 3.4.8 BASES ACTIONS A.1 and A.2 (continued)
A Note to the Required Actions of Condition A excludes the MODE change restriction of LCO 3.0.4. This exception a-iows entry into the applicable MODE(S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of a limiting event while exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to, power operation.
B.1, B.2.1. B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT 1-131 cannot be restored to r 0.2 pCi/gm within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or if at any time it is > 4.0 pCi/gm, it must be determined at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and all the main steam lines must be isolated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Isolating the main steam lines precludes the possibility of releasing radioactive material to the environment in an amount that is more than a small fraction of the requirements of 10 CFR/4WWring a postulated MSLB accident.
Alternately, the plant can be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).
In MODE 4, the requirements of the LCO are no longer applicable.
The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on the time needed to take and analyze a sample. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for bringing the plant to MODES 3 and 4 are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
RIVER BEND B 3.4-41 Revision No. 0
RCS Specific Activity B 3.4.8 BASES (continued)
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level.
This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
Qr-ý REFERENCES 1. 10 CFR 56,
T
~~2. USAR, Section 15.6.4.~~
RIVER BEND B 3.4-42 Revision No. 0
Primary Containment Air Locks B 3.6.1.2 BASES-BACKGROUND DBA. Not maintaining air lock integrity or leak tightness (continued) may result in a leakage rate in excess of that assumed in the unit safety analysis.
APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary-containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment Ji_ -2 designed with a maximum allowable leakage rate (L,) ofe by weight of the containment and drywell air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the calculated maximum peak containment pressure (P,) of 7.6 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.
Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.
irradi e ul iefulta has. ccupied par tof a /
t primary con inment (Ref. , to limit d es at the ite during ac vte ih eui mdn Primary containment air locks satisfy Criterion 3 of the NRC Policy Statement.
LCO As part of the primary containment, the air lock's safety function is related o contrQ of contai ent leaka e rates n i reac iv y or ater eve excursion. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such events.
(continued)
RIVER BEND B 3.6-6 Revision No. 2-3
Primary Containment Air Locks B 3.6.1.2 BASES LCO The primary containment air locks are required to be (continued) OPERABLE. For each air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be open at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from primary containment.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are.
reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining OPERABLE primary containment air locks in MODE 4 or 5 to ensure a control volume is only required during situations for which significant releases of radioactive material can be postulated; such as during operations with a potential for draining the reactorvessel DRVs r ur g f 1 move en of re!ý e tly ir adiated'leiassemn es in he pr miary men to ra ioactive ecay, pimar conta*
/con iinment. Due p imary c ntainment jvolving andling ecen y irr diatd uel (i. ., fuel th has occ pied par of criti a uareacto core wih the pre ous 11 ays).
ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component.
If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside primary containment by entering through the other OPERABLE air lock. However, if this is not practicable, or if repairs on either door must be performed from the barrel side of the door, then it is permissible to enter the air lock through the OPERABLE door, which means there is a short time during which the primary containment boundary is not intact (during access through the OPERABLE door). The ability to open the OPERABLE door, (conti nued)
ACTIONS even if it means the primary containment boundary is RIVER BEND B 3.6-7 Revision No. 2-3
Primary Containment Air Locks 8 3.6.1.2 BASES ACTIONS C. 2. and-C.3 (continued) failed a seal test or if the overall air within limits. In many instances (e.g., lock leakage is not door has failed) primary containment remains only one seal per only 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (according to LCO 3.6.1.1) OPERABLE, yet would be provided to restore the air lock door to requiring a plant shutdown. OPERABLE status prior to In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required requires that one door in the affected Action C.2 primary containment air locks must be verified closed. This must be completed within the I hour Completion Required Action specified time period is consistent with Time. This the ACTIONS of LCO 3.6.1.1, which require that primary containment be restored to OPERABLE status within I hour.
Additionally, the air lock must be restored to status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion OPERABLE reasonable for restoring an inoperable air Time is lock to OPERABLE status considering that at least one door is maintained closed in each affected air lock.
If the inoperable primary containment restored to OPERABLE status within the air lock cannot be Time while operating in MODE 1, 2, or 3,associated Completion the plant must be brought to a NODE in which the LCO does achieve this status, the plant must be not apply. To MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within brought to at least allowed Completion Times are reasonable, 36 hours. The based on operating experience, to reach the required plant power conditions in an orderly manner andconditions from full without challenging plant systems.
E.l. F-2, and F-3 If the inoperable primary containment restored to OPERABLE status within the airlock cannot be Time during operations with a potential associated Completion for draining the reactor vessel (OFOR s ^- - - r th.Sr action is reqired to iately represent a potential for releasingsuspend activities that radioactive material, (continuedl RIVER BENO B 3.6-11 Revision No. 2-3
-D
Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS E.1, E.2, and E.3 (continued) thus placing the unit in a Condition that minimizes*,
C</ ent-*m'ent bee murset*/.r~t e suspendedSues i-ne n f these 6f p*ioq ! f applicab~le, action must be immediately'iiitdt uspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.
SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of the Primary Containment Leakage Rate Testing Program (Ref. 5). This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The specified acceptance criteria i.e. , 5 K) containment air lock 1JRB*DRA1 and ensures that the combined leakage r*a */-o rate leakage paths is less than the specified leakage "rate. ollowing the removal of the fuel building as a secondary containment boundary in accordance with License Amendment 113, the leakage from primary containment air lock 1JRB*DRA2 represents secondary bypass lea Kage limit occ/hr. This provides assurance in MODES 1, 2, and 3 that the assumptions in the radiological evaluations are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (e.g., leakage through the air lock door with the highest leakage) unless the penetration is isolated by use of (for this Specification) one closed and locked air lock door. The leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation devices (e.g., air lock door).
If both air lock doors are closed, the actual leakage rate is the lesser leakage rate of the two barriers (doors). This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J, Option B, maximum pathway leakage limits used to evaluate Type A, B and C limits are to be quantified in accordance with Appendix J, Option B).
pM,6 O I-.fte 'Pl11P . .TR GsS, o OPDRVS, (continued)
RIVER BEND B 3.6-12 Revision No. 101
Primary Containment Air Locks B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.1 (continued)
REQUIREMENTS the reactor coolant system is not pressurized and specific primary containment leakage limits are not imposed. However, due to the size of the air lock penetration, leakage limits are imposed to assure an OPERABLE barrier. In these conditions the leakage limits are not related to radiological evaluations, but only reflect engineering judgment of an acceptable barrier. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program.
The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR, requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1.1 during operation in MODES 1,2, and 3. This ensures that air lock leakage is properly accounted for in determining the overall primary containment leakage rate. Since the overall primary containment leakage rate is only applicable in MODE 1, 2, and 3 operation, the Note 2 requirement is imposed only during these MODES.
SR 3.6.1.2.2 The seal air flask pressure is verified to be at __90 psig every 7 days to ensure that the seal system remains viable. It must be checked because it could bleed down during or following access through the air lock, which occurs regularly. The 7 day Frequency has been shown to be acceptable through operating experience and is considered adequate in view of the other indications available to operations personnel that the seal air flask pressure is low.
(continued)
RIVER BEND B 3.6-13 Revision No. 6-10
Primary Containment Air Locks B 3.6.1.2 BASES -
SURVEILLANCE SR 3.6.1.2.3 REQUIREMENTS (continued) The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure (Ref. 3), closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur. Oue to the nature of this interlock, and given that the interlock mechanism is only challenged when the primary containment airlock door is opened, this test is only required to be performed upon entering or exiting a primary containment air lock, but is not required more frequently than once per 184 days. The 184 day Frequency is based on engineering judgment and is" considered adequate in view of other administrative controls.
SR 3.6.1.2.4 A seal pneumatic system test to ensle that pressure does not decay at a rate equivalent to > ---- psig for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from an initial pressure o-0 psig is an effective leakage rate test to verify system performance.
The 18 month Frequency is based on the fact that operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. USAR, Section 3.8.
I 2. 10 CFR 50, Appendix J, Option B.
~~3. USAR, Table 6.2-1.~~
~~4. USAR, 15.7.4.~~
~~5. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.~~
RIVER BEND 8 3.6-14 Revision No. 2-3
PCIVs B 3.6.1.3 BASES (continued)
APPLICABLE The PCIVs LCO was derived from the assumptions related to SAFETY ANALYSES minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during major accidents. As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment. Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO.
The DBAs that result in a release f radioactive material for which the consequences are mi igated by PCIVs, are a loss of coolant accident (LOCA) -a main steam line break Re s. an 2 In the a ysis or each of these accidents, it is assumed that PCIVs are either closed or function to close within'the required isolation time following event initiation. This ensures that potential paths to the environment through PCIVs are minimized. Of the events analyzed in Reference 1, the LOCA is the most limiting event due to radiological consequences. It is assumed that the primary containment is isolated such that release of fission products to the environment is controlled.
PCIVs satisfy Criterion 3 of the NRC Policy Statement.
LCO PCIVs form a part of the primary containment boundary and some also form a part of the RCPB. The PCIV safety function is related to minimizing the loss of reactor coolant activity and establishing the primary containment boundary during a DBA.
The power operated isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal. Primary containment purge valves are not qualified to close under accident conditions, and therefore, are blocked to prevent full opening to be OPERABLE.
The normally closed PCIVs are considered OPERABLE when, as applicable, manual valves are closed or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, or blind flanges are in place. The valves covered by this LCO (continued)
B 3.6-16 Revision No. 0 RIVER BEND
PCIVs B 3.6.1.3 BAS ES ACTIONS C.'
(continued)
With the-. e;,hydrostatic ontainment bypass leakage rate leakage rate, or MSIV leakage rate not within limit, the assumptions of the safety analysis may not be met. Therefore, the leakage must be restored to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated power operated or automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolation penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration and the relative importance to the overall containment function.
D.1. D.2, and D.3 In the event one or more primary containment purge valves are not within the purge valve leakage limits, purge valve leakage must be restored to within limits or the affected penetration must be isolated. The method of isolation must be by the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated power operated or automatic valve, closed manual valve, and blind flange. If a purge valve with resilient seals is utilized to satisfy Required Action D.1, it must have been demonstrated to meet the leakage requirements of SR 3.6.1.3.5. The specified Completion Time is reasonable, considering that one primary containment purge valve remains closed so that a gross breach of primary containment does not exist.
In accordance with Required Action D.2, this penetration flow path must be verified to be isolated on a periodic basis. The periodic verification is necessary to ensure that primary containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be isolated should an event occur. This Required Action does not require any testing or valve manipulation. Rather, it involves (continued) 8 3.6-20 Revision No. 3-4 RIVER BEND
PCIVs B 3.6.1.3 BASES ACTIONS F.1 and F.2 (continued) vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
If suspending the OPDRVs would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valves to while OPERABLE status. This allows RHR to remain in service actions are being taken to restore the valve.
SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR verifies that the 36 inch primary containment purge valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of the limits.
The SR is also modified by a Note (Note 1) stating that primary containment purge valves are only required to be1, 2, closed in MODES 1, 2, and 3. At times other than MODE 3 when the purge valves are re orclo~singj*mn Mrradl**,=to be capable ol uired of pressuriation concilernls arenot preNse'nt and the
"-purge- vayes are allowed to be open (automatic isolation capability would be required by SR 3.6.1.3.4 and SR 3.6.1.3.7).
SR is The SR is modified by a Note (Note 2) stating that the not required to be met when the purge valves are open for may the stated reasons. The Note states that these valves be opened for pressure control, ALARA, or air quality or considerations for personnel entry or for Surveillances, special testing on the purge system that require the valves to be open (e.g., testing of the containment purge radiation monitors). These primary containment purge valves are capable of closing in the environment following a LOCA.
Therefore, these valves are allowed to be open for limited with periods of time. The 31 day Frequency is consistent other PCIV requirements.
(continued)
B 3.6-22 Revision No. 0 RIVER BEND
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.5 REQUIREMENTS (continued) For primary containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, A endix J (R _ is required to ensure OPERABILITY. . cffvc ,Ie ud e accep ance criterion for each purge exhaus va ye is < 0.01 La when pressurized to a P , 7.6 psig. Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation, and the importance of maintaining this penetration leak tight (due to the direct path between primary containment and the environment), a Frequency of 184 days was established. Additionally, this SR must be performed within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that which occurs to a valve that has not been opened). Thus, decreasing the interval (from 184 days) is a prudent measure after a Valve has been opened.
The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these MODES, purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria.
SR 3.6.1.3.6 Verifying that the full closure isolation time of each MSIV is within the specified limits is required to demonstrate
- '--'OPERABILITY. The full closure isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. The maximum
- closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges. The Frequency of this SR is in accordance with the Inservice Testing Program.
(continued)
RIVER BEND B 3.6-25 Revision No. 1
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.9 (continued)
REQUIREMENTS evaluations of Reference 4 are met. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated power operated or automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J, Option B maximum pathway leakage limits are to be quantified in accordance with Appendix J, Option B). The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 5).
A note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2 and 3. In the other conditions the Reactor Coolant System is not pressurized and primary containment leakage limits are not required.
SR 3.6.1.3.10 The analyses in References 2 and 3 are based on leakage out of th primary containment that is less than the specified leakage rate. eakage through the valves sealed in each division of MS-PLCS must be
< 150 scfh per division when tested at > Pa, 7.6 psig. The leakage rate must be verified to be in accordance with the leakage test requirements of Reference 4, as modified by approved exemptions.
A note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2 and 3. In the other conditions, the Reactor Coolant System is not pressurized and specific primary containment leakage limits are not required. The Frequency is required by the Primary Containment Leakage Rate Testing Program (Ref. 5).
(continued)
~~7~~~ý (14ZC dW$~~.i~ re -'5 C~~
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RIVER BEND B 3.6-27 Revision No. 3-4
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.11 REQUIREMENTS (continued) Surveillance of hydrostatically tested lines at a 1.1 P4, 8.36 psig provides assurance that the calculation assumptions of References 2 and 3 are met. The acceptance criteria for the combined leakage of all hydrostatically tested lines is 1.0 gpm times the total number of hydrostatically tested PCIVs when tested at 1.1 Pa The combined leakage rates must be demonstrated at the frequency of the leakage test requirements of the Primary Containment Leakage Rate Testing Program (Ref. 5).
A note is added to this SR which states that these valves are only required to meet the combined leakage rate in MODES 1, 2, and 3 since this is when the Reactor Coolant System is pressurized and primary containment is required.
In some instances, the valves are required to be capable of' automatically closing during MODES other than MODES 1, 2, and 3. However, specific leakage limits are not applicable in these other MODES or conditions.
"This SR e eof chbypas nedkage e ratepoh annm bus Sbypass le g ahs l s han e. specified lej~aoge Srate. T, s r vi e as u an e a the a s m t ,/ snin tnh j radiolic:a vl io rnce 4. are m t. The i' ea e aeo c by oak g pat is.ssumed be th maximum pa way leak e (leakage thro h the se of e two isol ion valv unless the p tratio is isolated
by use of e closed nd de-activate power o rated or automati valve, c sed manual va e, or b id flange. In thsce, tuh~e 1 kage rate of e isolat bypass leakage paths asumeto be the ac al pathwF leakage through th iso tion dev e. If both soilation alves in the petratlo are closed, e actual akgrteith esserfle-ng age rate of he two v.yes. This method quanti maximum athway le age is only to be ed for (conti nued)
RIVER BEND B 3.6-.28 Revision No. 3-4
Primary Containment-Shutdown B 3.6.1.10 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.10 Primary Containment-Shutdown BASES BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material to within limits. The primary containment consists of a steel lined, reinforced concrete vessel, which surrounds the Reactor Primary System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment.
Additionally, this structure provides shielding from the fission products that may be present in the primary containment atmosphere following accident conditions.
The isolation devices for the penetrations in the prinary containment boundary are a part of the primary containment leak tight barrier. To maintain this leak tight barrier for accidents during shutdown conditions:
a. All penetrations required to be closed during accident conditions are closed by manual valves, blind flanges, or de-activated power operated or automatic valves secured in their closed positions, or the equivalent, except as provided in LCO 3.6.1.3. "Primary Containment Isolation Valves (PCIVs)":
~~b. Primary containment air locks are OPERABLE, except as provided in LCO 3.6.1.2. "Primary Containment Air Locks":~~
~~and~~
~~c. All equipment hatches are closed.~~
This Specification ensures that the erformance of th'rImary criticality, or reactor vessel draindown, provides an acceptable leakage barrier to contain fission products, thereby minimizing offsite doses.
"I'll".
"I'l"I'll"I'll'll""I'll""I'lI APPLICABLE ý,-=-=rrerr design barsis for t-h- !entaimment SAFETY ANALYSES from a FHA imside rontin-wzd)
RIVER BEND B 3.6-50 Revision 6-13 A
Primary Containment-Shutdown B 3.6.1.10 BASES APPLICABLE rprimary containm t (Ref.2), to limit ses at the site undary to within SAFETY ANAL' limits. The privrary containment rforms no active fnction in respons (continued) to this evenlowever, its leak ghtness is requl d to ensure that t, release ofradioactive mater s from the prim containment is r tricte to thos eakage rates a med in safety a lyses.
Th FHA inside the rimary containm/t is assumed to oc r only after 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months since e reactor was la critical. The fissio product releas(
is, in turn, bas on an assumed akage rate from v t and drain val with a comb' ed flow rate of 70 cfm (ba=sed on a ssumed 0.367 itch water ga e differential pres re). This assume pressure refle the fact th the FHA does not oduce elevated c tainment pres res as i the c e for the DBA LO. However, as a added conse tism, the an sis assumes a no -mechanistic addi nal leakage o .26% of the Primary containment satisfies Criterion 3 of the NRC Policy Statement.
