View source for CNL-15-177, Submittal of Replacement Pages for Developmental and Final Revision J of the Technical Specification & Technical Specification Bases, and Developmental and Final Revision E of Technical Requirements Manual & Technical Manual.

{{Adams
| number = ML15247A564
| issue date = 09/04/2015
| title = Submittal of Replacement Pages for Developmental and Final Revision J of the Technical Specification & Technical Specification Bases, and Developmental and Final Revision E of Technical Requirements Manual & Technical Manual.
| author name = Shea J
| author affiliation = Tennessee Valley Authority
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000391
| license number = CPPR-0092
| contact person = 
| case reference number = CNL-15-177
| document type = Letter type:CNL, Manual, Technical Specifications
| page count = 82
}}

=Text=
{{#Wiki_filter:Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-177 September 4, 2015 10 CFR &sect;50.36(a)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Construction Permit No. CPPR-92 NRC Docket No. 50-391

==Subject:==
Watts Bar Nuclear Plant Unit 2 - Submittal of Replacement Pages for Developmental and Final Revision J of the Technical Specification &
Technical Specification Bases, and Developmental and Final Revision E of the Technical Requirements Manual & Technical Requirements Manual Bases

==Reference:==
TVA Letter to NRC, Watts Bar Nuclear Plant Unit 2 - Submittal of Developmental and Final Revision J of the Technical Specification &
Technical Specification Bases, and Developmental and Final Revision E of the Technical Requirements Manual & Technical Requirements Manual Bases, dated July 6, 2015 [ML15187A461]
The purpose of this letter is to provide replacement pages for changes to the Technical Specification (TS) Revision J and Technical Requirements Manual (TRM) Revision E that have not been previously submitted to the Nuclear Regulatory Commission (NRC). The referenced letter submitted TS Developmental Revision J and the associated TS Bases and TRM Developmental Revision E and associated Bases. Several unresolved issues necessitated revisions to the information provided in the referenced letter.
The enclosure provides the summary of changes to Developmental Revision J and E of the TS and TRM, respectively. The enclosure provides a basis for the new replacement pages.
Several of these changes have been previously submitted to the staff. These changes are also summarized in the enclosure. Attachments to the enclosure provide the new replacement pages for the TS and TS Bases and for the TRM and TRM Bases.

U.S. Nuclear Regulatory Commission CNL-15-177 Page 2 September 4, 2015 There are no new regulatory commitments contained in this submittal. Please contact Gordon Arent at 423-365-2004 if there are questions regarding this submittal.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 4th day of September 2015.
Respectfully, Digitally signed by J. W. Shea J. W. Shea DN: cn=J. W. Shea, o=Tennessee Valley Authority, ou=Nuclear Licensing, email=jwshea@tva.gov, c=US Date: 2015.09.04 21:16:44 -04'00' J. W. Shea Vice President, Nuclear Licensing

==Enclosure:==
Summary of Technical Specification and Technical Requirements Manual Changes cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant, Unit 2 NRC Project Manager - Watts Bar Nuclear Plant, Unit 2


ENCLOSURE Watts Bar Nuclear Plant Unit 2 Summary of Technical Specification and Technical Requirements Manual Changes


Tennessee Valley Authority (TVA) submitted a complete set of Technical Specifications (TS),
TS Bases, Technical Requirements Manual (TRM) and TRM Bases to the Nuclear Regulatory Commission (NRC) on July 6, 2015 by Reference 1. A limited number of changes have been identified since Reference 1 was submitted that change information previously provided.
Previously Submitted Replacement Pages By letter dated August 13, 2015, TVA provided replacement pages for TS Surveillance Requirements 3.6.11.2 and 3.6.11.3 regarding the amount of ice that must be maintained in the ice condenser. The mass of ice required was increased to support a revised containment loss of coolant accident pressure analysis that was submitted in Reference 2. These pages are not being resubmitted in this letter.
Previously Described Changes The NRC provided recommended changes to TS Bases 3.4.17, SG Tube Integrity. By letter dated September 3, 2014 (Reference 3), TVA agreed to make the changes and to provide TS Bases 3.4.17 pages showing the changes. These changes were not incorporated in the TS Developmental Revision J submittal. The appropriate changes are provided in Attachment 1 to this enclosure.
A proposed Environmental Protection Plan (TS Appendix B) was submitted by letter with WBN Unit 2 Developmental Revision H (Reference 4). The proposed Environmental Protection Plan (EPP) has been revised to address questions from the NRC staff provided in Reference 4 and a subsequent telephone conference between NRC and TVA staff on March 11, 2015. As discussed in the telephone conference, text from the WBN Unit 1 EPP Section 4.2, Unusual or Important Environmental Events, has been incorporated into the WBN Unit 2 EPP Section 4.1.1, Unusual or Important Environmental Events, in its entirety. Also as discussed during the telephone conference, Section 4.1.2 of the WBN Unit 1 EPP describing the maintenance of transmission line corridors is not included in the WBN Unit 2 EPP. The WBN Unit 2 EPP is provided in Attachment 2.
Changes Identified Since July 2015 Resulting in New Pages The resolution of the General Design Criteria (GDC) 5 issue has produced the largest number of changes to the TS and TS Bases. New TS 3.7.16, Component Cooling System (CCS) -
Shutdown and TS 3.7.17, Essential Raw Cooling Water (ERCW) System - Shutdown and the associated TS Bases were developed as part of this effort. These new TS and TS Bases sections were submitted on WBN Unit 1 as a license amendment request (Reference 6). The NRC conducted an audit of the WBN Unit 1 license amendment in support of resolution of GDC 5 issues. The questions and issues that the staff raised during the audit were addressed in Reference 7. Resolution of the GDC 5 issues resulted in modifications to the submitted CCS and ERCW TS and associated TS Bases.
TVA proposed to make additional changes to existing TS in response to discussions during the GDC 5 audit. A change has been made to the Mode of Applicability for Limiting Condition for Operation 3.3.2, Engineered Safety Feature Actuation Instrumentation, Function 6.f related to switching the source of water for the Auxiliary Feedwater System from the Condensate Storage 1


ENCLOSURE Watts Bar Nuclear Plant Unit 2 Summary of Technical Specification and Technical Requirements Manual Changes


Tank to the Essential Raw Cooling Water (ERCW) System. The table has been updated to note that this function is required to be operable in Mode 4 when the steam generators are being used for heat removal. The TS Bases for 3.3.2 Function 6.f was revised to describe the change in Mode applicability. Changes have been made to TS 3.4.6, RCS Loops - MODE 4, to add a requirement that the unit may not proceed to Mode 5 until at least seven hours have elapsed after entry into Mode 3 from Mode 1 or 2. These changes also require a minimum of two Reactor Coolant System (RCS) loops be operable, with one loop in operation for the first seven hours after entry into Mode 3 from Mode 1 or 2. This requirement is irrespective of the number of Residual Heat Removal Loops that are available or in operation. provides new TS and TS Bases pages for TS 3.3.2, 3.4.6, 3.7.16 and 3.7.17 to reflect the resolution of the GDC 5 audit issues.
Technical Requirements Manual Figure 3.1.6, Boric Acid Tank Limits, has been revised.
The figure submitted in TRM Developmental Revision E was the same figure that is in the WBN Unit 1 TRM. The WBN Unit 2 figure was revised to account for Unit 2 having the original steam generators, resulting in a smaller primary system volume than WBN Unit 1 has with the replacement steam generators. The revised figure is provided in Attachment 4 to this enclosure.

