BVY 13-043, Technical Specifications Proposed Change No. 304, Revision to Low Pressure Safety Limit

Technical Specifications Proposed Change No. 304, Revision to Low Pressure Safety Limit

ML13137A158
Person / Time
Site:	Vermont Yankee
Issue date:	05/14/2013
From:	Wamser C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 13-043
Download: ML13137A158 (24)

Text

Entergy Nuclear Operations, Inc.

Vermont Yankee 320 Governor Hunt Rd SEntergy Vernon, VT 05354 Tel 802 257 7711 Christopher J. Wamser Site Vice President BVY 13-043 May 14, 2013 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Technical Specifications Proposed Change No. 304 Revision to Technical Specification Low Pressure Safety Limit Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

REFERENCES:

1. GE Energy - Nuclear, 10 CFR Part 21 Communication, "Potential to Exceed Low Pressure Technical Specification Safety Limit," SC05-03, dated March 29, 2005

2. Letter, USNRC to Entergy Nuclear Operations, Inc. "Grand Gulf Nuclear Station Unit 1-Issuance of Amendment RE: Extended Power Uprate (TAC No. ME4679)," dated July 18, 2012

Dear Sir or Madam:

In accordance with 10CFR50.90, Entergy Nuclear Operations, Inc. (ENO) is proposing an amendment to Renewed Facility Operating License DPR-28 for Vermont Yankee Nuclear Power Station (VY).

The proposed change would revise the VY Technical Specifications (TS) to reduce the reactor pressure associated with Fuel Cladding Safety Limits from 800 psia to 700 psia. The proposed change would address the potential to exceed the low pressure TS Safety Limit associated with a Pressure Regulator Failure-Maximum Demand (Open) transient as reported by General Electric Nuclear Energy in Reference 1. The proposed changes are consistent with similar changes approved for Grand Gulf Nuclear Station in Reference 2.

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

BVY 13-043 / Page 2 of 2 ENO requests review and approval of the proposed license amendment by June 1, 2014 and a 60 day implementation period from the date of the amendment approval.

There are no new regulatory commitments made in this letter.

If you have any questions on this transmittal, please contact Mr. Robert Wanczyk at 802-451-3166.

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

Executed on May 14, 2013.

Sincerely, CJW/plc 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 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. Richard V. Guzman, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08C2A Washington, DC 20555 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 13-043 Docket No. 50-271 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Change 304 Description and Evaluation of Proposed Changes

BVY 13-043 / Attachment 1 / Page 1 of 9

1.

SUMMARY

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

The proposed change would revise the VY Technical Specifications (TS) to reduce the reactor pressure associated with the Fuel Cladding Integrity Safety Limits (SL) from 800 psia to 700 psia in SLs 1.1 .A and 1.1 .B. The proposed changes would address the potential to exceed the low pressure SL associated with a Pressure Regulator Failure-Maximum Demand (Open) (PRFO) transient as reported by General Electric Nuclear Energy (GE) in Reference 1. The proposed change is consistent with similar changes approved for Grand Gulf Nuclear Station in Reference 2.

On March 29, 2005, GE issued a Safety Communication (SC 05-03) in accordance with 10 CFR 21.21 (d). SC 05-03 documented a reportable condition for a potential to exceed the low pressure TS SL (Reference 1). GE identified an unanalyzed condition where a PRFO may cause a TS SL to be violated since reactor pressure could drop below the current VY TS SL 1.1 .B limit of 800 psia (785 psig) for a few seconds while reactor power is above 23% (the value is a plant specific number) of rated thermal power. GE identified that even plants with a main steam isolation valve (MSIV) low pressure isolation setpoint (LPIS) _>785 psig may experience an anticipated operational occurrence (AOO) that potentially could violate the SL (GE considers a PRFO to be an AOO). GE informed the affected licensees that recent calculations showed that during the PRFO transient, reactor pressure could fall below the SL. Depending upon the LPIS, the margin to the low pressure SL may not be adequate. Lowering the low pressure TS safety limit to 700 psia (685 psig) is supported by the expanded GEXL critical power correlation applicability range for GE14 and GNF2 fuels that are currently co-resident in the VY reactor core.

Entergy reviewed the GEXL14 and GEXL17 critical power correlations in NEDC-32851 P-A, Rev. 5 and NEDC-33292P, Rev. 3 (References 3 and 4), for GE14 and GNF2 fuel respectively and has determined that they are applicable to the GE14 and GNF2 fuel used in the VY core. The proposed reduction in the current 800 psia reactor pressure limit in SL 1.1 .B to 700 psia is within the range of applicable pressures in the GEXL correlations.