LCO Primary containment OPERABILITY is maintained by providing a contained volume to limit fission product escape following 4IIZ Q unanticipated reactivity or water level excursion. Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis. Since no credit is assumed for automatic isolation valve closure, and any leakage which would occur prior to valve closure is similarly not accounted for, all penetrations which could communicate gaseous fission products to the environment must remain closed.
However, a limited number of primary containment penetration vent and drain valves may remain opened, and the primary containment considered OPERABLE provided the calculated leakage flow rate through the open vent and drain valves is less < 70.2 cfm.
Leakage rates specified for the primary containment and air locks, addressed in LCO 3.6.1.1 and LCO 3.6.1.2 are not directly applicable during the shutdown conditions addressed in this LCO.
(continued)
RIVER BEND B 3.6-51 Revision 2-1
B3.6.1.10 "Primary Containment - Shutdown Insert A:
Containment integrity is not explicitly assumed in any of the shutdown analyses. An dose analysis for an operation with a potential for draining the reactor vessel (OPDRV) is not performed. However, an OPDRV could potentially uncover the core and lead to potential fuel damage. Containment is required to ensure that the LOCA remains the bounding accident.
Primary Containment-Shutdown B 3.6.1.10 BASES (continued)
APPLICABILITY In MODES 4 and 5. the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining an OPERABLE primary containment in MODE 4 or 5 to ensure a control volume, is only required during situations for which significant releases of radioactive material can be postulated; such as during operations with a potential for draining the reactor vessel (OPDRVs) or is only required to be OPERABLE during fuel handling (involeamtainment yrrpinofg rczcently irradiated fuel (i.e.. fuel that has acctipie&
...... !I days).
Requirements for ECCS OPERABILITY during MODES 1. 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1.
ACTIONS A.1 and A.2 In the event that primary containment is inoperable, action is required to immediately suspend activities that represent a potential for releasing radioactive material, thus p1acinjg th unit in a Condition that minimizes risk. flf appliable.-vme suzpended. Suspcnsien of these acti'.itic: shall net ppeeltde-
-Ifapplicable, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Action must continue until OPDRVs are suspended.
SURVEILLANCE REQUIREMENTS This SR verifies that each primary containment penetration that could communicate gaseous fission products to the environment during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive gases outside of the primary containment boundary is within design limits. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Single isolation barriers that meet this criterion are a closed and de-activated power operated or automatic valve, a closed manual valve, a blind flange. or equivalent. This does not preclude the use of two active (ie. power operated and/or automatic) valves in the closed position for a given penetration.
This SR does not require any testing or valve manipulation.
(continued)-
RIVER BEND B 3.6-52 Revision 6-13
Primary Containment-Shutdown B 3.6.1.10 BASES SURVEILLANCE SR 3.6.1.10.1 (continued)
REQUIREMENTS Rather, it involves verification, through a system walkdown, that the required valves are in the correct position. The 31 day Frequency was chosen to provide added assurance that the valves remain in the correct positions.
reuid obe mtovetad in lnepahy ovd.4 thle tlcal CuIe fl ow rate rouhopnv ania pa ays is <:5 .2 cfm. Adm' isrtv co ols ensur/
/(3topen ven and drain pa ~ways will: only be ,iened o()support 1 akage rate *lened requir ermonitoring no exceed vent(2) ansi/rain t ting; valv12 , lves, as w Sas the c Itainment-to- uxiliary buil ring differe iti
ýpressur) every 2 os; and (4) asr tleas one per on isa ndt eac~v oenpenet ra on (Ref.1 REFERENCES 1. -or
"" USAR, Section 15.77 RIVER BEND B 3.6-53 Revision No. 0
Secondary Containment- Operating B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment- Operating BASES BACKGROUND The function of the secondary containment is to contain, dilute.
and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT)
System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.
The secondary containment consists of the shield building and auxiliary building, and completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products.
It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g.. due to pump/motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2. "Secondary Containment
- so on Dampers (SCIDs) Rmr_
LCO 3.6.4.3. "Stand y Gas (S System. and
.reatment
'"--LCO 3.6.4.4. "Shield Building Annulus Mixing System."
The isolation devices for the penetrations in the secondary containment boundary are a part of the secondary containment barrier. To maintain this barrier:
~~a. All Auxiliary Building penetrations and Shield Building annulus penetrations required to be closed during accident conditions are either:~~
(continued)
RIVER BEND B 3.6-83 Revision No. 6-5
Secondary Containment- Operating B 3.6.4.1 BASES BACKGROUND 1. by an OPERABLE Capable of being closed solation secondary containmen ti signal. or (continued)
~~2. Closed by at least one manual valve, blind flange. or deactivated automatic valve or damper, as applicable.~~
~~secured in its closed position, except as provided in LCO 3.6.4.2:~~
~~b. All Auxiliary Building and Shield Building Annulus equipment hatches are closed and sealed:~~
~~c. The Standby Gas Treatment System is OPERABLE. except as provided in LCO 3.6.4.3: and~~
~~d. At least one door in each access to the Auxiliary Building and Shield Building Annulus is closed, except for routine entry and exit of personnel and equipment.~~
I APPLICABLE The principal accident for which credit is taken for secondary I SAFETY ANALYSES containment OPERABILITY is a LOCA (Ref. 1). The secondary I containment performs no active function in response to this limiting event: however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis. and that fission products entrapped within the secondary containment structures will be treated by the SGT System prior to discharge to the environment.
Secondary containment--operating satisfies Criterion 3 of the NRC Policy Statement.
LCO An OPERABLE secondary containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in the shield building or (continued)
RIVER BEND B 3.6-84 Revision No. 6-5
Secondary Containment- Operating B 3.6.4.1 BASES LCO auxiliary building, can be diluted and processed prior to (continued) release to the environment. For the secondary containment to be considered OPERABLE. it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.
APPLICABILITY In MODES 1, 2. and 3. a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES-. Therefore. maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume. tor-ogni, erxc~ 4e*nt
\~ j_._V pg! _;ti onSytemof Fueltp
. f,,:el aseble i."efe ACTIONS A.I If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1. 2. and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.
If the secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating (continued)
RIVER BEND B 3.6-85 Revision No. 6-5
Secondary Containment--Operating B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.-1.4 and SR 3.6.4.1.6 REQUIREMENTS The SGT System exhausts the shield building annulus and auxiliary building atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated. SR 3.6.4.1.4 verifies that the SGT System will rapidly establish and maintain a pressure in the shield building annulus and auxiliary building that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that one SGT subsystem will draw down the shield building annulus and auxiliary building to Ž 0.5 and 0.5 inches of vacuum water gauge in : 18.5 and econds. respectively. This cannot be accomplished if the condary containment boundary is not intact. SR 3.6.4.1.6 demonstrates that each SGT subsystem can maintain ; 0.5 and 0.25 inches of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months test C , period allows shield building annulus and auxiliary building to be in thermal equilibrium at steady state conditions. Therefore.
these two tests are used to ensure the integrity of this portion of the secondary containment boundary. Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS. however, to ensure that in addition to the requirements of LCO 3.6.4.3. either SGT subsystem will perform this test.
Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. USAR, Section 15.6.5.
~~2. USAR, Section 15.7.4.~~
RIVER BEND B 3.6-87 Revision No. 6-5
SC IDs*
B 3.6 .4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secon r ntalnment Isolation Dampers (SCIDs) -nd FPul Building
"" nm n"fT Isoltio BASES BACKGROUND The function of the SCIDs" . in combination with other accident mitigation systems. is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1). Secondary containment isolation within the time limits s ecified for those isolation dampers designed to close- -tmat4a , ensures that fission products that leak from primary contaiýýmet folowing a DBA, that are released during certain operations when primary containment is not required to be OPERABLE, or that take place outside primary containment, are maintained within the secondary containment boundary*.-ih-f,*--
-i*4e!i~hm h buaWilding bo....ary.
The OPERABILITY requirements for SCIDsd 4 help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. Isolation barrier(s) for the penetration are discussed in Reference 2. The isolation devices addressed by this LCO are either passive or active (automatic). Manual dampers, de-activated automatic dampers secured in their closed position, check dampers with flow through the damper secured, and blind flanges are considered passive devices. Check dampers and other automatic dampers designed to close without operator action following an accident are considered active devices.
SCIDs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary W!ig;G, G nn ,t . ; y, fe,,÷^met Other penetrations are isolated by the use of dampers or valves in the closed position or blind flanges.
APPLICABLE The SCIDs En_ý must be OPERABLE to ensure the barrier to SAFETY ANALYSES fission prod't releases is established. The principal accident for which the secondary containment boundary is required is a
- (continued)
RIVER BEND 8 3.6-89 Revision No. 6-5
SCIDst B 3.6.4.2 BASES APPLICABLE containment A 4 'l.j;A-j 4-,not perform an active function SAFETY ANALYSES in response toŽtIse limiting events. However. the boundary (continued) established by SCIDs re required to ensure that fission products are processed by the ventilation systems before being released to the environment.
Maintaining SCIDs L O OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SG System (following a LOCA) . -__ I.S j .
- , prior to discharge to the environment.
SCIDsAsatisfy Criterion 3 of the NRC Policy Statement.
LCO SCIDs form a part of the secondary containment boundary. The SCIO safety function is related to control of offsite radiation releases resulting from DBAs.
The power operated isolation dampers are considered OPERABLE when their isolation times are within limits. Additionally.
ertedampers are required to actuate on an isolation signal.
The normally closed isolation dampers or blind flanges are considered OPERABLE when manual dampers are closed or open in accordance with appropriate administrative controls, automatic dampers are de-activated and secured in their closed position.
or blind flanges are in place. The SCIDs -covered by this LCO. along with their associated strok-'tl mes, if applicable, are listed in Reference 4.
APPLICABILITY In MODES 1. 2. and 3. a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore. OPERABILITY of SCIDs is required.
In MODES 4 and 5. the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore. maintaining SCIDs OPERABLE is not equi red in MODE 4 or 5 e .0p - ...... .......
t_........ (cntine (continued)
RIVER BEND 8 3.6-90 Revision No. 6-5
SC IDsg B 3.6.4.2 BASES APPLICABILITY such as d ng movemen5of recently irrýlated fuel a mblies (continued) (i.e.. el that has ccupied part 0 critical e within t previ s 11 days) Moving irrad'ed fuel ass lies in th Pr* ary Contai ent is addre d adequately LCO 3.6.1 rimary Co inment-Shu wn."
Moving cently irr ated fuel ass blies in the uel buil*n will quire only,, e FBIDs asso 'ated with th fuel bui ng to be ERABLE.
ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way. the penetration can be rapidly isolated when the need for secondary containment isolation is indicated.
The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition pr vide appropriate compensatory actions for each inoperable SCID = Complying with the Required Actions may
~al for continued operation, and subsequent inoperable SCIDs49 izj*~ are governed by subsequent Condition entry and application of associated Required Actions.
The third Note ensures appropriate remedial actions are taken, if necessary. if the affected system(s) are rendered inoperable by an inoperable SCIDQ A.1 and A.2 In the event that there are one or more penetration flow paths with one SCID inoperable, the affected penetration flow path(s) musf isolated. The method of isolation must 6;
include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criteria are a closed and de-activated automatic damper. a closed manual damper or a blind flange. For penetrations isolated in accordance with Required Action A.1. the device used to isolate the penetration should be the closest available device to the applicable isolation boundary. This Required Action must be completed within the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion (continued)
RIVER BEND B 3.6-91 Revision No. 6-5
SC IDs u B 3.6.4.2 BASES ACTIONS A.1 and A.2 (continued)
Time. The specified time period is reasonable considering the time required to isolate the penetration and the low probability of a DBA. which requires the SCIDs (-40to close. occurring during this short time.
For affected penetrations that have been isolated in accordance with Required Action A.1. the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary containment or fuel building penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be isolated should an event occur. This Required Action does not require any testing or isolation device manipulation. Rather, it involves verification that the affected penetration remains isolated.
Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore.
the probability of misalignment once they have been verified to be in the proper position, is low.
With two SCIDs in one or more penetration flow paths inoperable (Con ion A is entered if one SCID "i s inoperable in each of two penetrations), the affected penetration flow path must be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic damper. a closed manual damper. and a blind flange. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable.
considering the time required to isolate the penetration and the low probability of a DBA, which requires the SCIDs(4 )to close, occurring during this short time.
(continued)
RIVER BEND B 3.6-92 Revision No. 6-5
SCIDsg B 3.6.4.2 BASES ACTIONS CA and C.2 (continued)
If any Required Action and associated Completion Time cannot be met for SCIDs. the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
Co 3.0 islynot appli cale. If emovi o recent y ir diatedfe r
Refanytquiredt 0c*Ion asesonm aied Ction a teista t that equassemb ie sswi ei n e 4ue or 5.et alCs..3 wol tuspency an ereOre.0 cthis UVILNESpende.6 . isnothe mo uis ing casei l rradiate Ss.so4o.hi2tviyshl.o1reldfu.it tosuen movemre sembl w entl rdae lein MODE corplethn h e et plantl mement isa mdendent oful mussoelaced time peion orpop in ac tofare less tion.
than or irradiated el assemblies ould not be sufficient re on to require a eactor shutd SURVEILLANCE SR 3.6I.4.2J1 REQU IREMENTS Verifying the 1 1ati n time of each required power operated
~j~SCID is within limits is required to demonstrate OPERABILITY. The isolation time test ensures that the SCIDst -frkill isolate in a time period less than or equal to tha assMumed in the safety analyses. The Frequency of this SR is 92 days.
(continued)
RIVER BEND B 3.6-93 Revision No. 6-5
SCIDs6B3.
B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.2 REQUIREMENTS (continued) Verifying that each required( SCID isolation signal is required to prevent leakage of I material from secondary containment DBA or other accidents. This SR ensures that eac will actuate to the isolation position on a seconda isolation signal, - -A perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
Therefore. the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. USAR, Section 15.6.5.
~~2. USAR, Section 6.2.3.~~
~~3. USAR. Section 15.7.4.~~
~~4. TRM. Table 3.6.4.2-1.~~
RIVER BEND B 3.6-94 Revision No. 6-5
SCIDs. ..
B 3.6.4.ý2 THIS PAGE INTENTIONALLY LEFT BLANK RIVER BEND B 3.6-95 Revision No. 6-5
SGT System B 3.6.4.3 BASES (continued)
SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT subsystem for a: 10 continuous hours ensures that both subsystems are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on (automatic heater cycling to maintain temperature) for z 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.
SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specified test frequencies and additional information are discussed in detail in the VFTP.
This SR requires verfication that -aenh SGT subsystem starts upon receji t 'i.A'initiation N signal.
Thhe in 3.3.6.2.5 overlaps this to p'rovi !ecompete testing of the safety function.
While this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle.
.Therefore, the Frequency was concluded to be acceptable from a 'reliability standpoint.
SR 3.6.4.3.4 This SR requires verification that the SGT filter cooling bypass damper can be opened and the fan started. This ensures that the ventilation mode of SGT System operation is (continued)
B 3.6-99 Revision No. 0 RIVER BEND
npFu(ef. 1).
Cl gh funcition of Fel the Building filteredof and System econsists adsoHbed The "bs FuelBuilding Fuel Ventilation (FHA)
HandlinFuAccident tw9fuly p*rior toexhaustigto the.envir~on n. ,
redundant su ystems, each wi4 its own set of ducork, daCmper ,.roal filtr. , tra ,fand controls. ......
order the directigton the aniro flow)
Each chaicoan lter trais consists of (compoaents listedin
~~a. A moisture separ tor; An electric ater;~~
~~d. A hig efficiency particul e air (HEPA) filter;~~
~~e. A arcoal adsorber;~~
~~f. second HEPA filte*,and g A centrifugal fa with inlet flow con 01 vanes.~~
n w*ve r in the air, whi e the a
The moisture sep ator isprovided me electric heat to emove entraine humidity of t airstream to less han 70% (Ref. 2).
prefilter re yes large particul e matter, filter is reduces the while vided to remove fe particulate mat r and protect t echarcoal from fo ing. The charcoal dsorber removes aseous elemeta I i e and organCic Iidan dsrea t
he HEPA RVER BED3(continued)-
RIVER BEND 8 3.6-112 Revision No. 0
Fuel"1uilding e 1sation-Sysem- Fu -Hindlin B 3.6.4.7 BASES BACKGROUND The Fuel Building ye ilation System /autom ically starts (continued) and operates in re onse to actuation si als indicative of conditions or an ccident that could r uire operation of the system.
APPLIC LEF The desi basis for the Fuel B iding Ventilation Syste is SAFE ANALYSES to mit ate the consequences a fuel handling accide (Ref. 3). For all events a yzed, the Fuel Buildin Ventilation Sy em is shown to reduce, via filtration and adsor tion. the rldioactive material rel ased to the environment. Since the system is assumed to f' ter all releases, with e analysis not accounting for any d ay in system startup, least one subsystem must be i operation while handl' g recently irradiated fuel (i.e., fuel at has occupied part a critical core within the previous 11 days).
Fuel Bui ing Ventilation Syste satisfies Criterion 3 of .te
/ ~ The~NRCPol ic citatement./
LCO Follo ng a FHA involving r ently irradiated fuel, ýa/nimum of /
one uel Building Ventilaion subsystem is requi redo maintain t fuel building at a n gative pressure with res ct to the nvironment and to pro ess gaseous releases. M ting the LCO requirements for two perable subsystems ensu s operation of at least one Fuel Bui ing Ventilation subsyst in the event of a single active faA ure. Requiring one sub stem to be in
/ operation ensu s no releases occur tha are not filtered and
~adsorbed.
