Text

Technical Specifications Proposed Change No. 306 Eliminate Certain ESF Requirements During Movement of Irradiated Fuel
	ML13323A518
Person / Time
Site:	Pilgrim, Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/14/2013
From:	Wamser C; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 13-097, TAC MC2705
	Download: ML13323A518 (44)
	v • d • e

SEntergy Entergy Nuclear Operations, Inc.

Vermont Yankee 320 Governor Hunt Rd Vernon, VT 05354 Tel 802 257 7711 Christopher J. Wamser Site Vice President BVY 13-097 November 14, 2013 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Technical Specifications Proposed Change No. 306 Eliminate Certain ESF Requirements during Movement of Irradiated Fuel Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

REFERENCES:

1.

Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

2.

Letter, USNRC to Entergy Nuclear Operations, Inc., "Pilgrim Nuclear Power Station - Issuance of Amendment RE: Alternative Source Term for the Fuel Handling Accident Dose Consequences (TAC No.

MC2705)," dated April 28, 2005

Dear Sir or Madam:

In accordance with 10CFR50.90, Entergy Nuclear Operations, Inc. (ENO) is proposing an amendment to Appendix A, Technical Specifications (TS), of the Vermont Yankee Nuclear Power Station (VY) Renewed Facility Operating License, DPR-28, to change the requirements associated with handling irradiated fuel and performing core alterations. Specifically, the changes would eliminate operability requirements for secondary containment when handling sufficiently decayed irradiated fuel or a fuel cask and while performing core alterations.

The changes proposed in this application are consistent with Technical Specification Task Force Traveler 51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations." In support of these changes, ENO has completed a calculation using a selective implementation of alternate source term guidance for evaluating the potential dose consequences of a fuel handling accident (FHA). This calculation uses the guidelines detailed in Regulatory Guide 1.183 (Reference 1). The calculation demonstrates that radiological doses at the exclusion area boundary, low population zone and in the control room from a FHA after a certain amount of time following shutdown are within allowable limits without crediting secondary containment operability and operation of the standby gas treatment system. The proposed changes are consistent with similar changes approved for Pilgrim Nuclear Station (Reference 2).

ENO has reviewed the proposed amendment in accordance with 10CFR50.92 and concludes it does not involve a significant hazards consideration.

-AUl

BVY 13-097 / Page 2 of 3 In accordance with 10CFR50.91, a copy of this application, with attachments, is being provided to the State of Vermont, Department of Public Service. to this letter provides a detailed description and evaluation of the proposed change. contains a markup of the current TS and Bases pages. Attachment 3 contains the retyped TS and Bases pages. Bases changes are provided for information only. Attachment 4 contains the new fuel handling accident calculation.

ENO requests NRC review and approval of the proposed changes by December 1, 2014 to support the anticipated fuel handling operations following final shutdown of the VY reactor, which is expected to occur in the fourth quarter of 2014. ENO requests a 60 day implementation period for the approved changes.

New regulatory commitments are described in Attachment 5 of this letter.

If you have any questions on this transmittal, please contact Mr. Coley Chappell at 802-451-3374.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 14, 2013.

Si nce r

CJW/pic Attachments:

1. Description and Evaluation of the Proposed Changes

2. Markup of the Current Technical Specifications and Bases Pages

3. Retyped Technical Specifications and Bases Pages

4. AREVA Document No. 32-9145461-001, "VYNPP - Re-analysis of AST/FHA Radiological Consequences with Open Containment"

5. List of Regulatory Commitments cc:

Mr. William M. Dean Region 1 Administrator U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Douglas V. Pickett, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08C2A Washington, DC 20555

BVY 13-097 / Page 3 of 3 cc list continued:

USNRC Resident Inspector Vermont Yankee Nuclear Power Station 320 Governor Hunt Road Vernon, VT 05354 Mr. Christopher Recchia Commissioner VT Department of Public Service 112 State Street, Drawer 20 Montpelier, VT 05620-2601

BVY 13-097 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Change 306 Description and Evaluation of Proposed Changes

BVY 13-097 / Attachment 1 / Page 1 of 17

1.

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Renewed Facility Operating License (OL) DPR-28 for Vermont Yankee Nuclear Power Station (VY).

The proposed change would revise the Technical Specifications (TS) contained in Appendix A of the VY OL to change the requirements associated with handling irradiated fuel, or a fuel cask, and while performing core alterations for selected specifications. Specifically, the changes would eliminate operability requirements for secondary containment (TS 3.7.C), the Standby Gas Treatment (SGT) system (TS 3.7.B.4), and reactor building ventilation isolation and SGT system initiation instrumentation (TS Table 3.2.3) when handling sufficiently decayed irradiated fuel, or a fuel cask, and while performing core alterations. The proposed changes would require subject systems to be operable only during movement of "recently" irradiated fuel assemblies in secondary containment and during operations with the potential to drain the reactor vessel (OPDRV). The period of sufficient radioactive decay was determined to be 13 days and will be defined in the TS Bases.

The proposed changes are consistent with the generic changes included in Technical Specification Task Force (TSTF) Traveler 51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," (Reference 1) and those previously approved for the Pilgrim Nuclear Power Station (Reference 14).

A revised fuel handling accident (FHA) (or refueling accident) analysis has been developed to support the proposed changes using the Alternate Source Term (AST) methodology described in Regulatory Guide 1.183 (Reference 2). Attachment 4 provides the supporting calculation for the revised FHA analysis.

Entergy Nuclear Operations, Inc. (ENO) requests NRC review and approval of the proposed changes by December 1, 2014 to support the anticipated fuel handling operations following final shutdown of the VY reactor, which is expected to occur in the fourth quarter of 2014 (Reference 3).

2.

DETAILED DESCRIPTION The VY TS currently impose restrictions on plant operations when handling irradiated fuel assemblies and while performing core alterations. These restrictions require that certain structures, systems or components (SSCs) be operable. These restrictions assure that the radiological consequences of a fuel handling accident do not exceed those estimated in design-basis analyses.

TSTF-51 removes TS requirements for engineered safeguard features (ESF) (e.g.,

primary/secondary containment, standby gas treatment, isolation capability) to be operable after sufficient radioactive decay has occurred to ensure off-site doses remain below the NUREG-0800 Standard Review Plan limits. TSTF-51 also deletes operability requirements during core alterations for ESF mitigation features.

The proposed TS changes are discussed below (additions are shown in underline format and deletions are shown in strikethrough format):

a. TS Table 3.2.3, Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation Footnotes applicable to the High Reactor Building Ventilation Radiation (Trip Function 3) and High Refuel Floor Zone Radiation (Trip Function 4) trip functions are revised as follows:

BVY 13-097 / Attachment 1 / Page 2 of 17 Current TS Proposed TS (c) During movement of irradiated fuel (c) During movement of recently assemblies or fuel cask in irradiated fuel assemblies eO-fuel secondary containment.

eask in secondary containment.

(d) During Alteration of the Reactor (d) During Altration, of th, o*cOt,-

Core.

Go% Deleted

b. TS 3.7.B.4, Standby Gas Treatment System TS 3.7.B.4 is revised as follows:

Current TS 3.7.B.4 Proposed TS 3.7.B.4

4. With two trains of the Standby Gas Treatment System inoperable, or as made applicable by Specification 3.7.B.3:

a. With the reactor in the run mode, startup mode, or hot shutdown condition, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. During movement of irradiated fuel assemblies or the fuel cask in the secondary containment, during core alterations, or during operations with the potential for draining the reactor vessel, immediately:

i.

Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and ii.

Suspend core alterations; and iii.

Initiate action to suspend operations with the potential for draining the reactor vessel.

4. With two trains of the Standby Gas Treatment System inoperable, or as made applicable by Specification 3.7.B.3:

a. With the reactor in the run mode, startup mode, or hot shutdown condition, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. During movement of recently irradiated fuel assemblies OF-the fuePeark in the secondary contain ment,-dUg-*F*--ee alteFatiORG, or during operations with the potential for draining the reactor vessel, immediately:

i.

Suspend movement of recently irradiated fuel assemblies and4he fuel eask-in secondary containment; and ii.

Suspend elet altedatienz; an~d Deleted iii.

Initiate action to suspend operations with the potential for draining the reactor vessel.

BVY 13-097 / Attachment 1 / Page 3 of 17

c. TS 3.7.C, Secondary Containment System TS 3.7.C.1 is revised as follows:

Current TS 3.7.C.1 1

ProDosed TS 3.7.C.1 1

1. Secondary Containment Integrity shall be maintained during the following modes or conditions:

a. Whenever the reactor is in the Run Mode, Startup Mode, or Hot Shutdown condition*; or

b. During movement of irradiated fuel assemblies or the fuel cask in secondary containment; or

c. During alteration of the Reactor Core; or

d. During operations with the potential for draining the reactor vessel.

