ML063000321

Changes to Technical Specifications Bases

ML063000321
Person / Time
Site:	Millstone
Issue date:	10/26/2006
From:	Gerald Bichof Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-837
Download: ML063000321 (18)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginid 23060 K/eh Addres: www.dom.com October 26, 2006 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 1 1555 Rockville Pike Rockville, Maryland 20852-2738 Serial No.06-837 NSS&UDF RO Docket Nos.

50-336 50-423 License Nos.

DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNITS 2 AND 3 CHANGES TO TECHNICAL SPECIFICATIONS BASES In accordance with the requirements of Millstone Power Station Unit 2 (MPS2)

Technical Specification 6.23.d and Millstone Power Station Unit 3 (MPS3), Technical Specification 6.1 8.d, Dominion Nuclear Connecticut, Inc. (DNC) is providing the Nuclear Regulatory Commission Staff with changes to MPS2 and MPS3 Technical Specifications Bases Sections. MPS2 changes affect Technical Specifications Bases Sections 314.4, 314.5, 314.6 and 314.7.

MPS3 changes affect Technical Specifications Bases Section 314.1. These changes are provided for information only. The changes to the bases sections were made in accordance with the provisions of 10 CFR 50.59.

These changes have been reviewed and approved by the Site Operations Review Committee.

Attachments 1 and 2 provide the retyped pages of the Technical Specifications Bases for MPS2 and MPS3 respectively.

If you have any questions or require additional information, please contact Mr. Paul R.

Willoug hby at (804) 273-3572.

Very truly yours, Gerald T. Bischof u Vice President - Nuclear Engineering

Serial No.06-837 Docket Nos. 50-336150-33-423 Changes To Technical Specifications Bases Page 2 of 2 Attachments:

1. Re-typed Bases Pages for Millstone Power Station Unit 2

2. Re-typed Bases Pages for Millstone Power Station Unit 3 Commitments made in this letter: None.

cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 1 9406-1 41 5 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 1 1 555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station

Serial No.06-837 Docket No. 50-336 ATTACHMENT 1 CHANGES TO TECHNICAL SPECIFICATIONS BASES RETYPED PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.06-837 Docket No. 50-336 Millstone Unit 2 Bases Paaes Section No.

Page No.

314.4 Reactor Coolant System B 314 4-6b B 314 4-7 B 314 4-7a B 314 4-7b B 314 4-7c 314.5 Emergency Core Cooling Systems I

314.7 Plant Svstems I B 314 7-7 B 314 5-3 B 314 5-4 B 314 5-5 B 314 5-6 314.6 Combustible Gas Control B 314 6-4

LBDCR 05-MP2-003 December 27,2005 REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table 4.6-1 of the Final Safety Analysis Report. Reactor operation and resultant fast neutron irradiation will cause an increase in the RTNDT-Therefore, an adjusted reference temperature, based upon the fluence, can be predicted using the methods described in Revision 2 to Regulatory Guide 1.99.

The heatup and cooldown limit curves shown on Figures 3.4-2a and 3.4-2b include predicted adjustments for this shift in RTmT at the end of the applicable service period, as well as adjustments for possible uncertainties in the pressure and temperature sensing instruments. The adjustments include the pressure and temperature instrument and loop uncertainties associated with the main control board displays, the pressure drop across the core (RCP operation), and the elevation differences between the location of the pressure transmitters and the vessel beltline region. In addition to these curve adjustments, the LTOP evaluation includes adjustments due to valve stroke times, PORV circuitry reaction times, and valve discharge backpressure.

The actual shift in RTWT of the vessel material is established periodically during operation by removing and evaluating, in accordance with 10CFR5O Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are similar, the measured transition shift for a sample can be correlated to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ARTmT determined from the surveillance capsule exceeds the calculated ARTmT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figures 3.4-2a and 3.4-2b for reactor criticality have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality. For inservice leak and hydrostatic testing, use of the heatup curve on Figure 3.4-2a and associated rates provide a conservative limit in lieu of a curve developed specifically for inservice leak and hydrostatic testing. Therefore, a separate leak and hydrostatic curve is not explicitly included on Figure 3.4-2a.

