View source for BVY 06-001, Vermont Yankee - Technical Specification Proposed Change No. 263 - Supplement No. 44 Extended Power Uprate - Revised License and Technical Specifications

{{Adams
| number = ML060170320
| issue date = 01/10/2006
| title = Vermont Yankee - Technical Specification Proposed Change No. 263 - Supplement No. 44 Extended Power Uprate - Revised License and Technical Specifications
| author name = Thayer J
| author affiliation = Entergy Nuclear Northeast, Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000271
| license number = DPR-028
| contact person = 
| case reference number = %dam200606, BVY 06-001, TAC MC0761
| document type = Letter type:BVY, Technical Specification, Bases Change
| page count = 37
| project = TAC:MC0761
| stage = Supplement
}}

=Text=
{{#Wiki_filter:Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
Vermont Yankee P.O. Box 0500
-E-          tery                                                               185 Old Ferry Road Brattleboro, VT 05302-0500 Tel 802 257 5271 January 10, 2006 Docket No. 50-271 BVY 06-001 TAC No. MC0761 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==Subject:==
Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 - Supplement No. 44 Extended Power Uprate - Revised License and Technical Specifications

==Reference:==
: 1) Entergy letter to U.S. Nuclear Regulatory Commission, "Vermont Yankee Nuclear Power Station, License No. DPR-28 (Docket No. 50-271), Technical Specification Proposed Change No. 263, Extended Power Uprate," BVY 03-80, September 10, 2003 This letter provides additional information regarding the application by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy) for a license amendment (Reference 1), as supplemented, to increase the maximum authorized power level of the Vermont Yankee Nuclear Power Station (VYNPS) from 1593 megawatts thermal (MWt) to 1912 MWt.
In support of the NRC staff granting the license amendment for extended power uprate (EPU),
Entergy is providing in Attachment 1 revised facility operating license (FOL) and Technical Specifications (TS) pages in final format. The revised pages incorporate the FOL and TS changes proposed as part of the Proposed Change No. 263 license amendment request. The re-typed pages are based on the current FOL and TS through amendment no. 228.
Additionally, during initial EPU power ascension, Entergy is planning to monitor offsite dose as part of its radiation measurement activities. To ensure that the radiation dose to an actual member of the public remains within the 40 CFR 190 annual limit of 25 mrem, as referenced by 10 CFR 20 paragraph 20.1301(e), during EPU power ascension testing, Entergy will monitor direct radiation dose at the highest offsite dose boundary location (i.e., the west site boundary) at each 5% power increment above currently licensed thermal power. Once power ascension testing up to approximately 1912 MWt is completed, the monitoring information will be documented in a report, reviewed by the VYNPS Operational Safety Review Committee and made available for inspection.
There are no new regulatory commitments contained in this submittal.
This supplement to the license amendment request provides additional information regarding Entergy's application for a license amendment and does not change the scope or conclusions in the original application, nor does it change Entergy's determination of no significant hazards consideration.

BVY 06-001 Docket No. 50-271 Page 2 of 2 If you have any questions or require additional information, please contact Mr. James DeVincentis at (802) 258-4236.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on January J/O, 2006.
Sincerely,
                                               ,/ayK.A-hayer
(/51e Vice President Vermont Yankee Nuclear Power Station Attachment (1) cc:     Mr. Samuel J. Collins (w/o attachment)
Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. Richard B. Ennis, Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0 8 B1 Washington, DC 20555 USNRC Resident Inspector (w/o attachment)
Entergy Nuclear Vermont Yankee, LLC P.O. Box 157 Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

)
BVY 06-001 Docket No. 50-271 Attachment 1 Vermont Yankee Nuclear Povwer Station Proposed Technical Specification Change No. 263 - Supplement No. 44 Extended Power Uprate Revised License and Technical Specifications Total number of pages in Attachment 1 (excluding this cover sheet) is 34.

IC Attachment I to BVY 06-001 Docket No. 50-271 Page 1 of 34 ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License Remove                        Insert Page 3                        Page 3 Page 9                        Page 9 Page 10 Page 11 Page 12 Page 13 Technical Specifications Remove                        Insert Page 3                        Page 3 Page 6                        Page 6 Page 7                        Page 7 Page 10                       Page 10 Page 12                       Page 12 Page 13                       Page 13 Page 14                       Page 14 Page 15                       Page 15 Page 17                       Page 17 Page 21                       Page 21 Page 24                       Page 24 Page 30                       Page 30 Page 83                       Page 83 Page 90                       Page 90 Page 92                       Page 92 Page 94                       Page 94 Page 97                       Page 97 Page 98                       Page 98 Page 135                      Page 135 Page 136                      Page 136 Page 137                      Page 137 Page 138                      Page 138 Page 142                      Page 142 Page 224                      Page 224 Page 225                      Page 225 Page 226                      Page 226 Page 228                      Page 228

E. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.
: 3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A. Maximum Power Level Entergy Nuclear Operations, Inc. is authorized to operate the facility at reactor core power levels not to exceed 1912 megawatts thermal in accordance with the Technical Specifications (Appendix A) appended hereto.
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.           , are hereby incorporated in the license. Entergy Nuclear Operations, Inc. shall operate the facility in accordance with the Technical Specifications.
C. Reports Entergy Nuclear Operations, Inc. shall make reports in accordance with the requirements of the Technical Specifications.
D. This paragraph deleted by Amendment No. 226.
E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:
Amendment No. 206, 208, 226
: b. Surety (i) The surety agreement must be in a form acceptable to the NRC and be in accordance with all applicable NRC regulations.
(ii)  The surety company providing any surety obtained to comply with the Order approving the transfer shall be one of those listed by the U.S. Department of the Treasury in the most recent edition of Circular 570 and shall have a coverage limit sufficient to cover the amount of the surety.
(iii)  Entergy Nuclear Vermont Yankee, LLC shall establish a standby trust to receive funds from the surety, if a surety is obtained, in the event that Entergy Nuclear Vermont Yankee, LLC defaults on its funding obligations for the decommissioning of Vermont Yankee. The standby trust agreement must be in a form acceptable to the NRC, and shall conform with all conditions otherwise applicable to the decommissioning trust agreement.
(iv) The surety agreement must provide that the agreement cannot be amended in any material respect, or terminated, without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.
Entergy Nuclear Vermont Yankee, LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for approval of the transfer of this license to Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., and the requirements of the Order approving the transfer, and consistent with the safety evaluation supporting the Order.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. shall take no action to cause Entergy Global Investments, Inc., or Entergy International Holdings Ltd. LLC, or their parent companies to void, cancel, or modify the lines of credit to provide funding for Vermont Yankee as represented in the application without prior written consent of the Director of the Office of Nuclear Reactor Regulation.
K. Minimum Critical Power Ratio When operating at thermal power greater than 1593 megawatts thermal, the safety limit minimum critical power ratio (SLMCPR) shall be established by adding 0.02 to the cycle-specific SLMCPR value calculated using the NRC-approved methodologies documented in General Electric Licensing Topical Report NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," as amended, and documented in the Core Operating Limits Report.
Amendment No. 208