1. Secondary Containment Integrity shall be maintained during the following modes or conditions:

a. Whenever the reactor is in the Run Mode, Startup Mode, or Hot Shutdown condition*; or

b. During movement of recently irradiated fuel assemblies er-the fuel-eask in secondary containment; or C.

DUring alteration of the Reactor Geie;-e Deleted

d. During operations with the potential for draining the reactor vessel.

TS 3.7.C.4 is revised as follows:

Current TS 3.7.C.4 Proposed TS 3.7.C.4

4. With Secondary Containment Integrity not maintained during movement of irradiated fuel assemblies or the fuel cask in secondary containment, during alteration of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and

b. Suspend alteration of the Reactor Core; and

c. Initiate action to suspend operations with the potential for draining the reactor vessel.

4. With Secondary Containment Integrity not maintained during movement of recently irradiated fuel assemblies eF the-fuel-eask in secondary containment; durfing altcration of the Roactor G9orc, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a. Suspend movement of recently irradiated fuel assemblies and-the fuel-eask in secondary containment; and

b. Suspond altoration of the Reactor Gore; Deleted

c. Initiate action to suspend operations with the potential for draining the reactor vessel.

BVY 13-097 / Attachment 1 / Page 4 of 17 The proposed changes would revise the Limiting Conditions for Operation in VY TS to relax secondary containment operability requirements when handling fuel that is not "recently" irradiated.

The proposed changes would allow for more efficient performance of reactor defueling and decommissioning activities, including transfer of spent reactor fuel from the spent fuel pool to dry fuel storage casks, while continuing to provide adequate controls against the release of fission product radioactivity to the outside atmosphere during fuel handling activities.

Current TS (Table 3.2.3, Specifications 3.7.B.4 and 3.7.C) require secondary containment, together with the SGT system to be operable:

(1) During movement of irradiated fuel or the fuel cask, (2) During core alterations, and (3) During operations with the potential for draining the reactor vessel (OPDRVs).

Changes are proposed to the secondary containment isolation requirements in TS Table 3.2.3 and the SGT system and Secondary Containment operability requirements specified in TS 3.7.B.4 and TS 3.7.C for refueling operations based on the revised FHA. The proposed changes will eliminate operability requirements during fuel handling activities that do not involve "recently" irradiated fuel, movement of a fuel cask, or core alterations. The systems will still be required to be operable during OPDRVs and during fuel handling activities involving recently irradiated fuel.

Based on TSTF-51, changes to the TS Bases define what time period must elapse before fuel is no longer considered "recently" irradiated. Based on the calculation performed to support the revised FHA analysis, the minimum time period that must elapse following reactor shutdown is 13 days. The time period was calculated using the AST methodology in accordance with Regulatory Guide 1.183 and the requirements in 10 CFR 50.67. This is included in the TS Bases to reflect the proposed changes to the TS. Proposed changes to the TS Bases are provided for information in. Upon approval of this amendment, changes to the Bases will be incorporated in accordance with TS 6.7.E, the TS Bases Control Program.

TSTF-51 states that licensees adding the term "recently" must make certain commitments which are consistent with draft NUMARC 93-01, Revision 3 (Reference 4). The following addresses compliance with this condition of the TSTF.

2.2 Administrative Controls The industry, through the Nuclear Energy Institute (NEI) and its predecessor, the Nuclear Management and Resources Council (NUMARC), has developed guidance to assess and manage the increase in risk that may result during outage activities. This guidance was issued as NUMARC 91-06 "Guidelines for Industry Actions to Assess Shutdown Management" and NUMARC 93-01 "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants". Sections 11.3.5 and 11.3.6 of NUMARC 93-01 Rev.

3 address risk assessments for shutdown conditions and include guidance for managing an open containment.

TSTF-51, "Reviewer's Note" states that licensees adding the term "recently" must make the following commitments which are consistent with draft NUMARC 93-01, Revision 3, Section 11.2.6 "Safety Assessment for Removal of Equipment from Service During Shutdown Conditions", subheading "Containment-Primary (PWR)/Secondary(BWR)":

During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity

BVY 13-097 / Attachment 1 1 Page 5 of 17 in the fuel decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure.

The purpose of the "prompt methods" mentioned above is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored."

NUMARC 93-01, Revision 3 was issued in July 2000 and has subsequently been superseded by NUMARC 93-01, Revision 4A (Reference 4). Section 11 of Rev. 4A has been endorsed by Regulatory Guide 1.160, Revision 3 (Reference 5), which superseded Regulatory Guide 1.182, Revision 0. Sections 11.3.5 and 11.3.6 of NUMARC 93-01, Rev.

4A address the scope and methods of assessing shutdown conditions. Section 11.3.6.5 in particular provides guidelines for assessing containment systems to be removed from service including the capability to achieve containment closure in sufficient time to mitigate potential fission product release. The guidelines are the same as those in Section 11.2.6 of draft NUMARC 93-01, Revision 3 as discussed in TSTF-51. ENO will implement these guidelines prior to implementation of the license amendment.

Contingency plans for closure of openings will include the following:

Equipment and tools needed to facilitate closure will be staged Personnel responsible for closure will be knowledgeable and trained in the procedures for establishing building integrity The closure response team will be accompanied by a Radiation Protection (RP) technician for radiation protection monitoring Hoses and cables routed through openings will employ a means to allow rapid, safe disconnect and removal One door in each airlock will be capable of expeditious closure 2.3 Background and Basis for the Proposed Changes Regulatory Guide 1.183 provides guidance for the implementation of ASTs. 10 CFR 50.67 requires licensees seeking to use ASTs to apply for a license amendment and include an evaluation of the consequences of the affected design-basis accidents. This application addresses these requirements by proposing limited scope application of the AST described in Regulatory Guide 1.183 in evaluating the radiological consequences of an FHA. As part of the implementation of the AST, the total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) is applied.

The NRC has already approved use of AST at VY by License Amendment 223 (Reference 6). The refueling accident described in Updated Final Safety Analysis Report (UFSAR)

Section 14.6.4 reflects a FHA analysis that implements the AST methodology.

BVY 13-097 / Attachment 1 / Page 6 of 17

3.

TECHNICAL EVALUATION 3.1 System Descriptions 3.1.1 Secondary Containment Secondary containment and the SGT system mitigate the potential effects of a FHA and are part of the primary success path for a design-basis FHA. The safety objectives of the secondary containment are to minimize ground level release of airborne radioactive materials and to provide a means for a controlled release of the building atmosphere should a design basis accident occur. There are two principal accidents for which credit is taken for secondary containment integrity. These are the loss-of-coolant accident (LOCA) and the refueling accident. The analysis of these accidents is discussed in the UFSAR Section 14.6.

The secondary containment performs a passive function in response to either of these limiting events. Leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission products entrapped in the secondary containment will be treated by the SGT system prior to discharge to the environment. In the COLD SHUTDOWN and Refueling modes (with reactor coolant temperature less than 212 0F), the probability of a LOCA event is reduced due to the pressure and temperature limitations in these conditions. Therefore, maintaining secondary containment integrity is not required in the cold shutdown condition or the refuel mode, except for other situations for which significant releases of radioactive material can be postulated.

UFSAR Section 5.3 describes the secondary containment requirements, including the SGT system (Section 5.3.4) and reactor building ventilation system (Section 5.3.5).

3.1.2 Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation The reactor building ventilation isolation and SGT system initiation instrumentation automatically initiates closure of the Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs) and starts the SGT system. The function of these components and systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated DBAs. Reactor Building (i.e., secondary containment) isolation and establishment of vacuum with the SGT system ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be operable, are maintained within applicable limits. The isolation and initiation signals generated by the reactor building ventilation isolation and SGT system initiation instrumentation are implicitly assumed in the safety analyses to initiate closure of the RBAVSIVs and start the SGT system to limit offsite doses.

The High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation Trip Functions are required to be operable in RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, Refuel (with reactor coolant temperature greater than 212 0F) where considerable energy exists in the reactor coolant system (RCS); thus, there is a possibility of pipe breaks resulting in significant releases of radioactive steam and gas. In COLD SHUTDOWN and Refuel (with reactor coolant temperature less than 212'F), the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these Modes; thus, these Trip Functions are not required. In addition, the Trip Functions are also required to be operable during OPDRVs, during movement of irradiated fuel assemblies or fuel cask in the secondary containment, and during Alteration of the

BVY 13-097 / Attachment 1 / Page 7 of 17 Reactor Core, because the capability of detecting radiation releases due to fuel failures must be provided to ensure that offsite dose limits are not exceeded.