==References:==
: 1. TVA Letter to NRC, Watts Bar Nuclear Plant Unit 2 - Submittal of Developmental and Final Revision J of the Technical Specification & Technical Specification Bases, and Developmental and Final Revision E of the Technical Requirements Manual & Technical Requirements Manual Bases, dated July 6, 2015 [ML15187A461]
: 2. TVA Letter to NRC, Revised FSAR Section 6.2.1 Containment Functional Design, dated August 13, 2015 [ML115225A382]
: 3. TVA Letter to NRC, Watts Bar Nuclear Plant Unit 2 - Response to NRC Requests Related to Granting an Operating License, dated September 3, 2014 [ML14246A546]
: 4. TVA Letter to NRC, Watts Bar Nuclear Plant Unit 2 - Submittal of Developmental Revision H of the Unit 2 Technical Specification and Technical Specification Bases, dated December 12, 2013 [ML13357A048]
: 5. NRC Letter to TVA, Watts Bar Nuclear Plant, Unit 2 - Environmental Protection Plan (Non-Radiological) Technical Specification Review, dated February 2, 2015
[ML15015A477]
: 6. TVA Letter to NRC, Watts Bar Nuclear Plant Unit 1 - Application to Revise Technical Specifications for Component Cooling Water and Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13), dated June 17, 2015 [ML15170A474]
: 7. TVA Letter to NRC, Responses to NRC Audit Review Questions for Watts Bar Nuclear Plant Unit 1 Essential Raw Cooling Water and Component Cooling Water System License Amendment Request, dated August 28, 2015 [ML15243A044]
2


Attachment 1 Watts Bar Nuclear Plant Unit 2 Technical Specification Bases 3.4.176*7XEH,QWHJULW\ Marked-up

SG TUBE INTEGRITY B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.17 Steam Generator (SG) Tube Integrity BASES BACKGROUND         Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.
The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary systems pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, RCS Loops - MODES 1 and 2, LCO 3.4.5, RCS Loops - MODE 3, LCO 3.4.6, RCS Loops - MODE 4, and LCO 3.4.7, RCS Loops - MODE 5, Loops Filled.
SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.
The SG performance criteria are used to manage SG tube degradation.
Specification 5.7.2.12, Steam Generator (SG) Program, requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.7.2.12, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.7.2.12. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.
The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).
(continued)
Watts Bar - Unit 2                        B 3.4-90 (developmental)                                                                                A

SG TUBE INTEGRITY B 3.4.17 BASES (continued)
APPLICABLE         The steam generator tube rupture (SGTR) accident is the limiting design SAFETY             basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES           Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.
The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from of 150 gallons per day (gpd) per unfaulted steam generator and 1 gallon per minute (gpm) in the faulted steam generator. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.
Steam generator tube integrity satisfies Criterion 2 of the NRC Commission Interim Policy Statement (Ref. 7).
LCO                The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, an SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.
The tube-to-tubesheet weld is not considered part of the tube.
(continued)
Watts Bar - Unit 2                          B 3.4-91 (developmental)                                                                                 J

SG TUBE INTEGRITY B 3.4.17 BASES LCO                An SG tube has tube integrity when it satisfies the SG performance (continued)     criteria. The SG performance criteria are defined in Specification 5.7.2.12, Steam Generator Program, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation. Tube collapse is defined as, For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero. The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term significant is defined as An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established. For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.
The division between primary and secondary classifications will be based on detailed analysis and/or testing.
Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section III, Service Level A (normal operating conditions), and Service Level B (upset or abnormal conditions) transients included in the design specification.
This includes safety factors and applicable design basis loads based on ASME Code, Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).
(continued)
Watts Bar - Unit 2                         B 3.4-92 (developmental)                                                                                 A

SG TUBE INTEGRITY B 3.4.17 BASES LCO                The accident induced leakage performance criterion ensures that the (continued)     primary to secondary LEAKAGE caused by a design basis accident, other than an SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpm in the faulted SG. The accident induced leakage rate includes any primary-to-secondary LEAKAGE existing prior to the accident in addition to primary-to-secondary LEAKAGE induced during the accident.
The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, RCS Operational LEAKAGE, and limits primary-to-secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to an SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
APPLICABILITY      Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.
RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary-to-secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.
ACTIONS            The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.
(continued)
Watts Bar - Unit 2                         B 3.4-93 (developmental)                                                                                A

SG TUBE INTEGRITY B 3.4.17 BASES ACTIONS            A.1 and A.2 (continued)
Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if an SG tube that should have been plugged ,
has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.
A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection.
This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.
B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
Watts Bar - Unit 2                        B 3.4-94 (developmental)                                                                                 J

SG TUBE INTEGRITY B 3.4.17 BASES (continued)
SURVEILLANCE       SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the as found condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.
Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.
The Steam Generator Program defines the Frequency of SR 3.4.17.1.
The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.7.2.12 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.7.2.12 until subsequent inspections support extending the inspection interval.
(continued)
Watts Bar - Unit 2                        B 3.4-95 (developmental)                                                                                J

SG TUBE INTEGRITY B 3.4.17 BASES SURVEILLANCE       SR 3.4.17.2 REQUIREMENTS (continued)     During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
The tube plugging criteria delineated in Specification 5.7.2.12 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
Steam Generator tube plugging is only performed using approved plugging methods as described in the Steam Generator Program.
The Frequency of prior to entering MODE 4 following an SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.
REFERENCES         1.      NEI 97-06, Steam Generator Program Guidelines.
: 2.      10 CFR 50 Appendix A, GDC 19, Control Room.
: 3.      10 CFR 100, Reactor Site Criteria.
: 4.      ASME Boiler and Pressure Vessel Code, Section III, Subsection NB.
: 5.      Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976.
: 6.      EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines.
(continued)
Watts Bar - Unit 2                         B 3.4-96 (developmental)                                                                               J

SG TUBE INTEGRITY B 3.4.17 BASES REFERENCES         7. NRC Commission, Interim Staff Policy Statement on Technical (continued)        Specification Improvements for Nuclear Power Reactors, Federal Register 52 FR 3788, dated February 6, 1987; Westinghouse Owners Group (R.A. Newton) letter to NRC Document Control Desk, Westinghouse Owners Group MERITS Program Phase II, Task 5, Criteria Application Topical Report, dated November 12, 1987; NRC (T.E. Murley to W.S. Wilgus) letter, NRC Staff Review of Nuclear Steam Supply System Vendor Owners, Groups Application of the Commissions Interim Policy Statement Criteria to Standard Technical Specification, dated May 9, 1988, ADAMS Accession No. ML11264A057; TVA letter, Watts Bar Nuclear Plant (WBN) Unit 1 - Proposed Technical Specifications (TS), dated August 27, 1992, ADAMS Accession No. ML073200281; and NRC letter, Issuance of Facility Operating License No. NPF-90, Watts Bar Nuclear Plant, Unit 1(TAC M94025), dated February 7, 1996, ADAMS Accession No. ML052930169.
Watts Bar - Unit 2                     B 3.4-97 (developmental)                                                                          J

Attachment 1 Watts Bar Nuclear Plant Unit 2 7HFKQLFDO6SHFLILFDWLRQ%DVHV
6*7XEH,QWHJULW\&OHDQ