The NRC has approved NEDC-32851 P-A for use. The GEXL1 7 correlation for GNF2 fuel, NEDC-33292P, is approved for use per NEDE-2401 1-P-A "General Electric Standard Application for Reactor Fuel (GESTAR II)" by reference. NEDE-24011 specifically states:

Fuel design compliance with the fuel licensing acceptance criteriaconstitutes USNRC acceptanceand approval of the fuel design without specific USNRC review. The fuel licensing acceptance criteriaare presented in the subsections that follow.

The fuel licensing acceptance criteria for a new critical power correlation can be found in GESTAR II subsection 1.1.7. NEDC-33270P (Reference 5) documents that GESTAR II subsection 1.1.7 criteria for a new correlation are met. Therefore, per GESTAR II, the GEXL17 correlation is approved for use.

Entergy has determined that changing the pressure limit in SL 1.1 .B to 700 psia provides greater margin for the PRFO transient, with reactor pressure expected to remain above the revised SL 1.11.B limit and that the change resolves the 10 CFR Part 21 condition.

BVY 13-043 / Attachment 1 / Page 2 of 9

2. DETAILED DESCRIPTION The following change is proposed to Fuel Cladding Integrity SL 1.1 .A "Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow >10% of Rated)" and SL 1.1 .B "Core Thermal Power Limit (Reactor Pressure <800 psia or Core Flow <10% of Rated)":

Current SL 1.1 .A Proposed SL 1.1.A A. Bundle Safety Limit (Reactor Pressure A. Bundle Safety Limit (Reactor Pressure

>800 psia and Core Flow >10% of Rated) >700 psia and Core Flow >10% of Rated)

When the reactor pressure is >800 psia When the reactor pressure is >700 psia and the core flow is greater than 10% of and the core flow is greater than 10% of rated: rated:

1. A Minimum Critical Power Ratio (MCPR) 1. A Minimum Critical Power Ratio (MCPR) of less than 1.09 (1.10 for Single Loop of less than 1.09 (1.10 for Single Loop Operation) shall constitute violation of the Operation) shall constitute violation of the Fuel Cladding Integrity Safety Limit Fuel Cladding Integrity Safety Limit (FCISL). (FCISL).

Current SL 1.1 .B Proposed SL 1.1.8 B. Core Thermal Power Limit (Reactor B. Core Thermal Power Limit (Reactor Pressure *800 psia or Core Flow 510% of Pressure *700 psia or Core Flow *10% of Rated) Rated)

When the reactor pressure is <800 psia or When the reactor pressure is *700 psia or core flow *10% of rated, the core thermal core flow 510% of rated, the core thermal power shall not exceed 23% of rated power shall not exceed 23% of rated thermal power. thermal power.

Associated wording in the VY TS Bases will also be revised after approval of this change, under the TS Bases Control Program specified in VY TS 6.7.E. Changes to TS Bases are provided for information only.

3. TECHNICAL EVALUATION 3.1 Pressure Regulator Failed Open Transient Section 7.11 of the VY UFSAR describes the pressure regulator and Turbine Generator Control System. The pressure regulator and Turbine Generator Control System control steam flow and pressure to the turbine and protect the turbine from overpressure or excessive speed. The main turbine generator controls work in conjunction with the Nuclear Steam System controls to maintain essentially constant reactor pressure and limit reactor transients during load variations. During normal planned operation, the steam admitted to the turbine is controlled by the pressure regulator which maintains essentially constant pressure at the turbine inlet, thus controlling reactor vessel pressure. This control scheme forces turbine generator output to follow reactor steam output. Changing recirculation flow or moving control rods changes the steam flow available from the reactor. The change in recirculation flow or rod motion directly changes the reactor steaming rate, and the controlling pressure regulator reacts by appropriately opening or closing the turbine

BVY 13-043 / Attachment 1 / Page 3 of 9 admission or bypass valves. Thus, the turbine and/or main condenser absorbs any change in reactor power.

Two pressure regulators, one electrical and one mechanical, are provided, one intended for use as a backup to the controlling regulator, either one of which can be used for control purposes with the unit at rated turbine inlet pressure. The controlling pressure regulator is used to control both the turbine admission valves and the turbine bypass valves.

Normally the bypass valves are held closed and the pressure regulator controls the admission valves using all the steam production to make electrical power. Ifthe speed controls or electrical load demand a reduced steam flow to the turbine, the pressure regulator functions to open the turbine bypass valves to send the excess steam flow to the main condenser. The backup pressure regulator functions to assure pressure control in the event of failure of the controlling regulator; its setpoint is normally a few psi above that of the controlling regulator.