3.2 Alternate Source Term ENO has completed a calculation (Reference 7) following 13 days of decay time evaluating the potential dose consequences of the FHA. A copy of the calculation is included as of this submittal. This calculation demonstrates that radiological doses at the exclusion area boundary (EAB), low population zone (LPZ) and in the control room (CR) are within allowable limits of 10 CFR 50.67 without crediting secondary containment operability, control room high efficiency air filtration or standby gas treatment systems, after a 13 day fuel decay period following reactor shutdown.

The postulated accident scenarios were based on the AST methodology in Regulatory Guide 1.183, Appendix B. Two main configurations of the Reactor Building during fuel movement were considered, as follows:

A closed containment, for the purpose of replicating the current licensing basis and thus provide a valid starting point for the re-analyses documented in the present calculation, and An open containment, for the purpose of identifying the required decay time prior to fuel movement and the various pre-and post-FHA Main Control Room (CR) ventilation configurations that would support refueling with open containment.

The VY design inputs for the FHA analysis are summarized in Table 3-1 through Table 3-3 of Attachment 4. The assumptions associated with each of the scenarios are described in subsections 5.1.1 (closed containment) and 5.2.1 (open containment) of Attachment 4. It should be noted that alternate scenarios described in the analysis considering upgrading the CR ventilation system with filtration or purging of the CR atmosphere following CR isolation prior to fuel movement are not considered in this license amendment request.

3.2.1 Acceptance Criteria The basic radiological acceptance criteria associated with the AST methodology are spelled out in 10 CFR 50.67(b)(2), and amount to 25 rem TEDE for offsite receptors and 5 rem for control room personnel. These criteria, however, are for evaluating certain postulated accidents of exceedingly low occurrence probability and low risk of public exposure to radiation. For events with higher probability of occurrence, such as a FHA, the acceptance criteria for the offsite receptors are more stringent, while that for the control room operators remains the same. The applicable AST criteria for an FHA are as follows (Regulatory Guide 1.183 and 10 CFR 50.67(b)(2)):

Exclusion Area Boundary (EAB): 6.3 rem TEDE Low Population Zone (LPZ): 6.3 rem TEDE Control Room (CR): 5.0 rem TEDE The EAB and LPZ criteria are referred to as being "well within" the regulatory limits (i.e., -25% below limit).

The LPZ doses were not addressed in the calculation since the release occurs within two hours, and as a result they are bounded by the corresponding doses at the EAB.

BVY 13-097 / Attachment 1 / Page 8 of 17 3.3 Atmospheric Dispersion Atmospheric dispersion factors (X/Qs) at the normal control room air intake were calculated using the ARCON 96 computer code. The following table (from Table 3-3 of Attachment 4) provides the X/Qs used in the analysis:

Post FHA xlQ No.

Release Point Receptor Point Pot FHA xIQ SInterval (sec/rm3) 1 Main stack EAB Instantaneous 1.35E-04 (Calculation release of record with closed 2lonen Control Room Instantaneous 2

containment, for Fresh Air Intake release 6.04E-05 confirmatory analyses) 3 Reactor Building EAB 0 - 2 hrs 1.69E-03 blowout panel 0 - 2 hrs 5.89E-03 (Calculation with 2 - 8 hrs 1.53E-03 4

open Control Room 8 -.24 hrs 6.41 E-04 containment for Fresh Air Intake 24 - 96 hrs 6.64E-04 present application) 96 - 720 hrs 5.10E-04 Primary assumptions used in the analysis for two release locations are summarized as follows:

Main Stack Release Point (Closed Containment Scenario)

" The reactor building was assumed to be closed during the refueling operations, such that all releases to the environment would be via the main stack, with no credit for any filtration by the SGT system, or any in-transit decay and plateout.

Transport of the released radioactivity to the receptors of interest is dictated by the applicable atmospheric dispersion factors in the above table.

" The CR ventilation system was assumed to remain in the normal operating mode during the entire exposure interval (30 days). The air intake flow is 3700 cubic feet per minute (cfm) (includes additional flow from surrounding areas as a result of ingress, egress and leakage) and is unfiltered.

Reactor Building Blowout Panel (Open Containment Scenario)

The reactor building was assumed to be open during the refueling operations, with all post-FHA releases to the environment assumed to be at ground level, via the RB blowout panels.

" Transport of the released radioactivity to the receptors of interest is dictated by the applicable atmospheric dispersion factors in above table for ground-level releases.

The CR ventilation system was assumed to remain in the normal operating mode during the entire exposure interval (30 days), with an intake flow of 3700 cfm, unfiltered. For demonstrative purposes, the CR envelope was considered to be fully isolated prior to fuel movement, with an unfiltered inleakage of 50 cfm.

As noted in the introduction of Regulatory Guide 1.194 (Reference 8), many of the positions in the guide represent significant changes. ARCON96 implements an improved building wake dispersion algorithm; assessments of ground level, building vent, elevated and

BVY 13-097 / Attachment 1 / Page 9 of 17 diffuse-source release models; use of hour-by-hour meteorological observations; sector averaging; and directional dependence of dispersion conditions. Therefore, no discussion of the comparison with current licensing basis X/Q values is presented.

3.4 Radiological Consequences of a Design-Basis Fuel Handling Accident The analysis was performed using the ELISA-2, Version 2.4 computer code. The dose conversion factors in ELISA-2 are from Federal Guidance Reports 11 and 12 (References 9 and 10). Dose rates and cumulative doses are computed for each organ, TEDE, skin and air. Of these, only the TEDE doses are presented in the main body of the calculation for comparing with regulations, which only specify TEDE limits.

For the closed containment scenario, the dose consequences for the FHA scenario described in subsection 5.1 of Attachment 4 are presented in Table 5-2 of Attachment 4. It is seen that there is very good agreement between the calculation of record and the analyses documented in the present calculation. The only noteworthy difference (about 10%, which is still acceptable) is due to the underestimation of the Xe-1 35m contribution to the dose, resulting from this noble-gas isotope not being identified in the calculation of record as a daughter product of 1-135 in the decay correction. The corresponding EAB dose in Case A3 is not impacted due to the long pre-FHA decay time for this isotope.

The CR doses in all cases are not affected by the Xe-1 35m underestimation primarily due to the finite-cloud correction to the submersion dose.

For the open containment scenario, the dose consequences for the FHA scenario described in subsection 5.2 of Attachment 4 are presented in Table 5-3 of Attachment 4 and shown graphically in Figures 5-1 and 5-2 of Attachment 4. It is seen that the EAB TEDE dose with open containment is less than the regulatory limit of 6.3 rem limit even with a 24-hr pre-FHA decay time. On the other hand, a decay time of about 11 to 13 days is needed to achieve a CR dose less than 5 rem, for intake flows ranging between 50 cfm (with the CR pre-isolated) and 3700 cfm (current CR configuration). For this license amendment request, the most conservative value of 13 days is applied throughout this application.

3.4.1 Results of Fuel Handling Accident Dose Consequences The dose evaluations of the postulated fuel handling accident demonstrates that the calculated TEDE values to the CR, EAB, and LPZ using a 13 day decay period for the reactor fuel with open containment and with the normal, unfiltered, control room ventilation system in operation are less than the regulatory limit of 6.3 rem TEDE even with a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pre-FHA decay time. The 13 day decay time is needed to achieve a CR dose of less than 5 rem.

3.5 Core Alterations The accidents postulated to occur during core alterations, in addition to the fuel handling accident, are inadvertent criticality due to control rod removal error and the inadvertent loading of, and subsequent operation with, a fuel assembly in an improper location.

These events are not postulated to result in fuel cladding integrity damage. Therefore, the only accident postulated to occur during core alterations that result in significant radioactive release is the FHA. Thus, the consequence of a FHA envelops the consequences of potential accidents postulated to occur during core alterations. Therefore, the proposed changes omitting core alterations are justified.

BVY 13-097 / Attachment 1 / Page 10 of 17 3.6

- Fuel Cask Movement The operability requirements during movement of a fuel cask for ESF mitigation are deleted as part of this proposed license amendment. The only accident postulated during handling of a fuel cask is the FHA. The FHA assumes an irradiated fuel assembly is dropped onto the reactor core or in the spent fuel pool from the maximum height allowed by the fuel handling equipment. The analysis assumes that the entire amount of potential energy is available for application to the fuel assemblies involved in the accident. Also, none of the energy associated with the dropped fuel assembly is absorbed by the fuel material.