L. Transient Testing
: 1. During the extended power uprate (EPU) power ascension test program and prior to exceeding 168 hours of plant operation at the nominal full EPU reactor power level, with feedwater and condensate flow rates stabilized at approximately the EPU full power level, Entergy Nuclear Operations, Inc. shall confirm through performance of transient testing that the loss of one condensate pump will not result in a complete loss of reactor feedwater.
: 2. Within 30 days at nominal full-power operation following successful performance of the test in (1) above, through performance of additional transient testing and/or analysis of the results of the testing conducted in (1) above, confirm that the loss of one reactor feedwater pump will not result in a reactor trip.
M. Potential Adverse Flow Effects This license condition provides for monitoring, evaluating, and taking prompt action in response to potential adverse flow effects as a result of power uprate operation on plant structures, systems, and components (including verifying the continued structural integrity of the steam dryer).
: 1. The following requirements are placed on operation of the facility above the original licensed thermal power (OLTP) level of 1593 megawatts thermal (MWt):
: a. Entergy Nuclear Operations, Inc. shall monitor hourly the 32 main steam line (MSL) strain gages during power ascension above 1593 MWt for increasing pressure fluctuations in the steam lines.
: b. Entergy Nuclear Operations, Inc. shall hold the facility for 24 hours at 105%,
110%, and 115% of OLTP to collect data from the 32 MSL strain gages required by Condition M.1.a, conduct plant inspections and walkdowns, and evaluate steam dryer performance based on these data; shall provide the evaluation to the NRC staff by facsimile or electronic transmission to the NRC project manager upon completion of the evaluation; and shall not increase power above each hold point until 96 hours after the NRC project manager confirms receipt of the transmission.
: c. If any frequency peak from the MSL strain gage data exceeds the limit curve established by Entergy Nuclear Operations, Inc. and submitted to the NRC staff prior to operation above OLTP, Entergy Nuclear Operations, Inc. shall return the facility to a power level at which the limit curve is not exceeded. Entergy Nuclear Operations, Inc. shall resolve the uncertainties in the steam dryer analysis, document the continued structural integrity of the steam dryer, and provide that documentation to the NRC staff by facsimile or electronic transmission to the NRC project manager prior to further increases in reactor power.
Amendment No.
: d. In addition to evaluating the MSL strain gage data, Entergy Nuclear Operations, Inc. shall monitor reactor pressure vessel water level instrumentation or MSL piping accelerometers on an hourly basis during power ascension above OLTP. If resonance frequencies are identified as increasing above nominal levels in proportion to strain gage instrumentation data, Entergy Nuclear Operations, Inc. shall stop power ascension, document the continued structural integrity of the steam dryer, and provide that documentation to the NRC staff by facsimile or electronic transmission to the NRC project manager prior to further increases in reactor power.
: e. Following start-up testing, Entergy Nuclear Operations, Inc. shall resolve the uncertainties in the steam dryer analysis and provide that resolution to the NRC staff by facsimile or electronic transmission to the NRC project manager. If the uncertainties are not resolved within 90 days of issuance of the license amendment authorizing operation at 1912 MWt, Entergy Nuclear Operations, Inc. shall return the facility to OLTP.
: 2. As described in Entergy Nuclear Operations, Inc. letter BVY 05-084 dated September 14, 2005, Entergy Nuclear Operations, lnc. shall implement the following actions:
: a. Prior to operation above OLTP, Entergy Nuclear Operations, Inc. shall install 32 additional strain gages on the main steam piping and shall enhance the data acquisition system in order to reduce the measurement uncertainty associated with the acoustic circuit model (ACM).
: b. In the event that acoustic signals are identified that challenge the limit curve during power ascension above OLTP, Entergy Nuclear Operations, Inc. shall evaluate dryer loads and re-establish the limit curve based on the new strain gage data, and shall perform a frequency-specific assessment of ACM uncertainty at the acoustic signal frequency.
: c. After reaching 120% of OLTP, Entergy Nuclear Operations, Inc. shall obtain measurements from the MSL strain gages and establish the steam dryer flow-induced vibration load fatigue margin for the facility, update the dryer stress report, and re-establish the steam dryer monitoring plan (SDMP) limit curve with the updated ACM load definition and revised instrument uncertainty, which will be provided to the NRC staff.
: d. During power ascension above OLTP, if an engineering evaluation is required in accordance with the SDMP, Entergy Nuclear Operations, Inc. shall perform the structural analysis to address frequency uncertainties up to +/-10% and assure that peak responses that fall within this uncertainty band are addressed.
: e. Entergy Nuclear Operations, Inc. shall revise the SDMP to reflect long-term monitoring of plant parameters potentially indicative of steam dryer failure; to reflect consistency of the facility's steam dryer inspection program with General Electric Services Information Letter 644, Revision 1; and to identify the NRC Project Manager for the facility as the point of contact for providing SDMP information during power ascension.
: f. Entergy Nuclear Operations, Inc. shall submit the final extended power uprate (EPU) steam dryer load definition for the facility to the NRC upon completion of the power ascension test program.
Amendment No.
: g. Entergy Nuclear Operations, Inc. shall submit the flow-induced vibration related portions of the EPU startup test procedure to the NRC, including methodology for updating the limit curve, prior to initial power ascension above OLTP.
: 3. Entergy Nuclear Operations, Inc. shall prepare the EPU startup test procedure to include the (a)stress limit curve to be applied for evaluating steam dryer performance; (b)specific hold points and their duration during EPU power ascension; (c) activities to be accomplished during hold points; (d) plant parameters to be monitored; (e) inspections and walkdowns to be conducted for steam, feedwater, and condensate systems and components during the hold points; (f) methods to be used to trend plant parameters; (g)acceptance criteria for monitoring and trending plant parameters, and conducting the walkdowns and inspections; (h) actions to be taken if acceptance criteria are not satisfied; and (i) verification of the completion of commitments and planned actions specified in its application and all supplements to the application in support of the EPU license amendment request pertaining to the steam dryer prior to power increase above OLTP. Entergy Nuclear Operations, Inc.
shall provide the related EPU startup test procedure sections to the NRC by facsimile or electronic transmission to the NRC project manager prior to increasing power above OLTP.
: 4. When operating above OLTP, the operating limits, required actions, and surveillances specified in the SDMP shall be met. The following key attributes of the SDMP shall not be made less restrictive without prior NRC approval:
: a. During initial power ascension testing above OLTP, each test plateau increment shall be approximately 80 MWt;
: b. Level 1 performance criteria; and
: c. The methodology for establishing the stress spectra used for the Level 1 and Level 2 performance criteria.
Changes to other aspects of the SDMP may be made in accordance with the guidance of NEI 99-04.
: 5. During each of the three scheduled refueling outages (beginning with the spring 2007 refueling outage), a visual inspection shall be conducted of all accessible, susceptible locations of the steam dryer, including flaws left 'as is' and modifications.
: 6. The results of the visual inspections of the steam dryer conducted during the three scheduled refueling outages (beginning with the spring 2007 refueling outage) shall be reported to the NRC staff within 60 days following startup from the respective refueling outage. The results of the SDMP shall be submitted to the NRC staff in a report within 60 days following the completion of all EPU power ascension testing.
Amendment No.
: 7. The requirements of paragraph 4 above for meeting the SDMP shall be implemented upon issuance of the EPU license amendment and shall continue until the completion of one full operating cycle at EPU. If an unacceptable structural flaw (due to fatigue) is detected during the subsequent visual inspection of the steam dryer, the requirements of paragraph 4 shall extend another full operating cycle until the visual inspection standard of no new flaws/fiaw growth based on visual inspection is satisfied.
: 8. This license condition shall expire upon satisfaction of the requirements in paragraphs 5, 6, and 7 provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw or unacceptable flaw growth that is due to fatigue.
: 4. This license is effective as of the date of issuance and shall expire at midnight on March 21, 2012.LI!i?2 FOR THE ATOMIC ENERGY COMMISSION Original Signed By Roger S. Boyd IfI A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing

==Enclosures:==

Appendix A Technical Specifications Date of Issuance:
Feb. 28, 1973 Amendment No. 208

i' VYNPS 1.0  DEFINITIONS or more plant parameters in order to initiate trip system action.
Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
: 3. Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level.
: 4. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.
P. Rated Neutron Flux - Rated neutron flux is the neutron flux that corresponds to a steady state power level of 1912 thermal megawatts.
Q. Rated Thermal Power - Rated thermal power means a steady state power level of 1912 thermal megawatts.
R. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup/Hot Standby" or "Run" position with the reactor critical and above 1% rated thermal power.
: 1. Startup/Hot Standby Mode - In this mode the low turbine condenser vacuum trip is bypassed when condenser vacuum is less than 12 inches Hg and both turbine stop valves and bypass valves are closed; the low pressure and the 10 percent closure main steamline isolation valve closure trips are bypassed; the reactor protection system is energized with IRM neutron monitoring system trips and control rod withdrawal interlocks in service and APRM neutron monitoring system operable.
: 2. Run Mode - In this mode the reactor system pressure is equal to or greater than 800 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service.
S. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detector.
T. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant subsequent to that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.
U. Deleted Amendment No. 74, 84, 4-65, 4-64, 4-93                                        3

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 >800 psia and Core Flow
          >10% of Rated)                                  The limiting safety system trip settings shall be as specified When the reactor pressure is                     below:
         >800 psia and the core flow is greater than 10% of rated:                       1. Neutron Flux Trip Settings
: 1. A Minimum Critical Power Ratio                   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 is Safety Limit (FCISL).                                  in the RUN position, the APRM flux scram Allowable Value shall be:
Two loop operation:
Ss 0.33W+ 53.7% for      0% < W 5 30.9%
SS 1.07W+ 30.8% for 30.9% < W
* 66.7%
S* 0.55W+ 65.5% for 66.7% < W   S 99.0%
With a maximum of 120.0% power  for W >
99.0%
Single loop operation:
SS 0.33W+ 51.1% for      0% < W 5 39.1%
S* 1.07W+ 22.2% for 39.1% < W 5 61.7%
S* 0.55W+ 54.3% for 61.7% < W 5 119.4%
With a maximum of 120.0% power for W >
119.4%
where:
S   =     setting in percent of rated thermal power (1912 MWt)         .1 Amendment No. 4&, 4%, 64-,  4, .94, 4.GI, -&5,4-9, 4-4,        2-1, 241                         6

VYNPS 1.1  SAFETY LIMIT                                2.1  LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit                         W   =   percent rated two loop (Reactor Pressure <800 psia                             drive flow where 100%
or Core Flow *10% of Rated)                              rated drive flow is that flow equivalent When the reactor pressure is                             to 48 x 106 lbs/hr core
            *800 psia or core flow *10%                              flow of rated, the core thermal I           power shall not exceed 23% of rated thermal power.                             In the event of operation at
                                                             > 23% Rated Thermal Power the    l C. Power Transient                                  APRM gain shall be equal to or greater than 1.0.
To ensure that the safety limit established in Specification 1.lA and l.lB 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. A4,  *3;, 4., 4, 464, 4s&, -94,     416, 1     -41                             7

VYNPS 1.1 SAFETY LIMIT               2.1 LIMITING SAFETY SYSTEM SETTING D. Reactor low-low water level Emergency Core Cooling System (ECCS) initiation shall be at least 82.5 inches above the top of the enriched fuel.
E. Turbine stop valve scram shall, when operating at greater than 25% of Rated Thermal Power, be less than or equal to 10% valve closure from full open.
F. Turbine control valve fast closure scram shall, when operating at greater than 25%
of Rated Thermal Power, trip upon actuation of the turbine control valve fast closure relay.
G. Main steam line isolation valve closure scram shall be less than or equal to 10% valve closure from full open.
H. Main steam line low pressure initiation of main steam line isolation valve closure shall be at least 800 psig.
Amendment No. 48, 84, i-7-1                                           10

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 < 800 psia or Core Flow
          <10% 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 800 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 that the core average mass flux is 0.3-0.4 Mlb/hr-ft2 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 800 psia is conservative.
Amendment No. 4E, 4,
                    .- 44, -94, 52                                              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 l.l.lA or l.l.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. 1&, -E@, 150-13                                                      13