APPLI ILI - Regardl sof the plant operati MODE, anytime recentl irra 'ated fuel (i.e., fuel a*Mt has occupied part ofa critical cor within the previous 1 days) is being handled ere is the ptential for a FHA and e Fuel Building Ventil ion System is equired to mitigate t consequences.
N Xx %-r.'
cM WA 6'AI&
Revision No. 6-5 B 3.6-113 RIVER RIVER BEND 8 3.6-113 Revision No. 6-5
Fuel Building Vtntilati on System--Fuel Ha ing ACTIONSA1 With one fuel ilding ventilation charc filtration subsystem inoperable. e inoperable subsystem m t be restored to OPERABL "status witrn 7 days. In this Condi on. the remaining OPERAB\
fuel buil ng ventilation charcoal iltration subsystem is adequa to perform the require radioactivity release co rol func on. However, the overa system reliability is r uced be use a single failure inte OPERABLE subsystem co d result Sthe radioactivity rele e control function not ing adequately performed. e 7 day Completion Time *sbased on consideration of such actors as the availabil y of the OPERABLE redundant fuel buil ng ventilation charcoal iltration subsystem and the low proba lity of a FHA occurrin during this period.
B.1 and B.
If the uel building ventilat* n charcoal filtration su stem cann be restored to OPERA E status within the requ ed Co letion Time the plant ust be brought to a cond' ion in which e LCO does not apply. Additionally, if both s systems are inoperable or if the ne required subsystem no in operation the system is incapabl of performing its requir accident mitigation funct' n and the plant must be rought to a condition I in which the L does not apply. To ac .eve this, recently irradiated fIe handling must be susp ded immnediately.
Suspensia shall not preclude compl ion of fuel movement a safe po*tion.
S VEILLANCE S 3.6.4.7.
QUIREMENTS .
olISuIVeill anice UIr ILoneL. a LI that u uu U ny venr*ilatio WIG
/* charcoal filtration ubsystem is in oper ion and f iltering the fuel building at phere. The Frequen of 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months "s is sufficient in v' of other visual d audible indi cations available to e operator for mon' oring the fuel bnuilding ventilation arcoal filtratio subsystem in the ccIntrol oom.
L / . .. (conti nued)
-S LEA- 'týox\ý'
RIVER BEND B 3.6-114 Revision No. 6-5
FuelBul g e -i-tio-- el Hand ng 3.6.4.7 BASESB SURVEILLA E SR 3.6.4.7.2 REQUIRE TS (con nued) Operating each uel building venV ation charcoal filtrati subsystem fo a 10 continuous urs ensures that both subsystems re OPERABLE and at all associated contro are functioni g properly. It so ensures that blockage fan or motor f'lure, or excessi evibration can be detec dfor corre ive action. Ope tion with the heaters o rating (au matic heater cyc ng to maintain temperat e) for 2: 10 c tinuous hours ev y 31 days eliminates mo& ture on the dsorbers and HEP filters. The 31 day Fr uency was developed in con deration of the known liability of fan motors and con ols and the redundancy vailable in the system.
SR 3.6. .7.3 This R verifies that the r quired fuel building venti ation ch coal filtration filte testing is performed in cordance with the Ven lation Filter Testing Pro am VFTP). The fuel bull ing ventilation charcoal ftration filter tests are in ccordance with Regulatory uide 1.52 (Ref. 4). The VFT includes testing HEPA fil er performance, cha oal adsorber efficiency, inimum system flow rate, and he physical properties of he activated charcoal (ge ral use and following sp ific operations).
Specified t st frequencies and additj nal information are discussed in detail in the VFTP.
SR .6.4.7.4 is SR requires verific ion that each fuel bui ing ventilation charcoal fi ration subsystem star upon receipt of an actual rsimulated initiatio signal. The LOGIC SYSTEM FUNCTI AL TEST in SR 3.3.6. . overlaps this SR to prov ide corn ete testing of the s ety function.
While this Surv ilance can be perfo dwith the reactor power, operati g experience has sho these components usually pass he Surveillance whe performed at the 18 onth Frequency, hich is based on the efueling cycle.
Therefor , the Frequency was c cluded to be accep ble from a reli ility standpoint.
Pck~cnntenvvX)
RIVER~~ta BEND B .615 evsonNo RIVER BEND B 3.6-115 Revision No. 0
I ui igVentilIatioS e- Fuel Hnln Buildng o 3.6.4.7
(
BASES*/*";
SURVEILLANCE .4.7.5 REQUIREMENTS (continued) This SR requires verif* ation that the fuel bui ding ventilation charcoal iltration filter cooli g bypass damper can be opened and e fan started. This sures that the ventilation mode f Fuel Building Venti tion System operation is av lable. While this S veillance can be performed wit the reactor at power operating experience has shown t se components usuall pass the Surveillance equency, which is bas on when perf med at the 18 month w
the ref ling cycle. Theref ethe Frequency was c cluded "to be cceptable from a reability standpoint.
REFERENCES 1. 10 CFR 50, Ap endix A, GDC 41.
~~2. USAR, Sec ion 6.2.3.~~
~~3. USAR, ection 15.6.5.~~
~~4. R ulatory Guide 1.52, 2.~~
...... .2
... ~
-- Tlatory Guide-B 3.6-116 Revision No. 0 RIVER BEND
B Drywel 3.6.5.11 BASES ACTIONS A.1 (continued)
Time drywell is inoperable is minimal. Also, the Completion primary is the same as that applied to inoperability of the containment in LCO 3.6.1.1, "Primary Containment-Operating."
8.1 and B.2 within If the drywell cannot be restored to OPERABLE status to a the required Completion Time, the plant must be brought this MODE in which the LCO does not apply. To achieve 3 within status, the plant must be brought to at least MODE 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating from full experience, to reach the required plant conditions power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.5.1.1 REQUIREMENTS at ? 75 psig The seal air flask pressure is verified to beremains viable.
every 7 days to ensure that the seal system during or It must be checked because it could bleed down The 7 day following access through the personnel door. operating Frequency has been shown to be acceptable through the other of experience and is considered adequate in view that the seal indications available to operations personnel air flask pressure is low.
SR 3.6.5.1.2 )
does A seal pneumatic system test to ens re that pressure period not decay at a rate equivalent to >" psig for a of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from an initial pressure of 75 psig is an effective leakage rate test to verify system toperformance.
perform this The 18 month Frequency is based on the need during a plant Surveillance under the conditions that apply transient if the outage and the potential for an unplanned at power.
Surveillance were performed with the reactor usually pass Operating experience has shown these components month Frequency, the Surveillance when performed'at the 18 the cycle. Therefore, which is based on the refueling from a reliability Frequency was concluded to be acceptable standpoint.
(continued)
Revision No. 0 RIVER BEND B 3.6-119
Drywell Air Lock B 3.6.5.2 BASES (continued)
SURVEILLANCE SR 3.6.5.2.5 (continued)
REQUI REMENTS system pressure does not decay at an unacceptable rate. The air lock seal will support drywell OPERABILITY down to a pneumatic pressure of 75 psig. Since the air lock seal air flask pressure i verified in SR 3.6.5.2.2 to be Ž 75 psig, a deca rate . psig over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable. The hour interval t24 is based on engineering judgment, considering that there is no postulated DBA where the drywell is still pressurized 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage when the air lock OPERABILITY is not required.
Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES 1. 10 CFR 50, Appendix J.
~~2. USAR, Chapters 6 and 15.~~
I RIVER BEND B 3.6-128 Revision No. 2-4
CRFA System B 3.7.2 B 3.7 PLANT SYSTEMS B 3.7.2 Control Room Fresh Air (CRFA) System BASES BACKGROUND The CRFA System provides a radiologically controlled environment from which the unit can be safely operated following a Design Basis Accident (DBA).
The safety related function of the CRFA System used to control radiation exposure consists of two independent and redundant high efficiency air filtration subsystems for treatment of recirculated air or outside supply air. Each subsystem consists of a demister, an electric heater, a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section, a second HEPA filter, a fan, and the associated ductwork and dampers.
Demisters remove water droplets from the airstream.
Prefilters and HEPA filters remove particulate matter that may be radioactive. The charcoal adsorbers provide a holdup period for gaseous iodine, allowing time for decay.
In addition to the safety related standby emergency filtration function, parts of the CRFA System are operated to maintain he control room environment during normal to prsonel-5he
ý*Pcur 4 t-co r
CRFA System 4-the isolation mode of operation to preven in tr ion of contaminated air into the control room. A system of dampers isolates the control room, and control room air flow is recirculated and processed through either of the two filter subsystems. I,5-T The CRFA System is designed to mai ain the control room environment for a 30 day contin s occupancy after a DBA, per the requirements of GDC 19. CRFA System operation in maintaining the control room habitability is discussed in the USAR, Sections 6.4.1 and 9.4.1 (Refs. 1 and 2, respectively).
APPLICABLE The ability of the CRFA System to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the USAR, Chapters 6 (continued)
Revision No. 0 RIVER BENDB B 3.7-10
CRFA System B 3.7.2 BASES ff i APPLICABLE and 15 (Refs. 3 and 4, respectivKy). The isolation mode SAFETY ANALYSES of the CRFA System is assum~edto operatedfollowin~. a loss (continued) of colog.o t accident, M "0 M r
, @ and control rod drop iEciden The radiological doses to control room personnel as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room.
The CRFA System satisfies Criterion 3 of the NRC Policy Statement.
LCO Two redundant subsystems of the CRFA System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.
Total system failure could result in a failure to meet the dose requirements of GDC 19 in t ve The CRFA System is considered when the individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:
~~a. Fan is OPERABLE;~~
~~b. HEPA filter and charcoal adsorber are not excessively restricting flow and are capable of performing their filtration functions; and~~
~~c. Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.~~
In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors.
APPLICABILITY In MODES 1, 2, and 3, the CRFA System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.
In MODES 4 and 5, the probability and consequences of a DBA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the CRFA System (continued)
Revision No. 0 RIVER BEND. 8 3.7-11
CRFA System B 3.7.2 BASES APPLICABILITY_ OPERABLE is not required in MODE 4 or 5. except for the (continued) following situations under which significant radioactive releases can be postulated:
~~a. During operations with a potential for draining the reactor vessel (OPDRVs) n~~
b. During movement of irradiated fiiel assemblles-rfLfl ACTIONS A.1 With one CRFA subsystem inoperable, the inoperable CRFA subsystem must be restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLE CRFA subsystem is adequate to perform control room radiation protection. However.
the cverall reliaoillty is recuced because a single failure in the OPERABLE subsystem could result in loss of CRFA System function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the required capabilities.
B.1 and B.2 In MODE 1. 2. or 3, if the inoperable CRFA subsystem cannot be restored to OPERABLE status within the associated Completion Time. the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience.
to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
C.l.,--and C.2 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not appl . 9 ctu (continued)
RIVER BEND B 3.7-12 Revision No. 6-13
CRFA System B 3.7.2 BASES ACTIONS C.1 .4and C.2(2 continued) fue mveentiýi i~iiie e~d'!!1'i'0 :f' r eatons. Ther1*fore.
Anabllternatveto o tofuel aismsediat Reu iren d
acti present a ActionfC en f is ot es is i
ncient tht requ n to require cnro tor sh td tun in a con diatio assemblties tm pririk cntainmenor fuel buildin orsuspnded.
IRFA subo Fstem cannot bems astored i o PEri LE status 2 o n the requirem Completion Ti e. the OPERABLEorFA s ubsystedmay be func in ng the ssytgiemergeu y OPERABLE.
mode. This t tion no f ensuresideao ocuand turethatw theaie t any a failre wl redly det eF d.e An alternative OPR~t activities that topresent inmieth Required Action C.r is vsslrrindown poabliyofad a potential to immediately suspend for releasing rand radioactivity that m;ght require isolat.ion of t.he control room. This, places the unit in a condition that minimizes risk, OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
D.E If both CRFA subsystems are inoperable in MODE 1. 2. or 3. the CRFA System may not be capable of performing the intended' function and the unit is in a condition outside of the accident analyses. Therefore. LCO 3.0.3 must be entered immediately.
(conti nued )
RIVER BEND B 3.7-13 Revision No. 6-13 I
CRFA System B 3.7.2 BASES SACTIONS E.1 and E.2 (continued) Durng emnt ira atd AA assembl-t* te mary conta' Iment or fuel buildrigRi. it w CRFA ,
sua stems perable, am c on must biaken imm t susy te subsequent potentialal ff releasing dand fm the cntr requi n isolationu the contr room.
thrdougaces HE theA in a c hritioa thad inimizes rS b
ý ria ryaplicabl mo°vement eirradiatn, fuelf ass lies in ti1e PRspn to minimize the pnrobblt f esldridw n n painmenr con aste buil must n d normaperatin conditi of ths son of th eactivere, shatin ea c ub e hpp eat e actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPcRVs are suspended.
shoul be chceseroialytsnurphttey trnde SURVEILLANCE theonc3.7.3.
hedarcoaly BSR F Ant frmSumidit i nadqut th vminir ytmith REQUIREMENTS evr'ont provie ch eon thissysem This SR verifies that a subsystem in a standby mode starts on demand from the control room and continues to operate with flow through the HEPA filters and charcoal adsorbers. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe. testing each subsystem once every month provides an adequate check on this system.
Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. Systems with heaters must be operated for ý!10 continuous hours with the heaters energized to demonstrate the function of the system.
Furthermore. the 31 day Frequency is based on the known reliability of the equipment and the two subsystem redundancy available.
(continued)
RIVER BEND B 3.7-14 Revision No. 6-13
CRFA System B 3.7.2 BASES (continued)
REFERENCES 1. USAR, Section 6.4.1.
~~2. USAR, Section 9.4.1.~~
~~3. USAR, Chapter 6.~~
~~4. USAR, Chapter 15.~~
~~5. Regulatory Guide 1.52, Revision 2, March 1978.~~
/. V =CFR5 ý-
RIVER BEND B 3.7-16 Revision No. 0
Control Room AC System B 3.7.3 BASES (continued)
LCO Two independent and redundant subsystems of the Control Room AC System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits.
The Control Room AC System is considered OPERABLE when the individual components necessary to maintain the control room temperature are OPERABLE in both subsystems. These components include the cooling coils, fans. chillers, compressors, ductwork.
dampers. and associated instrumentation and controls.
APPLICABILITY In MODE 1. 2. or 3. the Control Room AC System must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits.
In MODES 4 and 5. the probability and consequences of a Design Basis Accident are reduced due to the pressure ard t limitations in these MODES. Therefore, maintaining Lne Control Room AC System OPERABLE is not required in MODE 4 or 5. except for the following situations under which significant radioactive releases can be postulated:
~~a. During operations with a potential for draining the reactor vessel (OPDRVs) and:~~
~~b. During mevcment of irradiated fuel semleinthe pr-imaryV containlment or fuel building.~~
ACTIONS A.1 With one control room AC subsystem inoperable, the inoperable control room AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE control room AC subsystem is adequate to perform the control room air conditioning function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of the control room air conditioning (continued)
RIVER BEND B 3.7-18 Revision No. 6-13
AC Sources - Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."
APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 and 5 SAFETY ANALYSE .E,,H;.R.. 9....
fW9 13.0 sA~ t hat:
nures
~~a. The unit can be maintained in the shutdown or refueling condition for extended periods;~~
~~b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and~~
~~c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel :~~
In general, when the unit is shut down the Technical Specifications (TS) requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LCOs for required systems.
During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS.
This allowance is in recognition that (continued)
RIVER BEND B 3.8-34A Revision No. 0
AC Sources - Shutdown B 3.8.2 BASES APPLICABLE certain testing and maintenance activities must be conducted provided an SAFETY ANALYSES acceptable level of risk is not exceeded. During MODES 4 and 5, (continued) performance of a significant number of required testing and maintenance activities is also required. In MODES 4 and 5, the activities are generally planned and administratively controlled. Relaxations from typical MODE 1, 2, and 3 LCO requirements are acceptable during shutdown MODES based on:
~~a. The fact that time in an outage is limited. This is a risk prudent goal as well as utility economic consideration.~~
b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operating MODE analyses, or both.
~~c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.~~
d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODE 1, 2, and 3 OPERABILITY requirements) with systems assumed to function during an event.
In the event of an accident during shutdown, this LCO ensures the capability of supporting systems necessary to avoid immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite (diesel generator (DG)) power.
The AC sources satisfy Criterion 3 of the NRC Policy Statement.
LCO One offsite circuit supplying onsite Class 1E power distribution subsystem(s) of LCO 3.8.8, "Distribution Systems-Shutdown," ensures that all required loads are powered from offsite power. An OPERABLE DG, associated with a Division I or Division II Distribution System Engineered Safety Feature (ESF) bus required OPERABLE by LCO 3.8.10, ensures a diverse power source is available to provide (continued)
RIVER BEND B 3.8-35 Revision No. 0
AC Sources - Shutdown B 3.8.2 BASES LCO electrical power support, assuming a loss of the offsite circuit. Similarly, (continued) when the high pressure core spray (HPCS) is required to be OPERABLE, a separate offsite circuit to the Division III Class 1E onsite electrical power distribution subsystem, or an OPERABLE Division II DG, ensure an additional source of power for the HPCS. This additional source for Division III is not necessarily required to be connected to be OPERABLE.