The Reactor Building crane is a 110-ton capacity overhead bridge crane that provides services for the reactor and refueling area. The crane handles the spent fuel shipping cask.

The Reactor Building crane was modified in 1976 by replacing the original trolley with one that has a dual load path on the main hoist when used for cask handling operations. The design of the new trolley satisfies the criteria for dual load path or "single-failure-proof' cranes, and, with issuance of Amendment 29 to the VY operating license on January 28, 1977, the NRC staff accepted the crane as single-failure-proof (Reference 15).

In addition to the hardware improvements implemented to modify the crane with single-failure-proof components, VY implemented a number of other improvements to enhance conformance with the guidelines of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" (Reference 11). These improvements included:

revising maintenance procedures to define safe load paths for major loads.

revising procedures to include training and qualification requirements for crane operators, sling selection criteria, crane inspections prior to use, and supervisory oversight of heavy lift operations.

procurement of special lifting devices and performance of periodic non-destructive examinations to monitor the condition of lifting devices.

The NRC staff accepted these improvements through a safety evaluation transmitted by letter dated June 27, 1984 (Reference 12).

The single-failure-proof modifications, the procured special lifting devices, and the sling selection procedures provide reasonable assurance that the handling system used for heavy load movement of a fuel cask near the spent fuel pool will have designed-in features to prevent a load drop. As outlined in NUREG-0612, the NRC staff accepts that provision of a single-failure-proof handling system, in conjunction with other actions implemented at VY, provides defense-in-depth against drops of loads heavier than one fuel assembly and its associated handling tool. This would include a fuel cask loaded with spent fuel. Thus, actions and events necessary to result in a heavy load drop from the Reactor Building crane over spent fuel are not sufficiently credible that this event was included among design basis events.

The VY Technical Requirements Manual (TRM) contains the crane functionality and surveillance requirements, as well as requirements to install mechanical rail stops to prevent the movement of a fuel cask over irradiated fuel. License Amendment 239 relocated the TS requirements for the Reactor Building crane to the TRM (Reference 13).

Since the FHA resulting from a dropped fuel cask is shown to not be credible, the proposed TS changes omitting operability requirements during movement of a fuel cask for ESF mitigation is justified.

BVY 13-097 / Attachment 1 / Page 11 of 17

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria This section describes how the proposed changes and ENO's technical analyses satisfy applicable regulatory requirements and acceptance criteria.

10 CFR 50 Appendix A General Design Criterion 61, "Fuel Storage and Handling and Radiological Control" The general design criteria (GDC) in place today became effective after the VY construction permit was issued. A September 18, 1992 memorandum (ML003763736) to the NRC Executive Director of Operations from the Secretary of the NRC summarized the results of a Commissioners vote in which the Commissioners instructed the NRC staff not to apply the GDC to plants with construction permits issued prior to May 21, 1971. VY's construction permit was issued on December 11, 1967.

VY's design and licensing basis for fuel storage and handling and radiological controls is detailed in the Updated Final Safety Analysis Report (UFSAR) and other plant-specific licensing basis documents. UFSAR Appendix F provides a comparison of VY with the proposed GDC published by the Atomic Energy Commission (AEC) for public comment in the Federal Register dated July 11, 1967.

10 CFR 50.67 "Accident Source Term" 10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design-basis radiological consequence analyses. This document is part of a 10 CFR 50.90 license amendment application and evaluates the consequences of a design-basis fuel handling accident previously reported in the safety analysis report.

10 CFR 50.65 "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" 10 CFR 50.65(a)(4) requires licensees to assess and manage changes in risk that result from taking risk-significant systems out-of-service or during certain maintenance activities.

The NRC staff, in Regulatory Guide 1.160, endorsed NUMARC 93-01, Revision 4A as providing acceptable methods, with clarifications and provisions described in Part C of the Regulatory Guide, for complying with the requirements of 10 CFR 50.65. Section 11.3.6.5 "Containment - Primary (PWR)/Secondary (BWR)," of NUMARC 93-01, Revision 4A states:

Maintenance activities involving the need for open containment should include evaluation of the capability to achieve containment closure in sufficient time to mitigate potential fission product release. This time is dependent on a number of factors, including the decay heat level and the amount of RCS inventory available.

For BWRs, technical specifications may require secondary containment to be closed under certain conditions, such as during fuel handling and operations with a potential to drain the vessel.

In addition to the guidance in NUMARC 91-06, for plants which obtain license amendments to utilize shutdown safety administrative controls in lieu of Technical Specification requirements on primary or secondary containment operability and ventilation system operability during fuel handling or core alterations, the following guidelines should be included in the assessment of systems removed from service:

BVY 13-097 / Attachment 1 / Page 12 of 17 During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the RCS decays fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay, and to avoid unmonitored releases.

A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

To further limit the potential radiological consequences of a fuel handling accident at VY, ENO will revise the VY guidelines for assessing systems removed from service during the handling of recently irradiated fuel assemblies or core alterations to implement the provisions of Section 11.3.6.5 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 4A. These guidelines will address the capabilities to promptly close secondary containment and will be completed prior to the implementation of this license amendment. See Section 2.2 of this application for additional discussion on the administrative controls to be implemented upon approval of this amendment.

10 CFR 100, Paragraph 11, "Determination of Exclusion Area, Low Population Zone and Population Center Distance" This paragraph provides criteria for evaluating the radiological aspects of reactor sites. A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. A similar footnote appears in 10 CFR 50.67.

In accordance with the provisions of 10 CFR 50.67(a), the radiation dose reference values in 10 CFR 50.67(b)(2) were used in these analyses in lieu of those prescribed in 10 CFR 100. (Refer to footnote 5 on page 1.183-7 of Regulatory Guide 1.183, dated July 2000.)

Regulatory Guide 1.25, "Assumptions Used for Evaluatincq the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors". March 1972 Regulatory Guide 1.25 is not applicable to the application. Regulatory Guide 1.183 supersedes corresponding radiological assumptions provided in other regulatory guides and standard review plan chapters when used in conjunction with an approved alternate source term and the TEDE criteria provided in 10 CFR 50.67.

BVY 13-097 / Attachment 1 / Page 13 of 17 Re-gulatory Guide 1.183, "Alternative Radiological Source Terms for evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000 Regulatory Guide 1.183 outlines acceptable applications of ASTs, the scope, nature and documentation of associated analyses and evaluations, consideration of impacts on analyzed risk; and content of submittals. It also establishes acceptable ASTs and identifies the attributes of ASTs acceptable to the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the AST.

Regulatory Guide 1.183, Appendix B, Sec. 5.3, Footnote 3 states that the following provisions need to be implemented for refueling operations with open containment:

"The (NRC) staff will generally require that technical specifications allowing such operations (i.e., open containment during fuel handling operations) include administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative control will generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation."

The administrative controls that VY will put in place following approval of this amendment request are described in Section 2.2 of this application.

ENO used this regulatory guide extensively in the preparation of this "limited scope implementation" evaluation, the supported application and the supporting analyses. This application and the supporting analyses comply with this guidance to the extent practical.

NUREG-1 465, "Accident Source Terms for Light-Water Nuclear Power Plants" NUREG-1 465 provides more realistic estimates than Technical Information Document 14844 of "source term" releases into containment in terms of timing, nuclide types, quantities, and chemical form, given a severe core melt. NUREG-1465 provides much of the technical basis for the regulatory positions in Regulatory Guide 1.183.

NUREG-0800, Standard Review Plan, Section 15.7.4, "Radiological Conseguences of Fuel Handling Accidents" This SRP section covers the review of the radiological effects of a postulated fuel handling accident. Revision 1 does not reflect the guidance in Regulatory Guide 1.183 or the promulgation of 10 CFR 50.67.

NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36" NUREG-0612 provides guidance on the control of heavy loads that may be handled in several plant areas. If these loads were to drop in certain locations in the plant, they may impact spent fuel, fuel in the core, or equipment that may be required to achieve safe shutdown and continue decay heat removal. The guidelines proposed include definition of safe load paths, use of load handling procedures, training of crane operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives that include: use of a single failure proof handling system, use of mechanical stops or electrical interlocks to keep heavy loads away from fuel or safe shutdown equipment, or analyzing the consequences of postulated heavy load drops to show these are within acceptable limits.

BVY 13-097 / Attachment 1 / Page 14 of 17 4.2 Precedent The proposed VY TS changes follow TSTF-51 and similar changes approved for Pilgrim Nuclear Power Station. Even though VY's current TS appear different in style than the TSTF-51, the proposed changes are consistent with TSTF-51 requirements, and approved Pilgrim TS changes in scope and requirements.