VYNPS BASES:
2.1  FUEL CLADDING INTEGRITY A. Trip Settings The bases for individual trip settings are discussed in the following paragraphs.
: 1. Neutron Flux Trip Settings
: a. APRM Flux Scram Allowable Value (Run Mode)
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1912 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.
Analyses are performed to demonstrate that the APRM flux scram over the range of settings from a maximum of 120% to the minimum flow biased setting provide protection from the fuel safety limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.
An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. The relationship between recirculation drive flow and reactor core flow is non-linear at low core flows. Due to stability concerns, separate APRM flow biased scram trip setting equations are provided for low core flows.
The APRM flow biased flux scram Allowable Value is the limiting value that the trip setpoint may have when tested periodically, beyond which appropriate action shall be taken. For Vermont Yankee, the periodic testing is defined as the calibration. The actual scram trip is conservatively set in relation to the Allowable Value to ensure operability between periodic testing. For single recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation. The single loop Amendment No. 4A, .2-3-, 9, 4-7-,
64, 94, 44-6, 4-44, 2                         14

VYNPS BASES:   2.1  (Cont'd)
Flux Scram Trip Setting '(Refuel or Startup and Hot Standby Mode)
For operation in the startup mode while the reactor is at low pressure, the reduced APRM scram setting to 15% of rated power provides adequate thermal margin between the setpoint and the safety limit, 23% of the rated.    (During an outage when it is necessary to check refuel interlocks, the mode switch must be moved to the startup position. Since the APRM reduced scram may be inoperable at that time due to the disconnection of the LPRMs, it is required that the IRM scram and the SRM scram in noncoincidence be in effect.
This will ensure that adequate thermal margin is maintained between the setpoint and the safety limit.)    The margin is adequate to accommodate anticipated maneuvers associated with station startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.
Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The reduced APRM scram remains active until the mode switch is placed in the RUN position. This switch can occur when reactor pressure is greater than 800 psia.
The IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument, which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120/125 of full scale is active in each range of the IRM.
For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on range 5, the scram would be 120/125 of full scale on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded.
Amendment No. 4&, 'L&, 884, 94                                                  15

VYNPS BASES:    2.1  (Cont'd)
E. Turbine Stop Valve Closure Scram-Trip Setting The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of <10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed. This scram signal may be bypassed at < 25% of reactor Rated Thermal Power.
F. Turbine Control Valve Fast Closure Scram The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection coincident with failure of the bypass system. This transient is less severe than the turbine stop valve closure with failure of the bypass valves and therefore adequate margin exists. This scram signal may be bypassed at < 25% of reactor Rated Thermal Power.
G. Main Steam Line Isolation Valve Closure Scram The isolation scram anticipates the pressure and flux transients which occur during an isolation event and the loss of inventory during a pipe break.
This action minimizes the effect of this event on the fuel and pressure vessel.
H. Reactor Coolant Low Pressure Initiation of Main Steam Isolation Valve Closure The low pressure isolation of the main steam lines at 800 psig is provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide the reactor shutdown so that high power operation at low reactor pressure does not occur. Operation of the reactor at pressures lower than 800 psig requires that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram.
Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit.
Amendment No. A4,  2,  .84, a-64, 1-7-3, 4-87-,BVY- 00 1                            17

VYNPS TABLE 3.1.1 REACTOR PROTECTION SYSTEM      (SCRAM)       INSTRUMENT REQUIREMENTS Required ACTIONS When Modes in Which         Minimum Number     Minimum Functions Must be Operatiinig        Operating    Conditions Instrument   For Operation Channels Per      Are Not Trip Function                       Trip Settings                    Refuel       (1)     Startup(12) Run Trip System (2) Satisfied (3)
: 1. Mode Switch in Shutdown (5A-S1)                                                           X                   X      X           1            A
: 2. Manual Scram                                                               x                   X      X           1            A (5A-S3A/B)
: 3. IRM   (7-41(A-F))
High Flux                             <120/125                             x                   X                  2             A INOP                                                                 x                   X                  2             A
: 4. APRM    (APRM A-F)
High Flux            Two loop operation:      (4)                                                     X           2         A or B (flow bias)          S5 0.33W+ 53.7% for       0% < W  S 30.9%
S5 1.07W+ 30.8% for 30.9% < W     S 66.7%
SS 0.55W+ 65.5% for 66.7% < W     S 99.0%
With a maximum of 120.0% power     for W >
99.0%
Single loop operation:      (4)
S5 0.33W+ 51.1% for       0% < W  S 39.1%
SS 1.07W- 22.2% for 39.1% < W      S 61.7%
SS 0.55W+ 54.3% for 61.7% < W      S 119.4%
With a maximum of 120.0% power     for W >
119.4%
High Flux                                                                  X                  X                  2             A
                                              <15%
(reduced)
INOP                                                                                          X      X         2(5)         A or B
: 5. High Reactor                        <1055 psig                             X                  X      X           2             A Pressure (PT-2-3-55(A-D)
(M)
Amendment No. 2,      44, 64,    68, :-6, ;-4, :9, x98,     84, .64,   4-8-,    4843,     229                                          21

VYNPS TABLE 3.1.1 NOTES (Cont'd)
: 3. When the requirements in the column "Minimum Number of Operating Instrument Channels Per Trip System" cannot be met for one system, that system shall be tripped. If the requirements cannot be met for both trip systems, the appropriate ACTIONS listed below shall be taken:
a)    Initiate insertion of operable rods and complete insertion of all operable rods within four hours.
b)    Reduce power level to IRM range and place mode switch in the "Startup/Hot Standby" position within eight hours.
c)   Reduce turbine load and close main steam line isolation valves within 8 hours.
d)   Reduce reactor power to less than 25% of rated within 8 hours.
: 4. The specified APRM High Flux scram (flow bias) Trip Setting is an Allowable Value, which is the limiting value that the trip setpoint may have when tested periodically. The actual scram trip setting is conservatively set in relation to the Allowable Value. "W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow.
: 5. To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additional LPRM input and still be considered operable.
: 6. The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.
: 7. Deleted.
: 8. Deleted.
: 9. Channel signals for the turbine control valve fast closure trip shall be derived from the same event or events which cause the control valve fast closure.
: 10. Turbine stop valve closure and turbine control valve fast closure scram signals may be bypassed at < 25% of reactor Rated Thermal Power.
: 11. Not used.
: 12. While performing refuel interlock checks which require the mode switch to be in Startup, the reduced APRM high flux scram need not be operable provided:
: a. The following trip functions are operable:
: 1. Mode switch in shutdown,
: 2. Manual scram,
: 3. High flux IRM scram
: 4. High flux SRM scram in noncoincidence,
: 5. Scram discharge volume high water level, and;
: b. No more than two (2) control rods withdrawn. The two (2) control rods that can be withdrawn cannot be face adjacent or diagonally adjacent.
Amendment Po. a4, 2--l,  -2, 464, 64, -74, 4G, 44, 464, Gil, 3  6, 446, -14, 24-9   24