The maximum RTmT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 50°F. The Lowest Service Temperature limit is based upon this RTmT since Article NB-2332 (Summer Addenda of 1972) of Section I11 of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100°F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3 125 psia. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement.

MILLSTONE - UNIT 2 B 314 4-6b Amendment No. W,

LBDCR 05-MP2-003 December 27,2005 REACTOR COOLANT SYSTEM BASES Included in this evaluation is consideration of flange protection in accordance with 10 CFR 50, Appendix G The requirement makes the minimum temperature RTmT plus 90°F for hydrostatic test and RTNDT plus 120°F for normal operation when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure. Since the flange region RTNDT has been calculated to be 30°F, the minimum flange pressurization temperature during normal operation is 150°F (163°F with instrument uncertainty) when the pressure exceeds 20% of the preservice hydrostatic pressure. Operation of the RCS within the limits of the heatup and cooldown curves will ensure compliance with this requirement.

To establish the minimum boltup temperature, ASME Code Section XI, Appendix G, requires the temperature of the flange and adjacent shell and head regions shall be above the limiting RTmT temperature for the most limiting material of these regions. The RTNDT temperature for that material is 30°F. Adding 13"F, for temperature measurement uncertainty, results in a minimum boltup temperature of 43°F. For additional conservatism, a minimum boltup temperature of 70°F is specified on the heatup and cooldown curves. The head and vessel flange region temperature must be greater than 70°F, whenever any reactor vessel stud is tensioned..

The Low Temperature Overpressure Protection (LTOP) System provides a physical barrier against exceeding the 10CFR5O Appendix G pressureltemperature limits during low temperature RCS operation either with a steam bubble in the pressurizer or during water solid conditions. This system consists of either two PORVs (each PORV is equivalent to a vent of approximately 1.4 square inches) with a pressure setpoint 5 41 5 psia, or an RCS vent of sufficient size. Analysis has confmed that the design basis mass addition transient discussed below will be mitigated by operation of the PORVs or by establishing an RCS vent of sufficient size.

The LTOP System is required to be OPERABLE when RCS cold leg temperature is at or below 275°F (Technical Specification 3.4.9.3). However, if the RCS is in MODE 6 and the reactor vessel head has been removed, a vent of sufficient size has been established such that RCS pressurization is not possible. Therefore, an LTOP System is not required (Technical Specification 3.4.9.3 is not applicable).

Adjusted Referenced Temperature (ART) is the RTNDT adjusted for radiation effects plus a margin term required by Revision 2 of Regulatory Guide 1.99. The LTOP System is armed at a temperature which exceeds the limiting 114t ART plus 50°F as required by ASME Section XI, Appendix G. For the operating period up to 54 EFPY, the limiting 114t ART is 175°F which results in a minimum LTOP System enable temperature of at least 271°F when corrected for instrument uncertainty. The current value of 275°F will be retained.

MLLSTONE - UNIT 2 B 314 4-7 Amendment No. 58, %,44, ?I%,%%,

=,

LBDCR 05-MP2-003 December 27,2005 REACTOR COOLANT SYSTEM BASES The mass input analysis performed to ensure the LTOP System is capable of protecting the reactor vessel assumes that all pumps capable of injecting into the RCS start, and then one PORV fails to actuate (single active failure). Since the PORVs have limited relief capability, certain administrative restrictions have been implemented to ensure that the mass input transient will not exceed the relief capacity of a PORV. The analysis has determined two PORVs (assuming one PORV fails) are suflicient if the mass addition transient is limited to the inadvertent start of one high pressure safety injection (HPSI) pump and two charging pumps when RCS temperature is at or below 275°F and above 1 90°F, and the inadvertent start of one charging pump when RCS temperature is at or below 190°F.

The assumed active failure of one PORV results in an equivalent RCS vent size of approximately 1.4 square inches when the one remaining PORV opens. Therefore, a passive vent of at least 1.4 square inches can be substituted for the PORVs. However, a vent size of at least 2.2 square inches will be required when VENTING the RCS. If the RCS is depressurized and vented through at least a 2.2 square inch vent, the peak RCS pressure, resulting from the maximum mass input transient allowed by Technical Specification 3.4.9.3, will not exceed 300 psig (SDC System suction side design pressure).