The PRFO transient analysis for VY is described in Section 14.5.4.1 of the UFSAR. If either the controlling pressure regulator or the backup pressure regulator fails in an open direction, the turbine admission valves can be fully opened, and the turbine bypass valves can be partially opened. This action initially results in decreasing coolant inventory in the reactor vessel as the mass flow of steam leaving the vessel exceeds the mass flow of water entering the vessel. The total steam flow rate resulting from a pressure regulator malfunction is limited by the turbine controls to about 110% of design flow. The turbine bypass valves with 110% full capacity are assumed open for faster depressurization. The main steam line low pressure isolation setpoint is conservatively assumed to be 750 psig.

(The actual main steam line low pressure trip setting is a800 psig in accordance with TS 2.1 .H and Table 3.2.2, Function 1.d)

Initially, the vessel steam flow increases rapidly as the turbine control valves open because of the pressure regulator failure. With the high steam outflow, the vessel pressure decreases, which results in vessel water level swell as the bulk fluid void volume is increased. At 4.8 seconds, the water level reaches the high water level trip setpoint. This initiates Main Turbine Stop Valve Closure (MTSVC) and feedwater pump trip, followed by a reactor scram on MTSV position. At this time, the turbine bypass valves are fully opened, resulting in further depressurization of the vessel until the 750 psig turbine inlet pressure limit is reached, which initiates MSIV-closure. Following isolation, the system depressurization is terminated and the vessel water level begins to fall as the system pressure increases. The reactor vessel isolation limits the duration and severity of the depressurization. No significant thermal stresses are imposed on the nuclear system process barrier. Following isolation, the nuclear system safety relief valves operate intermittently to relieve the pressure rise resulting from decay heat generation. No reductions in fuel thermal margins occur. The nuclear system process barrier is not threatened by high internal pressure for this pressure regulator malfunction.

In SC05-03, GE concluded that during the PRFO, the Critical Power Ratio (CPR) increases during depressurization, the initial CPR is the limiting CPR condition during the entire transient, and that the conditions that exceed the low pressure SL exist for only a few seconds, fuel cladding integrity is not threatened. Nevertheless, GE considers the PRFO to be a known AOO that could contribute to the exceeding of a SL.

3.2 Safety Limits TS SLs are specified to ensure that acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and AQOs. Fuel Cladding Integrity

BVY 13-043 / Attachment 1 / Page 4 of 9 SLs are set such that fuel cladding integrity is maintained and no significant fuel damage would occur if the SLs are not exceeded.

Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur.

Although it is recognized that the onset of transition boiling would not result in damage to boiling water reactor fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties, in monitoring the core operating state and in the procedures used to calculate the critical power, result in an uncertainty in the value of the critical power. Therefore, the Fuel Cladding Integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The minimum critical power ratio (MCPR) SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved GE critical power correlations.

For VY SL 1.1 .B, the GE critical power correlation (also known as the GEXL critical power correlation) is applicable for operation at pressures greater than or equal to 800 psia and core flows greater than or equal to 10% of rated flow. A core power limit of 23% rated thermal power ensures consistency with the threshold for requiring thermal limit monitoring (i.e., average planar linear heat generation rate, linear heat generation rate, and MCPR).

This assures that for those power levels where thermal limit monitoring is required, the GE critical power correlation is applicable. This SL was introduced to ensure the validity of MCPR calculations when power is > 23% and the reactor pressure is within the validity range of the GEXL correlation. GE has updated the validity range of GEXL correlations via References 3 and 4, which allows the pressure to be reduced to 685 psig (700 psia) from 785 psig (800 psia). Therefore a wider pressure range is available for transients to demonstrate compliance with MCPR limits. Thus, the proposed change offers a greater pressure margin for a PRFO transient than what is currently available.

3.3 No Impact on the Low Main Steam Line Pressure Trip Function The Low Main Steam Line Pressure trip function is directly assumed in the analysis of the pressure regulator failure. The allowable value for Low Main Steam Line Pressure trip function (TS 2.1 .H and Table 3.2.2, Function 1.d) for VY is greater than or equal to 800 psig.

For the pressure regulator failure event, closure of the MSIVs ensures that the reactor pressure vessel temperature change limit (1000F/hr) is not reached. Also, as discussed in the TS Bases, this function supports actions to ensure that SL 1.1 .B is not exceeded. This function is described as closing the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to less than 23 percent of rated thermal power. The proposed change to reduce the reactor pressure from 800 psia to 700 psia does not change the function of the Low Main Steam Line Pressure trip, but does increase the margin between the setpoint and the safety limit the setpoint is protecting.