SG TUBE INTEGRITY B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.17 Steam Generator (SG) Tube Integrity BASES BACKGROUND         Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.
The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary systems pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, RCS Loops - MODES 1 and 2, LCO 3.4.5, RCS Loops - MODE 3, LCO 3.4.6, RCS Loops - MODE 4, and LCO 3.4.7, RCS Loops - MODE 5, Loops Filled.
SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.
The SG performance criteria are used to manage SG tube degradation.
Specification 5.7.2.12, Steam Generator (SG) Program, requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.7.2.12, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.7.2.12. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.
The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).
(continued)
Watts Bar - Unit 2                        B 3.4-90 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES (continued)
APPLICABLE         The steam generator tube rupture (SGTR) accident is the limiting design SAFETY             basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES           Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.
The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE of 150 gallons per day (gpd) per unfaulted steam generator and 1 gallon per minute (gpm) in the faulted steam generator. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.
Steam generator tube integrity satisfies Criterion 2 of the NRC Commission Interim Policy Statement (Ref. 7).
LCO                The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, an SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.
The tube-to-tubesheet weld is not considered part of the tube.
(continued)
Watts Bar - Unit 2                          B 3.4-91 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES LCO                An SG tube has tube integrity when it satisfies the SG performance (continued)      criteria. The SG performance criteria are defined in Specification 5.7.2.12, Steam Generator Program, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation. Tube collapse is defined as, For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero. The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term significant is defined as An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established. For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.
The division between primary and secondary classifications will be based on detailed analysis and/or testing.
Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section III, Service Level A (normal operating conditions), and Service Level B (upset or abnormal conditions) transients included in the design specification.
This includes safety factors and applicable design basis loads based on ASME Code, Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).
(continued)
Watts Bar - Unit 2                         B 3.4-92 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES LCO                The accident induced leakage performance criterion ensures that the (continued)      primary to secondary LEAKAGE caused by a design basis accident, other than an SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpm in the faulted SG. The accident induced leakage rate includes any primary-to-secondary LEAKAGE existing prior to the accident in addition to primary-to-secondary LEAKAGE induced during the accident.
The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, RCS Operational LEAKAGE, and limits primary-to-secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to an SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
APPLICABILITY      Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.
RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary-to-secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.
ACTIONS            The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.
(continued)
Watts Bar - Unit 2                         B 3.4-93 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES ACTIONS            A.1 and A.2 (continued)
Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if an SG tube that should have been plugged ,
has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.
A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection.
This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.
B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
Watts Bar - Unit 2                        B 3.4-94 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES (continued)
SURVEILLANCE       SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the as found condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.
Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.
The Steam Generator Program defines the Frequency of SR 3.4.17.1.
The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.7.2.12 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.7.2.12 until subsequent inspections support extending the inspection interval.
(continued)
Watts Bar - Unit 2                        B 3.4-95 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES SURVEILLANCE       SR 3.4.17.2 REQUIREMENTS (continued)      During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging.
The tube plugging criteria delineated in Specification 5.7.2.12 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
The Frequency of prior to entering MODE 4 following an SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.
REFERENCES         1.      NEI 97-06, Steam Generator Program Guidelines.
: 2.      10 CFR 50 Appendix A, GDC 19, Control Room.
: 3.      10 CFR 100, Reactor Site Criteria.
: 4.      ASME Boiler and Pressure Vessel Code, Section III, Subsection NB.
: 5.      Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976.
: 6.      EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines.
(continued)
Watts Bar - Unit 2                         B 3.4-96 Revision 0

SG TUBE INTEGRITY B 3.4.17 BASES REFERENCES         7. NRC Commission, Interim Staff Policy Statement on Technical (continued)         Specification Improvements for Nuclear Power Reactors, Federal Register 52 FR 3788, dated February 6, 1987; Westinghouse Owners Group (R.A. Newton) letter to NRC Document Control Desk, Westinghouse Owners Group MERITS Program Phase II, Task 5, Criteria Application Topical Report, dated November 12, 1987; NRC (T.E. Murley to W.S. Wilgus) letter, NRC Staff Review of Nuclear Steam Supply System Vendor Owners, Groups Application of the Commissions Interim Policy Statement Criteria to Standard Technical Specification, dated May 9, 1988, ADAMS Accession No. ML11264A057; TVA letter, Watts Bar Nuclear Plant (WBN) Unit 1 - Proposed Technical Specifications (TS), dated August 27, 1992, ADAMS Accession No. ML073200281; and NRC letter, Issuance of Facility Operating License No. NPF-90, Watts Bar Nuclear Plant, Unit 1(TAC M94025), dated February 7, 1996, ADAMS Accession No. ML052930169.
(continued)
Watts Bar - Unit 2                     B 3.4-97 Revision 0

Attachment 2 Watts Bar Nuclear Plant Unit 2 Technical Specifications Appendix B Environmental Protection Plan

APPENDIX B TO FACILITY OPERATING LICENSE ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
FOR WATTS BAR NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-391 TENNESSEE VALLEY AUTHORITY Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

WATTS BAR NUCLEAR PLANT UNIT 2 ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
TABLE OF CONTENTS Section                                                                                                                      Page 1.0            DEFINITIONS, ABBREVIATIONS, AND ACRONYMS .....................................1-1 2.0            LIMITING CONDITIONS FOR OPERATION ....................................................2-1 3.0            ENVIRONMENAL MONITORING .....................................................................3-1 3.1            Aquatic Monitoring .........................................................................................3-1 3.2            Terrestrial Monitoring .....................................................................................3-1 3.3            Maintenance of Transmission Line Corridors...............................................3-1 4.0            SPECIAL STUDIES AND REQUIREMENTS ....................................................4-1 4.1            Exceptional Occurrences ...............................................................................4-1 4.2            Special Studies ...............................................................................................4-2 5.0            ADMINISTRATIVE CONTROLS .......................................................................5-1 5.1            Responsibility .................................................................................................5-1 5.2            Review and Audit ............................................................................................5-1 5.3            Changes in Station Design or Operation.......................................................5-2 5.4            Station Reporting Requirements ...................................................................5-3 5.5            Changes in Environmental Protection Plan and Permits .............................5-6 5.6            Records Retention ..........................................................................................5-6 Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

1.0     DEFINITIONS, ABBREVIATIONS, AND ACRONYMS Annually                   Annually is once per calendar year at intervals of twelve (12) calendar months +/- 30 days Clean Water Act            Federal Water Pollution Control Act (FWPCA) as amended.
FES                        Final Environmental Statement (NUREG-0498) issued December 1978 by the NRC to the TVA (Control No. 7901100061).
FES Supplement 1           Final Environmental Statement (NUREG-0498 Supplement 1) issued April 1995 by the NRC to the TVA (ADAMS Accession No. ML081430592).
FES Supplement 2           Final Environmental Statement (NUREG-0498 Supplement 2, Vol. 1 &
Vol. 2) issued May 2013 by the NRC to the TVA (ADAMS Accession Nos.
ML13144A092 & ML13144A093).
FWS                        U.S. Fish and Wildlife Service NPDES Permit               NPDES permit is the National Pollutant Discharge Elimination System Permit No. TN0020168 issued by the U.S. Environmental Protection Agency to the Tennessee Valley Authority (TVA). This permit authorizes TVA to discharge controlled waste water, from the Watts Bar Plant Unit 2 into the Tennessee River.
NRC                        U.S. Nuclear Regulatory Commission Plant                      Plant refers to the Watts Bar Nuclear Plant, either Unit 1 or Unit 2.
Site                       Onsite includes any area within the property owned by the TVA specifically described in the WBN FES. Offsite includes all other areas.
Station                    Station refers to Watts Bar Nuclear Plant Unit 1 and Unit 2.
TVA                        Tennessee Valley Authority Unit                       Unit refers to Unit 1 or 2 (i.e., WBN Unit 1 or WBN Unit 2) of the Watts Bar Nuclear Plant, as defined by its usage.
WBN                        Watts Bar Nuclear Plant Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