VYNPS BASES:    3.1   (Cont'd)
Instrumentation is provided to detect a loss-of-coolant accident and initiate the core standby cooling equipment. This instrumentation is a backup to the water level instrumentation which is discussed in Specification 3.2.
The Control Rod Drive Scram System is designed so that all of the water that is discharged from the reactor by the scram can be accommodated in the discharge piping. This discharge piping is divided into two sections. One.
section. services the control rod drives on the north side of the reactor, the other serves the control rod drives of the south side. A part of the piping in each section is an instrument volume which accommodates in excess of 21 gallons of water and is at the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation, the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated, which would result in slow scram times or partial or no control rod insertion. To preclude this occurrence, level instrumentation has been provided for the instrument volume which scram the reactor when the volume of water reaches 21 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water, and precludes the situation in which a scram would be required but not be able to perform its function adequately.
The present design of the Scram Discharge System is in concert with the BWR Owner's Group criteria, which have previously been endorsed by the NRC in their generic "Safety Evaluation Report (SER) for Scram Discharge Systems",
dated December 1, 1980.
Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient without bypass.
Turbine stop valve (TSV) closure and turbine control valve (TCV) fast closure scram signals may be bypassed at < 25% of reactor Rated Thermal Power since, at low thermal power levels, the margins to fuel thermal-hydraulic limits and reactor primary coolant boundary pressure limits are large and an immediate scram is not necessary. This bypass function is normally accomplished automatically by pressure switches sensing turbine first stage pressure. The turbine first stage pressure setpoint controlling the bypass of the scram signals on TCV fast closure and TSV closure is derived from analysis of reactor pressurization transients.
Certain operational factors, such as turbine bypass valves open, can influence the relationship between turbine first stage pressure and reactor Rated Thermal Power. However, above 25% of reactor Rated Thermal Power, these scram functions must be enabled.
Amendment No. 21,    , 4-E, 30                                           30

VYNPS 3.3  LIMITING CONDITIONS FOR                 4.3 SURVEILLANCE REQUIREMENTS OPERATION positive coupling and the results of each test shall be recorded. The drive and blade shall be coupled and fully withdrawn. The position and over-travel lights shall be observed.
: 2. The Control Rod Drive                   2. The Control Rod Drive Housing Support System                      Housing Support System shall be in place when                      shall be inspected after the Reactor Coolant                         reassembly and the System is pressurized                       results of the inspection above atmospheric                           recorded.
pressure with fuel in the reactor vessel unless all operable control rods are fully inserted.
: 3. While the reactor is                    3. Prior to control rod I               below 17% power, the Rod                    withdrawal for startup Worth Minimizer (RWM)                       the Rod Worth Minimizer shall be operating while                     (RWM) shall be verified moving control rods                         as operable by performing except that:                                the following:
(a)    If after withdrawal                  (a)  Verify that the of at least 12                             control rod control rods during                        withdrawal sequence a startup, the RWM                         for the Rod Worth fails, the startup                         Minimizer computer may continue                               is correct.
provided a second licensed operator verifies that the operator at the reactor console is following the control rod program; or (b) If all rods, except                     (b)   The Rod Worth those that cannot be                       Minimizer diagnostic moved with control                         test shall be rod drive                                  performed.
Amendment No.   ,-3,  :4-, 4948                                                    83

VYNPS BASES:  3.3 & 4.3    (Cont'd)
: 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.
: 3. In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod. Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 17% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences.
: 4. Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).
: 5. The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration. It does provide the operator with a visual indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.
: 6. Deleted.
Amendment 2x, g3,    w A-G, 4a4, H.   ,99--_,244I,24-9                             90

VYNPS 3.4  LIMITING CONDITIONS FOR                4.4 SURVEILLANCE REQUIREMENTS OPERATION 3.4  REACTOR STANDBY LIQUID CONTROL         4.4 REACTOR STANDBY LIQUID CONTROL SYSTEM                                     SYSTEM Applicability:                             Applicability:
Applies to the operating status             Applies to the periodic testing of the Reactor Standby Liquid              requirement for the Reactor Control System.                            Standby Liquid Control System.
Objective:                                  Objective:
To assure the availability of an            To verify the operability of the independent reactivity control              Standby Liquid Control System.
mechanism.
Specification:                              Specification:
A. Normal Operation                       A. Normal Operation Except as specified in 3.4.B               The Standby Liquid Control below, the Standby Liquid                  System shall be verified Control System shall be                    operable by:
operable when the reactor mode switch is in either the               1. Testing pumps and valves "Startup/Hot Standby" or                        in accordance with "Run" position, except to                      Specification 4.6.E.
allow testing of                               A minimum flow rate of instrumentation associated                     35 gpm at 2 1325 psig     I with the reactor mode switch                   shall be verified for interlock functions                            each pump.
provided:
: 2. Verifying the continuity
: 1. Reactor coolant                             of the explosive charges temperature is less than                    at least monthly.
or equal to 2120 F; In addition, at least once
: 2. All control rods remain                 during each operating cycle, fully inserted in core                  the Standby Liquid Control cells containing one or                 System shall be verified more fuel assemblies;                   operable by:
and
: 3. Deleted
: 3. No core alterations are in progress.
: 4. Initiating one of the standby liquid control loops, excluding the primer chamber and inlet fitting, and verifying that a flow path from a pump to the reactor vessel is available.
Both loops shall be tested over the course of two operating cycles.
Amendment No. 440,  4-2-&, 4-64, 17,
                                   -    2-24                                    92

VYNPS 3.4  LIMITING CONDITIONS FOR                              4.4 SURVEILLANCE REQUIREMENTS OPERATION
: 2. The solution                                           2. Sodium pentaborate temperature, including                                 concentration shall be that in the pump suction                               determined at least once piping, shall be                                       a month and within maintained above the                                    24 hours following the curve shown in                                         addition of water or Figure 3.4.2.                                          boron, or if the solution temperature
: 3. The combination of                                        drops below the limits Standby Liquid Control                                  specified by System pump flow rate,                                  Figure 3.4.2.
boron concentration, and boron enrichment shall satisfy the following relationship for the                                 3. The boron-10 enrichment Standby Liquid Control                                  of the borated solution System to be considered                                 required by Specification operable:                                               3.4.C.3 shall be tested and verified once per Q        M251         C          E                        operating cycle.
I             -      x --       x   -    x   -    2.1.29 86         M          13      19.8 where:
C =     the concentration of sodium pentaborate solution (weight percent) in the Standby Liquid Control System tank E =the          boron-10 enrichment (atom percent) of the sodium pentaborate solution I            Q 2    35 gpm M251
                       = a constant (the M          ratio of mass of water in the reference plant compared to VY)
D. If Specification 3.4.A or B is not met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours.
E. If Specification 3.4.C is not met, action shall be immediately initiated to correct the deficiency.                  If at the end of 12 hours the system has not been restored to full operability, then a shutdown shall be initiated with the reactor in cold shutdown within 24 hours of initial discovery.
Amendment No.     :-&,  *-7&,   4-G-S,  4-Add9                                                  94