Either the circuit required by LCO Item a, or a circuit required to meet LCO Item c may be connected, with the second source available for connection. Together, OPERABILITY of the required offsite circuit(s) and DG(s) ensure the availability of sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g.,(4 reactor vessel draindown).
The qualified offsite circuit(s) must be capable of maintaining rated frequency and voltage while connected to their respective ESF bus(es),
and accepting required loads during an accident. Qualified offsite circuits are those that are described in the USAR and are part of the licensing basis for the plant. The offsite circuit consists of incoming breaker and disconnect to the respective preferred station service transformers 1C and 1D, the 1 C and 1D preferred station service transformers, and the respective circuit path including feeder breakers to all 4.16 kV ESF buses required by LCO 3.8.10.
The required DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 10 seconds for DG 1A and DG 1B and 13 seconds for DG 1C. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as: DG in standby with the engine hot and DG in standby with the engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillance, e.g., capability of the DG to revert to standby status on an ECCS signal while operating in parallel test mode.
Proper sequencing of loads, including tripping of (continued)
RIVER BEND B 3.8-36 Revision No. 0
AC Sources - Shutdown B 3.8.2 BASES LCO nonessential loads, is a required function for DG OPERABILITY. In (continued) addition, proper load sequence operation is an integral part of offsite circuit and DG OPERABILITY since its inoperability impacts the ability to start and maintain energized any loads required OPERABLE by LCO 3.8.10.
It is acceptable for divisions to be cross tied during shutdown conditions, permitting a single offsite power circuit to supply all required AC electrical power distribution subsystems.
As described in Applicable Safety Analyses, in the event of an accident during shutdown, the TS are designed to maintain the plant in a condition such that, even with a single failure, the plant will not be in immediate difficulty.
APPLICABILITY The AC sources required to be OPERABLE in MODES 4 and 5 ai, d 9r ding provide assurance that:
a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel; 0 ,/ Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.
The AC power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.1.
ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not to me...1 S- hutri
,*q ,,61'. .. . . .f'- ..
i~r~radated el. . .
as~sem w ier r- .. ... et. ,,continued) sn.DEVrmV
,2 ,,VQ (continued)
RIVER BEND B 3.8-37 Revision No. 0
AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.1 (continued)
An offsite circuit is considered inoperable if it is not available to one required ESF division. If two or more ESF 4.16 kV buses are required per LCO 3.8.10, division(s) with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS ,andNe operations with a potential for draining the reactor vese. By the allowance of the option to declare required features inoperable which are not powered from offsite power, appropriate restrictions can be implemented in accordance with the required feature(s) LCOs' ACTIONS. Required features remaining powered from offsite power (even though that circuit may be inoperable due to failing to power other features) are not declared inoperable by this Required Action.
A.2.1, A.2.2, A.2.3,ýQ .. 1, B3.2,2a*
With the offsite circuit not available to all required divisions, the option still exists to declare all required features inoperable. Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With the required DG inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, activities that could potentially result in inadvertent draining of the reactor vessel.
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize probability of the occurrence of postulated events. It is further required to initiate action immediately to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to (continued)
RIVER BEND B 3.8-38 Revision No. 0
AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.2.1, A.2.2, A.2.3 . B.1, B.2, B.3a (continued) minimize the time during which the plant safety systems may be without sufficient power.
Pursuant to LCO 3.0.6, the Distribution System ACTIONS are not entered even if all AC sources to it are inoperable, resulting in de-energization.
Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESF bus, ACTIONS for LCO 3.8.10 must be immediately entered. This Note allows Condition A to provide requirements for the loss of the offsite circuit whether or not a division is de-energized.
LCO 3.8.10 provides the appropriate restrictions for the situation involving a de-energized division.
C.1 When the HPCS is required to be OPERABLE, and the additional required Division III AC source is inoperable, the required diversity of AC power sources to the HPCS is not available. Since these sources only affect the HPCS, the HPCS is declared inoperable and the Required Actions of the affected Emergency Core Cooling Systems LCO entered.
In the event all sources of power to Division III are lost, Condition A will also be entered and direct that the ACTIONS of LCO 3.8.10 be taken. If only the Division III additional required AC source is inoperable, and power is still supplied to HPCS, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the additional required AC source to OPERABLE. This is reasonable considering HPCS will still perform its function, absent an additional single failure.
SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LCO 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, and 3. SR 3.8.1.8 is not required to be met since only one offsite circuit is required to be OPERABLE. SR 3.8.1.17 is not required to be met because the required OPERABLE DG(s) is not required to undergo periods (continued)
RIVER BEND B 3.8-39 Revision No. 102
AC Sources- Shutdown B 3.8.2 BASES SURVEILLANCE SR 3.8.2.1 (continued)
REQUIREMENTS of being synchronized to the offsite circuit. SR 3.8.1.20 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.
This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during the performance of SRs, and preclude de-energizing a required 4.16 KV ESF bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources available, a single event could compromise both the required circuit and the DG. It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG is required to be OPERABLE.
REFERENCES None.
RIVER BEND B 3.8-40 Revision No. 102
DC Sources - Shutdown B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources-Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LCO 3.8.4, "DC Sources- Operating."
APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY ANALYSES the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control and switching during all MODES of operation.
The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.
The OPERABILITY of the minimum DC electricalted power sources during MODES 4 and 55*oRdrdi_*F*9 f
~~a. The facility can be maintained in the shutdown or refueling condition for extended periods;~~
~~b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and~~
c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel The DC sources satisfy Criterion 3 of the NRC Policy Statement.
LCO One DC electrical power subsystem consisting of one battery, one battery charger, and the corresponding control equipment and interconnecting cabling supplying power to the associated bus within the division, associated with Division (continued)
RIVER BEND B 3.8-59 Revision No. 0
DC Sources - Shutdown B 3.8.5 BASES LCO I or II onsite Class 1 E DC electrical power distribution subsystem(s)
(continued) required by LCO 3.8.10, "Distribution Systems-Shutdown" is required to be OPERABLE. Similarly, when the High Pressure Core Spray (HPCS) system is required to be OPERABLE, the Division III DC electrical power subsystem associated with the Division III onsite Class 1E DC electrical power distribution subsystem required to be OPERABLE by LCO 3.8.10 is required to be OPERABLE. In addition to the preceding subsystems required to be OPERABLE, a Class 1 E battery or battery charger and the associated control equipment and interconnecting cabling capable of supplying power to the remaining Division I or II onsite Class 1E DC electrical power distribution subsystem(s), when portions of both Division I and II DC electrical power distribution subsystems are required to be OPERABLE by LCO 3.8.10. This ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g.,
f inadvertent reactor vessel draindown).
APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4
<I -t_*.nAýt" - @ r' I .*-rfro'vidie assurance* that:
!.L_b,
~~a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;~~
~~b. ogurodfooturco noeded te m~itigate a ftiel han1dling meeident ar~~
, *. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and
/* Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.
The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4.
(continued)
RIVER BEND B 3.8-60 Revision No. 0
DC Sources - Shutdown B 3.8.5 BASES (continued)
ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does no A.1, A.2.1, A.2.2, .. ,an If more than one DC distribution subsystem is required according to LCO 3.8.10, the DC subsystems remaining OPERABLE with one or more DC power sources inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS,E noKn and operations with a potential for draining the reactor vessel.
By allowing the option to declare required features inoperable with associated DC power source(s) inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. In many instances this option may involve undesired administrative efforts.
Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS RP and any activities that could result in inavertent draining of the reactor vessel).
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.
SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see (continued)
RIVER BEND B 3.8-61 Revision No. 0
DC Sources - Shutdown B 3.8.5 BASES SURVEILLANCE SR 3.8.5.1 (continued)
REQUIREMENTS the corresponding Bases for LCO 3.8.4 for a discussion of each SR.
This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DC sources from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required.
REFERENCES 1. USAR, Chapter 6.
~~2. USAR, Chapter 15.~~
RIVER BEND B 3-8-62 Revision No. 0
Inverters - Shutdown B 3.8.8 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Inverters-Shutdown BASES BACKGROUND A description of the inverters is provided in the Bases for LCO 3.8.7, "Inverters - Operating."
APPLICABLE The initial conditions of Design Basis Accident (DBA) and transient SAFETY ANALYSES accident analyses in the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature systems are OPERABLE.
The DC to AC inverters are designed to provide the required capacity, capability, redundancy, and reliability to ensure the availability of necessary power to portions of the ESF instrumentation and controls so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.
The OPERABILITY of the inverters is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.
The OPERABILITY of the minimum inverters to each AC vital bus during
~~a. The facility can be maintained in the shutdown or refueling condition for extended periods;~~
~~b. Sufficient instrumentation and control capability are available for monitoring and maintaining the unit status; and~~
c. Adequate power is available to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel, The inverters were previously identified as part of the Distribution System and, as such, satisfy Criterion 3 of the NRC Policy Statement.
(continued)
RIVER BEND B 3.8-74 Revision No. 6-1
Inverters - Shutdown B 3.8.8 BASES (continued)
LCO The inverters ensure the availability of electrical power for the instrumentation for systems required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or postulated DBA. The battery powered inverters provide uninterruptible supply of AC electrical power to the AC vital buses even if the 4.16 kV safety buses are de-energized. OPERABLE inverters require the associated AC vital bus be powered by the inverter through inverted DC voltage from the required Class 1 E battery, or from an internal AC source via a rectifier with the battery available as backup. This ensures the availability of sufficient inverter power sources to operate the plant in a safe manner and to miti t the consequences of postulated events during shutdown (e.g., '.dc. . -5 inadvertent reactor vessel draindown).
APPLICABILITY The inverters required to be OPERABLE in MODES 4 and 5 aisey fue buil-ag prvide assurance that:
a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel; b.sctc ,Rmcnoded to mFitigato a fuel han*di*n accid*e*nRtr Available
/. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and monitoring C /e. Instrumentation and control capability is available for or refueling and maintaining the unit in a cold shutdown condition condition.
Inverter requirements for MODES 1, 2, and 3 are covered in LCO 3.8.7.
ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not appEly. , 0~-tF I,oer..er; . . . . Y.. = 1l 2'e"*(on 4..................-
Ile'n* s!deed of reactoroe-tes hrfr,*"t (continued)
RIVER BEND B 3.8-75 Revision No. 0
Inverters - Shutdown B 3.8.8 BASES ACTIONS A.1, A.2.1, A.2.2. A.2.3 and (continued)
Iftwo divisions are required by LCO 3.8.10, "Distribution Systems-Shutdown," the remaining OPERABLE inverters may be capable of supporting sufficfeature(s) to allow continuation of CORE ALTERATIONS1m isand operations with a potential for draining the reactor vessel--gyt-Re a owance of the option to declare required feature(s) inoperable with the associated inverter(s) inoperable, appropriate restrictions are implemented in accordance with the affected required feature(s) of the LCOs' ACTIONS. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspenjd..
CORE ALTEPATIONS *-mn f!rqitdtJ' _
/.*, , r*,,., ,,i aln d an activities that could result in inadvertent draining of the reactor vessel).
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters and to continue this action until restoration is accomplished in order to provide the necessary inverter power to the plant safety systems.
Notwithstanding performance of the above conservative Required Actions, the unit is still without sufficient AC vital power sources to operate in a safe manner. Therefore, action must be initiated to restore the minimum required AC vital power sources and continue until the LCO requirements are restored.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power or powered from a constant voltage source transformer.
SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers (continued)
RIVER BEND B 3.8-76 Revision No. 0
Inverters - Shutdown B 3.8.8 BASES SURVEILLANCE SR 3.8.8.1 (continued)
REQUIREMENTS closed and AC vital buses energized from the inverter. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the AC vital buses.
The 7 day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions.
REFERENCES 1. USAR, Chapter 6.
~~2. USAR, Chapter 15.~~
RIVER BEND B 3.8-77 Revision No. 0
Distribution Systems-Shutdown B 3.8.10 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.10 Distribution Systems-Shutdown BASES BACKGROUND A description of the AC, DC, and AC vital bus electrical power distribution systems is provided in the Bases for LCO 3.8.9, "Distribution Systems-Operating."
APPLICABLE The initial conditions of Design Basis Accident and transient analyses in SAFETY ANALYSES the USAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC, DC, and AC vital bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded.
The OPERABILITY of the AC, DC, and AC vital bus electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY.
The OPERABILITY of the minimum AC, DC, and AC vital bus electrical power sources and associated power distribution subsystems during
( i- A;;,,,,R*R@R Q.. f . .. .. . ensures.. .. t.ha..
~~a. The facility can be maintained in the shutdown or refueling condition for extended periods;~~
~~b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and~~
c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel~f/ 7i The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement.
(continued)
RIVER BEND B 3.8-89 Revision No. 0
Distribution Systems-Shutdown B 3.8.10 BASES (continued)
LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LCO explicitly requires energization of the portions of the electrical distribution system necessary to support OPERABILITY of Technical Specifications' required systems, equipment, and components - both specifically addressed by their own LCOs, and implicitly required by the definition of OPERABILITY.
Maintaining these portions of the distribution system energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,
dtinadvertent reactor vessel draindown).
APPLICABILITY The AC, DC, and AC vital bus electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the primary containment or fuel building provide assurance that:
~~a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;~~
~~b. Syat-9nee~cded to mnitigate a fuel handling@ occident aro available;:~~
c' Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and C ,d- Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown or refueling condition.
The AC, DC, and AC vital bus electrical power distribution subsystem requirements for MODES 1, 2, and 3 are covered in LCO 3.8.9.
(continued)
RIVER BEND B 3.8-90 Revision No. 0
Distribution Systems-Shutdown B 3.8.10 BASES (continued)
ACTIONS ap lyACTIONS The M8 mH 11"', modified ln are rd14c by a Note indicating A-,.ssc'm,,,4.- _ wi',l E LCO 3.0.3 does not R_ that vemnt s ndeenent of reactor-operations", in topfbo
-,pd movement of irradiated fujel n.s*mhlle;is otsufiint reason- p_
\ to require reactor shu-tdown.
A.1, A.2.1, A.2.2, A.2.3,.24, and.
Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS ind operations with a potential for draining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend t',asmbc nt, CORE ALTERATIONS*t f .*,n
~~, *,n~~
~~,, *n,* .,..q ,,*,nrand any activities that could result~~
_ýý ýýh reac-or vessel.
in inadvertent ddrainingg oof the reac or vessel).
Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.
Notwithstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR-SDC) subsystem may be inoperable. In this case, Required Actions A.2.1 through A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the RHR-SDC ACTIONS would not be entered. Therefore, Required Action A.2.5 is provided to direct declaring RHR-SDC inoperable, which results in taking the appropriate RHR-SDC ACTIONS.
(continued)
RIVER BEND B 3.8-91 Revision No. 0
Distribution Systems-Shutdown B 3.8.10 BASES ACTIONS A.1, A.2.1, A.2.2, A.2.3, A.2.4, ancontinued)
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.
SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the required AC, DC, and AC vital bus electrical power distribution subsystems are functioning properly, with the buses energized. The verification of proper voltage availability on the required buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The 7 day Frequency takes into account the redundant capability of the electrical power distribution subsystems, as well as other indications available in the control room that alert the operator to subsystem malfunctions.
REFERENCES 1. USAR, Chapter 6.
~~2. USAR, Chapter 15.~~
RIVER BEND B 3.8-92 Revision No. 0
Attachment 4 Letter RBG-45930 List of Regulatory Commitments Letter RBG-45930 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check one) SCHEDULED ONE- CONTINUING COMPLETION COMMITMENT TIME COMPLIANCE DATE (If ACTION Required)
Technical Requirements Manual TLCO 3.3.7.1 X provides the requirements for the remote air intake monitors. This TLCO will also be applied to the main air intake monitors since the main air intake monitors have the same licensing function as the remote air intake monitors, i.e., provide indication to operators so that they can select the more favorable intake.
Attachment 5 Letter RBG-45930 POST-LOCA SUPPRESSION POOL pH EVALUATION
~~SUMMARY~~
~~Letter RBG-45930 Page 1 of 3 POST-LOCA SUPPRESSION POOL pH EVALUATION~~
~~SUMMARY~~
SCOPE The NUREG-1465 accident isotopic release specification allows deposition of iodine in the suppression pool. The iodine is assumed to remain in solution as long as the pool pH is maintained above 7. Upon detection of symptoms indicating that core damage is occurring, RBS procedures shift over to severe accident scenarios. Operators are directed to manually initiate the Standby Liquid Control System (SLCS) upon initiation of severe accident procedures.
No credit is taken for any operator action during the first 10 minutes of an event. If an accident were to occur which would create the fuel damage conditions assumed in the analyses, it is reasonable to assume that manual initiation of SLCS injection would be initiated promptly. For the purposes of the analysis, however, this action is conservatively delayed by releasing the system's 1657 gallon inventory of sodium pentaborate solution into the RPV 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the postulated accident. The buffering effect of the SLCS solution is shown to maintain the suppression pool pH above 7 for the 30-day duration of the postulated LOCA and prevent iodine re-evolution.
INPUTS AND ASSUMPTIONS The inputs and assumptions used in the evaluation of the post-LOCA suppression pool pH are listed in Tables 1. These inputs are consistent with the requirements of RG 1.183 and the approach in NUREG/CR-5950, "Iodine Evolution and pH Control." The analysis conservatively maximizes the acid contributions from Hydroiodic Acid from core halogens, Nitric Acid from radiolysis of water, and Hydrochloric Acid from radiolysis of chloride bearing cables inside containment while not crediting the core Cesium Hydroxide.