2.0      LIMITING CONDITIONS FOR OPERATION (N/A)
None required 3.0      ENVIRONMENTAL MONITORING1 Environmental monitoring programs are conducted in accordance with the guidance and controls of agencies outside of the NRC. The NRC will relay on decisions made by the U.S.
Environmental Protection Agency, U.S. Fish and Wildlife Service, and the State of Tennessee for any requirements on environmental monitoring. Therefore, no specific environmental monitoring is required by the NRC under this EPP.
3.1      Aquatic Monitoring The certifications and permits required under the Clean Water Act provide mechanisms for protecting water quality and, indirectly, aquatic biota. The NRC will rely on the decision made by the U.S. Environmental Protection Agency and the State of Tennessee under the authority of the Clean Water Act for any requirements for aquatic monitoring.
3.2      Terrestrial Monitoring Terrestrial monitoring is not required.
_________________________
1 In consideration of the provisions of the Clean Water Act (33 USC &sect;1251, et seq.) and in the interest of avoiding duplication of effort, the conditions and monitoring requirements related to water quality and aquatic biota are specified in the National Pollution Discharge Elimination System (NPDES) Permit No. TN0020168 issued by the U.S. Environmental Protection Agency to the Tennessee Valley Authority (TVA). This permit authorizes TVA to discharge controlled waste water from the Watts Bar Nuclear Plant Unit 2 into the Tennessee River.
The Nuclear Regulatory Commission will be relying on the NPDES permit for protection of the aquatic environment from non-radiological effluents.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

4.0     SPECIAL STUDIES AND REQUIREMENTS 4.1     Exceptional Occurrences 4.1.1   Unusual or Important Environmental Events Requirements Any occurrence of an unusual event or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC within 24 hours followed by a written report in accordance with Subsection 5.4.2. If an event is reportable under 10 CFR 50.72, then a duplicate immediate report under this subsection is not required. However, a follow-up written report is required in accordance with Subsection 5.4.2.
The following are examples: excessive bird impact events, onsite plant or animal disease outbreaks, mortality of, or unusual occurrence involving any species protected by the Endangered Species Act of 1973 (ESA), the identification of any threatened or endangered species for which the NRC has not initiated consultation with the FWS, fish kills, increase in nuisance organisms or conditions in excess of levels anticipated in station environmental impact appraisals, and unanticipated or emergency discharge of waste water or any other chemical substance that exceeds the limits of, or is not authorized by, the NPDES permit and requires 24-hour notification to the State of Tennessee.
The licensee shall also notify the FWS Cookeville Field Office Field Supervisor or his designee when an unusual or important event results in the taking of, or could result in an adverse impact to, any species protected by the ESA. TVA should also notify the FWS law enforcement agent in Nashville, Tennessee if an unusual or important event involves the death, injury, or illness of any individual of a species protected by the ESA. Initial notification must be completed within 24 hours of the unusual or important event, followed by a written report per subsection 5.4.2 No routine monitoring programs are required to implement this condition.
Action Should an Unusual or Important Environmental Event occur, the licensee shall make a prompt report to the NRC in accordance with the provisions of Subsections 5.4.2.a and 5.4.2.c, or Subsection 5.4.2.d.
4.1.2   Exceeding Limits of Other Relevant Permits Requirement The licensee shall notify the NRC of occurrences in which the limits specified, in relevant permits and certificates issued by other Federal, State, and local governments are exceeded and which are reportable to those agencies. This requirement shall commence with the date of issuance of the operating license and continue for the life of the plant, unless changed in accordance with Subsection 5.5.1.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

Action The licensee shall make a report to the NRC in accordance with the provisions of Subsections 5.4.2.b and 5.4.2.c, or Subsection 5.4.2.d in the event of a reportable occurrence in which a limit specified in a relevant permit or certificate issued by another Federal, State, or local agency is exceeded.
4.2       Special Studies None required at the present time.
5.0       ADMINISTRATIVE CONTROLS 5.1       Responsibility The Plant Manager has responsibility for operating the plant in compliance with this Environmental Protection Plan.
5.2       Review and Audit The licensee shall provide for review and audit of compliance with the Environmental Protection Plan. The audits shall be conducted independently of the Individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.
5.2.1     Review The licensee is responsible for the review of procedures for meeting the Environmental Protection Plan.
The above mentioned review shall be conducted on the following:
A.        Proposed changes to the Environmental Protection Plan and evaluated impact of the change.
B.        Proposed changes to station operating procedures, which affect the environmental effects of the station.
C.        Proposed changes, construction, or modifications to station or unit equipment, or systems which might have an environmental impact, in order to determine the environmental impact of the change2.
___________________________________
2 Activities are excluded from this requirement if all measurable environmental effects are confined to on-site areas previouslydisturbedduringsitepreparationandplantconstruction.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

D.        All routine reports prior to their submittals to NRC (described in Subsection 5.4.1).
E.        All nonroutine reports prior to submittal of the written report (described in Subsections 5.4.a, b, and c).
F.        Investigations of all reported instances of noncompliance with the Environmental Protection Plan, associated corrective actions, and measures taken to prevent recurrence.
5.2.2    Audit The licensee shall conduct an audit on the environmental monitoring program. The audits shall be conducted independently of the individual or group responsible for performing the specific activity. Results of the audit activities shall be maintained and made available for inspection.
5.3      Changes in Station Design or Operation Changes in station design or operation may be made subject to the following conditions:
A.        The licensee may (1) make changes in the station design and operation, and (2) conduct tests and experiments not described in this document without prior Commission approval, unless the proposed change, test or experiment involves a change in the objectives of the Environmental Protection Plan3 and/or an unreviewed environmental question of significant impact.
B.        A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental impact statement as modified by testimony to the Atomic Safety and Licensing Board, supplements thereto, environmental impact appraisals, or in initial or final adjudicatory decisions; or (2) a matter not previously reviewed and evaluated in the documents specified in (1) of this section which may have a significant adverse environmental impact.
C.        The licensee shall maintain records of changes in facility design or operation made pursuant to this subsection. The licensee shall also maintain records of tests ad experiments carried out pursuant to paragraph A of this Subsection. These records shall include a written change, test, or experiment does not involve an unreviewed environmental question or substantive impact or constitute a change in the objectives of the Environmental Protection Plan. The licensee shall furnish to the Commission, annually or at such shorter intervals as may be specified in the license, a report containing description, analyses, interpretations, and evaluations of such changes, tests, and experiments.
___________________________________
3 This provision does not relieve the licensee of the requirements of 10 CFR 50.59.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