No changes are required or proposed to any instrumentation settings associated with the Low Main Steam Line Pressure trip function, including the TS allowable value. The TS Bases description for this function is revised to indicate the change in reactor pressure.

BVY 13-043 / Attachment 1 / Page 5 of 9 The trip on low main steam line pressure will occur as previously specified, at the assumed instrument settings discussed above. The Low Main Steam Line Pressure trip function setpoint provides added assurance that with the revised reactor pressure of 700 psia, that SL 1.1 .B would not be violated under realistic conditions.

3.4 Conclusion The proposed change to the Fuel Cladding Integrity SLs continues to ensure that a valid CPR calculation is performed for the AQOs described in the VY UFSAR, including the PRFO transient. VY has determined that with the value of 700 psia proposed for the reactor pressure, a PRFO transient would not result in a violation of SL 1.1..B. Since this approach follows, and is consistent with, the way the reactor pressure has been established and valid CPR calculations will continue to be performed, it is a safe and appropriate method to address the 10 CFR Part 21 issue.

4. REGULATORY EVALUATION 4.1 APPLICABLE REGULATORY REQUIREMENT/CRITERIA The proposed change addresses an issue identified in a 10 CFR Part 21 communication regarding the potential for boiling water reactors to experience reactor pressure below the low pressure SL of 785 psig defined in standard Improved Technical Specifications SL 2.1.1.1 under certain transient conditions.

The construction permit for VY was issued by the Atomic Energy Commission (AEC) on December 11, 1967. As discussed in Appendix F of the VY UFSAR, the plant was designed and constructed based on the proposed General Design Criteria (GDC) published by the AEC in the FederalRegister (32 FR 10213) on July 11, 1967 (referred to as "draft GDC").

The AEC published the final rule that added Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR 50 in the FederalRegister (36 FR 3255) on February 20, 1971 (referred to as "final GDC").

Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. As discussed in the NRC Staff Requirements Memorandum for SECY-92-223, dated September 18, 1992 (ML003763736), the NRC decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971. At the time of issuance of 10 CFR 50 Appendix A, the NRC stressed that the final GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. As such, VY was not licensed to the final GDC.

Appendix F of the UFSAR provides a plant comparative evaluation with the proposed AEC 70 design criteria. VY has determined that draft GDC 6, "Reactor Core Design" is applicable to this proposed change. The draft GDC 6 requirements are as follows:

Draft GDC 6 - Reactor Core Design The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified.

The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of off-site power.

BVY 13-043 / Attachment 1 / Page 6 of 9 Attachment 2 to Entergy letter BVY 03-090, dated October 1, 2003 (ML032810447),

provides a matrix of the draft GDCs versus the corresponding final GDCs. Based on Attachment 2 of letter BVY 03-090, final GDC 10 corresponds to draft GDC 6. Conformance with the fuel licensing criteria of final GDC 10, "Reactor Design," is achieved by preventing the violation of fuel design limits. Final GDC 10 states:

GDC 10 - Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

VY has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. As long as the core pressure and flow are within the range of validity of the specified critical power correlation, in this case the GEXL14 and GEXL 17 critical power correlations, the proposed reactor pressure change to SLs 1.1 .A and 1.1..B will continue to ensure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition. This satisfies the requirements of GDC 10 regarding acceptable fuel design limits and continues to assure that the underlying criteria of the safety limit is met. Based on this, there is reasonable assurance that the health and safety of the public, following approval of this TS change, is unaffected.

10 CFR 50.36(c)(1) requires that SLs be included in the TS. SLs for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any SL is exceeded, the reactor must be shut down. The proposed change modifies an existing SL.

4.2 PRECEDENT The proposed change is consistent with similar changes approved for Grand Gulf Nuclear Station in Reference 2.

4.3 NO SIGNIFICANT HAZARDS CONSIDERATION 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.

The proposed change would revise the Vermont Yankee (VY) Technical Specifications (TS) to reduce the reactor pressure associated with the Fuel Cladding Integrity Safety Limits from 800 psia to 700 psia in Safety Limits 1.1 .A and 1.1 .B. The proposed changes would address the potential to exceed the low pressure TS safety limit associated with a Pressure Regulator Failure-Maximum Demand (Open) (PRFO) transient as reported by General Electric Nuclear Energy (GE) in Reference 1. The proposed change is consistent with similar changes approved for Grand Gulf Nuclear Station in Reference 2.

BVY 13-043 / Attachment 1 / Page 7 of 9 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 significant increase in the Probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to the reactor pressure in Fuel Cladding Integrity Safety Limits 1.1 .A and 1.1.B does not alter the use of the analytical methods used to determine the safety limits that have been previously reviewed and approved by the NRC. The proposed change is in accordance with NRC approved critical power correlation methodologies and as such maintains required safety margins. The proposed change does not adversely affect accident initiators or precursors nor does it alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated and maintained.