4.3 No Significant Hazards Consideration Determination Pursuant to 10CFR50.92, Entergy Nuclear Operations, Inc. (ENO) has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c). These criteria require that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

ENO proposes changes to the (a) Vermont Yankee Nuclear Power Station (VY) licensing basis of the Fuel Handling Accident (FHA) as described in the VY Updated Final Safety Analysis Report (UFSAR) and (b) VY Technical Specifications (TS). The proposed changes to the TS would eliminate secondary containment operability requirements when handling sufficiently decayed irradiated fuel, or fuel casks, and performing core alterations, and clarify requirements associated with operations with potential to drain the reactor vessel (OPDRVs). These changes revise secondary containment isolation requirements, and standby gas treatment (SGT) system, and secondary containment operability requirements for refueling operations.

The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

1.

Does the Proposed amendment involve a siqnificant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes do not modify the design or operation of equipment used to move spent fuel or to perform core alterations. The proposed changes cannot increase the probability of any previously analyzed accident because they are based on changes in Source Term, atmospheric dispersion and dose consequence analysis methodology, not in procedures or equipment used for fuel handling.

The conservative re-analysis of the FHA concludes that the radiological consequences are within the regulatory limits established 10 CFR 50.67.

This conclusion is based on the Alternate Source Term and guidance provided in Appendix B of Regulatory Guide 1.183 and analyses of fission product release and transport path that does not take credit for dose mitigation provided by engineered safeguards including secondary containment and the SGT system. The results of the core alteration events, other than the FHA, remain unchanged from the original design-basis that showed these events do not result in fuel cladding damage or radioactive release.

BVY 13-097 / Attachment 1 / Page 15 of 17 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

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

Response: No.

The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant.

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

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Regulation 10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design-basis radiological consequence analyses. This license amendment application evaluates the consequences of a design-basis fuel handling accident in accordance with this regulation and Regulatory Guide 1.183. The revised analysis concludes that the radiological consequences of the fuel handling accident are less than the regulatory allowable limits. Safety margins and analytical conservatisms are retained to ensure the analysis adequately bounds all postulated event scenarios. The selected assumptions and release models provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensates for large uncertainties in facility parameters, accident progression, radioactive material transport and atmospheric dispersion. The proposed TS applicability statements continue to ensure that the total effective dose equivalent (TEDE) at the boundaries of the control room, the exclusion area, and low population zone boundaries are below the corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).

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

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

4.4 Conclusion Analyses of the radiological dose of a postulated FHA involving irradiated fuel assembles that have been allowed to decay for at least 13 days show that the calculated TEDE values to the control room occupants and at the exclusion area boundary, without crediting secondary containment and SGT system operation, are below the allowable TEDE limits established in Regulatory Guide 1.183 and 10 CFR 50.67. Therefore, after 13 days, movement of irradiated fuel assemblies can commence and continue without the operability requirements for the SGT system and secondary containment. This conclusion forms the basis for the proposed TS changes, with the exception of the deletion of operability

BVY 13-097 / Attachment 1 / Page 16 of 17 requirements during movement of a fuel cask for ESF mitigation features. The basis for this change is described in Section 3.6 of this request.

5.

ENVIRONMENTAL CONSIDERATIONS This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 1 OCFR51.22(c)(9) as follows:

(i)

The amendment involves no significant hazards determination.

As described in Section 4 of this evaluation, the proposed change involves no significant hazards consideration.

(ii)

There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed amendment does not involve any physical alterations to the plant configuration that could lead to a change in the type or amount of effluent release offsite.

(iii)

There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above, ENO concludes that the proposed change meets the eligibility criteria for categorical exclusion as set forth in 1 0CFR51.22(c)(9). Pursuant to 1 OCFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6.

REFERENCES

1. TSTF-51, Rev. 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," Excel Services Corporation

2. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000

3. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Permanent Cessation of Operations," BVY 13-079, dated September 23, 2013

4. NUMARC 93-01, Revision 4A, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," April 2011

5. Regulatory Guide 1.160, Revision 3, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," May 2012

6. Letter, USNRC to Entergy Nuclear Operations, Inc., "Vermont Yankee Nuclear Power Station Issuance of Amendment RE: Alternative Source Term (TAC No.

MC0253)," NVY 05-045, dated March 29, 2005

7. AREVA Document No. 32-9145461-001, "VYNPP - Re-analysis of AST/FHA Radiological Consequences with Open Containment," October 2010

BVY 13-097 / Attachment 1 / Page 17 of 17

8. Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants", June 2003

9. EPA 520/1-88-020, Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration, and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," September 1988

10. EPA 402-R-93-081, Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil," September 1993

11. NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36," July 1980

12. Letter, USNRC to Vermont Yankee Nuclear Power Corporation, "Control of Heavy Loads (Phase I)," NVY 84-139, dated June 27, 1984

13. Letter, USNRC to Entergy Nuclear Operations, Inc., "Vermont Yankee Nuclear Power Station Issuance of Amendment RE: Relocation of Reactor Building Crane Technical Specification (TAC No. MD9725)," NVY 09-077, dated July 13, 2009

14. Letter, USNRC to Entergy Nuclear Operations, Inc., "Pilgrim Nuclear Power Station

- Issuance of Amendment RE: Alternative Source Term for the Fuel Handling Accident Dose Consequences (TAC No. MC2705)," dated April 28, 2005

15. Letter, USNRC to Vermont Yankee Nuclear Power Corporation, Amendment No. 29, dated January 28, 1977

BVY 13-097 Docket 50-271 Vermont Yankee Nuclear Power Station Proposed Change 306 Markup of the Current Technical Specifications and Bases Pages

VYNPS Table 3.2.3 (page 1 of 1)

Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation ACTIONS WHEN REQUIRED REQUIRED APPLICABLE MODES OR CHANNELS CHANNELS OTHER SPECIFIED PER TRIP ARE TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE TRIP SETTING

1. Low Reactor

RUN, STARTUP/HOT 2

Note 1

Ž 127.0 inches Vessel Water

STANDBY, HOT SHUTDOWN, Level Refuella),

(b)

2.

High Drywell

RUN, STARTUP/HOT 2

Note 1

2.5 psig Pressure

STANDBY, HOT SHUTDOWN, Refuel (a)

3.

High Reactor

RUN, STARTUP/HOT 1

Note 1

14 mR/hr Building

STANDBY, HOT SHUTDOWN, Ventilation Refuel(a),

(b),

(c),

(d)

Radiation

4.

High Refueling

RUN, STARTUP/HOT 1

Note 1 1 100 mR/hr Floor Zone

STANDBY, HOT SHUTDOWN, Radiation Refuel(a),

(b),

(c),

(d)

(a)

With reactor coolant temperature > 212 'F.

(b)

During operations with potential for draining the reactor vessel.

(c)

During movement containment.

(d)

Du4ng Alterati<

of irradiated fuel assemblies or fuiel cask in secondary

---recen~tly AP ~

t~

Amendment No.

4-64, 2-34 51

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION shutdown condition, the actions and completion times of Specification 3.7.B.4.b shall apply.

After seven days with an inoperable train of the Standby Gas Treatment System during refueling or cold shutdown conditions requiring secondary containment integrity, the operable train of the Standby Gas Treatment System shall be placed in operation and its associated diesel generator shall be operable, or the actions and completion times of Specification 3.7.B.4.b shall apply.

4.

With two trains of the Standby Gas Treatment System inoperable, or as made applicable by Specification 3.7.B.3:

a.

With the reactor in the run mode, startup mode, or hot shutdown condition, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. During movement of

-recently irradiated fuel assemblies or-the fuieil eak in the secondary containment, during eer.e alteratien-s, or during operations with the potential for draining the reactor vessel, immediately:

Amendment No.

4-9-1 155

VYNPS 3.7 LIMITING CONDITIONS FOR

-OPERATION 4.7 SURVEILLANCE REQUIREMENTS

i.

Suspend movement of irecently I >irradiated fuel assemblies a444 the fuci] rz;sk in secondary containment; and IDeleted Suspend core altcrations; iii.

Initiate action to suspend operations with the potential for draining the reactor vessel.

C.

Secondary Containment System C.

Secondary Containment System

1. Secondary Containment Integrity shall be maintained during the following modes or conditions:

a.

Whenever the reactor is in the Run Mode, Startup Mode, or Hot Shutdown condition*;

or 1.

Secondary containment capability to maintain a 0.15 inch of water vacuum under calm wind (2<0<5 mph) conditions with a filter train flow rate of not more than 1,550 cfm, shall be demonstrated at least quarterly.

NOTE:

The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup

Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:

1. Reactor coolant temperature is

< 212 0 F;

2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and

3.