D.      Changes in the special studies, if required in Section 4.2, which affects sampling frequency, location, gear, or replication shall be reported to the NRC within 30 days after their implementation, unless otherwise reported in accordance with Subsection 5.4.2. These reports shall describe the changes made, the reasons for making the changes, and an evaluation of the effectiveness of the revised program in assessing environmental impacts.
5.4     Station Reporting Requirements 5.4.1   Routine Reports Annual Environmental Operation Report A WBN dual-unit report on the environmental monitoring program for the previous year shall be submitted to the NRC separate from other NRC reporting requirements within 90 days following each anniversary of issuance of the WBN Unit 1 operating license. The WBN Unit 1 operating license anniversary date is utilized as the basis for the WBN dual-unit anniversary date, since it was the basis for the initial and subsequent reports. The report shall include summaries, analyses, interpretations, and statistical evaluation of the results of the environmental monitoring required by special studies and requirements (Section 4) for the report period, including a comparison with preoperational studies, operating controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the station operation on the environment. If harmful effects or evidence of irreversible damage are suggested by the monitoring programs, the licensee shall provide a more detailed analysis of the data and a proposed course of action to alleviate the problem.
For those programs concerned with water quality or protection of aquatic biota, which are regulated under the Clean Water Act, the requirements of this section shall be satisfied by submitting to the NRC copies of the reports as required by the NPDES permit (or otherwise required pursuant to the Clean Water Act), and in accordance with the frequency, content and schedules set forth by the agencies responsible for implementing the Clean Water Act.
In the event that some results are not available by the report date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The Annual Report shall also include a summary of:
: 1.       All Environmental Protection Plan noncompliances and the corrective actions taken to remedy them.
: 2.       Changes made to applicable State and Federal permits and certifications.
: 3.       Changes to station design which could involve a significant environmental impact or change the findings of the FES.
: 4.       All nonroutine reports submitted per Environmental Protection Plan Section 4.1.
: 5.       Changes in the approved Environmental Protection Plan.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

5.4.2       Nonroutine Reports A report shall be submitted in the event that an Unusual or Important Environmental Event, as specified in Subsection 4.1.1 occurs, or if another relevant permit is violated as specified in Subsection 4.1.2. The schedule and content for these nonroutine reports are described below.
5.4.2.a     Prompt Report Those events specified as requiring prompt reporting shall be reported within 24 hours by telephone, telegraph, or facsimile transmission to the NRC followed by a written report to the NRC within 30 days.
5.4.2.b     Thirty Day Report Those events not requiring a prompt report as described in Subsection 5.4.2.a shall be reported to the NRC within 30 days of their occurrence.
5.4.2.c     Content of Nonroutine Reports Written 30-day reports and, to the extent possible, the preliminary telephone, telegraph, or facsimile reports shall (a) describe, analyze, and evaluate the occurrence, including extent and magnitude of the impact, (b) describe the cause of the occurrence, (c) indicate the action taken to correct the reporting occurrence, and (d) indicate the corrective action taken (including any significant changes made in procedures) to preclude repetition of the occurrence and to prevent similar occurrences involving similar components or systems.
5.4.2.d     Exceptions for Matters Regulated Under the Clean Water Act For matters regulated under the Clean Water Act, the report schedules and content requirements described in Subsections 5.4.2.a, 5.4.2.b, and 5.4.2.c shall be satisfied by submitting, to the NRC copies of the reports as required b the NPDES permit (or other regulations pursuant to the Clean Water Act) and in accordance with the schedules and content requirements imposed thereby.
5.5     Changes in the Environmental Protection Plan and Permits 5.5.1   Changes in the Environmental Protection Plan Requests for change to the Environmental Protection Plan shall be submitted to the NRC for review and authorization per 10 CFR 50.90. The request shall include an evaluation of the environmental impact of the proposed change and a supporting justification. Implementation of such requested changes to the Environmental Protection Plan shall not commence prior to incorporation by the NRC of the specifications in the license.
5.5.2   Changes in Permits and Certifications Changes and additions to required Federal (other than NRC), State, local, and regional authority permits and certificates for the protection of the environment shall be reported to the NRC within 30 days. In the event that the licensee initiates or becomes aware of a request for changes to any of the water quality requirements, limits, or values stipulated in any certification or permit issued pursuant to the Clean Water Act, the NRC shall be notified within 30 days.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days. If, as a result of the appeal process, the permit or certification requirements are changed, the change shall be dealt with as described in the previous paragraph of this section.
5.6     Records Retention Records and logs relative to the environmental aspects of station operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to NRC on request.
5.6.1   The following records shall be retained for the life of the station:
(a)       Record of changes to the Environmental Protection Plan including, when applicable, records of NRC approval of such changes.
(b)       Record of modifications to station structures, systems, and components determined to potentially affect the continued protection of the environment.
(c)      Record of changes to permits and certifications required by Federal (other than the NRC), State, local, and regional authorities for the protection of the environment.
(d)      Routine reports submitted to the NRC.
5.6.2   Records of the following shall be retained for a minimum of six (6) years:
(a)      Review and audit activities.
(b)      Events, and the reports thereon, which are the subjects of non-routine reports to the NRC.
5.6.3   Records associated with requirements of Federal (other than the NRC), State, local, and regional authorities permits and certificates for the protection of the environment shall be retained for the period established by the respective permit or certificate.
Watts Bar Nuclear Plant - Unit 2 Environmental Protection Plan - Appendix B

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:DWWV%DU8QLW                                             
5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES

APPLICABLE         f. Auxiliary Feedwater - Pump Suction Transfer on Suction SAFETY                Pressure - Low
: ANALYSES, LCO, and              A low pressure signal in the AFW pump suction line protects APPLICABILITY         the AFW pumps against a loss of the normal supply of water (continued)           for the pumps, the CST. Three pressure switches are located on each motor driven AFW pump suction line from the CST. A low pressure signal sensed by two switches of a set will cause the emergency supply of water for the respective pumps to be aligned. ERCW (safety grade) is then lined up to supply the AFW pumps to ensure an adequate supply of water for the AFW System to maintain at least one of the SGs as the heat sink for reactor decay heat and sensible heat removal.
Since the detectors are located in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the NTSP reflects only steady state instrument uncertainties.
Thisese Functions must be OPERABLE in MODES 1, 2, and 3, and 4, when the steam generators are relied on to remove decay heat from the reactor, to ensure a safety grade supply of water for the AFW System to maintain the SGs as the heat sink for the reactor. Thisese Functions does not have to be OPERABLE in MODES 5 and 6 because there is not enough heat being generated in the reactor to require the SGs as a heat sink.
(continued)
Watts Bar - Unit              B 3.3-94 5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            When the number of inoperable channels in a trip function exceed those (continued)        specified in one or other related Conditions associated with a trip function, then the unit is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation.
A.1 Condition A applies to all ESFAS protection functions.
Condition A addresses the situation where one or more channels or trains for one or more Functions are inoperable at the same time. The Required Action is to refer to Table 3.3.2-1 and to take the Required Actions for the protection functions affected. The Completion Times are those from the referenced Conditions and Required Actions.
B.1, B.2.1 and B.2.2 Condition B applies to manual initiation of:
x    SI; x    Containment Spray; x    Phase A Isolation; and x    Phase B Isolation.
Condition B also applies to the Auxiliary Feedwater Pump Suction Transfer on Suction Pressure - Low.
For the manual initiation Functions, tThis action addresses the train orientation of the SSPS for the functions listed above. For the AFW System pump suction transfer channels, this action recognizes that placing a failed channel in trip during operation is not necessarily a conservative action. Spurious trip of this function could align the AFW System to a source that is not immediately capable of supporting pump suction. If a channel or train is inoperable, 48 hours is allowed to return it to an OPERABLE status. Note that for containment spray and Phase B isolation, failure of one or both channels in one train renders the train inoperable. Condition B, therefore, encompasses both situations.
(continued)
Watts Bar - Unit                          B 3.3-10
                                                                                    5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            B.1, B.2.1 and B.2.2 (continued)
For the manual initiation Functions, Tthe specified Completion Time is reasonable considering that there are two automatic actuation trains and another manual initiation train OPERABLE for each Function, and the low probability of an event occurring during this interval. For the AFW System pump suction transfer channels, the specified Completion Time is reasonable considering the nature of this Function, the available redundancy, and the low probability of an event occurring during this interval. If the channel or train cannot be restored to OPERABLE status, the plant must be placed in a MODE in which the LCO does not apply.
This is done by placing the plant in at least MODE 3 within an additional 6 hours (54 hours total time) and in MODE 5 within an additional 30 hours (84 hours total time). The allowable 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. For the AFW System pump suction transfer channels, aligning the RHR System for decay heat removal, so that the steam generators are not relied on for heat removal, places the plant in a MODE in which the LCO no longer applies. Therefore, per LCO 3.0.2, completion of the Required Action to place the unit in MODE 5 is not required.
For the manual initiation Functions, Tthe allowance of 48 hours is justified in Reference 7.
(continued)
Watts Bar - Unit                        B 3.3-10
                                                                                   5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            F.1, F.2.1, and F.2.2 (continued)
Condition F applies to:
x    Manual Initiation of Steam Line Isolation; x    Loss of Offsite Power; x    Auxiliary Feedwater Pump Suction Transfer on Suction Pressure -
Low; and x    P-4 Interlock.
For the Manual Initiation and the P-4 Interlock Functions, this action addresses the train orientation of the SSPS. For the Loss of Offsite Power Function, this action recognizes the lack of manual trip provision for a failed channel. For the AFW System pump suction transfer channels, this action recognizes that placing a failed channel in trip during operation is not necessarily a conservative action. Spurious trip of this function could align the AFW System to a source that is not immediately capable of supporting pump suction. If a train or channel is inoperable, 48 hours is allowed to return it to OPERABLE status. The specified Completion Time is reasonable considering the nature of these Functions, the available redundancy, and the low probability of an event occurring during this interval. If the Function cannot be returned to OPERABLE status, the plant must be placed in MODE 3 within the next 6 hours and MODE 4 within the following 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and without challenging plant systems. In MODE 4, the plant does not have any analyzed transients or conditions that require the explicit use of the protection functions noted above.
(continued)
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0DLQ)HHGZDWHU                                                                     65
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           /HYHO/RZ                                                                        65
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           /HYHO+LJK                                                                       65
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:DWWV%DU8QLW                                             
5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES

APPLICABLE         f. Auxiliary Feedwater - Pump Suction Transfer on Suction SAFETY                Pressure - Low
: ANALYSES, LCO, and              A low pressure signal in the AFW pump suction line protects APPLICABILITY         the AFW pumps against a loss of the normal supply of water (continued)           for the pumps, the CST. Three pressure switches are located on each motor driven AFW pump suction line from the CST. A low pressure signal sensed by two switches of a set will cause the emergency supply of water for the respective pumps to be aligned. ERCW (safety grade) is then lined up to supply the AFW pumps to ensure an adequate supply of water for the AFW System to maintain at least one of the SGs as the heat sink for reactor decay heat and sensible heat removal.
Since the detectors are located in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the NTSP reflects only steady state instrument uncertainties.
This Function must be OPERABLE in MODES 1, 2, 3, and 4, when the steam generators are relied on to remove decay heat from the reactor, to ensure a safety grade supply of water for the AFW System to maintain the SGs as the heat sink for the reactor. This Function does not have to be OPERABLE in MODES 5 and 6 because there is not enough heat being generated in the reactor to require the SGs as a heat sink.
(continued)
Watts Bar - Unit              B 3.3-94
                                                                        5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            When the number of inoperable channels in a trip function exceed those (continued)        specified in one or other related Conditions associated with a trip function, then the unit is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation.
A.1 Condition A applies to all ESFAS protection functions.
Condition A addresses the situation where one or more channels or trains for one or more Functions are inoperable at the same time. The Required Action is to refer to Table 3.3.2-1 and to take the Required Actions for the protection functions affected. The Completion Times are those from the referenced Conditions and Required Actions.
B.1, B.2.1 and B.2.2 Condition B applies to manual initiation of:
x    SI; x    Containment Spray; x    Phase A Isolation; and x    Phase B Isolation.
Condition B also applies to the Auxiliary Feedwater Pump Suction Transfer on Suction Pressure - Low.
For the manual initiation Functions, this action addresses the train orientation of the SSPS for the functions listed above. For the AFW System pump suction transfer channels, this action recognizes that placing a failed channel in trip during operation is not necessarily a conservative action. Spurious trip of this function could align the AFW System to a source that is not immediately capable of supporting pump suction. If a channel or train is inoperable, 48 hours is allowed to return it to an OPERABLE status. Note that for containment spray and Phase B isolation, failure of one or both channels in one train renders the train inoperable. Condition B, therefore, encompasses both situations.
(continued)
Watts Bar - Unit                          B 3.3-10
                                                                                    5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            B.1, B.2.1 and B.2.2 (continued)
For the manual initiation Functions, the specified Completion Time is reasonable considering that there are two automatic actuation trains and another manual initiation train OPERABLE for each Function, and the low probability of an event occurring during this interval. For the AFW System pump suction transfer channels, the specified Completion Time is reasonable considering the nature of this Function, the available redundancy, and the low probability of an event occurring during this interval. If the channel or train cannot be restored to OPERABLE status, the plant must be placed in a MODE in which the LCO does not apply.
This is done by placing the plant in at least MODE 3 within an additional 6 hours (54 hours total time) and in MODE 5 within an additional 30 hours (84 hours total time). The allowable 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. For the AFW System pump suction transfer channels, aligning the RHR System for decay heat removal, so that the steam generators are not relied on for heat removal, places the plant in a MODE in which the LCO no longer applies. Therefore, per LCO 3.0.2, completion of the Required Action to place the unit in MODE 5 is not required.
For the manual initiation Functions, the allowance of 48 hours is justified in Reference 7.
(continued)
Watts Bar - Unit                        B 3.3-10
                                                                                   5HYLVLRQ

ESFAS Instrumentation B 3.3.2 BASES ACTIONS            F.1, F.2.1, and F.2.2 (continued)
Condition F applies to:
x    Manual Initiation of Steam Line Isolation; x    Loss of Offsite Power; and x    P-4 Interlock.
For the Manual Initiation and the P-4 Interlock Functions, this action addresses the train orientation of the SSPS. For the Loss of Offsite Power Function, this action recognizes the lack of manual trip provision for a failed channel. If a train or channel is inoperable, 48 hours is allowed to return it to OPERABLE status. The specified Completion Time is reasonable considering the nature of these Functions, the available redundancy, and the low probability of an event occurring during this interval. If the Function cannot be returned to OPERABLE status, the plant must be placed in MODE 3 within the next 6 hours and MODE 4 within the following 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power in an orderly manner and without challenging plant systems. In MODE 4, the plant does not have any analyzed transients or conditions that require the explicit use of the protection functions noted above.
(continued)
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'HYHORSPHQWDO                                                                                                               -

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'HYHORSPHQWDO                                                                           -

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'HYHORSPHQWDO                                                                             -