The proposed change does not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed change does not require any physical change to any plant SSCs nor does it require any change in systems or plant operations. The proposed change is consistent with the safety analysis assumptions and resultant consequences.

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.

There are no hardware changes nor are there any changes in the method which any plant systems perform a safety function. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change.

The proposed change does not introduce any new accident precursors, nor does it involve any physical plant alterations or changes in the methods governing normal plant operation. Also, the change does not impose any new or different requirements or eliminate any existing requirements. The change does not alter assumptions made in the safety analysis.

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.

Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to

BVY 13-043 / Attachment 1 / Page 8 of 9 perform their design functions during and following postulated accidents.

Evaluation of the 10 CFR Part 21 issue by General Electric determined that the PRFO transient provides additional margin to the Minimum Critical Power Ratio Safety Limit and is not a threat to fuel cladding integrity.

The proposed change to Fuel Integrity Cladding Safety Limits 1.1 .A and 1.1 .B is consistent with, and within the capabilities of the applicable NRC approved critical power correlations, and thus continues to ensure that valid critical power calculations are performed. No setpoints at which protective actions are initiated are altered by the proposed change. The proposed change does not alter the manner in which the safety limits are determined.

This change is consistent with plant design and does not change the TS operability requirements; thus, previously evaluated accidents are not affected by this proposed change.

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.

5. ENVIRONMENTAL CONSIDERATIONS This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10CFR51.22(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 off site.

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 10CFR51.22(c)(9). Pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6. REFERENCES

1. GE Energy - Nuclear, 10 CFR Part 21 Communication, "Potential to Exceed Low Pressure Technical Specification Safety Limit," SC05-03, dated March 29, 2005

BVY 13-043 / Attachment 1 / Page 9 of 9

2. Letter, USNRC to Entergy Nuclear Operations, Inc. "Grand Gulf Nuclear Station Unit 1-Issuance of Amendment RE: Extended Power Uprate (TAC No. ME4679)," dated July 18, 2012

3. NEDC-32851P-A, Rev. 4, "GEXL14 Correlation for GE14 Fuel", dated September 15, 2008

4. NEDC-33292P, Rev 3, "GEXL17 Correlation for GNF2 Fuel", dated June 2009

5. NEDC-33270P, Rev. 4, "GNF2 Advantage Generic Compliance with NEDE-2401 1-P-A (GESTAR II)", dated October 2011

BVY 13-043 Docket 50-271 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Change 304 Markup of the Current Technical Specifications and Bases Pages

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the interrelated Applies to trip setting of the variable associated with fuel instruments and devices which are thermal behavior. provided to prevent the nuclear system safety limits from being exceeded.

Objective: Objective:

To establish limits below which the To define the level of the process integrity of the fuel cladding is variable at which automatic preserved. protective action is initiated.

Specification: Specification:

A. Bundle Safety Limit (Reactor A. Trip Settings Pressure >BHla psia and Core Flow

>10% of Rated) - - The limiting safety system trip settings shall be as When the reactor pressure is specified below:

>**0a psia and the core flow is grea than 10% of rated: 1. Neutron Flux Trip Settinqs 2 1. A MidraruTrt Critical Power Ratio (MCPR) of less than 1.09 (1.10 for Single Loop Operation) shall constitute violation of the Fuel Cladding Integrity Safety Limit (FCISL).

a. APRM Flux Scram Allowable Value (Run Mode)

When the mode switch is in the RUN position, the APRM flux scram Allowable Value shall be:

Two loop operation:

S* 0.33W+ 50.45% for 0% < W

• 30.9%

S* 1.07W+ 27.23% for 30.9% < W

• 66.7%

S5 0.55W+ 62.34% for 66.7% < W

• 99.0%

With a maximum of 117.0% power for W >

99.0%

Single loop operation:

S* 0.33W+ 48.00% for 0% < W

• 39.1%

S5 1.07W+ 19.01% for 39.1% < W

• 61.7%

S* 0.55W+ 51.22% for 61.7% < W ! 119.4%

With a maximum of 117.0% power for W >

119.4%

where:

S = setting in percent of rated thermal power (1912 MWt)

Amendment No. 4-4, 4-, -64, 44, -94, 4-0-9, 4-5-5, 4*-9, 4-74, a-44, 24-9-, 2-2-9,46 6

VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit W percent rated two loop (Reactor Pressure :ý444 psia drive flow where 100%

or Core Flow I10% of 7ed) rated drive flow is that flow equivalent When the reactor pressure is to 48 x 106 lbs/hr core

<go) ia or core flow *10% flow of ra the core thermal power sha -,not exceed 23% of In the event of operation at rated thermaI wer.

