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| number = ML11193A224
| number = ML11193A224
| issue date = 07/05/2011
| issue date = 07/05/2011
| title = Millstone Unit 3 License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3, Attachment 3 Through Attachment 6, Marked-up TS Pages and TS Bases Pages
| title = License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3, Attachment 3 Through Attachment 6, Marked-up TS Pages and TS Bases Pages
| author name =  
| author name =  
| author affiliation = Dominion Nuclear Connecticut, Inc
| author affiliation = Dominion Nuclear Connecticut, Inc
Line 14: Line 14:
| page count = 92
| page count = 92
| project =  
| project =  
| stage = Other
| stage = Request
}}
}}


=Text=
=Text=
{{#Wiki_filter:fbi- ul iy 28, 2600 PLANT SYSTEMS AUXILIARY FEED WATER SYSTEM LIMITING CONDITION FOR OPERATION'4/ACTION: (Continued)
{{#Wiki_filter:fbi- ul iy 28, 2600 PLANT SYSTEMS AUXILIARY FEED WATER SYSTEM
Required ACTION A'Inoperable Equipment Required ACTION.'r+e. Three auxiliary feedwater e.pumps in MODE 1, 2, or 3.--------NOTE -LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status. /Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status.I/SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE: a.b.At 1--st ofte r 31 days by:frequency specified in the Surveillance Frequency Control Progqram...--. .- ..- ..- N kU I t -. .... ...- -........ ..-Auxiliary fjedwater pumps may be considered OPERABLE during alignment and operation l/r steam generator level control, if they are capable of being manually realgnedo the auxiliary feedwater mode of operation.
                                                                                                                      '4/
raligne? .. .... ..... ...Verifyiot each auxiliary feedwater manual, power operated, and automatic valve in/each w ter flow path and in each required steam supply flow path to the steam turbin drriven auxiliary feedwater pump, that is not locked, sealed, or otherwise sec d in position, is in the correct positin. .....Spciiaton, 40, b P2 cays Specification 4.0.5, by:-n go I t4 , tested pursuant to 1) Verifying that on recirculation flow each motor-driven pump develops a total head of greater than or equal to 3385 feet;2) Verifying that on recirculation flow the steam turbine-driven pump develops a total head of greater than or equal to 3780 feet when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.MILLSTONE  
LIMITING CONDITION FOR OPERATION ACTION: (Continued)
-UNIT 3 3/4 7-5 Amendment No. 96, 400, 4-2-7, 4-39,-206,-435-Febmuary 28, 2007 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
A'
: c. At p by verifying that each auxiliary feedwater pump starts as igned automatically upon receipt of an Auxiliary Feedwater Actuation test signsna. For the steam turbine-driven auxiliary feedwater pump, the provisions of Specifi tion 4.0.4 are not applicable for entry into MODE 3.4.7.1.2.2 An auxiliary fe dwater flow path to each steam generator shall be demonstrated OPERABLE following each OLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to ach steam generator.
                                                                                                                      .'r Inoperable Equipment                       Required ACTION Required
Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE  
                                          +
-UNIT 3 3/4 7-5a Amendment No. 244 September 11, 1997 PLANT SYSTEMS DEMINERALIZED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The demineralized water storage tank (DWST) shall be OPERABLE with a water volume of at least 334,000 gallons.APPLICABILITY:
: e. Three auxiliary feedwater               e.                                                                         I pumps in MODE 1, 2, or 3.
MODES 1, 2, and 3.ACTION: With the DWST inoperable, within 4 hours either: a. Restore the DWST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, or b. Demonstrate the OPERABILITY of the condensate storage tank (CST) as a backup supply to the auxiliary feedwater pumps and restore the DWST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST shall be demonstrated OPERABLE at last pe 12 hu,, by verifying the water volume is within its limits when the tank is the sup y source for the auxiliary feedwater pumps.4.7.1.3.2 The CST shall be demonstrated OPERABLE least once per 12 hours by verifying that the combined volume of both the DWST and CST is at least 384,000 gallons of water whenever the CST and DWST are the supply source for he auxiliary feedwater pumps.Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE  
                                                - -   -   -----       NOTE         -
-UNIT 3 3/4 7-6 Amendment No. +50-Jftnutftry 31,986 PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microCurie/gram DOSE EQUIVALENT 1-131.APPLICABILITY:
LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status.                           /
MODES 1,2, 3, and 4.ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microCurie/gram DOSE EQUIVALENT 1-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by ........ o t, sampling a on o ifFa, 4. -.determining the Gross Radioactivity and DOSE EQUIVALENT 1-131 Concentration at the frequency specified in the Surveillance Frequency Control Program MILLSTONE  
                                                                                                                        /
-UNIT 3 3/4 7-7 Delete Table 4.7-1 JaT ..a .y 3 1 , 1 9 8 6 TABLE 4.7-1 N iLE U IN A K Y kL IU U L A IN 1 L)1I V1 ,l IJV I A L 11V 11 SAMPLE AND ANALYSTS PR ' MRAM TWY-P MEASUREMENT SAMPL D ANALYSIS AN DNALYSIS FRE NCY 1. Gross Radioacti At least once per 72 hour Determination
Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status.