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CCS - Shutdown 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Component Cooling System (CCS) - Shutdown LCO 3.7.16          Two CCS trains shall be OPERABLE with one pump powered from Train A and aligned to the Train A header, and two pumps powered from Train B and aligned to the Train B header.
APPLICABILITY:      MODES 4 and 5.
                    ----------------------------------------------NOTE------------------------------------------------
This LCO is not applicable more than 48 hours after entry into MODE 3 from MODE 1 or 2.
                    ------------------------------------------------------------------------------------------------------
ACTIONS CONDITION                              REQUIRED ACTION                           COMPLETION TIME A.      One CCS train                A.1      Be in MODE 5.                              24 hours inoperable in MODE 4.
AND Complying with Required Actions to be in MODE 5.
B.      One CCS train                B.1      Verify two OPERABLE                        Once per 12 hours inoperable in                          reactor coolant system (RCS)
MODE 4 for reasons                     loops and one RCS loop in other than                             operation.
Condition A.
AND B.2      Verify Tavg > 200&#xba;F.                       Once per 12 hours (continued)
Watts Bar-Unit                                      3.7-33
'HYHORSPHQWDO                                                                                                          -

CCS - Shutdown 3.7.16 ACTIONS (continued)
CONDITION                   REQUIRED ACTION                        COMPLETION TIME C.      Two CCS trains         C.1 ---------------NOTES--------------
inoperable in MODE 4.      1. LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one CCS train is restored to an OPERABLE status.
: 2. Enter Conditions and Required Actions of LCO 3.4.6, RCS Loops -
MODE 4, for residual heat removal (RHR) loops made inoperable by CCS.
                                     ----------------------------------------
Initiate action to restore one          Immediately CCS train to OPERABLE status.
D.      One or more CCS        D.1   ---------------NOTE----------------
train(s) inoperable in       Enter applicable Conditions MODE 5.                      and Required Actions of LCO 3.4.7, "RCS Loops -
MODE 5, Loops Filled," for RHR loops made inoperable by CCS.
                                     ----------------------------------------
Initiate action to restore CCS          Immediately train(s) to OPERABLE status.
Watts Bar-Unit                          3.7-34
'HYHORSPHQWDO                                                                                    J

CCS - Shutdown 3.7.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                 FREQUENCY SR 3.7.16.1     Verify correct breaker alignment and indicated power 12 hours available to the required pump(s) that is not in operation.
SR 3.7.16.2     Verify two CCS pumps are aligned to CCS Train B. 12 hours
 
Watts Bar-Unit                             3.7-35
'HYHORSPHQWDO                                                                           -

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'HYHORSPHQWDO                                                                                     -

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'HYHORSPHQWDO                                                                      -

ERCW - Shutdown B 3.7.17

B 3.7 PLANT SYSTEMS B 3.7.17 Essential Raw Cooling Water (ERCW) System BASES BACKGROUND           The general description of ERCW is provided in TS Bases 3.7.8, Essential Raw Cooling Water (ERCW) System. The descriptions of Applicable Safety Analyses, LCOs, Applicability, ACTIONS and Surveillance Requirements for applicable MODES are also described in TS Bases 3.7.8. The following discussion applies to the specific Applicability in TS 3.7.17 during the first 48 hours after shut down when the Residual Heat Removal (RHR) System is being used for residual and decay heat removal. The ERCW System provides a heat sink for the removal of process and operating heat from safety related components during a design basis accident (DBA) or transient. During normal operation, and a normal shutdown, the ERCW System also provides this function for various safety related and non-safety related components.
The major post-accident heat load on the ERCW System is the Component Cooling System (CCS) heat exchangers (HXs), which are used to cool RHR and the containment spray HXs. The major heat load on the ERCW System when a unit is shut down on RHR is the CCS HX associated with the train(s) of RHR in service.
During a normal shutdown, decay heat removal is via the reactor coolant system (RCS) loops until sometime after the unit has been cooled down to RHR entry conditions (Tcold < 350&#xba;F). Therefore, as LCO 3.7.17 becomes Applicable (entry into Mode 4) the RCS loops are still OPERABLE. After the RHR System is aligned as the principle method of decay heat removal, the heat loads on the ERCW System are increased.
Normally, two ERCW pumps are sufficient to handle the cooling needs for maintaining one unit in normal operation while mitigating a DBA on the other unit. However, in the unlikely event of a loss of coolant accident (LOCA) on Unit 2 with a concurrent loss of offsite power and a single failure that results in the loss of both Train A or both Train B 6.9 kV shutdown boards while Unit 1 is on RHR shutdown cooling and has been shutdown for less than 48 hours, three ERCW pumps may be required.
This LCO controls the availability of ERCW pumps necessary to support mitigation of a LOCA on Unit 2 when Unit 1 has been shut down for 48 hours and is utilizing RHR for heat removal.
Additional information about the design and operation of the ERCW System, along with a list of the components served, is presented in the FSAR, Section 9.2.1 (Ref. 1).
(continued)
Watts Bar-Unit                               B 3.7-85
'HYHORSPHQWDO                                                                                      -

ERCW - Shutdown B 3.7.17 BASES (continued)

APPLICABLE        The design basis of the ERCW System is for one ERCW train, in SAFETY            conjunction with the CCS and a 100% capacity Containment Spray ANALYSES          System and RHR, to remove core decay heat following a design basis LOCA as discussed in the FSAR, Section 9.2.1 (Ref. 1). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System (RCS) by the Emergency Core Cooling System (ECCS) pumps.
The ERCW System is designed to perform its function with a single failure of any active component, assuming a loss of offsite power.
The ERCW System, in conjunction with the CCS, also cools the unit, as discussed in the FSAR, Section 5.5.7 (Ref. 2) from RHR entry conditions to MODE 5 during normal and post accident operations. The time required to enter MODE 5 is a function of the number of CCS and RHR System trains that are operating. One ERCW train is sufficient to remove heat during subsequent operations in MODES 5 and 6. This assumes a maximum ERCW inlet temperature of 85&#xba;F occurring simultaneously with maximum heat loads on the system. In the first 48 hours after the shutdown of Unit 1 assuming a DBA LOCA on Unit 2 with the loss of offsite power and the concurrent loss of two 6.9 kV shutdown boards on the same power train as a single failure. Three ERCW pumps are required to provide the heat removal capacity assumed in the safety analysis for Unit 2 while continuing the cooldown of Unit 1.
ERCW - Shutdown satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
LCO               This LCO provides ERCW train OPERABILITY requirements beyond the requirements of LCO 3.7.8 during the first 48 hours after reactor shutdown, when the heat loads are at sufficiently high levels that the normal pump requirement of two ERCW pumps on one train may not be sufficient to support shutdown cooling of Unit 1, concurrent with a LOCA on Unit 2, an assumed loss of offsite power, and a single failure that affects both 6.9 kV shutdown boards in one power train.
Two ERCW trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to support a cooldown to MODE 5.
An ERCW train is considered OPERABLE during the first 48 hours after shutdown when:
: a. Two pumps per train, aligned to separate shutdown boards, are OPERABLE; and (continued)
Watts Bar-Unit                           B 3.7-86
'HYHORSPHQWDO                                                                                    -