> 23% Rated Thermal Power the 700 APRM gain shall be equal to or C. Power Transient greater than 1.0.

To ensure that the safety limit established in Specification 1.1A and 1.1B is not exceeded, each required scram shall be initiated by its expected scram signal. The safety limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

D. Whenever the reactor is shutdown with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the enriched fuel when it is seated in the core.

Amendment No. 4-4, -,4Z,-4:, 6 4&, *64, -4, 4-i4, 4-8&4, 3-4, 2299 7

VYNPS BASES:

1.1 FUEL CLADDING INTEGRITY A. Refer to General Electric Company Licensing Topical Report, "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (most recent revision).

The fuel cladding integrity Safety Limit (SL) is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations.

The MCPR fuel cladding integrity SL is increased for single loop operation in order to account for increased core flow measurement and TIP reading uncertainties.

B. Core Thermal Power Limit (Reactor Pressure

• SfLL psia or Core Flow

=10% of Rated)

The General Electric critical power correlationý nown as the GEXL critical power correlation) is applicable for operation at pressures greater than or equal to BLL psia and core flows greater than or equal to 10% of rated flow. For opw)ation at lower pressures or core flows, the following basis is used:

At power levels at or below the low pressure, low flow (low power) thermal limit, the minimum core flow occurs for natural circulation, and as the power to flow ratio in natural circulation increases with increasing power, the maximum and most limiting power to flow ratio occurs for natural circulation at the low power thermal limit. This condition is therefore also the condition with the minimum margin to critical power. Analysis of the natural circulation flow rate at the low power thermal limit has shown that the core average mass flux is 0.3-0.4 Mlb/hr-ft 2 and the corresponding core pressure drop is 5-6 psi. For these conditions, full scale ATLAS test data have shown a critical power of 4-5 MWt. Analysis has also shown that a maximum radial peaking factor of 2 is expected at the low power thermal limit condition. Since the low power thermal limit basis corresponds to a maximum average bundle power of 1.2 MWt or less, fuel bundles with radial peaking factor as high as 3 will have margin to critical power. This bounds any radial peaking, and therefore the low power thermal limit is conservative. An average bundle power of 1.2 MWt occurs at 23% rated thermal power. Thus, a limit of 23% rated thermal power for operation with reactor pressure less than or equal to .LL psia is conservative.

Amendment No. 4-4, -34, 4-4, 94, 4-&&, 292_9 12

VYNPS BASES: 1.1 (Cont'd)

With no reactor coolant recirculation loops in operation, the plant must be brought to a condition in which the LCO does not apply. Operation of at least one reactor coolant recirculation loop provides core flow greater than natural circulation, so the margin to a critical power condition is significantly greater than this bounding example for all normal operating conditions with power less than the low power thermal limit. Therefore, a low power thermal limit of 23% rated thermal power is conservative.

Additionally, a core thermal power limit of 23% rated thermal power ensures consistency with the threshold for requiring thermal limit monitoring (i.e.,

average planar linear heat generation rate, linear heat generation rate, and minimum critical power ratio). This assures that for those power levels where thermal limit monitoring is required, the General Electric critical power correlation is applicable.

C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.1A or 1.1.1B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

The computer provided with Vermont Yankee has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.

This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.

D. Reactor Water Level (Shutdown Condition)

During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat.

If reactor water level should drop below the top of the enriched fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the enriched fuel provides adequate margin. This level will be continuously monitored.

Amendment No. 9, &&, 4--G, 292.9 13

VYNPS BASES: 3.2.B/4.2.B PRIMARY CONTAINMENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) the isolation is initiated on high flow to prevent or minimize core damage.

The High Main Steam Line Flow Trip Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 4). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 50.67 limits.

The MSL flow signals are initiated from 16 differential pressure transmitters that are connected to the four MSLs (the differential pressure transmitters sense differential pressure across a flow restrictor) . The differential pressure transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of High Main Steam Line Flow Trip Function for each MSL (two channels per trip system) are available and are required to be operable so that no single instrument failure will preclude detecting a break in any individual MSL.

The Trip Setting is chosen to ensure that fuel peak cladding temperature and offsite dose limits are not exceeded due to the break.

This Trip Function isolates the Group 1 valves.

l.d. Low Main Steam Line Pressure Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100'F/hr if the pressure loss is allowed to continue. The Low Main Steam Line Pressure Trip Function is directly assumed in the analysis of the pressure regulator failure (Ref. 5).