No core alterations are in progress.

Amendment No.

4-4, 4-44, 4-9-,

4, 2-2-&,

155a

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

b.

During movement of irecentwly

> irradiated fuel assemblies er-te fuel eask in secondary containment; or I---l-----

c-1c.

Dur-ing alteration of

[~~~

2 J the Reaeter Ccrei er

d.

During operations with the potential for draining the reactor vessel.

Amendment No.

4-4, 49 2-2-3, 2*-6 5

156

VYNPS 3.7 LIMITING CONDITIONS FOR OPERATION

2.

With Secondary Containment Integrity not maintained with the reactor in the Run Mode, Startup Mode, or Hot Shutdown condition, restore Secondary Containment Integrity within four (4) hours.

4.7 SURVEILLANCE REQUIREMENTS

2.

Intentionally blank.

3.

Intentionally blank.

4.

Intentionally blank.

3.

4.

If Specification 3.7.C.2 cannot be met, place the reactor in the Hot Shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in the Cold Shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

recentlyl With Secondary Containment Int rity not maintained d inng movement of irradiated fuel assemblies er-th fuel eak in secondary containment, "duri*

alteration of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a.

Suspend movement of Irecently irradiated fuel assemblies and the fuel eask in secondary containment; and

b.

Suspend alteratio feleted[

the Re. tr CGr-;

c.

Initiate action to suspend operations with the potential for draining the reactor vessel.

Amendment No.

4-4-,

2-9,5 157

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES,

LCO, and APPLICABILITY (continued) instrumentation are implicitly assumed in the safety analyses of References 2, 3,

and 4, to initiate closure of the RBAVSIVs and start the SGT System to limit offsite doses.

Reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The operability of the reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions specified in Table 3.2.3.

Each Trip Function must have the required number of operable channels in each trip system, with their trip setpoints within the calculational as-found tolerances specified in plant procedures.

Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.3.

As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures.

The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions. and recently In genera, the individual Trip Functions are required to be OPERABLE in

RUN, STARTUP/H T STANDBY, HOT SHUT OWN, Refuel (with reactor coolant temperature 212 0 F),

uring operations th the potential for draining the reactor vessel (OPDRVs)T during movement of irradiated fuel assemblies or fuel cask in secondary containment, and during Alteratien of the Reactor Care; consistent with the Applicability for the SGT System and secondary containment requirements in Specifications 3.7.B and 3.7.C.

Trip Functions that have different Applicabilities are discussed below in the individual Trip Functions discussion.

Thk specific Applicable Safety Analyses,

LCO, and Applicability discussions are lis ed below on a Trip Function by Trip Function basis.

ý-Insert I

1. Low Reactor Vessel Water Level Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened.

Should RPV water level decrease too far, fuel damage could result.

An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite release.

The Low Reactor Vessel Water Level Trip Function is one of the Trip Functions assumed to be operable and capable of providing isolation and initiation signals.

The isolation and initiation of systems on Low Reactor Vessel Water Level support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

Low Reactor Vessel Water Level signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Low Reactor Vessel Water Level Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation and initiation function.

Amendment No.

2-34 76p

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES,

LCO, and APPLICABILITY (continued) have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident.

When High Reactor Building Ventilation Radiation or High Refueling Floor Zone Radiation is detected, secondary containment isolation and actuation of the SGT System are initiated to support actions to limit the release of fission products as assumed in the UFSAR safety analyses (Ref.

4).

The High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation signals are initiated from radiation detectors that are located on the ventilation exhaust duct coming from the reactor building and the refueling floor zones, respectively.

Two channels of High Reactor Building Ventilation Radiation Trip Function and two channels of High Refueling Floor Radiation Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation and initiation function.

The Trip Settings are chosen to promptly detect gross failure of the fuel cladding.

The High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation Trip Functions are required to be operable in

RUN, STARTUP/HOT

STANDBY, HOT SHUTDOWN, Refuel (with reactor coolant temperature > 212'F) where considerable energy exists in the RCS; thus, there is a possibility of pipe breaks resulting in significant releases of radioactive steam and gas.

In COLD SHUTDOWN and Refuel (with reactor coolant temperature < 212'F),

the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these Modes; thus, these Trip Functions are not required.

In addition, the Trip Functions are also required to be operable during OPDRVsr during movement of irradiated fuel assemblies er fuel cqask in the secondary ntainment, odA,-n of tc,acr Coro, because the capability f detecting radiat on releases due to fuel failures (due to fuel uncovery o dropped fuel asse lies) must be provided to ensure that offsite dose liS ts are not exceede.

and recently ACTION Insert 1 Table 3.2.

Note 1 Because pf the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Trip Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Trip Functions that have channel components common to RPS instrumentation, i.e.,

Trip Functions 1 and 2, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Trip Functions that do not have channel components common to RPS instrumentation, i.e.,

all other Trip Functions),

has been shown to be acceptable (Refs.

5 and 6) to permit restoration of any inoperable channel to operable status.

This out of service time is only acceptable provided the associated Trip Function is still maintaining isolation capability (refer to next paragraph).

If the inoperable channel cannot be restored to operable status within the allowable out of service time, the channel must be placed in the tripped condition per Table 3.2.3 Note l.a.l) or l.a.2),

as applicable.

Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

Alternately, Amendment No.

2-4 76r

VYNPS BASES:

3.7 (Cont'd) surveillances such as monthly torus to drywell vacuum breaker tests.

Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to prevent overpressurization of the SBGT system or the reactor building ductwork, should a LOCA occur.

For this and similar analyses performed, a

spurious opening of AC-6 or AC-7 (one of the closed containment isolation valves) is not assumed as a failure simultaneous with a postulated LOCA.

Analyses demonstrate that for normal plant operation system alignments, including surveillances such as those described above, that SBGT integrity would be maintained if a LOCA was postulated.

Therefore, during normal plant operations, the 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock does not apply.

Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per calendar year (except for performance of the subject valve stroke time surveillances -

in which case the appropriate corresponding valves are closed to protect equipment should a LOCA occur).

This restriction will apply whenever primary containment integrity is required.

The 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock will apply anytime purge and vent evolutions can not assure the integrity of the SBGT trains or related equipment.

B. and C.

Standby Gas Treatment System and Secondary Containment System The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident.

The Reactor Building provides secondarycontainment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.

Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required except, however, for initial fuel loading and low power physics testing.

In the Cold Shutdown condition or the Refuel Mode,the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these conditions.

Therefore, maintaining Secondary Containment Integrity is not required in the Cold Shutdown condition or the Refuel Mode, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel, during alt.rati.n of the Rca.t.r Core, or during movement of irradiated fuel assemblies or the fuel cask in the secondary containmeni" nt-In order for secondary containment integrity to be met, the secondary containment must function properly in conjunction with the operation of the Standby Gas Treatment System to ensure that the required vacuum can be established and maintained.

This means that the reactor building is intact with at least one door in each access opening closed, and all reactor building automatic ventilation system isolation valves are operable or the affected penetration flow path is isolated.

With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdown condition, if Secondary Containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition.

This time period also ensures that the probability of an accident (requiring Secondary Containment Integrity) occurring during periods where Secondary Containment Integrity is not maintained, is minimal.

Amendment No.

4-4,

-44, 4*-4, 197 165a

VYNPS BASES:

3.7 (Cont'd) recently If Secondary ontainment Integrity cannot be restored within the required tim period, the plant must be brought to a mode or condition in which the LCO does not apply.

Movement of irradiated fuel assemblies or the fuel cask in the secondary containment, altcratien of the Reactor Cor*, and operations Movement of h the potential for draining the reactor vessel can be postulated to recently caus-efission product release to the secondary containment.

In such

Icases, secondary containment is the only barrier to release of fission pro cts to the environment.

At-eraticn of thA Reactor Core oment irradiated fuel assemblies and the fuel cash must be immediately suspended if Secondary Containment Integrity is not maintained.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position.

Also, action must be immediately initiated to suspend operations with the potential for draining the reactor vessel to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Actions must continue until operations with the potential for draining the reactor vessel are suspended.

Amendment No.

4-4-3-,

14?-, 5 165b

VYNPS BASES:

3.7 (Cont'd)

The Standby Gas Treatment System (SGTS) is designed to filter and exhaust the Reactor Building atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the Reactor Building to the environs.

To insure that the standby gas treatment system will be effective in removing radioactive contaminates from the Reactor Building air, the system is tested periodically to meet the intent of ANSI N510-1975.

Laboratory charcoal testing will be performed in accordance with ASTM D3803-1989, except, as allowed by GL 99-02, testing can be performed at 70%

relative humidity for systems with humidity control.