: 2. Isotopic Analysis for DO a) Once per 31 days, when-EQUIVALENT 1-131 ever the gross radio-Concentration activity determination indicates concentrations ct ,,eater than 10% of the"a wavl ta b l e l im i t f o r" a d i o i n s .b) Once per 6 nths, when-~ever the gross r~a-.-.activity determ inatio .indicates concentrations less than or equal to 10%of the allowable limitj for radioiodines.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
MILLSTONE  
: a. At 1--st ofte       r 31 days     by: specified in the Surveillance Frequency Control Progqram
-UNIT 3 3/4 7-8 1O79 99 PLANT SYSTEMS STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES LIMITING CONDITION FOR OPERATION 3.7.1.6 Each steam generator atmospheric relief bypass valve (SGARBV) line shall be OPERABLE, with the associated main steam atmospheric relief isolation (block) valve in the open position.APPLICABILITY:
                            *7--1the      frequency
MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.ACTIONS a. With one required SGARBV line inoperable, restore required SGARBV line to OPERABLE status within 7 days or be in at least MODE 3 within the next 6 hours and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours. LCO 3.0.4 is not applicable.
                    - .             .- .     .- .   .-N kU I t -. . . .. .   .   - . . . .
: b. With two or more required SGARBV lines inoperable, restore all but one required SGARBV line to OPERABLE status within 24 hours or be in at least MODE 3 within the next 6 hours and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours.SURVEILLANCE REQUIREMENTS 4.7.1.6.1 Verify one complete cycle of each SGARBV , o fi,-tths.4.7.1.6.2 Verify one complete cycle of each mai eam atmospheric relief isolation (block)valve ...18 lat the frequency specified in the Surveillance Frequency Control Program I MILLSTONE  
Auxiliary fjedwater pumps may be considered OPERABLE during alignment and operation l/r steam generator level control, if they are capable of being manually realgnedo the auxiliary feedwater mode of operation.
-UNIT 3 3/4 7-9a Amendment No. 14ý PLANT SYSTEMS 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent reactor plant component cooling water safety loops shall be OPERABLE.APPLICABILITY:
raligne? ..           .       .     ..
MODES 1,2, 3, and 4.ACTION: With only one reactor plant component cooling water safety loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.7.3 At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE: a. At by verifying that each valve (manual, power-operated, or autobtic) servicing safety-related equipment that is not locked, sealed, or otherwis cured in position is in its correct position; and I_ .__ _ 'b. At --- by verifying that: 1) Eac tornatic e actuates to its correct position on its associated Engineere fety F e actuation signal, and 2) Each Component Coo .r System pump starts automatically on an SIS test signal.Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE  
Verifyiot each auxiliary feedwater               . power manual,   . operated,
-UNIT 3 3/4 7-11 Amendment No. 4-2-4, -M&6-ly 24.-, -VV2-PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE.APPLICABILITY:
                                                                            .        .. and automatic
MODES 1,2,3, and 4.ACTION: With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.7.4 At least two service water loops shall be demonstrated OPERABLE: a. At by verifying that each valve (manual, power-operated, or aut tic) servicing safety-related equipment that is not locked, sealed, or otherwi secured in position is in its correct position; and b. At`- --,, r 24 mo-tts by verifying that: 1) Ea utoma valve servicing safety-related equipment actuates to its correct tion its associated Engineered Safety Feature actuation signal, and 2) Each Service Water Sys pump starts automatically on an SIS test signal.Ithe frequency specified in the Surveillance Frequency Control Proqram I MILLSTONE  
                                                                                                  .      . valve
-UNIT 3 3/4 7-12 Amendment No. 4-2-4, 2e6 August 29, 1995 PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with an average water temperature j, of less than or equal to 75°F.APPLICABILITY:
                                                                                                              . in/
MODES 1,2, 3, and 4.ACTION: If the UHS temperature is above 75'F, monitor the UHS temperature once per hour for 12 hours.If the UHS temperature does not drop below 75'F during this period, place the plant in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. During this period, if the UHS temperature increases above 77°F, place the plant in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.7.5 The UHS shall be determined OPERABLE: a. A+ 1-f per 24 hoofs by verifying the average water temperature to be within lim b. At least oncer 6 hours by verifying the average water temperature to be within limits w the erage water temperature exceeds 70 0 F.Ithe freauencv swecified in the Surveillance Freauencv Control Proaram MILLSTONE  
each wter flow path and in each required steam supply flow path to the steam turbin drriven auxiliary feedwater pump, that is not locked, sealed, or otherwise sec d in position, is in the correct positin. .....
-UNIT 3 3/4 7-13 Amendment No. ++9-S~pt@Mb@r 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION  
: b. Spciiaton, Specification 40,           cays -n go I 4.0.5,P2 bby:
/ACTION: (Continued)
t4
                                                                                          , tested pursuant to
: 1)       Verifying that on recirculation flow each motor-driven pump develops a total head of greater than or equal to 3385 feet;
: 2)       Verifying that on recirculation flow the steam turbine-driven pump develops a total head of greater than or equal to 3780 feet when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.
MILLSTONE - UNIT 3                                 3/4 7-5               Amendment No. 96, 400, 4-2-7, 4-39,
                                                                                                          -206,-435-
 