RESULTS The results of the RBS post-LOCA suppression pool pH evaluation, assuming full injection, are provided in Table 2. The results show that the suppression pool pH will be maintained above 7.0 throughout the duration of the accident, thus, preventing re-evolution of elemental iodine dissolved in the pool water.
A parametric study was also performed. The results of the parametric study are presented in Figure 1. The study demonstrates that only 30% of the sodium pentaborate is required to reach the suppression pool to ensure that the pH remains above 7.0.
Letter RBG-45930 Page 2 of 3 TABLE 1 RBS POST-LOCA SUPPRESSION POOL pH EVALUATION PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption
~~1. Suppression Pool Volume 1,045,560 gallons~~
~~2. Minimum Initial pH value 5.3~~
~~3. Minimum SLCS injection rate 41.2 gpm~~
~~4. Minimum SLCS available volume 1657 gallons~~
~~5. Minimum sodium pentaborate weight percent 7.13%~~
~~in SLCS solution~~
6. Start time for SLCS injection Analysis assumes total inventory released at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This value bounds initiation within the first hour and -40 minutes of injection.
~~7. Suppression Pool Mixing The suppression pool is assumed to be well mixed.~~
~~8. Core Thermal Power 3100 MWt~~
~~9. Core Activity Release 25% alkalis (5% gap, 20% early in vessel release) 30% iodines (5% gap, 25% early in vessel release)~~
~~10. Total Exposed Chloride Bearing Cable Mass 38007 Ibm~~
~~11. HCI production from cables Per Appendix B of NUREG/CR-5950.~~
~~Beta dose to cables are based on current TID dose rates. Beta doses to cables in trays reduced by a factor of 2 per NUREG-0588, Section 1.4~~
12. CsOH No credit for CsOH reaching the Suppression Pool Letter RBG-45930 Page 3 of 3 TABLE 2 RBS POST-LOCA SUPPRESSION POOL pH EVALUATION Time(hrs) HI (moles) HNO 3 (moles) HCI (moles) pH 2 0.76 43.3 58.3 8.59 4 0.78 67.9 93.4 8.57 8 0.78 100.2 143.1 8.54 16 0.78 136.8 206.3 8.50 24 0.78 158.0 247.3 8.48 48 0.78 221.5 346.4 8.42 72 0.78 259.5 419.6 8.38 120 0.78 309.3 526.2 8.33 168 0.78 343.2 602.0 8.29 240 0.78 379.9 682.9 8.25 480 0.78 453.4 827.7 8.17 720 0.78 497.3 898.9 8.13 pH as a Function of Standby Liquid Control Core Injection Volume 8.40 . . .. . . . . . . . . . .
8.20 8-00 7.80 S7.760 7.40 7.20 7.00 6.80 300
....i ........I........Iý.... . . . .
700 900 1100 1300 1500 1700 Standby Liquid Control Core Injection (Gallons)
Figure 1
Attachment 6 Letter RBG-45930 Atmospheric Dispersion Factors ( /Q) Calculations Letter RBG-45930 Page 1 of 10 Atmospheric Dispersion Factors (,/Q) Calculations In support of the Alternate Source Term project at RBS the Atmospheric Dispersion Factors (T/Q) were reviewed. The X/Q values currently used in RBS accident dose analyses are based on methodologies from Murphy and Campe's, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19," for the Main Control Room and Regulatory Guide 1.145, Revision 1 for off-site locations.
Also, the data for the X/Q values currently used was taken between 3/17/77 and 3/16/79 (USAR Section 2.3). The X/Qs were recalculated using current NRC codes and methodologies. Specifically, the NRC codes ARCON96 and PAVAN were used to recalculate the x/Q values for the Main Control Room (MCR) and off-site locations, respectively. The meteorological data used (1/95 through 12/00) is more recent than that used in the original analysis.
Main Control Room Calculations The Main Control Room x/Q values were originally determined using the methodology found in Murphy and Campe and Regulatory Guide 1.145, Revision 1. The recalculated values were generated using the computer code ARCON96 as described in NUREG/CR-6331, Atmospheric Relative Concentrationsin Building Wakes, Revision 1.
A number of input assumptions are consistent with DG-1 111, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," Table A-1. A summary of input data and assumptions may be found in Table 1.
Atmospheric stability was classified according to the temperature gradient values listed for the seven Pasquill stability categories in Regulatory Guide 1.23. Environmental data was obtained from the permanently installed meteorological tower and associated equipment. Joint wind speed, wind direction, atmospheric stability summaries, based on wind speed and wind direction at the 30-ft tower level, and the temperature difference between the 30-ft and 150-ft levels were obtained by the plant's permanently installed meteorological tower. Five years' worth of meteorological data was used in the evaluation. This data was taken from two separate time periods 1 : January 1, 1995 through December 31, 1998; and January 1, 2000 through December 31, 2000. Use of these dates provided over 41,400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months of meteorological data for the 5 year period which exceeds the minimum requirement of 90%.
1 The hourly average data required by ARCON96 could not be recovered for 1999 due to the failure of the tapes containing the data. This was not of concern for the off-site calculations since PAVAN uses the joint frequency tables which were available elsewhere. Two separate time periods were chosen to ensure that the most recent meteorological data was used in the evaluation.
Letter RBG-45930 Page 2 of 10 The X/Q values were calculated for both the main Control Room main air intake and the remote air intake. The main intake is located on the east side of the roof of the control building and the remote intake is located on the west side of the standby cooling tower.
Six release points were considered. Each release path was modeled as a ground level point source with zero vent velocity and zero vent flow. The bounding release points were evaluated:
"* the Main Plant Stack (Standby Gas Treatment System),
"* the Main Steam Blowout Panel,
"* the Containment Building,
"* the Turbine Building,
"* the Radwaste Building, and
"* the Fuel Building.
The results of the analyses may be viewed in Table 2 below. Review indicates that the x/Q calculated using ARCON96 and the 1995-2000 data yields much more favorable results that the previous evaluations. For example, the 0-2 hour Containment/SGTS X/Q values were calculated to be roughly 50% and 20% of the current values for the main and remote intakes, respectively. Section 3.2 of NUREG/CR-6331, Revision 1 discusses the differences between the Murphy-Campe model and ARCON96 modeling. The NUREG states that "the Murphy-Campe model did not predict the variations of the concentrations in the vicinity of buildings particularly well. The studies also showed that one of the primary reasons that the Murphy-Campe modes did not predict concentration well was that it over-predicted concentrations during low speed conditions." The Murphy Campe evaluations for RBS used a 1.3 mph (0.58 m/s) for the containment release for the 0-8 hour time period. Figure 27 from the NUREG shows Murphy-Campe/ARCON concentrations by wind speed. The figure implies that for 0.6 m/s the ratio can range between 20 to 200. The ratio between values calculated for RBS is on the order of 2 - 5.
Further comparison between the Murphy-Campe and ARCON methodology indicated that several key assumptions were significantly different. For example, the Murphy Campe evaluations assumed a diffuse release whereas the ARCON96 evaluations assumed a point source. Also, the occupancy factors for the MCR are inherent in the Murphy-Campe values. Therefore, based on this review it was concluded that the results calculated by ARCON96 were reasonable and appropriate for use at RBS.
Letter RBG-45930 Page 3 of 10 Off-Site Calculations: Exclusion Area Boundary and Low Population Zone Atmospheric dispersion factors were calculated for two locations, the exclusion area boundary (EAB) and the low population zone (LPZ). The off-site evaluations used the computer code PAVAN as described in NUREG/CR-2858, An Atmospheric Dispersion Program for Evaluation Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations." The joint frequency tables for six years of data (1994 through 1999) were used in this evaluation. These tables were generated utilizing data obtained from the site meteorological equipment discussed above. The bounding release points were evaluated:
"* the Main Plant Stack (Standby Gas Treatment System),
"* the Main Steam Blowout Panel,
"* the Containment Building,
"* the Standby Cooling Tower,
"* the Turbine Building, and
"* the Radwaste Building.
The EAB is designated by a 3,000-ft (914 m) radius circle drawn about the reactor center. Although it is deemed "off-site," the EAB is entirely within RBS property, as shown in USAR Figure 2.1-2. The LPZ surrounding River Bend Station encompasses an area within a distance of 2.5 miles. The distance for the LPZ was chosen based on the requirements of 10CFR1 00.11. USAR Figure 2.1-8 shows roads and facilities within the LPZ.
Table 3 contains a summary of the meteorological data used in the off-site atmospheric dispersion factor analyses. The distances used in the analyses were based on the shortest distance from the release point to the EAB and LPZ. The values used are contained in Table 4 below. The top of the containment structure is 51.5 m above grade.
The PAVAN results for the intermediate time steps were verified by hand calculations.
The EAB values are found in Table 5, and the results of the LPZ calculations are presented in Table 6.
Letter RBG-45930 Page 4 of 10 Table 1 Main Control Room Atmospheric Dispersion Factors - ARCON96 Inputs Parameter Main Main Contain. Turbine Radwaste Fuel Comments Plant Steam Eq. Building Building Handling Stack Tunnel Hatch (worst Building Blowout point)
Panel Lower Meas. Height, m 9.1 9.1 9.1 9.1 9.1 9.1 Lower instrument is 30' above grade.
Upper Meas. Height, m 45.7 45.7 45.7 45.7 45.7 45.7 Upper instrument is 150' above grade.
Wind Speed Units m/s m/s m/s m/s m/s m/s Release Height - MAI, m 58.8 22.5 2.5 28.3 37.6 22.9 Release Height - RAI, m 58.8 22.5 2.5 28.3 37.6 3.0 Building Area - MAI, m' 2,121 1,006 2,121 909.5 2,121 838 Building Area - RAI, m2 2,121 911.5 2,121 911.5 911.5 838 Vertical Velocity, m/s 0 0 0 0 0 0 Point releases - set to 0 per DG-1 111, Table A-1 Stack Flow, ms/s 0 0 0 0 0 0 Flow not credited - Set to 0 per DG-1 111, I Table A-1 Stack Radius, m 0 0 0 0 0 0 Set to 0 per DG-1111, Table A-1 Distance to Main Intake, 61.9 61.7 56.7 42.8 100.0 67.4 See Figure 1 for release points (Note 1).
m Distance to Remote 118.2 151.2 119.7 136 96.3 72.7 See Figure 1 for release points (Note 1).
Intake, m Main Intake Height, m 18.0 18.0 18.0 18.0 18.0 18.0 Remote Intake Height, m 9.0 9.0 9.0 9.0 9.0 9.0 Elevation Difference 0.0 0.0 0.0 1.1 0.9 1.1 (MAI), m I- I Elevation Difference 1.1 1.1 1.1 0.0 0.2 0.0 (RAI), m I I Letter RBG-45930 Page 5 of 10 Parameter Main Main Contain. Turbine Radwaste Fuel Comments Plant Steam Eq. Building Building Handling Stack Tunnel Hatch (worst Building Blowout point)
Panel Direction to Source (MAI), 255 214 274 177 246 281 Direction to Source (RAI), 099 111 090 127 118 091 Surface Roughness 0.2 0.2 0.2 0.2 0.2 0.2 DG-1 111, Table A-1 Length, m Wind Direction Window, 90 90 90 90 90 90 DG-1 111, Table A-1 Min. Wind Speed, m/s 0.5 0.5 0.5 0.5 0.5 0.5 DG-1 111, Table A-1 Avg. Sector Width 4.3 4.3 4.3 4.3 4.3 4.3 DG-1 111, Table A-1 Constant Initial Diffusion 0 0 0 0 0 0 Point releases - set to 0 per DG-1 111, Coefficients, m Table A-1 Hours in Averages Default Default Default Default Default Default Default values used per DG-1 111, Table A Minimum Number of Default Default Default Default Default Default Default values used per DG-1 111, Table A Hours I IIII1 1 Note 1: The fuel building assumes releases through the FB ventilation system for the Main Air Intake and assumes a release through the truck bay doors for the remote air intake. Radwaste building releases are assumed through the building ventilation system.
Letter RBG-45930 Page 6 of 10 Table 2 Control Room 5% Probability Level /Q Values (Sec/m 3)
Time Standby Gas Treatment Main Steam Tunnel Blowout Containment Equipment Period System Panel Hatch Main Air Intake Remote Air Main Air Intake Remote Air Main Air Intake Remote Air Intake Intake Intake 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.08E-03 4.29E-04 1.42E-03 2.65E-04 1.21 E-03 3.44E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 7.65E-04 3.46E-04 1.08E-03 2.16E-04 7.46E-04 2.27E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.43E-04 1.37E-04 4.56E-04 9.19E-05 3.39E-04 9.62E-05 I to 4 days 2.47E-04 1.18E-04 3.50E-04 6.67E-05 2.65E-04 7.75E-05 4 to 30 days 2.18E-04 8.59E-05 2.58E-04 4.85E-05 2.20E-04 5.78E-05 Time Turbine Building Vent Radwaste Building Fuel Handling Building Door Period Main Air Intake Remote Air Main Air Intake Remote Air Main Air Intake Remote Air Intake Intake Intake 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.02E-03 3.69E-04 5.66E-04 6.63E-04 1.09E-03 8.56E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.47E-03 3.08E-04 4.27E-04 5.47E-04 6.83E-04 5.53E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.05E-03 1.30E-04 1.88E-04 2.23E-04 3.17E-04 2.41 E-04 1 to 4 days 9.01 E-04 1.07E-04 1.37E-04 1.82E-04 2.34E-04 1.89E-04 4 to 30 days 6.74E-04 7.15E-05 1.21 E-04 1.24E-04 1.95E-04 1.42E-04 Letter RBG-45930 Page 7 of 10 Table 3 Total Valid Hours in Stability Class YEAR CLASS CLASS CLASS CLASS CLASS E CLASS F CLASS TOTAL A B C D G 1994 330 545 380 3180 2540 851 774 8600 1995 381 626 400 2920 2375 881 958 8541 1996 523 814 423 2757 2367 907 866 8657 1997 545 708 415 2679 2053 765 888 8053 1998 189 347 293 3922 2192 777 852 8572 1999 474 575 368 2849 2367 927 1116 8676 Table 4 Off-Site - Minimum Distances and Sector LOCATION SECTOR EAB (m) LPZ (m)
Turbine Building S 759 3868 Radwaste Building SSW 822 3931 Main Steam Tunnel S, ESE, SE, SSE 871 3980 Edge of Containment All 894 4003 SGTS (plant stack) SSE 914 4023 Standby Cooling Tower W 807 3916 Letter RBG-45930 Page 8 of 10 Table 5 Exclusion Area Boundary IQ Results Letter RBG-45930 Page 9 of 10 Table 6 Low Population Zone IQ Results Wind Frequency Distribution Wind Frequency Distribution Wind Frequency Distribution Turbine Building Radwaste Building Main Steam Tunnel Time 0.5% Sector/Dist 5% Site 0.5% Sector/Dist 5% Site 0.5% Sector/Dist 5% Site Period Max /Q Limit Max /Q Limit Max /Q Limit 0-8 Hrs 7.79E-05 SW / 3868 6.78E-05 7.67E-05 SW / 3931 6.74E-05 7.57E-05 SW / 3980 6.65E-05 8-24 Hrs 5.23E-05 Same 4.64E-05 5.14E-05 Same 4.60E-05 5.08E-05 Same 4.54E-05 1-4 Days 2.21E-05 Same 2.03E-05 2.16E-05 Same 2.01E-05 2.13E-05 Same 1.98E-05 4-30 Days 6.40E-06 WSW 6.23E-06 6.36E-06 WSW / 3931 6.12E-06 6.24E-06 WSW / 3980 6.01E-06 Notes:
(1) Units for relative concentration, /Q, values are in seconds per cubic meter (sec/m 3 ).
(2) 0.5% /Q values represent the maximum selected from among all sector-dependent values (as indicated). Distance units are in meters.
Wind Frequency Distribution Wind Frequency Distribution Wind Frequency Distribution SGTS (plant stack) Edge of Containment Standby Cooling Tower Time 0.5% Sector/Dist 5% Site 0.5% Sector/Dist 5% Site 0.5% Sector/Dist 5% Site Period Max /Q Limit Max /Q Limit Max /Q Limit 0-8 Hrs 7.49E-05 SW / 4023 6.59E-05 7.53E-05 SW / 4003 6.62E-05 7.69E-05 SW / 3919 6.76E-05 8-24 Hrs 5.02E-05 Same 4.49E-05 5.04E-05 Same 4.5 1E-05 5.17E-05 Same 4.62E-05 1-4 Days 2.10E-05 Same 1.95E-05 2.11E-05 SW, WSW 1.96E-05 2.17E-05 Same 2.02E-05 4-30 Days 6.13E-06 WSW/4023 5.91E-06 6.18E-06 WSW 5.95E-06 6.39E-06 WSW 6.16E-06 Notes:
3 (1) Units for relative concentration, /Q, values are in seconds per cubic meter (sec/m ).
(2) 0.5% /Q values represent the maximum selected from among all sector-dependent values (as indicated). Distance units are in meters.