When the RCS is at or below 1 90°F, additional pumping capacity can be made capable of injecting into the RCS by establishing an RCS vent of at least 2.2 square inches. Removing a pressurizer PORV or the pressurizer manway will result in a passive vent of at least 2.2 square inches. Additional methods to establish the required RCS vent are acceptable, provided the proposed vent has been evaluated to ensure the flow characteristics are equivalent to one of these.

Establishing a pressurizer steam bubble of sufficient size will be sufficient to protect the reactor vessel from the energy addition transient associated with the start of an RCP, provided the restrictions contained in Technical Specification 3.4.1.3 are met. These restrictions limit the heat input £?om the secondary system. They also ensure sufftcient steam volume exists in the pressurizer to accommodate the insurge. No credit for PORV actuation was assumed in the LTOP analysis of the energy addition transient.

The restrictions apply only to the start of the first RCP. Once at least one RCP is running, equilibrium is achieved between the primary and secondary temperatures, eliminating any significant energy addition associated with the start of the second RCP.

The LTOP restrictions are based on RCS cold leg temperature. This temperature will be determined by using RCS cold leg temperature indication when RCPs are running, or natural circulation if it is occurring. Otherwise, SDC return temperature indication will be used.

MILLSTONE - UNIT 2 Amendment No. £44,

LBDCR 05-MP2-003 December 27,2005 REACTOR COOLANT SYSTEM BASES Restrictions on RCS makeup pumping capacity are included in Technical Specification 3.4.9.3. These restrictions are based on balancing the requirements for LTOP and shutdown risk.

For shutdown risk reduction, it is desirable to have maximum makeup capacity and to maintain the RCS full (not vented). However, for LTOP it is desirable to minimize makeup capacity and vent the RCS. To satisfl these competing requirements, makeup pumps can be made not capable of injecting, but available at short notice.

A charging pump can be considered to be not capable of injecting into the RCS by use of any of the following methods and the appropriate administrative controls.

1. Placing the motor circuit breaker in the open position.

2. Removing the charging pump motor overload heaters from the charging pump circuit.

3. Removing the charging pump motor controller from the motor control center.

A HPSI pump can be considered to be not capable of injecting into the RCS by use of any of the following methods and the appropriate administrative controls.

I. Racking down the motor circuit breaker from the power supply circuit.

2. Shutting and tagging the discharge valve with the key lock on the control panel (2-SI-654 or 2-SI-656).

3. Placing the pump control switch in the pull-to-lock position and removing the breaker control power fuses.

4. Placing the pump control switch in the pull-to-lock position and shutting the discharge valve with the key lock on the control panel (2-SI-654 or 2-SI-656).

These methods to prevent charging pumps and HPSI pumps from injecting into the RCS, when combined with the appropriate administrative controls, meet the requirement for two independent means to prevent pump injection as a result of a single failure or inadvertent single action.

These methods prevent inadvertent pump injections while allowing manual actions to rapidly restore the makeup capability if conditions require the use of additional charging or HPSI pumps for makeup in the event of a loss of RCS inventory or reduction in SHUTDOWN MARGIN.

MILLSTONE - UNIT 2 Amendment No. M, 227, ?43,

LBDCR 05-MP2-003 December 27,2005 REACTOR COOLANT SYSTEM BASES If a loss of RCS inventory or reduction in SHUTDOWN MARGIN event occurs, the appropriate response will be to correct the situation by starting RCS makeup pumps. If the loss of inventory or SHUTDOWN MARGIN is significant, this may necessitate the use of additional RCS makeup pumps that are being maintained not capable of injecting into the RCS in accordance with Technical Specification 3.4.9.3. The use of these additional pumps to restore RCS inventory or SHUTDOWN MARGIN will require entry into the associated ACTION statement. The ACTION statement requires immediate action to comply with the specification.

The restoration of RCS inventory or SHUTDOWN MARGIN can be considered to be part of the immediate action to restore the additional RCS makeup pumps to a not capable of injecting status.