Febmuary 28, 2007 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
: c. At             p             by verifying that each auxiliary feedwater pump starts as     igned automatically upon receipt of an Auxiliary Feedwater Actuation test signsna. For the steam turbine-driven auxiliary feedwater pump, the provisions of Specifi tion 4.0.4 are not applicable for entry into MODE 3.
4.7.1.2.2 An auxiliary fe dwater flow path to each steam generator shall be demonstrated OPERABLE following each OLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to ach steam generator.
Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3                       3/4 7-5a                           Amendment No. 244-  4-
 
September 11, 1997 PLANT SYSTEMS DEMINERALIZED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3   The demineralized water storage tank (DWST) shall be OPERABLE with a water volume of at least 334,000 gallons.
APPLICABILITY:         MODES 1, 2, and 3.
ACTION:
With the DWST inoperable, within 4 hours either:
: a.     Restore the DWST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, or
: b.     Demonstrate the OPERABILITY of the condensate storage tank (CST) as a backup supply to the auxiliary feedwater pumps and restore the DWST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST shall be demonstrated OPERABLE at last               pe 12 hu,,   by verifying the water volume is within its limits when the tank is the sup y source for the auxiliary feedwater pumps.
4.7.1.3.2 The CST shall be demonstrated OPERABLE least once per 12 hours by verifying that the combined volume of both the DWST and CST is at least 384,000 gallons of water whenever the CST and DWST are the supply source for he auxiliary feedwater pumps.
Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 3                         3/4 7-6                           Amendment No. +50-
 
Jftnutftry 31,986 PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4     The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microCurie/gram DOSE EQUIVALENT 1-131.
APPLICABILITY:         MODES 1,2, 3, and 4.
ACTION:
With the specific activity of the Secondary Coolant System greater than 0.1 microCurie/gram DOSE EQUIVALENT 1-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.4     The specific activity of the Secondary Coolant System shall be determined to be within the limit by ........     o t, sampling a         ono          ifFa, 4. -.
determining the Gross Radioactivity and DOSE EQUIVALENT 1-131 Concentration at the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                         3/4 7-7
 
JaT. .a .y 31 , 1 98 6 Delete Table 4.7-1 TABLE 4.7-1 N iLEU IN A K Y kL IUU L A IN   1 L)1I V1 l*
                                                          ,l               IJVA I L 11V 11 SAMPLE AND ANALYSTS PR                       '     MRAM TWY-P       MEASUREMENT                             SAMPL                   D ANALYSIS AN     DNALYSIS                                           FRE             NCY
: 1. Gross Radioacti                             At least once per 72 hour Determination
: 2. Isotopic Analysis for DO                   a)   Once per 31 days, when-EQUIVALENT 1-131                               ever the gross radio-Concentration                                   activity determination indicates concentrations "a
                "*                                          ,,eater ctta bthan wavl   l e l im10%      of the i t fo r ad io i n s .
              ~ever b) Oncethe    pergross 6       nths, when-r~a-.-
                                                        .activity determ inatio indicates concentrations less than or equal to 10%
of the allowable limitj for radioiodines.
MILLSTONE - UNIT 3                       3/4 7-8
 