Both standby gas treatment fans are designed to automatically start upon containment isolation and to maintain the Reactor Building pressure to approximately a negative 0.15 inch water gauge pressure; all leakage should be in-leakage.

Should the fan fail to start, the redundant alternate fan and filter system is designed to start automatically.

Each of the two fans has 100% capacity.

This substantiates the availability of the operable train and results in no added risk; thus, reactor operation or refueling operation can continue.

If neither train is operable, the plant is brought to a condition where the system is not required.

recently When the reactor is i

cold shutdown or refueling the drywell may be open and the Reactor Building becomes the only containment system.

During cold shutdown the probability and consequences of a DBA LOCA are substantially reduc due to the pressure and temperature limitations in this mode.

Howe er, for other situations under which significant radioactive release can be postulated, such as during operations with a potential for drai *ng the reactor vessel, during core alterations, or during movement of irradiated fuel in the secondary containment, operability of standby gas treatment is required.

Both trains of the Standby Gas Treatment System are normally operable when secondary containment integrity is required.

However, Specification 3.7.B.3 provides Limiting Conditions for Operation when one train of the Standby Gas Treatment System is inoperable.

Provisional, continued operation is permitted since the remaining operable train is adequate to perform the required radioactivity release control function.

If the applicable conditions of Specification 3.7.B.3 cannot be met, the plant must be placed in a mode recently or condition where the Limiting Conditions for Operation do not apply.

E try into a refueling condition with one train of SBGTS inoperable is ceptable and there is no prohibition on mode or condition entry in his situation.

In this case, the requirements of TS 3.7.B.3.b are uufficient to ensure that adequate controls are in place.

During Lefueling conditions, accident risk is significantly reduced, and the rrimary activities of concern involve core alterations, movement of irradiated fuel assemblies-- and OPDRVs.

During refueling and cold shutdown conditions Specification 3.7.B.3.b provides for the indefinite continuance of refueling operations with one train of the Standby Gas Treatment System inoperable.

When the seven-day completion time associated with Specification 3.7.B.3.b is not met and secondary containment integrity is required, the operable train of the Standby Gas Treatment System should immediately be placed into operation.

This action ensures that the remaining train is operable, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.

An alternative to placing the operable train of Standby Gas Treatment in operation is to immediately suspend activities that represent a potential for releasing radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk.

Amendment No.

4-&,

4-9,,

444, 19-7 166

Insert 1 "Recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core within the previous 13 days, i.e. reactor fuel that has decayed less than 13 days following reactor shutdown. This minimum decay period is enforced to maintain the validity of the Fuel Handling Accident dose consequence analysis.

BVY 13-097 Docket 50-271 Vermont Yankee Nuclear Power Station Proposed Change 306 Retyped Technical Specifications and Bases Pages

VYNPS Table 3.2.3 (page 1 of 1)

Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation Instrumentation ACTIONS WHEN REQUIRED REQUIRED APPLICABLE MODES OR CHANNELS CHANNELS OTHER SPECIFIED PER TRIP ARE TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE TRIP SETTING

1. Low Reactor

RUN, STARTUP/HOT 2

Note 1

Ž 127.0 inches Vessel Water

STANDBY, HOT SHUTDOWN, Level Refuel(a),

(b)

2.

High Drywell

RUN, STARTUP/HOT 2

Note 1

2.5 psig Pressure

STANDBY, HOT SHUTDOWN, Refuel (a)

3.

High Reactor

RUN, STARTUP/HOT 1

Note 1

14 mR/hr Building

STANDBY, HOT SHUTDOWN, Ventilation Refuel(a),

(b),

(c),

(d)

Radiation

4.

High Refueling

RUN, STARTUP/HOT 1

Note 1 1

100 mR/hr Floor Zone

STANDBY, HOT SHUTDOWN, Radiation Refuel(a),

(b),

(c),

(d)

II (a)

With reactor coolant temperature > 212 OF.

(b)

During operations with potential for draining the reactor vessel.

(c)

During movement of recently irradiated fuel assemblies in secondary containment.

(d) Deleted Amendment No.

4-64, 2-51 51

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES,

LCO, and APPLICABILITY (continued) instrumentation are implicitly assumed in the safety analyses of References 2, 3, and 4, to initiate closure of the RBAVSIVs and start the SGT System to limit offsite doses.

Reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The operability of the reactor building ventilation isolation and Standby Gas Treatment System initiation instrumentation is dependent on the operability of the individual instrumentation channel Trip Functions specified in Table 3.2.3.

Each Trip Function must have the required number of operable channels in each trip system, with their trip setpoints within the calculational as-found tolerances specified in plant procedures.

Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that, under worst case design basis conditions, the associated trip will occur within the Trip Settings specified in Table 3.2.3.

As a result, a channel is considered inoperable if its actual trip setpoint is not within the calculational as-found tolerances specified in plant procedures.

The actual trip setpoint is calibrated consistent with applicable setpoint methodology assumptions.

In general, the individual Trip Functions are required to be OPERABLE in RUN, STARTUP/HOT STANDBY, HOT SHUTDOWN, Refuel (with reactor coolant temperature

> 212'F), during operations with the potential for draining the reactor vessel (OPDRVs) and during movement of recently irradiated fuel assemblies in secondary containment; consistent with the Applicability for the SGT System and secondary containment requirements in Specifications 3.7.B and 3.7.C.

Trip Functions that have different Applicabilities are discussed below in the individual Trip Functions discussion.

"Recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core within the previous 13 days, i.e. reactor fuel that has decayed less than 13 days following reactor shutdown.

This minimum decay period is enforced to maintain the validity of the Fuel Handling Accident dose consequence analysis.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Trip Function by Trip Function basis.

1. Low Reactor Vessel Water Level Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened.

Should RPV water level decrease too far, fuel damage could result.

An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite release.

The Low Reactor Vessel Water Level Trip Function is one of the Trip Functions assumed to be operable and capable of providing isolation and initiation signals.

The isolation and initiation of systems on Low Reactor Vessel Water Level support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

Low Reactor Vessel Water Level signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Low Reactor Vessel Water Level Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation and initiation function.

Amendment No.

2-3-p 76p

VYNPS BASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GAS TREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident.

When High Reactor Building Ventilation Radiation or High Refueling Floor Zone Radiation is detected, secondary containment isolation and actuation of the SGT System are initiated to support actions to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 4).

The High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation signals are initiated from radiation detectors that are located on the ventilation exhaust duct coming from the reactor building and the refueling floor zones, respectively.

Two channels of High Reactor Building Ventilation Radiation Trip Function and two channels of High Refueling Floor Radiation Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation and initiation function.

The Trip Settings are chosen to promptly detect gross failure of the fuel cladding.

The High Reactor Building Ventilation Radiation and High Refueling Floor Zone Radiation Trip Functions are required to be operable in

RUN, STARTUP/HOT

STANDBY, HOT SHUTDOWN, Refuel (with reactor coolant temperature > 2120 F) where considerable energy exists in the RCS; thus, there is a possibility of pipe breaks resulting in significant releases of radioactive steam and gas.

In COLD SHUTDOWN and Refuel (with reactor coolant temperature < 212 0F),

the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these Modes; thus, these Trip Functions are not required.

In addition, the Trip Functions are also required to be operable during OPDRVs and during movement of recently irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded.

"Recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core within the previous 13 days, i.e. reactor fuel that has decayed less than 13 days following reactor shutdown.

This minimum decay period is enforced to maintain the validity of the Fuel Handling Accident dose consequence analysis.

ACTIONS Table 3.2.3 ACTION Note 1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months depending on the Trip Function (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for those Trip Functions that have channel components common to RPS instrumentation, i.e.,

Trip Functions 1 and 2, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those Trip Functions that do not have channel components common to RPS instrumentation, i.e., all other Trip Functions),

has been shown to be acceptable (Refs.

5 and 6) to permit restoration of any inoperable channel to operable status.

This out of service time is only acceptable provided the associated Trip Function is still maintaining isolation capability (refer to next paragraph).

If the inoperable channel cannot be restored to operable status within the allowable out of service time, the channel must be placed in the tripped condition per Table 3.2.3 Note l.a.l) or l.a.2), as applicable.

Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

Alternately, Amendment No.

2-3-r 76r

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION shutdown condition, the actions and completion times of Specification 3.7.B.4.b shall apply.

After seven days with an inoperable train of the Standby Gas Treatment System during refueling or cold shutdown conditions requiring secondary containment integrity, the operable train of the Standby Gas Treatment System shall be placed in operation and its associated diesel generator shall be operable, or the actions and completion times of Specification 3.7.B.4.b shall apply.