While recovering RCS inventory or SHUTDOWN MARGIN, RCS pressure will be maintained below the Appendix G limits. After RCS inventory or SHUTDOWN MARGIN has been restored, the additional pumps should be immediately made not capable of injecting and the ACTION statement exited.

An exception to Technical Specification 3.0.4 is specified for Technical Specification 3.4.9.3 to allow a plant cooldown to MODE 5 if one or both PORVs are inoperable. MODE 5 conditions may be necessary to repair the PORV(s).

3/4.4.10 DELETED MlLLSTONE - UNIT 2 B 3/4 4-7c Amendment No. ?H, 230,243,264,

February 10,2005 LBDCR 05-MP2-001 EMERGENCY CORE COOLING SYSTEMS BASES 314.5.5 TRISODIUM PHOSPHATE (TSP)

BACKGROUND Trisodium phosphate (TSP) is placed on the floor or in the sump of the containment building to ensure that iodine, which may be dissolved in the recirculated reactor cooling water following a loss of coolant accident (LOCA), remains in solution. TSP also helps inhibit stress corrosion cracking (SCC) of austenitic stainless steel components in containment during the recirculation phase following an accident.

Fuel that is damaged during a LOCA will release iodine in several chemical forms to the reactor coolant and to the containment atmosphere. A portion of the iodine in the containment atmosphere is washed to the sump by containment sprays. The emergency core cooling water is borated for reactivity control. This borated water causes the sump solution to be acidic. In a low pH (acidic) solution, dissolved iodine will be converted to a volatile form. The volatile iodine will evolve out of solution into the containment atmosphere, significantly increasing the levels of airborne iodine. The increased levels of airborne iodine in containment contribute to the radiological releases and increase the consequences from the accident due to containment atmosphere leakage.

After a LOCA, the components of the core cooling and containment spray systems will be exposed to high temperature borated water. Prolonged exposure to the core cooling water combined with stresses imposed on the components can cause SCC. The SCC is a fbnction of stress, oxygen and chloride concentrations, pH, temperature, and alloy composition of the components. High temperatures and low pH, which would be present after a LOCA, tend to promote SCC. This can lead to the failure of necessary safety systems or components.

Adjusting the pH of the recirculation solution to levels above 7.0 prevents a significant fraction of the dissolved iodine from converting to a volatile form. The higher pH thus decreases the level of airborne iodine in containment and reduces the radiological consequences from containment atmosphere leakage following a LOCA. Maintaining the solution pH above 7.0 also reduces the occurrence of SCC of austenitic stainless steel components in containment. Reducing SCC reduces the probability of failure of components.

MILLSTONE - UNIT 2 Amendment No. W,

February 10,2005 LBDCR 05-MP2-001 EMERGENCY CORE COOLING SYSTEMS BASES 314.5.5 TRISODIUM PHOSPHATE (TSP)

BACKGROUND (continued)

TSP is employed as a passive form of pH control for post LOCA containment spray and core cooling water. Baskets of TSP are placed on the floor or in the sump of the containment building to dissolve from released reactor coolant water and containment sprays after a LOCA.

Recirculation of the water for core cooling and containment sprays then provides mixing to achieve a uniform solution pH. The hydrated form (45-57% moisture) of TSP is used because of the high humidity in the containment building during normal operation. Since the TSP is hydrated, it is less likely to absorb large amounts of water from the humid atmosphere and will undergo less physical and chemical change than the anhydrous form of TSP.

APPLICABLE SAFETY ANALYSES The LOCA radiological consequences analysis takes credit for iodine retention in the sump solution based on the recirculation water pH being 2 7.0. The radionuclide releases from the containment atmosphere and the consequences of a LOCA would be increased if the pH of the recirculation water were not adjusted to 7.0 or above.

LIMITING CONDITION FOR OPERATION The TSP is required to adjust the pH of the recirculation water to 2 7.0 after a LOCA. A pH 2 7.0 is necessary to prevent significant amounts of iodine released from he1 failures and dissolved in the recirculation water from converting to a volatile form and evolving into the containment atmosphere. Higher levels of airborne iodine in containment may increase the release of radionuclides and the consequences of the accident. A pH 2 7.0 is also necessary to prevent SCC of austenitic stainless steel components in containment. SCC increases the probability of failure of components.