1O79 99 PLANT SYSTEMS STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES LIMITING CONDITION FOR OPERATION 3.7.1.6     Each steam generator atmospheric relief bypass valve (SGARBV) line shall be OPERABLE, with the associated main steam atmospheric relief isolation (block) valve in the open position.
APPLICABILITY:           MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTIONS
: a.       With one required SGARBV line inoperable, restore required SGARBV line to OPERABLE status within 7 days or be in at least MODE 3 within the next 6 hours and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours. LCO 3.0.4 is not applicable.
: b.       With two or more required SGARBV lines inoperable, restore all but one required SGARBV line to OPERABLE status within 24 hours or be in at least MODE 3 within the next 6 hours and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.6.1   Verify one complete cycle of each SGARBV         , o fi,-tths.
4.7.1.6.2   Verify one complete cycle of each mai   eam atmospheric relief isolation (block) valve ... 18 ~n.*
lat the frequency specified     in the Surveillance Frequency Control Program I MILLSTONE - UNIT 3                         3/4 7-9a                         Amendment No. 14ý
 
PLANT SYSTEMS 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3     At least two independent reactor plant component cooling water safety loops shall be OPERABLE.
APPLICABILITY:         MODES 1,2, 3, and 4.
ACTION:
With only one reactor plant component cooling water safety loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.3     At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE:
: a. At                           by verifying that each valve (manual, power-operated, or autobtic) servicing safety-related       equipment that is not locked, sealed, or I_
otherwis     _ ' in position is in its correct position; and
                      . __ cured
: b. At *.          ---               by verifying that:
: 1)     Eac     tornatic       e actuates to its correct position on its associated Engineere       fety F     e actuation signal, and
: 2)     Each Component Coo             . r System pump starts automatically on an SIS test signal.
Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3                     3/4 7-11                               Amendment No. 4-2-4, -M&6-
 
ly 24.-, -VV2-PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4       At least two independent service water loops shall be OPERABLE.
APPLICABILITY:         MODES 1,2,3, and 4.
ACTION:
With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.4       At least two service water loops shall be demonstrated OPERABLE:
: a.     At                         by verifying that each valve (manual, power-operated, or aut   tic) servicing safety-related equipment that is not locked, sealed, or otherwi secured in position is in its correct position; and
: b.     At`-     -- ,,   r 24 mo-tts by verifying that:
: 1)     Ea     utoma valve servicing safety-related equipment actuates to its correct     tion   its associated Engineered Safety Feature actuation signal, and
: 2)     Each Service Water Sys         pump starts automatically on an SIS test signal.
Ithe frequency specified in the Surveillance Frequency Control Proqram             I MILLSTONE - UNIT 3                         3/4 7-12                     Amendment No. 4-2-4, 2e6
 
August 29, 1995 PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5       The ultimate heat sink (UHS) shall be OPERABLE with an average water temperature       j, of less than or equal to 75°F.
APPLICABILITY:         MODES 1,2, 3, and 4.
ACTION:
If the UHS temperature is above 75'F, monitor the UHS temperature once per hour for 12 hours.
If the UHS temperature does not drop below 75'F during this period, place the plant in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. During this period, if the UHS temperature increases above 77°F, place the plant in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.5         The UHS shall be determined OPERABLE:
: a. A+ 1-f         per 24 hoofs by verifying the average water temperature to be within lim
: b. At least oncer 6 hours by verifying the average water temperature to be within limits w     the erage water temperature exceeds 70 0 F.
Ithe freauencv swecified in the Surveillance Freauencv Control Proaram MILLSTONE - UNIT 3                         3/4 7-13                         Amendment No. ++9-
 