Letter RBG-45930 Page 10 of 10 Figure 1 Main Control Room Atmospheric Dispersion Factor Release and Receptor Points 4N. INTAKE FUEL BLDG
[]
EXHAUST DUCTr CASK HANDLING DOORS CNDS STOR RADWASTE BLDG LEGEN TURBINE BUILDING RELEASE 91 POINT FOR MAIN AIR INTAKE TURBINE BUILDING RELEASE NOTE POINT FOR REMOTE AIR INTAKE THIS FIGURE IS INTENDED TO SHOW GENERAL BUILDING LAYOUT 1 FUEL BUILDING RELEASE AIR INTAKE LOCATION FOR REMOTE MAIN PLANT STACK (SGTS)
RELEASE LOCATION E MAIN STEAM PRESSURE TUNNELPANELS RELEASE
Attachment 7 Letter RBG-45930 Loss of Coolant Accident (LOCA) Dose Analysis Summary Letter RBG-45930 Page 1 of 26 Loss of Coolant Accident (LOCA) Dose Analysis Summary SCOPE The Loss of Coolant Accident (LOCA) is postulated to occur as a consequence of a double ended guillotine break of a recirculation line. The LOCA is assumed to occur concurrently with a Safe Shutdown Earthquake (SSE), a Loss of Off-site Power (LOP), and a Single Active Failure (SAF) of an Emergency Diesel Generator (EDG). Additionally, a Main Steam Isolation Valve (MSIV) is assumed to fail to close which represents a second SAF. Traditionally only one SAF is required, however, for some doses the EDG failure would be the bounding assumption and for others the MSIV failure is likely to be bounding. Both SAF were assumed to prevent the necessity of performing detailed sensitivity analyses for each receptor location.
The computer code RADTRAD version 3.02 was used to determine the dose consequences in all of the analyses summarized in this submittal. The radiological consequences of the LOCA event are determined for the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ),
and the Main Control Room (MCR). The calculated results are then evaluated against the acceptance criteria of 1 0CFR50.67.
INPUTS AND ASSUMPTIONS The LOCA analysis evaluates four release points:
" Containment is assumed to leak at the proposed technical specification limit of 0.325 volume percent per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The containment leakage rate (La) is reduced to 55% of that value at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on the power uprate containment pressure response analysis as allowed per Regulatory Guide 1.183, Appendix A, Section 3.7. Since this term is in volume % per day, this leakage is assumed for both the drywell and containment. Manual initiation of the Standby Gas Treatment System (SGTS) is assumed, therefore, secondary containment releases were not credited for the first 30 minutes of the event. During this time period leakage from the primary containment is released directly to the environment.
After 30 minutes containment leakage is directed to the annulus building which is treated by SGTS.
" The second contributor considered is Secondary Containment Bypass (SCB) leakage. SCB leakage is independent of La. SOB is assumed to leak at the proposed Technical Specification limit of 580,000 cc/hr (at Pa=7.6 psig) for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the event. Since containment pressure is the driving force for this leakage term, SCB is also reduced to 55%
of the original value after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on the containment pressure response. This leakage is assumed to originate from the containment and is released directly to the environment via the turbine building for the duration of the event.
Letter RBG-45930 Page 2 of 26
" The third contributor is 50 scfh through one main steam line in accordance with the proposed TS value. This leakage would only occur as a result of a failed MSIV. RBS currently has administrative limits of 30 scfh per MSIV and 10 scfh per drain line. The current TS limit is 150 scfh per division for all four main steam lines. The three remaining steam lines would not leak due to the fact that the pressure from the steam trapped between the MSIVs is significantly greater than the maximum containment pressure during an accident. This release is terminated 25 minutes into the event when the Main Steam Positive Leakage Control System (MSPLCS) becomes fully operational. This leakage is assumed to originate from the drywell and is released directly to the environment via the turbine building. Note that the current (i.e., TID) LOCA analysis neglects MSIV leakage in its entirety due to the trapped steam between the MSIVs and the fact that a failed MSIV is not assumed.
" The final contributor is the liquid leakage from Engineered Safety Features (ESF) cooling systems. Specifically, 1 gpm of suppression pool water is assumed to leak for the duration of the event. This leakage is assumed to be released directly to the environment for the first 30 minutes of the event. After 30 minutes the leakage is directed to the auxiliary building where it is treated by SGTS prior to release to the environment.
The fission product inventory used was based on that used in the power uprate submittal and is presented in Table 1. The release fractions assumed for each release phase are consistent with Regulatory Guide 1.183, Table 1. The release phases' start time and duration assumed are consistent with Regulatory Guide 1.183, Table 4. The dose conversion factors used are taken from Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion," and FGR 12, "External Exposure to Radionuclides in Air, Water and Soil."
The activity released from the fuel is assumed to mix instantaneously with the drywell volume.
Natural deposition of core halogens is credited in the drywell using the Powers (10%) model for BWRs from RADTRAD. Deposition for elemental iodine was also credited with a deposition coefficient of 1.01 hr-1. RBS does not have containment sprays, therefore, none were credited in the analysis. Also, reduction of airborne radioactivity in the containment by suppression pool scrubbing was not credited.
The transport mechanisms between the drywell and containment vary as the event progresses.
Early in the event the drywell has a significantly higher pressure than the containment due to the initial pipe break. However, roughly 10 minutes after the pipe break the drywell pressure drops below the containment pressure (See USAR Figure 6.2-5). The flow from the drywell to containment during this time period was calculated to be 4.74E+05 cfm based on an average drywell overpressure of 20 psid (from 2 - 10 minutes) and a flow area (AR/k) of 1.0 ft 2. After the drywell becomes "negative" no flow was modeled. Experience with the RBS model indicates that calculated doses are not sensitive to minor changes in this flow rate. The fuel damage postulated in the analysis requires a significant amount of core damage. This would also generate a significant amount of hydrogen from water-cladding reactions. The hydrogen mixing system is assumed to be initiated 25 minutes into the event (same start time as MS-PLCS). This would equalize the drywell and containment atmospheres. The flow rate for the hydrogen mixing system is 600 cfm. An additional 3000 cfm is assumed from the drywell to the containment to account for core steaming. This value is identical to that used by Grand Gulf and Perry which should bound RBS since both those BWR/6 plants are larger than RBS. Finally, just before 2 Letter RBG-45930 Page 3 of 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months a very high mixing rate between the drywell and containment is assumed. Regulatory Guide 1.183, Appendix A, Section 3.7 states "After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment."
The Control Room Fresh Air (CRFA) ESF charcoal filters are credited in the LOCA dose analysis. The system is assumed to operate in normal mode for the first 20 minutes of the event, at which time manual initiation of the system is assumed to occur. Breathing rates and occupancy factors are based on Regulatory Guide 1.183 guidance. The normal air intake is 2,000 cfm, however, an unfiltered inleakage of 300 cfm is conservatively assumed so the corrected flow intake flow rate is 1,700 cfm. A schematic of the actual and simulated CRFA system configuration is presented in Figures 1 and 2 for the normal and emergency modes, respectively. RBS has dual manual air intakes which meet the criteria set forth in Murphy and Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19." SRP 6.4.111.4.ii states that plants with dual inlet designs limited to manual selection may use the more favorable inlet location. Further the SRP allows the more favorable value to be reduced by a factor of 4. This credit will be taken for the normal air intake. The main air intake for the control room is located on the control building just above the control room envelope. The main air intake (MAI) X/Qs will be assumed to apply to the unfiltered inleakage since the MAI is located on the roof of the Control Building just above the MCR envelope. The LOCA analysis utilized a flow weighted average of the applicable X/Q values. These values are presented in Table 2.
Q 1700cfmi* Qot fav + 300cfiN(MAI)
Qeffective 0 0f 42000cfmn
~~where, X/Qmost-fav = More favorable X/Q value (main air intake or remote air intake),~~
X/QMA = X/Q value corresponding to the main air intake, and X/Qeff = The effective X/Q value used in the LOCA analysis.
Three separate RADTRAD input decks were required. One analyzed the dose consequences of primary containment leakage from containment and secondary containment. The postulated accident activity is released as a "ground level release" from the main plant stack (since the stack is not 2.5x the height of containment as would be required for a "stack" release). The second file evaluated the doses from MSIV and SCB leakage. Both of these leakage terms are assumed to originate from the Turbine Building. The third and final file evaluated the liquid ESF leakage which is also assumed to be released via the main plant stack.
Letter RBG-45930 Page 4 of 26 RESULTS The radiological consequences for the postulated LOCA event are summarized in Table 4. The LOCA event results in offsite and control room doses within the regulatory limits of 10CFR50.67.
One exception to Regulatory Guide 1.183 guidance was taken:
Engineered Safety Features Liquid Leakage: RG 1.183, Appendix A, Section 3.8 states that ESF leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specifications would require declaring such systems inoperable. The leakage rate assumed in this analysis is 1 gpm based on the initial licensing of the plant. Specifically, the RBS SER (NUREG 0989), Section 15.6.5, states that "The leak rate chosen for ESF equipment is that used by the staff in past evaluations; it represents the largest flow that could plausibly exist over the duration of the accident without discovery and isolation of this source." In addition to the initial licensing of the plant, this value was used in the Technical Specification Amendment 98, 113, and 114 submittals which were approved by the NRC.
It should be noted that the Vital Area Access (VAA) were reviewed using AST assumptions. The current values documented in the USAR were found to be bounding. Equipment Qualification (EQ) doses were not reevaluated as Regulatory Guide 1.183, Section 6, states that "licensees may continue to use either the AST or TID-14844 assumptions to perform the required EQ analyses." Design changes are evaluated by the River Bend EQ Program to ensure that all equipment will operate as required during postulated accidents.
Letter RBG-45930 Page 5 of 26 Table 1 BWR CORE INVENTORIES AST Isotope RADTRAD Power Uprate Assumed Inventory Group Inventory Inventory (Ci/MWt)
(CilMWt) (CiIMWt) 1 Kr-85 2.51 E+02 3.02E+02 3.02E+02 1 Kr-85m 9.11 E+03 6.73E+03 6.73E+03 1 Kr-87 1.66E+04 1.29E+04 1.29E+04 1 Kr-88 2.24E+04 1.83E+04 1.83E+04 1 Xe-133 5.43E+04 5.53E+04 5.53E+04 I Xe-135 1.29E+04 7.15E+03 7.15E+03 2 1-131 2.58E+04 2.63E+04 2.63E+04 2 1-132 3.79E+04 3.85E+04 3.85E+04 2 1-133 5.42E+04 5.50E+04 5.50E+04 2 1-134 5.93E+04 6.06E+04 6.06E+04 2 1-135 5.1OE+04 5.19E+04 5.19E+04 3 Rb-86 1.40E+01 4.70E+01 4.70E+01 3 Cs-I 34 4.23E+03 5.36E+03 5.36E+03 3 Cs-136 1.13E+03 1.18E+03 1.18E+03 3 Cs-1 37 2.53E+03 3.32E+03 3.32E+03 4 Sb-1 27 2.32E+03 2.28E+03 2.28E+03 4 Sb-1 29 8.07E+03 8.08E+03 8.08E+03 4 Te-127 2.25E+03 2.25E+03 2.25E+03 4 Te-127m 3.03E+02 3.41 E+02 3.41E+02 4 Te-129 7.57E+03 7.60E+03 7.60E+03 4 Te-129m 1.99E+03 2.06E+03 2.06E+03 4 Te-131 m 3.82E+03 3.73E+03 3.73E+03 4 Te-132 3.74E+04 3.79E+04 3.79E+04 5 Sr-89 2.77E+04 2.47E+04 2.47E+04 5 Sr-90 1.96E+03 2.58E+03 2.58E+03 5 Sr-91 3.60E+04 3.16E+04 3.16E+04 5 Sr-92 3.77E+04 3.37E+04 3.37E+04 6 Ba-1 39 4.99E+04 4.93E+04 4.93E+04 6 Ba-140 4.93E+04 4.75E+04 4.75E+04 Letter RBG-45930 Page 6 of 26 TABLE 1 - cont.
BWR CORE INVENTORIES AST Isotope RADTRAD Power Uprate Assumed Inventory Group Inventory Inventory (CilMWt)
(Ci/awt) (Ci/Mwt) 7 Co-58 1.53E+02 Not listed 1.53E+02 7 Co-60 1.83E+02 Not listed 1.83E+02 7 Mo-99 4.86E+04 5.01 E+04 5.01 E+04 7 Tc-99m 4.20E+04 4.32E+04 4.32E+04 7 Ru-1 03 3.69E+04 4.24E+04 4.24E+04 7 Ru-1 05 2.46E+04 2.99E+04 2.99E+04 7 Ru-106 1.00E+04 1.51 E+04 1.51 E+04 7 Rh-105 1.84E+04 2.52E+04 2.52E+04 8 Ce-141 4.47E+04 4.40E+04 4.40E+04 8 Ce-143 4.36E+04 4.15E+04 4.15E+04 8 Ce-144 2.90E+04 3.53E+04 3.53E+04 8 Np-239 5.68E+05 Not listed 5.68E+05 8 Pu-238 3.95E+01 Not listed 3.95E+01 8 Pu-239 1.00E+01 Not listed 1.OOE+01 8 Pu-240 1.25E+01 Not listed 1.25E+01 8 Pu-241 2.16E+03 Not listed 2.16E+03 9 Y-90 2.1OE+03 2.79E+03 2.79E+03 9 Y-91 3.39E+04 3.22E+04 3.22E+04 9 Y-92 3.78E+04 3.39E+04 3.39E+04 9 Y-93 4.30E+04 3.91E+04 3.91E+04 9 Zr-95 4.46E+04 4.42E+04 4.42E+04 9 Zr-97 4.59E+04 4.54E+04 4.54E+04 9 Nb-95 4.22E+04 4.42E+04 4.42E+04 9 La-140 5.03E+04 5.03E+04 5.03E+04 9 La-141 4.64E+04 4.44E+04 4.44E+04 9 La-142 4.47E+04 4.34E+04 4.34E+04 9 Pr-143 4.26E+04 4.11E+04 4.11E+04 9 Nd-147 1.91 E+04 1.81E+04 1.81E+04 9 Am-241 2.19E+00 Not listed 2.19E+00 9 Cm-242 5.79E+02 Not listed 5.79E+02 9 Cm-244 3.13E+01 Not listed 3.13E+01 Note: RADTRAD inventory values were used for isotopes which were not listed in the power uprate source term (Co, Pu, Am, and Cm isotopes).
Letter RBG-45930 Page 7 of 26 TABLE 2 MAIN CONTROL ROOM FLOW BIASED X/Q VALUES Time Period Main Air Remote Air More MFI4 Effective Intake Intake favorable Main Plant Stack 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.08E-03 4.29E-04 4.29E-04 1.07E-04 2.53E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 7.65E-04 3.46E-04 3.46E-04 8.65E-05 1.88E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.43E-04 1.37E-04 1.37E-04 3.43E-05 8.06E-05 1 to 4 days 2.47E-04 1.18E-04 1.18E-04 2.95E-05 6.21 E-05 4 to 30 days 2.18E-04 8.59E-05 8.59E-05 2.15E-05 5.10E-05 Turbine Building 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.02E-03 3.69E-04 3.69E-04 9.23E-05 5.31 E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.47E-03 3.08E-04 3.08E-04 7.70E-05 4.36E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.05E-03 1.30E-04 1.30E-04 3.25E-05 1.85E-04 1 to 4 days 9.01 E-04 1.07E-04 1.07E-04 2.68E-05 1.58E-04 4 to 30 days 6.74E-04 7.15E-05 7.15E-05 1.79E-05 1.16E-04 TABLE 3 XIQ VALUES USED IN LOCA ANALYSIS Release Point EAB* LPZ MCR SGTS/Containment 0-2 hours 6.05E-4 7.49E-5 2.53E-4 2-8 hours 6.05E-4 7.49E-5 1.88E-4 8-24 hours 6.05E-4 5.02E-5 8.06E-5 1-4 days 6.05E-4 2.1OE-5 6.21E-5 4-30 days 6.05E-4 6.13E-6 5.1OE-5 Turbine Building 0-2 hours 7.51 E-4 7.79E-5 5.31 E-4 2-8 hours 7.51 E-4 7.79E-5 4.36E-4 8-24 hours 7.51 E-4 5.23E-5 1.85E-4 1-4 days 7.51E-4 2.21E-5 1.58E-4 4-30 days 7.51 E-4 6.40E-6 1.16E-4 Note *: The 0 - 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months values was conservatively applies for the duration of the accident to ensure the "maximum" 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose is calculated as required per RG 1.183.
Letter RBG-45930 Page 8 of 26 TABLE 4 RBS LOCA RADIOLOGICAL CONSEQUENCE ANALYSIS PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption I. Data and assumptions used to estimate radioactive source from postulated accident.
~~1. Power Level 3100 MWt~~
~~2. Core Activity available for release Table 1~~
~~3. Gap Activity Release Fractions Per Table 1 of RG 1.183~~
~~4. Release fission product species and chemical Per RG 1.183, Section 3.5 form II. Release Rates~~
1. Primary Containment Leakage Rate 0 0- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.325 volume % per day
~~1 - 30 days 0.179 volume % per day~~
~~2. Secondary Containment Bypass Leakage Rate 580,000 cc/hr @ Pa (0.341 cfm)~~
~~0 -24 hours 319,000 cc/hr @ Pa (0.188 cfm) 0 1 - 30 days~~
~~3. Main Steam Line Leakage~~
0- 25 minutes 50 scfh = 0.85cfm*
25 minutes- 30 days 0 scfh Note *: 50 scfh was converted based on a maximum DW temperature of 330 0F. The pressure assumed was 7.6 psig (power uprate reports show that the drywell pressure is 22.8 psia @ 121 seconds decreases steadily to -19 psia at 10 minutes).