The required amount of TSP is based upon the extreme cases of water volume and pH possible in the containment sump after a large break LOCA. The minimum required volume is the volume of TSP that will achieve a sump solution pH of 2 7.0 when taking into consideration the maximum possible sump water volume and the minimum possible pH. The amount of TSP needed in the containment building is based on the mass of TSP required to achieve the desired pH. However, a required volume is specified, rather than mass, since it is not feasible to weigh the entire amount of TSP in containment. The minimum required volume is based on the manufactured density of TSP. Since TSP can have a tendency to agglomerate from high humidity in the containment building, the density may increase and the volume decrease during normal plant operation. Due to possible agglomeration and increase in density, estimating the minimum volume of TSP in containment is conservative with respect to achieving a minimum required pH.

MILLSTONE - UNIT 2 B 314 5-4 Amendment No.

February 10,2005 LBDCR 05-MP2-001 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 TRISODIUM PHOSPHATE [TSP) (continued)

APPLICABILITY In MODES 1, 2, and 3, the RCS is at elevated temperature and pressure, providing an energy potential for a LOCA. The potential for a LOCA results in a need for the ability to control the pH of the recirculated coolant.

In MODES 4, 5, and 6, the potential for a LOCA is reduced or nonexistent, and TSP is not required.

ACTIONS If it is discovered that the TSP in the containment building sump is not within limits, action must be taken to restore the TSP to within limits. During plant operation the containment sump is not accessible and corrections may not be possible.

The completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for restoring the TSP within limits because 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is the same time allowed for restoration of other ECCS components.

If the TSP cannot be restored within limits within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time, the plant must be brought to a MODE in which the LC0 does not apply. The specified completion times for reaching MODES 3 and 4 were chosen to allow reaching the specified conditions from full power in an orderly manner without challenging plant systems.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.5.5.1 Periodic determination of the volume of TSP in containment must be performed due to the possibility of leaking valves and components in the containment building that could cause dissolution of the TSP during normal operation. A frequency of 18 months is required to determine visually that a minimum of 282 cubic feet is contained in the TSP baskets. This requirement ensures that there is an adequate volume of TSP to adjust the pH of the post LOCA sump solution to a value L 7.0.

The periodic verification is required every 18 months, since access to the TSP baskets is only feasible during outages, and normal he1 cycles are scheduled for 18 months. Operating experience has shown this surveillance frequency acceptable due to the margin in the volume of TSP placed in the containment building.

MILLSTONE - UNIT 2 B 3/4 5-5 Amendment No.

February 10,2005 LBDCR 05-MP2-001 EMERGENCY CORE COOLING SYSTEMS BASES 314.5.5 TRISODIUM PHOSPHATE {TSP) (continued)

Surveillance Requirement 4.5.5.2 Testing must be performed to ensure the solubility and buffering ability of the TSP after exposure to the containment environment. Passing this test verifies the TSP is active and provides assurance that the stored TSP will dissolve in borated water at postulated post-LOCA temperatures. This test is performed by submerging a sample of 0.6662 +_ 0.0266 grams of TSP from one of the baskets in containment in 250 + 10 milliliters of water at a boron concentration of 2482 +_ 20 ppm, and a temperature of 77 + 5°F. Without agitation, the solution is allowed to stand for four hours. The liquid is then decanted, mixed, and the pH measured. The pH must be 2 7.0.

The TSP sample weight is based on the minimum required TSP mass of 12,042 pounds, which at the manufactured density corresponds to the minimum volume of 223 ft3 (The minimum Technical Specification requirement of 282 ft3 is based on 223 ft3 of TSP for boric acid neutralization and 59 ft3 of TSP for neutralization of hydrochloric and nitric acids.), and the maximum sump water volume (at 77OF) following a LOCA of 2,046,441 liters, normalized to buffer a 250 + 10 milliliter sample. The boron concentration of the test water is representative of the maximum possible concentration in the sump following a LOCA. Agitation of the test solution is prohibited during TSP dissolution since an adequate standard for the agitation intensity cannot be specified. The dissolution time of four hours is necessary to allow time for the dissolved TSP to naturally diffuse through the sample solution. In the containment sump following a LOCA, rapid mixing will occur, significantly decreasing the actual amount of time before the required pH is achieved. The solution is decanted after the four hour period to remove any undissolved TSP prior to mixing and pH measurement. Mixing is necessary for proper operation of the pH instrument.