VYNPS BASES:
3.4 & 4.4    REACTOR STANDBY LIQUID CONTROL SYSTEM A. Normal Operation The design objective of the Reactor Standby Liquid Control System (SLCS) is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the Standby Liquid Control System is designed to inject a quantity of boron which produces a concentration of 800 ppm of natural boron in the reactor core in less than 138 minutes. An 800 ppm natural boron concentration in the reactor core is required to bring the reactor from full power to a 5% Ak subcritical condition. An additional margin (25%
of boron) is added for possible imperfect mixing of the chemical solution in the reactor water. A minimum quantity of 3850 gallons of solution having a 10.1% natural sodium pentaborate concentration is required to meet this shutdown requirement.
The time requirement (138 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak. For a required minimum pumping rate of 35 gallons per minute, the maximum net storage volume of the boron solution is established as 4830 gallons.
In addition to its original design basis, the Standby Liquid Control System also satisfies the requirements of 10CFR50.62(c)(4) on anticipated transients without scram (ATWS) by using enriched boron.
The ATWS rule adds hot shutdown and neutron absorber (i.e., boron-10) injection rate requirements that exceed the original Standby Liquid Control System design basis. However, changes to the Standby Liquid Control System as a result of the ATWS rule have not invalidated the original design basis.
With the reactor mode switch in the "Run" or "Startup/Hot Standby" position, shutdown capability is required. With the mode switch in "Shutdown," control rods are not able to be withdrawn since a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. With the mode switch in "Refuel,"
only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate shutdown margin by Specification 3.3.A ensures that the reactor will not become critical.
Therefore, the Standby Liquid Control System is not required to be operable when only a single control rod can be withdrawn.
Pump operability testing (by recirculating demineralized water to the test tank)in accordance with Specification 4.6.E is adequate to detect if failures have occurred. Flow, circuitry, and trigger assembly testing at the prescribed intervals assures a high reliability of system operation capability. The maximum SLCS pump discharge pressure during the limiting ATWS event is 1325 psig. This value is based on a reactor vessel lower plenum pressure of 1292 psia that occurs during the limiting ATWS event at the time of SLCS initiation, i.e., 120 seconds into the event. There is adequate margin to prevent the SLCS relief valve from lifting. Recirculation of the borated solution is done during each operating cycle to ensure one suction line from the boron tank is clear. In addition, at least once during each operating cycle, one of the standby liquid control loops will be initiated to verify that a flow path from a pump to the reactor vessel is available by pumping demineralized water into the reactor vessel.
Amendment No. _42., -4,  .-,&,4-7-&t BVa 0_ ;8, g24, A4c97                    97

C.x I                           VYNPS BASES:  3.4 & 4.4 (Cont'd)
B. Operation With Inoperable Components Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. Assurance that the system will perform its intended function is obtained from the results of the pump and valve testing performed in accordance with the Requirements of Specification 4.6.E.
C. Standby Liquid Control System Tank - Borated Solution The solution saturation temperature varies with the concentration of sodium pentaborate. The solution shall be kept at least 100 F above the saturation temperature to guard against boron precipitation. The 100 F margin is included in Figure 3.4.2. Temperature and liquid level alarms for the system are annunciated in the Control Room.
Once the solution has been made up, boron concentration will not vary unless more boron or water is added. Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change. Considering these factors, the test interval has been established.
Sodium pentaborate concentration is determined within 24 hours following the addition of water or boron, or if the solution temperature drops below specified limits. The 24-hour limit allows for 8 hours of mixing, subsequent testing, and notification of shift personnel.
Boron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Isotopic tests of the sodium pentaborate are performed periodically to ensure that the proper boron-10 atom percentage is being used.
10CFR50.62(c) (4)requires a Standby Liquid Control System with a minimum flow capacity and boron content equivalent to 86 gpm of 13 weight percent natural sodium pentaborate solution in the 251-inch reactor pressure vessel reference plant. Natural sodium pentaborate solution is 19.8 atom percent boron-10. The relationship expressed in Specification 3.4.C.3 also contains the ratio M251/M to account for the difference in water volume between the reference plant and Vermont Yankee.   (This ratio of masses is 628,300 lbs./401,247 lbs.)
To comply with the ATWS rule and the plant-specific ATWS analysis, the combination of three Standby Liquid Control System parameters must be considered: boron concentration, Standby Liquid Control System pump flow rate, and boron-10 enrichment. If the product of the expression in Specification 3.4.C.3 is equal to or greater than 1.29, the Standby Liquid Control System satisfies the requirements of IOCFR50.62(c)(4) and the plant-specific ATWS analysis.
Amendment No. -142, 44&, 4-7&, Q4-, 219, 2-2-9                                 98