For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100 0 F/hr) is not reached. In addition, this Trip Function supports actions to ensure that Safety Limit 1.1.B is not exceeded. (This Trip Function closes the MSIVs at Ž800 psig prior to pressure decreasing below 2a- psig [Ann. psia], which results in a scram due to MSIV closure, thus reducing r pq to < 23% RATED THERMAL POWER.)

The MSL low ý7ýure s +/-graims are initiated from four pressure switches that are connected to the MSL header. The switches are arranged such that, even though physically separated from each other, each pressure switch is able to detect low MSL pressure. Four channels of Low Main Steam Line Pressure Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Setting was selected to be high enough to prevent excessive RPV depressurization.

The Low Main Steam Line Pressure Trip Function is only required to be operable in the RUN Mode since this is when the assumed transient can occur (Ref. 5).

This Trip Function isolates the Group 1 valves.

Amendment No. 2_3*r 76d

BVY 13-043 Docket 50-271 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Change 304 Retyped Technical Specifications and Bases Pages

VYN PS L. I :-SAFETY ti, iT 2 1.l. MFT[H'.,  :]AFETY SYSTEM SETTING L. I FUEL CLADDING INTEGRITY 2. 1 FUEL CLADDING [NTEGRITY Applicability: AppLicability:

Applies to the interreLated Applies to trip setting of the variable associated with fuel instruments and devices which are thermal behavior. provided to prevent the nuclear system safety limits from being exceeded.

Objective: Objective:

To establish limits below which the To define the level of the process integrity of the fuel cladding is variable at which automatic preserved. protective action is initiated.

Specification: Specification:

A. Bundle Safety Limit (Reactor A. Trip Settings Pressure >700 psia and Core Flow

>10% of Rated) The limiting safety system trip settings shall be as When the reactor pressure is specified below:

>700 psia and the core flow is greater than 10% of rated: 1. Neutron Flux Trip Settings

1. A Minimum Critical Power Rati O a. APRM Flux Scram (MCPR) of less than 1.07 (1.09 Allowable Value for Single Loop Operation) (Run Mode) shall constitute violation of the Fuel Cladding Integrity When the mode switch Safety Limit (FCISL). is in the RUN position, the APRM flux scram Allowable Value shall be:

Two loop operation:

S5 0.33W+ 50.45% for 0% < W 30.9%

S* 1.07W+ 27.23% for 30.9% < W 66.7%

S* 0.55W+ 62.34% for 66.7% < W 99.0%

With a maximum of 117.0% power for W >

99.0%

Single loop operation:

Ss 0.33W+ 48.00% for 0% < W

• 39.1%

S* 1.07W+ 19.01% for 39.1% < W

• 61.7%

S* 0.55W+ 51.22% for 61.7% < W

• 119.4%

With a maximum of 117.0% power for W >

119.4%

where:

S setting in percent of rated thermal power (1912 MWt)

Amendment No. 4-&, 4-4, 4&, 4G, -4, 444-, 4-2-1, 4-7-, 4--, 2-24-, 24-3, 6

VYNPS L. I SAFETY LIMIT 2. 1 L[,M[TING SAFETY SYSTEM SETTING B. Core Thermal Power Limit W = percent rated two loop (Meactor Pressure *1700 Dsia drive flow where 100%

or Core Flow <10% of Rated) rated drive flow is When the reactor pressure is that flow equivalent to 48 x 106 lbs/hr core

• 700 psia or core flow *10% flow of rated, the core thermal power shall not exceed 23% of rated thermal power. In the event of operation at

> 23% Rated Thermal Power the C. Power Transient APRM gain shall be equal to or greater than 1.0.

To ensure that the safety limit established in Specification 1.1A and 1.1B is not exceeded, each required scram shall be initiated by its expected scram signal. The safety limit shall be assumed to be exceeded when scram is accomplished by means other than the expected scram signal.

D. Whenever the reactor is shutdown with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the enriched fuel when it is seated in the core.

Amendment No. 4-4, -4, 4-, 4, 4&4-,4*&, 44, 44-1-, 4-84-, 2-4-, 2-24, 7

VYN PS BASES:

. I FUEL CLADDING iNTEGRITY A. Refer to General Electric Company Licensing Topical Report, "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (most recent revision).

The fuel cladding integrity Safety Limit (SL) is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations.

The MCPR fuel cladding integrity SL is increased for single loop operation in order to account for increased core flow measurement and TIP reading uncertainties.