MILLSTONE - UNIT 2 Amendment No.

LBDCR 04-MP2-004 July 5,2004 CONTAINMENT SYSTEMS BASES 314.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipinent and systems required for control of hydrogen gas 1

ensures that this equipment will be available to maintain the hydrogen concentratioil within containment below its flammable limit during post-LOCA conditions.

I The post-incident recirculation systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

MILLSTONE - UNIT 2 Amendment No.*,

LBDCR 06-MP2-009 May 4,2006 PLANT SYSTEMS BASES 314.7.10 DELETED 314.7.1 1 ULTIMATE HEAT SINK The limitations on the ultimate heat sink temperature ensure that sufficient cooling capacity is available to either,

1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.

The limitations on maximum temperature are based on a 30-day cooling water supply to safety related equipment without exceeding their design basis temperature.

Various indications are available to monitor the temperature of the ultimate heat sink (UHS). The following guidelines apply to ensure the UHS Technical Specification limit is not exceeded.

The control room indications are normally used to ensure compliance with this specification.

Control room indications are acceptable because of the close correlation between control room indications and local Service Water System (SWS) header indications (historically within approximately 2°F). The highest reading valid temperature obtained fiom the Unit 2 intake structure and the inlets to the Circulating Water System water boxes shall be used to veri@ the UHS temperature is 1 70°F.

I When the highest reading valid control room indication indicates the temperature of the UHS is

> 70°F, local SWS header indications must be used. The highest reading valid local SWS header temperature shall be used to verify the UHS temperature limit of 75°F is not exceeded. Normally, local SWS header temperature will be taken at the inlet to the vital AC switchgear room cooling coils. If the local SWS header temperature cannot be taken at the inlet to the vital AC switchgear room cooling coils, the inlet to the Reactor Building Closed Cooling Water heater exchangers, or other acceptable instrumentation should be used to determine SWS header temperature.

If the UHS temperature exceeds 75"F, plant operations may continue provided the LC0 recorded water temperatures averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, are at or below 75°F. This verification is required to be performed once per hour when the water temperature exceeds 75°F.

If the UHS temperature, averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, exceeds the 75°F Technical Specification limit, or if the UHS temperature exceeds 77"F, a plant shutdown in accordance with I

the ACTION requirements will be necessary.

MILLSTONE - UNIT 2 B 314 7-7 Amendment No. ~

~

243,247,257,

Serial No.06-837 Docket No. 50-423 ATTACHMENT 2 CHANGES TO TECHNICAL SPECIFICATIONS BASES RETYPED PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.06-837 Docket No. 50-423 Millstone Unit 3 Bases Paaes I Section No.

I Page No.

314.1 Reactivity Control Systems B 314 1-1

LBDCR 06-MP3-013 April 5,2006 314.1 REACTIVITY CONTROL SYSTEMS BASES 314.1.1 BORATION CONTROL 314.1.1.1 and 314.1.I.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (I) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. In MODES 1 and 2, the most restrictive condition occurs at EOL with Tavg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of I

this accident, a minimum SHUTDOWN MARGIN as defined in Specification 314.1.1.I.I is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis I

assumptions. In MODES 3,4 and 5, the most restrictive condition occurs at BOL, associated with a boron dilution accident. In the analysis of this accident, a minimum SHUTDOWN MARGIN as defined in Specification 314.1.1.1.2 is required to allow the operator 15 minutes from the initiation of the Shutdown Margin Monitor alarm to total loss of SHUTDOWN MARGIN. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting requirement and is consistent with the accident analysis assumption.

The locking closed of the required valves in MODE 5 (with the loops not filled) will preclude the possibility of uncontrolled boron dilution of the Reactor Coolant System by preventing flow of unborated water to the RCS.

314.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

MILLSTONE - UNIT 3 B 314 1-1 Amendment No. W,&, 44,447,