~~4. Engineered Safety Features Leakage 1 gpm Ill. Dispersion Data~~
~~1. EAB X/Q Data See Table 3~~
~~2. LPZ X/Q Data See Table 3~~
~~3. Control Room X/Q Data See Table 3 IV. Control Room Parameters~~
~~1. Unfiltered In-Leakage Rate 300 cfm~~
~~2. Outside Air Ventilation Rate~~
"* Actual 2000 cfm
"* Assumed 2000- 300 = 1700 cfm
~~3. Filter Initiation Time 20 minutes~~
~~4. CR ESF Iodine Filter Efficiency~~
"* Elemental/Organic (Charcoal) 98%
"* Particulate (HEPA) 99%
5. Control Room Breathing Rates and Per RG 1.183 Occupancy Factors Letter RBG-45930 Page 9 of 26 Description of Input/Assumption Design Basis Input and/or Assumption V. Standby Gas Treatment Parameters
~~1. Positive Pressure Period 30 minutes~~
~~2. Flow Rates~~
"* Annulus 2,500 cfm
"* Auxiliary Building 10,000 cfm
~~3. SGTS Iodine Filter Efficiency~~
"* Elemental/Organic (Charcoal) 90%
"* Particulate (HEPA) 99%
Vl. Building Volumes
~~1. Drywell 2.36E+05 ftW~~
~~2. Containment 1.19E+06 ft' ý~~
~~3. Annulus* 3.57E+05 ftt~~
4. Auxiliary Building* 1.1 6E+06 ftW Note *: Only 50% of the annulus and auxiliary building volumes were credited in the actual analysis (values listed are actual volumes)
~~5. Control Room 1.88E+05 ftW~~
~~6. Suppression Pool** 1.25E+05 ftW Note **: The actual analysis conservatively assumed a volume of 120,000 ft3.~~
~~VII. Containment Mixing Data~~
~~1. Blowdown Data (Drywell = Containment)~~
~~0-10 minutes 4.74E+05 cfm~~
~~10 minutes + 0 cfm~~
~~2. Hydrogen Mixing Data (Drywell <~~
~~Containment) 0 cfm~~
0-25 minutes 600 cfm 0 25 minutes - 1.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 1.OE+08 cfm 0 1.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 30 days (Infinite Mixing)
~~3. Steaming Data (Drywell : Containment)~~
~~0-25 minutes 0 cfm~~
~~25 minutes - 1.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 3000 cfm~~
~~1.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months - 30 days (Infinite Mixing) (Included in Hydrogen Mixing)~~
~~VIII. Misc. Data~~
~~1. Dose Conversion Factors Based on FGR 11 & 12~~
~~2. Off-Site Breathing Rates Based on RG 1.183, Section 4.1.3~~
~~3. Drywell Plateout Coefficients~~
"* Elemental 1.01 hr-1
"* Particulate RADTRAD default - Powers (10) Model
4. ESF Leakage - Halogen Flashing Fraction 0.10 Letter RBG-45930 Page 10 of 26 TABLE 5 AST SCENARIO TIMING Time AST Event Time 0 sec -121 sec.
~~DER of one of the two Recirculation Lines in the reactor occurs.~~
~~Plant experiences a SSE.~~
~~Loss of Offsite Power occurs.~~
~~Release of all reactor coolant halogens in AST scenario.~~
~~0+ sec. -121 sec.~~
~~LOCA signal from high drywell pressure occurs effectively instantaneously.~~
~~Reactor low water level reached.~~
~~All Emergency Diesel Generators (EDG) signaled to start.~~
~~ECCS systems signaled to start.~~
~~Main Control Room Ventilation Signaled to start.~~
1.1 sec.* -119.9 Drywell peak pressure reached.
sec.
5.5 sec.* -114.5 7 of 8 MSIVs are closed (one MSIV assumed to fail open) sec.
10 sec.* -111 sec. 2 of 3 EDG start (1 assumed to fail to start) and ready to load.
23 sec.* -98 sec. Top of Active Fuel Uncovered (USAR Figure 6.3-11) 27 sec.* -94 sec. Initiation of HPCS.
37 sec. * -84 sec. Initiation of LPCS and LPSI.
56 sec.* -65 sec. Low Pressure Core Spray flow begins (initiating ESF leakage) 121 sec. 0 sec. Failure of fuel cladding begins in AST scenario. This results in the release of gap activity. The following is assumed based on the above:
"* Containment leaking at proposed La
"* ESF leakage at 1 gpm
"* SCB leakage at proposed TS limit.
293 sec.* 172 sec. Vessel is reflooded (Not in AST scenario which has no cooling).
584 sec. 463 sec. Drywell pressure < Containment pressure. Suppression pool bypass flow assumed to terminate.
20 min. 18 min.
~~Operators assumed to initiate the Main Control Room Fresh Air Emergency Filters.~~
~~Operators assumed to initiate Main Steam Positive Leakage Control System.~~
~~Operators assumed to initiate the SGTS system.~~
~~25 min. 23 min.~~
~~MS-PLCS becomes fully operational terminating failed MSIV release.~~
0 Operators initiate the hydrogen mixing system.
30 min. 28 min. Annulus and Auxiliary Building assumed to reach -0.25" w.g.,
therefore, secondary containment is established.
32 min. 30 min.** Core geometry is compromised in AST scenario - fuel melting begins. Early In-Vessel Release begins.
<1 hour <1 hour Operators initiate Standby Liquid Control System.
Letter RBG-45930 Page 11 of 26 Time AST Event Time 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Fuel melting terminated.
8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months MCR and LPZ XIQ changed.
24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 
~~MCR and LPZ X/Q changed.~~
~~0 Off-site breathing rate reduced.~~
~~Containment and SCB leakage reduced to 55% of proposed TS allowable value based on containment pressure analyses.~~
0 MCR occupancy reduced.
96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 96 hours
~~MCR and LPZ X/Q changed.~~
~~Off-site breathing rate reduced.~~
~~MCR occupancy reduced.~~
30 days 30 days End of Dose Calculation.
Note *: Minor changes in these times will not have an appreciable affect the dose calculation.
Note **" Times >30 minutes were no longer adjusted to account for the 2 minutes for fuel damage.
Letter RBG-45930 Page 12 of 26 TABLE 6 LOCA RADIOLOGICAL CONSEQUENCES Release Descriptions EAB LPZ MCR Containment/Secondary Containment 2.6 1.7 0.4 Secondary Containment Bypass/MSIV 12.3 5.4 2.9 ESF Liquid Leakage <0.1 0.4 <0.1 Total 15.0 7.5 3.4 RegulatoryLimit 25.0 25.0 5.0 Figure la Primary and Secondary Containment Releases: 0 - 10 Minutes 0.341 cfm Note *: The times referred to in Figures 1 through 3 are post-break times rather that AST time (i.e., time after the onset of fuel damage). This was done for convenience and to avoid confusion. In reality the input decks generated by RBS were done in AST time and neglect the first 2 minutes.
Note **: The "sink" node was used to contain radioactivity which is evaluated elsewhere. This was necessary to apply the appropriate dispersion factors to different release paths.
Specifically, for the primary and secondary containment releases the SCB and MSIV leakage paths are evaluated in a separate input deck (as depicted in Figures 2a through 2e).
Letter RBG-45930 Page 13 of 26 Figure 1 b Primary and Secondary Containment Releases: 10 - 25 Minutes Letter RBG-45930 Page 14 of 26 Figure Ic Primary and Secondary Containment Releases: 25 - 30 Minutes Letter RBG-45930 Page 15 of 26 Figure Id Primary and Secondary Containment Releases: 30 Minutes - 1.9 Hours Letter RBG-45930 Page 16 of 26 Figure le Primary and Secondary Containment Releases: 1.9 - 24 Hours Letter RBG-45930 Page 17 of 26 Figure If Primary and Secondary Containment Releases: I - 30 Days Letter RBG-45930 Page 18 of 26 Figure 2a Secondary Containment Bypass and MSIV Releases: 0 - 10 Minutes Letter RBG-45930 Page 19 of 26 Figure 2b Secondary Containment Bypass and MSIV Releases: 10 - 25 Minutes Letter RBG-45930 Page 20 of 26 Figure 2c Secondary Containment Bypass and MSIV Releases: 25 Minutes - 1.9 Hours V1.
Containment Letter RBG-45930 Page 21 of 26 Figure 2d Secondary Containment Bypass and MSIV Releases: 1.9 - 24 Hours Containment Letter RBG-45930 Page 22 of 26 Figure 2e Secondary Containment Bypass and MSIV Releases: 1 - 30 Days tSink Io.179%fday cl D~veD O.188cfm Containment Letter RBG-45930 Page 23 of 26 Figure 3a Engineered Safety Features Liquid Leakage: 0 - 30 Minutes 111 Suppression 1 om= 0. 1337 cfm 0
Pool Flash Fration = 0.10 Letter RBG-45930 Page 24 of 26 Figure 3b Engineered Safety Features Liquid Leakage: 30 Minutes - 30 Days Letter RBG-45930 Page 25 of 26 Figure 4a Control Room Ventilation Model - Normal Mode LII 2,000 cfin Main Control Room 2,000 cfm ->
0 cfm vt Letter RBG-45930 Page 26 of 26 Figure 4b Control Room Ventilation Model - Emergency Mode
Attachment 8 Letter RBG-45930 Fuel Handling Accident (FHA) and Light Load Drop Accident (LLA) Summary Letter RBG-45930 Page 1 of 5 Fuel Handling Accident (FHA) and Light Load Drop Accident (LLA) Summary SCOPE The Fuel Handling Accident (FHA) is postulated to occur as a consequence of a failure of the fuel assembly lifting mechanism, resulting in a drop of a raised fuel assembly onto stored fuel bundles. Several FHA events have been evaluated including postulated scenarios for containment upper pool refueling operations as well as fuel building activities. The postulated FHA inside containment results in a larger number of fuel rods damaged as a result of a drop of a spent fuel bundle over the reactor core.
A bounding RBS FHA event involving the drop of a spent fuel assembly onto fuel assemblies in either the spent fuel pool or the reactor core has been analyzed. All fuel types used by RBS were evaluated (GE8, GEl1, and Framatone Atrium 10 fuel). Based on conservative and limiting assumptions consistent with USNRC Regulatory Guide 1.183, the RBS bounding FHA event occurs in containment and results in a total of 122 GE 9x9 fuel rods damaged. This value was conservatively increased to 150 GE 9x9 rods. The failed rods gap activity is immediately released to the fuel building or containment atmosphere. Over a period of two hours, this accident activity is released into the atmosphere without crediting building mixing or dilution. No credit is taken for operation of either building's charcoal filtration units. Also, no credit is taken for filtration by the Main Control Room ESF charcoal filters. However, credit is taken for control room operations manual selection of the more favorable air intake in accordance with SRP 6.4 guidelines (i.e., the more favorable IQ is divided by four since RBS has dual air intake with manual selection).
The postulated accident activity is released as a "ground level release" from the fuel building or containment ventilation vent and dispersed to offsite and control room receptors according to plant specific atmospheric dispersion factors demonstrated as bounding for both release points.
These plant specific offsite and control room dispersion factors have been shown to be conservative in comparison to those estimated per the criteria in Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and NUREG/CR-6331, "Atmospheric Relative Concentrations in Building Wakes" (ARCON96) respectively, as demonstrated in Table 1.
The Light Load Drop Accident (LLA) is closely related to the FHA. "Light loads" for RBS are 1200 lbs. This scenario involves dropping a load, other than fuel, onto irradiated fuel in either the containment or fuel buildings. Typically LLA scenarios are bounded by FHA results, however, RBS performed a conservative analysis to demonstrate that a drop of a light load could not cause dose limits to be exceeded. The bounding scenario was the drop of a light load in primary containment from the polar crane onto the reactor core. This drop results in a maximum of 247 GE 9x9 fuel rods damaged, however, the dose analysis conservatively used 300 rods to ensure that all fuel types are bounded. All other assumptions are consistent with the FHA analysis.
The radiological consequences of the event are evaluated against the acceptance criteria of 10CFR50.67 and GDCQ19 of 10CFR50 Appendix A.
Letter RBG-45930 Page 2 of 5 INPUTS AND ASSUMPTIONS The inputs and assumptions used in the radiological consequence analysis of the FHA are listed in Tables 1 and 2. The assumptions used in the LLA dose analysis are presented in Table 4.
These inputs are consistent with the requirements of RG 1.183.
RESULTS The radiological consequences for the postulated RBS FHA event are summarized in Table 3.
The FHA event result in offsite and control room doses within the regulatory limits of 10CFR50.67 and GDC 19 of 10CFR50 Appendix A. No exceptions to Regulatory Guide 1.183 guidance were taken.
Letter RBG-45930 Page 3 of 5 TABLE 1 COMPARISONS OF ATMOSPHERIC DISPERSION FACTORS Location Values Based ARCON961PAVAN ARCON96/PAVAN on M&CI Fuel Building Containment RG 1.145 Ventilation/Truck Equipment Hatch (Note 1)
Methodology Bay Doors (Notes 1,4)
EAB 8.58E-04 sec./m 3 Note 2 6.52E-05 sec./m3 9 0- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months LPZ
~~0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.13E-04 sec./m 3 6.62E-05 sec./m3~~
~~8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.89E-05 sec./m3 Note 2 4.51 E-05 sec./m3~~
~~1 - 4 days 3.65E-05 sec./m 3 1.96E-05 sec./m 3~~
~~4 - 30 days 1.21 E-05 sec./m 3 5.95E-06 sec./m 3 Main Control Room Normal Intake~~
~~0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.62E-03 sec./m 3 1.09E-03 sec./m 3 1.21 E-03 sec./m 3~~
~~2- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.62E-03 sec./m 3 6.83E-04 sec./m3 7.46E-04 sec./m 3~~
~~8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.20E-03 sec./m 3 3.17E-04 sec./m 3 3.39E-04 sec./m 3~~
~~1 - 4 days 4.05E-04 sec./m 3 2.34E-04 sec./m3 2.65E-04 sec./m3~~
~~4 - 30 days(Note 3) 6.48E-05 sec./m 3 1.95E-04 sec./m 3 2.20E-04 sec./m3 Main Control Room Remote Intake~~
~~0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.90E-03 sec./m3 8.56E-04 sec./m 3 3.44E-04 sec./m 3~~
~~2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.90E-03 sec./m 3 5.53E-04 sec./m 3 2.27E-04 sec./m 3~~
~~8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.46E-03 sec./m 3 2.41 E-04 sec./m 3 9.62E-05 sec/iM3~~
~~1 - 4 days 6.08E-04 sec./m 3 1.89E-04 sec./m 3 7.75E-05 sec./m 3~~
~~4 - 30 days 1.52E-04 sec./m 3 1.42E-04 sec./m 3 5.78E-05 sec./m 3 Note 1: The 5% X/Q values are presented for comparison.~~
Note 2: Fuel Building Cask Handling Door values were not calculated for off-site locations.
Note 3: The Murphy & Campe values bound those recalculated using ARCON96 with the exception of the 4-30 day values. Since the FHA/LLA analyses assume a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months release use of the higher values during this time frame would not impact the calculated results.
Note 4: The fuel building ventilation system exhaust vent was assumed for the RBS Control Room's Main Air Intake, and the truck bay doors was assumed for the remote air intake. Each release point was the most conservative location for the respective air intake.
Letter RBG-45930 Page 4 of 5 TABLE 2 RBS FHA RADIOLOGICAL CONSEQUENCEANALYSIS PARAMETERS Description of InputlAssumption Design Basis Input and/or Assumption I. Data and assumptions used to estimate radioactive source from postulated accident.
~~1. Power Level 3100 MWt~~
~~2. Number of damaged rods (GEl1) 150~~
~~All fuel types used by RBS were evaluated.~~
~~GEl 1 (9x9 array) was determined to be bounding for RBS.~~
~~3. Total number of rods in core 46176 (Limiting GE 9x9 Fuel)~~
~~4. Core Activity available for release Table 3~~
~~5. Radial peaking factor 2.00~~
~~6. Gap Activity Release Fractions* Per Table 3 of RG 1.183~~
~~RBS has verified that the fuel does not exceed the burnup requirements to meet the release fractions of Regulatory Guide 1.183.~~
~~7. Release fission product species and chemical Per RG 1.183, Appendix B form I1. Data and assumptions used to estimate activity released to the environment.~~
~~1. 1. Building Release Rate 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months linear release rate~~
~~2. Halogen Decontamination Factor 200 Ill. Dispersion Data~~
~~1. EAB /Q Data 0-2 hrs 8.58E-04 sec/m3~~
~~2. LPZ /Q Data 3 0-8 hrs 1.13E-04 sec/m 3~~
8-24 hrs 7.89E-05 sec/m 3
1-4 days 3.65E-05 sec/m 4-30 days 1.21 E-05 sec/m3
~~3. Control Room /Q Data 0-20 min 1.62E-03 sec/m3 3~~
20 min-8 hrs 4.05E-04 sec/m 3
8-24 hrs 3.OOE-04 sec/m 3
1-4 days 1.01 E-04 sec/im 4-30 days 1.62E-05 sec/m3 RBS has dual Control Room air intakes and redundant radiation detectors within each intake.
This dispersion data credits operator action at 20 minutes to select most favorable air intake.