ERCW - Shutdown B 3.7.17 BASES

LCO              b.      One additional Train A pump and one additional Train B pump are (continued)            capable of being aligned to their respective Unit 1 6.9 kV shutdown board (1A-A and 1B-B) and manually placed in service.
APPLICABILITY    Prior to aligning the RHR System for RCS heat removal in MODE 4, one additional ERCW pump must be capable of being powered by its respective Unit 1 6.9 kV shutdown board (1A-A and 1B-B) and manually placed in service to ensure adequate heat removal capability.
The Applicability is modified by a Note stating the LCO does not apply after the initial 48 hours after the unit enters MODE 3 from MODE 1 or MODE 2. Following extended operation in MODE 1, the heat loads are at sufficiently high levels that the normal pump requirement of LCO 3.7.8 for two ERCW pumps may not be sufficient to support shutdown cooling of Unit 1, concurrent with a design basis LOCA on Unit 2 with loss of offsite power and a single failure of Train A power to 6.9 kV Shutdown Boards 1A-A and 2A-A. However, after the initial 48 hours following shutdown of the unit, the heat removal capability of both units is within the capabilities of the ERCW System without the need for an additional ERCW pump in each train.
ACTIONS          A.1 In MODE 4, if one ERCW train is inoperable, and the unit is required to be placed in MODE 5 to comply with Required Actions, action must be taken to place the unit in MODE 5 within 24 hours. When the Required Actions of an LCO direct the unit to be placed in MODE 5, either a loss of safety function has occurred or the Required Action and Completion Time for restoring a safety-related component has not been met. Therefore, it is prudent to place the unit in a condition of lower energy with a lower potential for a postulated event. In this Condition, the remaining OPERABLE ERCW train is adequate to perform the heat removal function. The 24 hour Completion Time is consistent with LCO 3.4.6, RCS Loops - MODE 4, Required Action B.1 for the Condition of one required RHR loop inoperable and no RCS loops OPERABLE.
B.1 and B.2 In MODE 4, if one ERCW train is inoperable, and the unit is not required to be placed in MODE 5 to comply with Required Actions, actions are taken to verify LCO 3.4.6 is being met with two OPERABLE RCS loops with one loop in operation, and that the unit remains in MODE 4 (Tavg > 200&#xba;F). These actions indicate the preference to maintain the unit (continued)
Watts Bar-Unit                           B 3.7-87
'HYHORSPHQWDO                                                                                   -

ERCW - Shutdown B 3.7.17 BASES

ACTIONS          B.1 (continued) in a condition with multiple methods of decay heat removal available, i.e.,
maintain the unit in MODE 4 with two RCS loops operable in addition to the remaining OPERABLE RHR loop. This action precludes entry into the LCO 3.4.6 Actions, as LCO 3.4.6 is met with two OPERABLE RCS loops and one RCS loop in operation. This Action is conservative to the Required Actions of LCO 3.4.6 when there are two OPERABLE RCS loops.
Maintaining the unit in MODE 4 with additional methods of decay heat removal available minimizes the likelihood of a situation where the decay heat and residual heat of the unit exceeds the capability of the available RHR loop resulting in the possibility of an unintentional MODE change.
The Frequency of once per 12 hours ensures that the systems being relied on for heat removal are operating properly and are maintaining the unit in MODE 4. The 12 hour Frequency is reasonable, considering the low probability of a change in system operation during this time period.
If the Required Actions and Completion Times of Condition B are not met, no actions are specified. Therefore, LCO 3.0.3 applies, requiring the unit to be placed in MODE 5 in 37 hours. With one ERCW train inoperable and Required Actions require the unit to be placed in MODE 5, Condition A applies, requiring the unit to be placed in MODE 5 in 24 hours. This Action is consistent with the Required Actions of LCO 3.4.6 Condition B (no OPERABLE RCS loops and one inoperable RHR loop).
Although LCO 3.7.17 provides requirements in addition to those of LCO 3.7.8, the additional requirements of LCO 3.7.17 are not required for DG OPERABILITY. There is sufficient flow to the DGs from ERCW without a third ERCW in each train to support DG OPERABILITY.
Although the requirement of LCO 3.7.17 may not be met (i.e., a third pump capable of being aligned to each ERCW Train) the requirement of LCO 3.7.8 is still met. If the requirement of LCO 3.7.8 is not met, the Actions of LCO 3.7.8 include the requirement to enter the Conditions and Required Actions of LCO 3.8.1 for DGs made inoperable by ERCW.
C.1 In MODE 4, if two ERCW trains are inoperable, immediate action must be taken to restore one of the ERCW trains to an OPERABLE status, as no ERCW train is available to support the heat removal function. Required Action C.1 is consistent with LCO 3.4.6, RCS Loops - MODE 4, (continued)
Watts Bar-Unit                         B 3.7-88
'HYHORSPHQWDO                                                                                 -

ERCW - Shutdown B 3.7.17 BASES

ACTIONS          C.1 (continued)
Required Action D.1 for the Condition of required RCS or RHR loops inoperable and no RCS or RHR loop in operation.
Required Action C.1 is modified by two Notes. Note 1 indicates that all required MODE changes or power reductions are suspended until one ERCW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the plant into a less safe condition. Note 2 indicates that the applicable Conditions and Required Actions of LCO 3.4.6 be entered for RHR loops made inoperable by the inoperable ERCW trains. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
D.1 Required Action D.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.7, RCS Loops - MODE 5, Loops Filled, be entered for RHR loops made inoperable by one or more inoperable ERCW train(s).
This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
In MODE 5, if one or more ERCW train(s) is inoperable, action must be initiated immediately to restore the ERCW train(s) to an OPERABLE status to restore heat removal paths. The immediate Completion Time reflects the importance of maintaining the capability of heat removal.
SURVEILLANCE     SR 3.7.17.1 REQUIREMENTS Verifying the availability of the ERCW pumps provides assurance that adequate ERCW flow is provided for heat removal. Verification that each required ERCW pump that is not in operation is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal. Verification is performed by verifying proper breaker alignment and power available to the ERCW pump(s). The ERCW pump Interlock Bypass Switches do not need to be in Bypass in order to meet this SR. The associated ERCW pump Interlock Bypass Switch is positioned by procedure when the third ERCW pump in the respective train is required to be started. The 12 hour Frequency is based on engineering judgment.
REFERENCES       1.      Watts Bar FSAR, Section 9.2.1, "Essential Raw Cooling Water."
: 2.      Watts Bar FSAR, Section 5.5.7, "Residual Heat Removal System."
Watts Bar-Unit                          B 3.7-89
'HYHORSPHQWDO                                                                                     -

Attachment 4 Watts Bar Nuclear Plant Unit 2 Technical Requirements Manual Bases Section 3.1.6, BorLF Acid Tank/LPLWV Clean

Borated Water Sources, Operating TR 3.1.6 TECHNICAL REQUIREMENTS FIGURE 3.1.6 BORIC ACID TANK LIMITS BASED ON RWST BORON CONCENTRATION 11500 MINIMUM INDICATED STORED VOLUME IN BAT - GALLONS REGION OF ACCEPTABLE OPERATION 11000                             RWST = 3100 ppm B RWST = 3200 ppm B 10500 RWST  3300 ppm B 10000 9500 9000 REGION OF UNACCEPTABLE OPERATION 6120 ppm MINIMUM           6990 ppm MAXIMUM 8500 Indicated values include additional 1,910 gal compensation for unusable vol, instrument error, and conservatism.
8000 6000 6100 6200 6300 6400 6500 6600 6700 6800 6900 7000 7100 CONCENTRATION IN BAT - PPM BORON RWST Concentration 3100 ppm B         3200 ppm B         3300 ppm B Watts Bar - Unit 2                                                                           3.1-2 Technical Requirements                                                                                                                 Revision 0}}