A VYNPS Figure 3.6.1 Reactor Vessl Presure-Temperature Umitations Hydrostatic Pressure and Leak Tests, Core Not Critical 40'FJhr HeatuplCooldown Lmit I1                                                                                      Valid Through 4.827E8 MWH(t) 1200-                                                                                    ,                                                         ,,
                    .                Bottom HeadCuv '                                   -T-                   -            ___                   - -n           a      a                              _ *e_         __a
__Use                nipnmofbotom                                               L_                     .r_.            _J _                 L !fr L
_ _ and                         ____---                 -   _Use minim"              of downwmer        _       _     _
                              --       tUud rperatue and                                                 ~                          ~                      -rgo                  erprli 1100                    bom head saxtace                                                                            -                                 rei-o-n and lemp-rre               -
                            -r: ternperatufe.                                            - a-r;                                        T           -            ftg egion Ot"                      -
                                                                                                                                                     - - - - surfacetemperaItue,except                _
_            _                   \L        _                         whenlangelerperatureis                JJ-I__L_
1000           .        .               :,.--,                                   . .                I             .                           _ greater Om 1 1I we*                           .            I.
                        .IJ_              -L.               J-L _                                  LI...L                                 _J        -              fthedowncomnerregion A               J...I           L._
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700     '     I       I        I      a                I      '                  I                 I                I          *g          &sect;             I   I         I    I     I          I      I       I
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                                                                                         -  -I-    -.
                                                                                                                               -4,r                                                  Bottom Head Upper Regions I        I         I                        .- T--                                        Temperature
                        -------              r-
                         - I
_- 4      _ I I
a g-   _       _ J     -  _;-          L-_
                                                                                        -~-I--
(.F)                  (psig)             (psig) o               I                                                              - T-                  I         I                                           --
0 600                      IX--r                     I--       I---r              _I_                            I      r-r-80                    0                     0 0                                                                                                                                                          80                   665                  253 C*
hi es                   712                  253 90                   764                  253 hi tc 4 00          a                                 a        a          a            a                 a                a        a 95                   821                  253 r-r-                           --L-
                                                                                                            -WJ      -      r....L...      I   -                   100                   885                  253 105                   954                  253 110                    1032                  253 110                    1032                  842 30020- ;;;                                     ;;                               ;                T
                                                                                                                 ;                   -   -r
                              -a----a--                -a---,-I---                     _-         ------                -        -r                              115                    1117                  885 120                  1211                  932
_ a             a                 a _                 a            a        a        a                a        a-                      125                  1316                  984 20 01     n, 7  .           a                 a      a           a            a         a        a               a           a 130                                   .1042
_   .1   -.. I                - t.235                                         J1105-*
_ s - I- -r-                                                                 -l     --f-          --     a---                                                                       1175
                                                       -   J-      -I-      -_L -_     _-I         -fL _ 1 -_                                                      140 I      I           I                                                                                  145                                        1253 I-,Ta ----a         r-             -'         .I       -                 --- r.         T,-
                                                                                                                           --     a-r--
a- I U -        -               -      -,
60                                                                100                              120                              140                    160                 180                       200 TEMPERATURE (7)
Amendment No. -4,                         6,             B,               64, 4,,                           20,                24-3                                                                                       135

(5 VYNPS FIGU                  3.6.2 Reactor Vessel Pressure-Temperature Limdtations Normnal Operatfon, Core Not Critical UmoerBgkims 10WF~tr HealupJCooldown Limit                                                                          Use minimrn of Valid Through 4.827E8 MWH(t)                                                                         downrr mon IMid I'                                                                                                                                                                                                    temperAtre and ftv .
regonout" wrsufae I_I r Is                           I I                               I            II                    4II. I. +2/.*
tentteratUre. except *hen w  eo             .       .         .                .     .        .                . _ .        .
                                                                                                                                                                  -1       -         _     .
fan ternferawre i a       I         a                                                I       I        I
_  _:_ _!_ _ t/                     LI _J_                 -
                                                                                                                                                                                             -          I- greater than 140'F. use
                           -r-IL I I-I
                                                     -        _ _I_ _I_
_ I_        ,    _I I   _
_ I _1 I     I J_I I
I8--r
                                                                                                                                             -_y-I a
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Amendment No.                    - ,                 3-,, Q                          2+,                                                                                                                                                         136 136

VyNPS FIGURE 3.6.3 Reactor Vessel Pressure-Temperature Limitations Normal Operation, Core Critical 100WF/hr HeatuplCooldown Limit If Pressure < 253 pstg, Water Level must be within Normal Range for Power Operation I.                                                                             Valid Through 4.827E8 MWH(t)
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Amendment No. -3,                   9-3-,           -L               4                                                                                                                                                                      137 137

5 VYNPS BASES:
3.6 and 4.6   REACTOR COOLANT SYSTEM A. Pressure and Temperature Limitations All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The Pressure/Temperature (P/T) curves included as Figures 3.6.1, 3.6.2, and 3.6.3 were developed using 10CFR50 Appendix G, 1995 ASME Code, Section XI, Appendix G (including the Summer 1996 Addenda), and ASME Code Case N-640. These three curves provide P/T limit requirements for Pressure Test, Core Not Critical, and Core Critical. The P/T curves are not derived from Design Basis Accident analysis. They are prescribed to avoid encountering pressure, temperature or temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the reactor pressure boundary, a condition that is unanalyzed.
During heating events, the thermal gradients in the reactor vessel wall produce thermal stresses that vary from compressive at the inner wall to tensile at the outer wall. During cooling events the thermal stresses vary from tensile at the inner wall to compressive at the outer wall. The thermally induced tensile stresses are additive to the pressure induced tensile stresses. In the flange region, bolt preload has a significant affect on stress in the flange and adjacent plates.
Therefore heating/cooling events and bolt preload are used in the determination of the pressure-temperature limitations for the vessel.
The guidance of Branch Technical Position - MTEB 5-2, material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTWT) for all pressure boundary components. For the plates and welds adjacent to the core, fast neutron (E > 1 Mev) irradiation will cause an increase in the RTNDT.
For these plates and welds an adjusted RTNDT (ARTNDT) of 890 F and 730 F (d and h thickness locations) was conservatively used in development of these curves for core region components. Based upon plate and weld chemistry, initial RTmn values, predicted peak fast neutron fluence (3.18 x 1017 n/cm2 at the reactor vessel inside surface) for a gross power generation of 4.827 x 108 MWH(t), these core region ART= values conservatively bound the guidance of Regulatory Guide 1.99, Revision 2.
There were five regions of the reactor pressure vessel (RPV) that were evaluated in the development of the P/T Limit curves: (1) the reactor vessel beltline region, (2) the bottom head region, (3) the feedwater nozzle, (4) the recirculation inlet nozzle, and (5) the upper vessel flange region. These regions will bound all other regions in the vessel with respect to considerations for brittle fracture.
Two lines are shown on each P/T limit figure. The dashed line is the Bottom Head Curve. This is applicable to the bottom head area only and includes the bottom head knuckle plates and dollar plates. Based on bottom head fluid temperature and bottom head surface temperature, the reactor pressure shall be maintained below the dashed line at all times.
Amendment No. 33, 62, .4, 4a,  -94, 4A2-C, 4-4-&, 2.4, 243                          138