IV. Control Room Parameters
~~1. Free Air Volume 188,000 ft3~~
~~2. Unfiltered In-leakage Rate 300 cfm~~
~~3. Outside Air Ventilation Rate 1700 cfm~~
~~4. CR ESF Iodine Filter Efficiency 0% (Not credited)~~
5. Control Room Breathing Rates and Per RG 1.183 Occupancy Factors Letter RBG-45930 Page 5 of 5 TABLE 3 RBS FHA CORE ACTIVITY Isotope EOC Core Inventory (CilMWt) 1-131 2.631E+04 1-132 3.845E+04 1-133 5.502E+04 1-134 6.056E+04 1-135 5.195E+04 Kr-85 3.015E+02 Kr-85m 6.734E+03 Kr-87 1.292E+04 Kr-88 1.830E+04 Xe-1 33 5.528E+04 Xe-135 7.148E+03 TABLE 4 LIGHT LOAD DROP PARAMETERS Description of InputlAssumption Design Basis Input and/or Assumption Maximum Load Height- Polar Crane El. 214.3' Top of Fuel Bundles in Reactor Core El. 133.5' Maximum Drop Distance 80.8' Load Weight 1200 lbs Cladding Yield Strength 200 ft-lbs/GE 9x9 rod Clad / Non-Fuel Mass Ratio 0.510 Kinetic Energy Absorbed by Fuel 100%
Total Kinetic Energy Available 96,600 ft-lb Kinetic Energy Absorbed by Cladding 49,500 ft-lb Number of Rods Damaged 247 GE 9x9 rods TABLE 5 FHA AND LLA RADIOLOGICAL CONSEQUENCES Receptor Regulatory Limit FHA Dose LLA Dose (REM TEDE) (REM TEDE) (REM TEDE)
EAB 6.3 2.5 5.0 LPZ 6.3 0.4 0.7 Control Room 5 1.7 3.3
Attachment 9 Letter RBG-45930 Control Rod Drop Accident (CRDA) Summary Letter RBG-45930 Page 1 of 5 CONTROL ROD DROP ACCIDENT (CRDA)
~~SUMMARY~~
SCOPE The plant design basis Control Rod Drop Accident (CRDA) is a postulated event in which a high worth control rod drops from its fully inserted or intermediate position in the core. The removal of large negative reactivity from the core results in a localized power excursion. For the CRDA accident scenario GE8 fuel was found to be bounding. Based on conservative and limiting assumptions consistent with USNRC Regulatory Guide 1.183, the RBS postulated CRDA at full power results in a total of 850 GE8 fuel rods damaged.
The postulated accident activity is released at ground level from the plant condenser and dispersed to offsite and control room receptors according to plant specific atmospheric dispersion factors. These plant specific offsite and control room dispersion factors were determined in accordance with Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and NUREG/CR 6331, "Atmospheric Relative Concentrations in Building Wakes" (ARCON96) guidance, respectively.
Control room doses are estimated for two CRDA events:
1. A design basis case based on Regulatory Guide 1.183 assumptions. Source term is conservatively based on 100% power operations with activity released via the plant condenser. No credit is taken for the Control Room ESF charcoal filters. Activity is released instantaneously from the reactor fuel to the coolant.
2. Low power operation. Limited CRDA that does not result in Main Steam Line Radiation Monitor (MSLRM) Trip. Activity release via the condenser Mechanical Vacuum Pumps at 4000 cfm until isolated per operator action at 20 minutes. Manual initiation of the Control Room ESF charcoal filters is credited 20 minutes into the event. A 10 second "burst" release is assumed from the reactor fuel in accordance with NEDO-31400A, Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Radiation Monitor, assumptions.
The radiological consequences of the event are evaluated against the acceptance criteria of 10CFR50.67.
INPUTS AND ASSUMPTIONS The inputs and assumptions used in the radiological consequence analysis of the RBS CRDA are listed in Tables 1 and 2. These inputs are consistent with the requirements of RG 1.183.
In the low power operation CRDA event, activity release assumptions are established to maximize both the offsite and Control Room doses. Offsite doses for this case are estimated based cladding damage only. The dose rate to the MSLRM was calculated to be 8.4 R/hr which corresponds to a "design basis" N-16 concentration of 300 uCi/g (See USAR Table 11.1-1) which only occurs during Hydrogen Water Chemistry (HWC) operation (the maximum N-16 concentration Normal Water Chemistry, NWC, is 50 uCi/g). The fuel damage is based on a Main Steam Line Radiation Monitor setpoint of 30 R/hr which corresponds to 25 R/hr (3.0 x HWC dose rates) plus 20% for instrument uncertainty. The dose rate from 1 GE 9x9 rod (assuming a Letter RBG-45930 Page 2 of 5 "burst" release) was determined to be 0.6 R/hr. Background radiation is neglected this corresponds to 50 GE 9x9 rods being damaged. The MSLRM setpoints are based on 1.5x and 3.Ox "full power background" values. The HWC system affects dose rates in the steam affected areas of the plant by as much as a factor of 6 (50 uCi/g vs. 300 uCi/g). The assumptions used justify basing the MSLRM isolation setpoint (HIGH-HIGH) on 6x design basis HWC dose rates which correspond to 18.Ox NWC dose rates. Similarly, the HIGH setpoint will be based on 1.5x HWC design basis dose rates which corresponds to 9.Ox NWC values.
RESULTS The radiological consequences for the postulated RBS CRDA events are summarized in Table 3.
The Full Power and Low Power CRDA events result in offsite and control room doses within the regulatory limits of 10CFR50.67. No exceptions to Regulatory Guide 1.183 guidance were taken.
Letter RBG-45930 Page 3 of 5 TABLE 1 RBS CRDA RADIOLOGICAL CONSEQUENCEANALYSIS PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption I. Data and assumptions used to estimate radioactive source from postulated accident.
~~1. Power Level 3100 MWt~~
~~2. Number of damaged rods 100% Power Event 850 GE 8x8 Low Power Event (gap release) 50 GE 9x9~~
~~3. Total rods in core GE 8x8 38,688 (62 rods per assembly)~~
~~GE 9x9 46,176 (74 rods per assembly)~~
~~4. Number of assemblies damaged Design Basis - Maximum Fuel Damage 850/62 = 13.7 (based on 8x8)~~
Limited CRDA - No MVP trip (Based on 50 1 74 = 0.68 9x9)
Note: For the CRDA scenario GE8 fuel is bounding. This is confirmed each reload cycle.
~~5. Core Activity available for release Table 2~~
~~6. Radial Peaking Factor 2.00~~
~~7. Assumed % fuel melt Design Basis - Maximum Fuel Damage 100%~~
~~Limited CRDA - No MVP trip 0%~~
~~8. Gap Activity Release Fractions Per RG 1.183, Table 3 and Appendix C 10% noble gases, 10% iodines, 12% alkali metals~~
~~9. Fuel Melt Release Fractions Per RG 1.183, Appendix C 100% noble gases, 50% lodines~~
10. Fuel Release Duration Design Basis - Maximum Fuel Damage Instantaneous Limited CRDA - No MVP trip 10 sec. burst II. Data and assumptions used to estimate activity released to the environment.
1. Condenser Leak Rate Design Basis CRDA Per RG 1.183, Appendix C. 1% per day for 24 Limited CRDA - No MVP trip hours 4000 cfm for 20 minutes, 1% per day for next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 
~~2. Condenser Iodine Release Fractions Per RG 1.183, Appendix C 97% Elemental, 3% Organic~~
~~3. Condenser Radioactive Decay During Holdup Credited~~
4. Condenser Volume 106,460 ft' Letter RBG-45930 Page 4 of 5 Description of Input/Assumption Design Basis Input and/or Assumption Ill. Dispersion Data
~~1. EAB /Q Data 0-2 hrs 7.51 E-04 sec/mr~~
2. LPZ /Q Data 0-8 hrs 7.79E-05 sec/m33 8-24 hrs 5.23E-05 sec/m 1-4 days 2.21E-05 sec/m 3 6.40E-06 sec/m 3 4-30 days
3. Control Room /Q Data Main Air Intake 3.03E-03 sec/m 3 0-2 hrs 2.47E-03 sec/m 3 1.05E-03 sec/m 3 2-8 hrs 8-24 hrs 9.01 E-04 sec/m 3 1-4 days 6.74E-04 sec/m 3 4-30 days Remote Air Intake 0-2 hrs 3.69E-04 sec/m3 2-8 hrs 3.08E-04 sec/m 3 8-24 hrs 1.30E-04 sec/m 3 1-4 days 1.07E-04 sec/m3 4-30 days 7.15E-05 sec/m 3 Flow Biased Values 0-2 hrs 5.31 E-04 sec/m 3 3
2-8 hrs 4.63E-04 sec/m 8-24 hrs 1.85E-04 sec/m 3 1-4 days 1.58E-04 sec/m33 4-30 days 1.16E-05 sec/m The "design basis" case assumes the main air intake values for the duration of the event since no credit is taken for operators initiating the ESF filters.
RBS has dual Control Room air intakes and redundant radiation detectors within each intake.
This dispersion data credits operator action and weights the values based on assumed flow rates (similar to the LOCA analysis).
IV. Control Room Parameters
~~1. Free Air Volume 188,000 ft 3~~
~~2. Unfiltered In-leakage Rate 300 cfm~~
~~3. Outside Air Ventilation Rate 1700 cfm Letter RBG-45930 Page 5 of 5 Description of Input/Assumption Design Basis Input and/or Assumption~~
~~4. Intake Iodine Filter Efficiency Design Basis CRDA Aerosol 0%~~
Elemental and Organic 0%
Limited CRDA Aerosol 99%
Elemental and Organic 98%
~~5. Time for Control Room Ventilation Isolation per Operator Action.~~
~~Design Basis CRDA Not credited Limited CRDA 20 minutes~~
~~6. Emergency Mode Recirculation Rate (Post- 2000 cfm isolation Mode)~~
7. Control Room Breathing Rates and Per RG 1.183 Occupancy Factors TABLE 2 RBS CRDA CORE ACTIVITY Isotope EOC Core Inventory (CilMWt) 1-131 2.631E+04 1-132 3.845E+04 1-133 5.502E+04 1-134 6.056E+04 1-135 5.195E+04 Kr-85 3.015E+02 Kr-85m 6.734E+03 Kr-87 1.292E+04 Kr-88 1.830E+04 Xe-1 33 5.528E+04 Xe-1 35 7.148E+03 Cs-134 5.357E+03 Cs-136 1.179E+03 Cs-1 37 3.324E+03 Rb-86 4.696E+01 TABLE 3 CONTROL ROD DROP ACCIDENT RADIOLOGICAL CONSEQUENCES Receptor Regulatory Limit 100% Power Event Low Power Event (TEDE) Dose Dose (TEDE) (TEDE)
EAB 6.3 0.9 4.7 LPZ 6.3 0.4 0.5 Control Room 5 0.5 1.3
Attachment 10 Letter RBG-45930 Main Steam Line Break (MSLB) Outside of Primary Containment Summary 0
Letter RBG-45930 Page 1 of 4 MAIN STEAM LINE BREAK ACCIDENT (MSLB) OUTSIDE OF PRIMARY CONTAINMENT
~~SUMMARY~~
SCOPE The plant design basis Main Steam Line Break Accident (MSLB) assumes an instantaneous guillotine break at one of the four main steam lines outside containment at a location downstream of the outermost isolation valve. The break results in mass loss from both ends of the break. The flow from the upstream side of the break is initially limited by the flow restrictor upstream of the inboard isolation valve. Flow from the downstream side is initially limited by the total area of the flow restrictor in the three unbroken lines. Subsequent closure of the Main Steam Isolation Valves (MSIVs) further limits the flow when the valve flow area becomes less than the limiter area and finally terminates the mass loss when the full closure is reached. The mass release used is based on initial licensing of the plant and bounds power uprated conditions.
No credit was taken for the Control Room ESF charcoal filter trains. No fuel damage is predicted and thus the postulated release halogen activity is based on the maximum coolant activity allowed by Technical Specifications as required by USNRC Regulatory Guide 1.183, Appendix D. The noble gas activity is based on an offgas release rate of 310,000 uCi/sec. (after 30 minutes decay) which conservatively bounds the 290 mCi/sec. allowed per Technical Specifications. The alkali metals design basis coolant concentrations were also increased by 2% to account for uncertainties in the core thermal power level. The analysis conservatively assumed that 100% of the halogen and alkali metal activity present in the released coolant was transported to the environment.
The postulated accident activity is released as a "ground level release" from the Main Steam Tunnel (MST) blowout panel and dispersed to offsite and control room receptors according to plant specific atmospheric dispersion factors. These plant specific offsite and control room dispersion factors were determined in accordance with Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants,"
and NUREG/CR-6331, "Atmospheric Relative Concentrations in Building Wakes" (ARCON96) respectively.
Consistent with RG 1.183, Appendix D, Offsite and Control room doses are estimated for two MSLB events:
~~1. Primary coolant iodine concentration corresponding to an assumed maximum pre-accident spike of 4 Ci/gm dose equivalent 1-131 per plant Technical Specifications.~~
~~2. Primary coolant iodine concentration corresponding to the Technical Specifications' maximum equilibrium value of 0.2 Ci/gm dose equivalent 1-131.~~
The radiological consequences of the event are evaluated against the acceptance criteria of 10CFR50.67.
0 Letter RBG-45930 Page 2 of 4 INPUTS AND ASSUMPTIONS The inputs and assumptions used in the radiological consequence analysis of the RBS MSLB are listed in Tables 1 and 2. These inputs are consistent with the requirements of RG 1.183.
RESULTS The radiological consequences for the postulated RBS MSLB events are summarized in Table 3.
The MSLB events result in offsite and control room doses within the regulatory limits of 10CFR50.67. No exceptions to Regulatory Guide 1.183 guidance were taken.
0 Letter RBG-45930 Page 3 of 4 TABLE 1 RBS MSLB RADIOLOGICAL CONSEQUENCEANALYSIS PARAMETERS Description of Input/Assumption Design Basis Input and/or Assumption I. Data and assumptions used to estimate radioactive source from postulated accident.
~~1. Power Level 3100 MWt~~
~~2. Maximum Pre-accident Spike Iodine 4 Ci/gm DE 1-131 Concentration~~
~~3. Maximum Equilibrium Iodine Concentration 0.2 Ci/gm DE 1-131~~
~~4. Noble Gas Source Term Based on 310,000 Ci/sec at 30 minutes, corrected to time equal zero.~~
~~5. Alkali Metals Reactor coolant activity design concentration ratioed to account for 102% power.~~
~~II. Data and assumptions used to estimate activity released to the environment.~~
1. Mass Release (Note this data corresponds to Steam, 11620 Ibm that calculated for initial licensing of the plant. Liquid, 68942 Ibm Analyses demonstrate these values bound the hot standby conditions for Power Uprated conditions.)
~~2. Iodine Carryover Fraction 4%~~
~~3. Break Isolation Time 5.5 seconds~~
~~4. Building Release Rate Instantaneous ground level release with no credit for plateout, holdup, or dilution.~~
~~5. Iodine Species Release Fractions to Per RG 1.183, Appendix D, Environment 95% Aerosol, 4.85 % Elemental, 0.15% Organic~~
~~6. Activity Released to Environment Table 2 III. Dispersion Data~~
~~1. EAB /Q Data 0-2 hrs 8.25E-04 sec/m3~~
2. LPZ /Q Data 0-8 hrs 1.11E-04 sec/m3 8-24 hrs 7.79E-05 sec/m33 1-4 days 3.60E-05 sec/m 3
~~4-30 days 1.19E-05 sec/m~~
3. Control Room /Q Data 0-8 hrs 3.64E-03 sec/m3 8-24 hrs 2.69E-03 sec/m33 1-4 days 1.46E-03 sec/m 4-30 days 2.73E-04 sec/m3 Since no credit is taken for the ESF filter trains the X/Q values assumed in this analysis were based on the Main Air Intake.
~~IV. Control Room Parameters~~
1. Free Air Volume 188,000 ft*
~~2. Unfiltered In-leakage Rate 300 cfm 0~~
~~Letter RBG-45930 Page 4 of 4 Description of InputlAssumption Design Basis Input and/or Assumption~~
~~3. Outside Air Ventilation Rate 1700 cfm~~
~~4. Emergency Mode Filtered Intake/Unfiltered 2000 cfm Inleakage Rate (1700 cfm ventilation rate +~~
~~300 inleakage rate)~~
5. Control Room Breathing Rates and Per RG 1.183 Occupancy Factors TABLE 2 RBS MSLB ACTIVITY RELEASED TO ENVIRONMENT Isotope 0.2 Cilgm DE 1-131 4 Cilgm DE 1-131 Case (Ci) Case (Ci) 1-131 1.70E+00 3.40E+01 1-132 2.51E+01 5.02E+02 1-133 2.27E+01 4.53E+02 1-134 3.97E+01 7.94E+02 1-135 2.19E+01 4.37E+02 Cs-1 34 5.42E-03 5.42E-03 Cs-136 3.51E-03 3.51E-03 Cs-137 1.40E-02 1.40E-02 Kr-85m 5.22E-02 5.22E-02 Kr-85 1.63E-04 1.63E-04 Kr-87 1.79E-01 1.79E-01 Kr-88 1.79E-01 1.79E-01 Xe-133 6.58E-02 6.58E-02 Xe-135 1.95E-01 1.95E-01 TABLE 3 MAIN STEAM LINE BREAK ACCIDENT RADIOLOGICAL CONSEQUENCES Case EAB Dose LPZ Dose Regulatory Limit (REM TEDE) (REM TEDE) (REM TEDE) 4 Ci/gm DE 1-131 1.4 0.2 25 0.2 Ci/gm DE 1-131 <0.1 <0.1 2.5 Case Control Room Dose Regulatory Limit (REM TEDE) (REM TEDE) 4 Ci/gm IDE 2.2 5 0.2 Ci/gm IDE 0.2 5

RBG-45930, License Amendment Request Re Full-Scope Application of NUREG-1465 Alternative Source Term Insights

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RBG-45930, License Amendment Request Re Full-Scope Application of NUREG-1465 Alternative Source Term Insights

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SUBJECT:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

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