4.

With two trains of the Standby Gas Treatment System inoperable, or as made applicable by Specification 3.7.B.3:

a.

With the reactor in the run mode, startup

mode, or hot shutdown condition, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b.

During movement of recently irradiated fuel assemblies in the secondary containment or during operations with the potential for draining the reactor vessel, immediately:

Amendment No.

4-71 155

VYNPS 3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS

i.

Suspend movement of recently irradiated fuel assemblies in secondary containment; and ii.

Deleted iii.

Initiate action to suspend operations with the potential for draining the reactor vessel.

C.

Secondary Containment System C.

Secondary Containment System

1. Secondary Containment Integrity shall be maintained during the following modes or conditions:

a.

Whenever the reactor is in the Run Mode, Startup Mode, or Hot Shutdown condition*;

or

1. Secondary containment capability to maintain a 0.15 inch of water vacuum under calm wind (2<Q<5 mph) conditions with a filter train flow rate of not more than 1,550 cfm, shall be demonstrated at least quarterly.

NOTE:

The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:

1. Reactor coolant temperature is < 212 0 F;

2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and

3.

No core alterations are in progress.

Amendment No.

44-4, 4-44, 4-94, 2-,L-,

2-2 155a

VYNPS 3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS I

b.

During movement of recently irradiated fuel assemblies in secondary containment; or

c.

Deleted

d.

During operations with the potential for draining the reactor vessel.

Amendment No.

4-4-,

4:-9, 2,L3, 2-2-6 156

VYNPS 3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS

2.

With Secondary Containment Integrity not maintained with the reactor in the Run Mode, Startup Mode, or Hot Shutdown condition, restore Secondary Containment Integrity within four (4) hours.

3.

If Specification 3.7.C.2 cannot be met, place the reactor in the Hot Shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in the Cold Shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

With Secondary Containment Integrity not maintained during movement of recently irradiated fuel assemblies in secondary containment of the Reactor Core, or during operations with the potential for draining the reactor vessel, immediately perform the following actions:

a.

Suspend movement of recently irradiated fuel assemblies in secondary containment; and

b.

Deleted

c.

Initiate action to suspend operations with the potential for draining the reactor vessel.

2.

Intentionally blank.

3.

Intentionally blank.

4.

Intentionally blank.

Amendment No.

4-9-7-,

2-2-1 157

VYNPS BASES:

3.7 (Cont'd) surveillances such as monthly torus to drywell vacuum breaker tests.

Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to prevent overpressurization of the SBGT system or the reactor building ductwork, should a LOCA occur.

For this and similar analyses performed, a spurious opening of AC-6 or AC-7 (one of the closed containment isolation valves) is not assumed as a failure simultaneous with a postulated LOCA.

Analyses demonstrate that for normal plant operation system alignments, including surveillances such as those described above, that SBGT integrity would be maintained if a LOCA was postulated.

Therefore, during normal plant operations, the 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock does not apply.

Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per calendar year (except for performance of the subject valve stroke time surveillances -

in which case the appropriate corresponding valves are closed to protect equipment should a LOCA occur).

This restriction will apply whenever primary containment integrity is required.

The 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock will apply anytime purge and vent evolutions can not assure the integrity of the SBGT trains or related equipment.

B. andC.

Standby Gas Treatment System and Secondary Containment System The secondary containment is designed to minimize any ground level release of radioactive materials which might result from a serious accident.

The Reactor Building provides secondary containment during reactor operation, when the drywell is sealed and in service; the Reactor Building provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.

Because the secondary containment is an integral part of the complete containment system, secondary containment is required at all times that primary containment is required except, however, for initial fuel loading and low power physics testing.

In the Cold Shutdown condition or the Refuel Mode,the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these conditions.

Therefore, maintaining Secondary Containment Integrity is not required in the Cold Shutdown condition or the Refuel Mode, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel, or during movement of recently irradiated fuel assemblies in the secondary containment.

"Recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core within the'previous 13 days, i.e. reactor fuel that has decayed less than 13 days following reactor shutdown.

This minimum decay period is enforced to maintain the validity of the Fuel Handling Accident dose consequence analysis.

In order for secondary containment integrity to be met, the secondary containment must function properly in conjunction with the operation of the Standby Gas Treatment System to ensure that the required vacuum can be established and maintained.

This means that the reactor building is intact with at least one door in each access opening closed, and all reactor building automatic ventilation system isolation valves are operable or the affected penetration flow path is isolated.

With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdown condition, if Secondary Containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition.

This time period also ensures that the probability of an accident (requiring Secondary Containment Integrity) occurring during periods where Secondary Containment Integrity is not maintained, is minimal.

Amendment No.

4-4, 4-4-3, 94-4, 4-45-, 4-946 165a

VYNPS BASES:

3.7 (Cont'd)

If Secondary Containment Integrity cannot be restored within the required time period, the plant must be brought to a mode or condition in which the LCO does not apply.

Movement of recently irradiated fuel assemblies in the secondary containment and operations with the potential for draining the reactor vessel can be postulated to cause fission product release to the secondary containment.

In such cases, the secondary containment is the only barrier to release of fission products to the environment.

Movement of recently irradiated fuel assemblies must be immediately suspended if Secondary Containment Integrity is not maintained.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position.

Also, action must be immediately initiated to suspend operations with the potential for draining the reactor vessel to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Actions must continue until operations with the potential for draining the reactor vessel are suspended.

Amendment No.

4-4-3, 44-4, 4-6-5 165b

VYNPS BASES:

3.7 (Cont'd)

The Standby Gas Treatment System (SGTS) is designed to filter and exhaust the Reactor Building atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the Reactor Building to the environs.

To insure that the standby gas treatment system will be effective in removing radioactive contaminates from the Reactor Building air, the system is tested periodically to meet the intent of ANSI N510-1975.

Laboratory charcoal testing will be performed in accordance with ASTM D3803-1989, except, as allowed by GL 99-02, testing can be performed at 70%

relative humidity for systems with humidity control.

Both standby gas treatment fans are designed to automatically start upon containment isolation and to maintain the Reactor Building pressure to approximately a negative 0.15 inch water gauge pressure; all leakage should be in-leakage.

Should the fan fail to start, the redundant alternate fan and filter system is designed to start automatically.

Each of the two fans has 100% capacity.

This substantiates the availability of the operable train and results in no added risk; thus, reactor operation or refueling operation can continue.

If neither train is operable, the plant is brought to a condition where the system is not required.

When the reactor is in cold shutdown or refueling the drywell may be open and the Reactor Building becomes the only containment system.

During cold shutdown the probability and consequences of a DBA LOCA are substantially reduced due to the pressure and temperature limitations in this mode.

However, for other situations under which significant radioactive release can be postulated, such as during operations with a potential for draining the reactor vessel or during movement of recently irradiated fuel in the secondary containment, operability of standby gas treatment is required.

Both trains of the Standby Gas Treatment System are normally operable when secondary containment integrity is required.

However, Specification 3.7.B.3 provides Limiting Conditions for Operation when one train of the Standby Gas Treatment System is inoperable.

Provisional, continued operation is permitted since the remaining operable train is adequate to perform the required radioactivity release control function.

If the applicable conditions of Specification 3.7.B.3 cannot be met, the plant must be placed in a mode or condition where the Limiting Conditions for Operation do not apply.

Entry into a refueling condition with one train of SBGTS inoperable is acceptable and there is no prohibition on mode or condition entry in this situation.

In this case, the requirements of TS 3.7.B.3.b are sufficient to ensure that adequate controls are in place.

During refueling conditions, accident risk is significantly reduced, and the primary activities of concern involve movement of recently irradiated fuel assemblies and OPDRVs.

During refueling and cold shutdown conditions Specification 3.7.B.3.b provides for the indefinite continuance of refueling operations with one train of the Standby Gas Treatment System inoperable.

When the seven-day completion time associated with Specification 3.7.B.3.b is not met and secondary containment integrity is required, the operable train of the Standby Gas Treatment System should immediately be placed into operation.

This action ensures that the remaining train is operable, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected.

An alternative to placing the operable train of Standby Gas Treatment in operation is to immediately suspend activities that represent a potential for releasing radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk.

Amendment No. 4*-, 4-9, 3, 4-8-4, 1 166

v t e Letter Sequence Request
TAC:MC2705, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Open)
Initiation Request Acceptance... Administration Questions 13 September 2004 Answers Results Approval... Other: ML043650060, ML051040065, ML051220095
MONTHYEAR2004-09-13 [Table View]

ML13323A518

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

SUMMARY

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