B. Core Thermal Power Limit (Reactor Pressure

• 700 psia or Core Flow

ý10% of Rated)

The General Electric critical power correlation (also known as the GEXL critical power correlation) is applicable for operation at pressures greater than or equal to 700 psia and core flows greater than or equal to 10% of rated flow. For operation at lower pressures or core flows, the following basis is used:

At power levels at or below the low pressure, low flow (low power) thermal limit, the minimum core flow occurs for natural circulation, and as the power to flow ratio in natural circulation increases with increasing power, the maximum and most limiting power to flow ratio occurs for natural circulation at the low power thermal limit. This condition is therefore also the condition with the minimum margin to critical power. Analysis of the natural circulation flow rate at the low power thermal limit has shown 2

that the core average mass flux is 0.3-0.4 Mlb/hr-ft and the corresponding core pressure drop is 5-6 psi. For these conditions, full scale ATLAS test data have shown a critical power of 4-5 MWt. Analysis has also shown that a maximum radial peaking factor of 2 is expected at the low power thermal limit condition. Since the low power thermal limit basis corresponds to a maximum average bundle power of 1.2 MWt or less, fuel bundles with radial peaking factor as high as 3 will have margin to critical power. This bounds any radial peaking, and therefore the low power thermal limit is conservative. An average bundle power of 1.2 MWt occurs at 23% rated thermal power. Thus, a limit of 23% rated thermal power for operation with reactor pressure less than or equal to 700 psia is conservative.

Amendment No. -4, 3-4, 4-4, -94, 1-&-&, ,19, 12

'IY N P S BASES: 1.1 (Cont'd)

With no reactor coolant recirculation loops in operation, the plant must be brought to a condition in which the LCO does not apply. Operation of at Least one reactor coolant recirculation Loop provides core flow greater than natural circulation, so the margin to a crit Lca i power condition i-s significantLy greater than this bounding example for all normal operating conditions with power less than the low power thermal limit. Therefore, a low power thermal limit of 23% rated thermal power is conservative.

Additionally, a core thermal power Limit of 23% rated thermal power ensures consistency with the threshold for requiring thermal limit monitoring (i.e.,

average planar linear heat generation rate, linear heat generation rate, and minimum critical power ratio). This assures that for those power levels where thermal limit monitoring is required, the General Electric critical power correlation is applicable.

C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.IA or 1.1.lB will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

The computer provided with Vermont Yankee has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur.

This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.

D. Reactor Water Level (Shutdown Condition)

During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat.

If reactor water level should drop below the top of the enriched fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the enriched fuel provides adequate margin. This level will be continuously monitored.

Amendment No. 4-1, &8-, I-&G, 2-2-4, 13

VYNPS BASES: 3.2.13/4.2.3 PRIMARY CONTAIN'MENT ISOLATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) the isolation is initiated on high flow to prevent or minimize core damage.

The High Main Steam Line Flow Trip Function is directly assumed in the analysis off the main steam line break (MSLB) (Ref. 4). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 50.67 limits.

The MSL flow signals are initiated from 16 differential pressure transmitters that are connected to the four MSLs (the differential pressure transmitters sense differential pressure across a flow restrictor). The differential pressure transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of High Main Steam Line Flow Trip Function for each MSL (two channels per trip system) are available and are required to be operable so that no single instrument failure will preclude detecting a break in any individual MSL.

The Trip Setting is chosen to ensure that fuel peak cladding temperature and offsite dose limits are not exceeded due to the break.

This Trip Function isolates the Group 1 valves.

l.d. Low Main Steam Line Pressure Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100*F/hr if the pressure loss is allowed to continue. The Low Main Steam Line Pressure Trip Function is directly assumed in the analysis of the pressure regulator failure (Ref. 5).

For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100 0 F/hr) is not reached. In addition, this Trip Function supports actions to ensure that Safety Limit l.1.B is not exceeded. (This Trip Function closes the MSIVs at Ž800 psig prior to pressure decreasing below 685 psig [700 psia], which results in a scram due to MSIV closure, thus reducing reactor power to < 23% RATED THERMAL POWER.)

The MSL low pressure signals are initiated from four pressure switches that are connected to the MSL header. The switches are arranged such that, even though physically separated from each other, each pressure switch is able to detect low MSL pressure. Four channels of Low Main Steam Line Pressure Trip Function are available and are required to be operable to ensure that no single instrument failure can preclude the isolation function.

The Trip Setting was selected to be high enough to prevent excessive RPV depressurization.

The Low Main Steam Line Pressure Trip Function is only required to be operable in the RUN Mode since this is when the assumed transient can occur (Ref. 5).

This Trip Function isolates the Group 1 valves.

Amendment No. a--3-6 76d