VYNPS BASES:   3.6 and 4.6 (Cont'd)
C. Coolant Leakage The 5 gpm limit for unidentified leaks was established assuming such leakage was coming from the reactor coolant system. Tests have been conducted which demonstrate that a relationship exists between the size of a crack and the probability that the crack will propagate. These tests suggest that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly. Leakage less than the limit specified can be detected within a few hours utilizing the available leakage detection systems. If the limit is exceeded and the origin cannot be determined in a reasonably short time the plant should be shutdown to allow further investigation and corrective action.
The 2 gpm increase limit in any 24 hour period for unidentified leaks was established as an additional requirement to the 5 gpm limit by Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping."
The removal capacity from the drywell floor drain sump and the equipment drain sump is 50 gpm each. Removal of 50 gpm from either of these sumps can be accomplished with considerable margin.
D. Safety and Relief Valves Safety analyses have shown that only three of the four relief valves are required to ensure compliance with the MCPR safety limit for the analyzed transients.
The setpoint tolerance value for as-left or refurbished valves is specified in Section III of the ASME Boiler and Pressure Vessel Code as
          +/-1% of set pressure. However, the code allows a larger tolerance value for the as-found condition if the supporting design analyses demonstrate that the applicable acceptance criteria are met. For the purposes of this limiting condition, a relief valve that is unable to actuate within tolerance of its set pressure is considered to be as inoperable as a mechanically malfunctioning valve. Safety analysis has been performed which shows that with all safety and safety relief valves within +/-3% of the specified set pressures in Table 2.2.1 and with one inoperable safety relief valve, the reactor coolant pressure safety limit of 1375 psig and the MCPR safety limit are not exceeded during the limiting overpressure transient.
Ch~ange 16,41areh 28, 1:974,- 44,-44, 128, 449,444, 4-,  4-,4-9 64,                     142

VYNPS 3.11 LIMITING CONDITIONS FOR                    4.11    SURVEILLANCE REQUIREMENTS OPERATION 3.11 REACTOR FUEL ASSEMBLIES                    4.11    REACTOR FUEL ASSEMBLIES Applicability:                                     Applicability:
The Limiting Conditions for                        The Surveillance Requirements Operation associated with the                      apply to the parameters which fuel rods apply to these                           monitor the fuel rod operating parameters which monitor the                       conditions.
fuel rod operating conditions.
Objective:                                         Objective:
The Objective of the Limiting                      The Objective of the Conditions for Operation is to                     Surveillance Requirements is to assure the performance of the                      specify the type and frequency fuel rods.                                         of surveillance to be applied to the fuel rods.
Specifications:                                    Specifications:
A. Average Planar Linear Heat                     A. Average Planar Linear Heat Generation Rate (APLHGR)                            Generation Rate (APLHGR)
During operation at                                  The APLHGR for each type of I           2 23% Rated Thermal                                   fuel as a function of Power, the APLHGR for                                average planar exposure, each type of fuel as a                              power, and flow shall be function of average                                 determined once within 12 planar exposure, power,                              hours after 2 23% Rated and flow shall not                                  Thermal Power and daily       I exceed the limiting                                  during operation at 2 23%
values provided in the                               Rated Thermal Power Core Operating Limits                                thereafter.
Report. For single recirculation loop operation, the limiting values shall be the values provided in the Core Operating Limits Report listed under the heading "Single Loop Operation." If at any time during operation I            at 2 23% Rated Thermal Power it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, APLHGR(s) shall be returned to within prescribed limits within two (2) hours; otherwise, the reactor shall be I            brought to < 23% Rated Thermal Power within 4 hours. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
Amendment No.  -I-&, 4-, 64, G, -94, 100G, @4&-, 414,   48-,    24-9                  224

VYNPS 3.11 LIMITING CONDITIONS FOR                4.11 SURVEILLANCE REQUIREMENTS OPERATION B. Linear Heat Generation Rate            B. Linear Heat Generation Rate (LHGR)                                      (LHGR)
I            During operation at 2 23%                  The LHGR as a function of Rated Thermal Power, the                   core height shall be checked linear heat generation rate                once within 12 hours after (LHGR) of any rod in any                   2 23% Rated Thermal Power fuel assembly at any axial                 and daily during operation   I location shall not exceed                  at 2 23% Rated Thermal Power the maximum allowable LHGR                 thereafter.
provided in the Core Operating Limits Report.
If at any time during I            operation at 2 23% Rated Thermal Power it is determined by normal surveillance that the limiting value for LHGR is being exceeded, LHGR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor shall be brought I            to < 23% Rated Thermal Power within 4 hours.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
Amendment No. 4, A4g,.64, 4-&8, 4--, 42                                     225

I 11V
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VYNPS 3.11 LIMITING CONDITIONS FOR               1 4.11 SURVEILLANCE REQUIREMENTS OPERATION C. Minimum Critical Power Ratio            C. Minimum Critical Power Ratio (MCPR)                                       (MCPR)
: 1. During operation at                    MCPR shall be determined I                2 23% Rated Thermal                     once within 12 hours after Power the MCPR operating               2 23% Rated Thermal Power    I value shall be equal to                 and daily during operation or greater than the MCPR                at 2 23% Rated Thermal Power limits provided in the                 thereafter.
Core Operating Limits Report. For single recirculation loop operation, the MCPR Limits at rated flow are also provided in the Core Operating Limits Report. If at any time during operation at I                 2 23% Rated Thermal Power it is determined by normal surveillance that the limiting value for MCPR is being exceeded, MCPR(s) shall be returned to within the prescribed limits within two (2) hours; otherwise, the reactor power shall be brought I                 to < 23% Rated Thermal Power within 4 hours.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
Amendment No. 44, ., 43, x94, 16 4.8,   244                                226

VYNPS BASES:
4.11   FUEL RODS A. The APLHGR, LHGR and MCPR shall be checked daily when operating at l 23% Rated Thermal Power to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below 23% of rated thermal power, an unreasonably large peaking factor would be required, which is not the case for operating control rod sequences. The 12 hour allowance after thermal power l 23% Rated Thermal Power is achieved is acceptable given the large inherent margin to operating limits at low power levels.
B. At certain times during plant startups and power changes the plant technical staff may determine that surveillance of APLHGR, LHGR and/or MCPR is necessary more frequently than daily. Because the necessity for such an augmented surveillance program is a function of a number of interrelated parameters, a reasonable program can only be determined on a case-by-case basis by the plant technical staff. The check of APLHGR, LHGR and MCPR will normally be done using the plant process computer. In the event that the computer is unavailable, the check will consist of either a manual calculation or a comparison of existing core conditions to those existing at the time of a previous check to determine if a significant change has occurred.
If a reactor power distribution limit is exceeded, an assumption regarding an initial condition of the DBA analysis, transient analyses, or the fuel design analysis may not be met. Therefore, prompt action should be taken to restore the APLHGR, LHGR or MCPR to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour completion time is sufficient to restore the APLHGR, LHGR, or MCPR to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR, LHGR, or MCPR out of specification.
C. Minimum Critical Power Ratio (MCPR) -  Surveillance Requirement At core thermal power levels less than or equal to 23%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 23% thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 23% rated thermal power is sufficient since power distribution shifts are very slow during normal operation.
Amendment No. 4.,  i4&,  -422                                               228}}