S~pt@Mb@r 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION                                                                       /
ACTION:     (Continued)
: e. With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency Air Filtration System required to be in the emergency mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, or with one or more Control Room Emergency Air Filtration System Trains inoperable due to an inoperable CRE boundary, immediately suspend the movement of recently irradiated fuel assemblies.
: e. With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency Air Filtration System required to be in the emergency mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, or with one or more Control Room Emergency Air Filtration System Trains inoperable due to an inoperable CRE boundary, immediately suspend the movement of recently irradiated fuel assemblies.
SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE:
SURVEILLANCE REQUIREMENTS 4.7.7     Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE:                     the frequency specified in the Surveillance Frequency Control Proqram ]
the frequency specified in the Surveillance Frequency Control Proqram ]a. At m by verifying that the control room air temperature is less than ual to 95'F;b. At pen by initiating, from the contr room, flow through the HEPA filters and charcoal adsorbers and verifying a sys m flow rate of 1,120 cfm +/- 20% and that the system operates for at least 10 cont uous hours with the heaters operating;
: a. At                       m by verifying that the control room air temperature is less than       ual to 95'F;
: c. At -or following painting, fire, or chemical release in any ventilation zone communicating with the system by: 1) Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the system flow rate is 1,120 cfm +/- 20%;2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 54 ft/min; and 3) Verifying a system flow rate of 1,120 cfm +/- 20% during system operation when tested in accordance with ANSI N510-1980.
: b. At             pen                                             by initiating, from the contr room, flow through the HEPA filters and charcoal adsorbers and verifying a sys m flow rate of 1,120 cfm +/- 20% and that the system operates for at least 10 cont uous hours with the heaters operating;
MILLSTONE  
: c. At                 -         or following painting, fire, or chemical release in any ventilation zone communicating with the system by:
-UNIT 3 3/4 7-16 Amendment No. 2, --23, 44, 4-84, 203, 206, 23-7, 243 Septem.be 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
: 1)       Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the system flow rate is 1,120 cfm +/- 20%;
: d. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5%when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86 0 F), and a relative humidity of 70%, and a face velocity of 54 ft/min.,T Ethe frequency specified in the Surveillance Frequency Control Proqram e. At' least .. pr"24 .mnth by: 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm + 20%;2) Deleted 3) Verifying that the heaters dissipate 9.4 +1 kW when tested in accordance with ANSI N510-1980.
: 2)       Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 54 ft/min; and
: f. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1120 cfm +/- 20%; and g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1120 cfm + 20%.h. By performance of CRE unfiltered air inleakage testing in accordance with the CRE Habitability Program at a frequency in accordance with the CRE Habitability Program.* ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.MILLSTONE  
: 3)       Verifying a system flow rate of 1,120 cfm +/- 20% during system operation when tested in accordance with ANSI N510-1980.
-UNIT 3 3/4 7-17 Amendment No. 2, 23, 4-8-1, 20-, 2-20, 24-3 29, 2007 PLANT SYSTEMS 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.9 Two independent Auxiliary Building Filter Systems shall be OPERABLE.APPLICABILITY:
MILLSTONE - UNIT 3                         3/4 7-16         Amendment No. 2, --23, 44, 4-84, 203, 206, 23-7, 243
MODES 1,2, 3, and 4.ACTION: With one Auxiliary Building Filter System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. In addition, comply with the ACTION requirements of Specification 3.6.6.1.SURVEILLANCE REQUIREMENTS the frequency specified in the Surveillance Frequency Control Proqram 4.7.9 Each Au ary Building Filter System shall be demonstrated OPERABLE: a. At p, -r 31 day on a S B-,,6ERBD TEST BASIS by initiating, from the contr 1 room, flow through the HEPA filters and charcoal adsorbers and verifying a sy em flow rate of 30,000 cfm +/-10% and that the system operates for at least 10 con nuous hours with the heaters operating;
 
: b. AtEa 2- p2 4 rntnhs or following painting, fire, or chemical release in any , ventilation zone communicating with the system by: 1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is 30,000 cfm I+10%;2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl MILLSTONE  
Septem.be 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
-UNIT 3 3/4 7-20 Amendment No. 2, 84, +-2-, 4-84, 2404, 206,--2-34--
: d.     After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5%
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86°F), a relative humidity of 70%, and a face velocity of 52 ft/min; and 3) Verifying a system flow rate of 30,000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1980.
when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86 0 F),
: c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 52 ft/min;the frequency specified in the Surveillance Frequency Control Program d. A ..- .1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.8 inches Water Gauge while operating the system at a flow rate of 30,000 cfm +/-10%, 2) Verifying that the system starts on a Safety Injection test signal, and 3) Verifying that the heaters dissipate 180 +/-18 kW when tested in accordance with ANSI N510-1980.
and a relative humidity of 70%, and a face velocity of 54 ft/min.
: e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 30,000 cfm +/- 10%;and f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 30,000 cfm +/-1 0%.* ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.MILLSTONE  
                  ,T Ethe frequency specified in the Surveillance Frequency Control Proqram
-UNIT 3 3/4 7-21 Amendment No. 2, 87, --2--, 4-84, 206-Rme 24, 1997ý-PLANT SYSTEMS 3/4.7.14 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.14 The temperature limit of each area shown in Table 3.7-6 shall not be exceeded.APPLICABILITY:
: e.     At' least .. pr"24 .mnth by:
Whenever the equipment in an affected area is required to be OPERABLE.ACTION: With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6: a. By less than 20'F and for less than 8 hours, record the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s).b. By less than 20'F and for greater than or equal to 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment.
: 1)     Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm + 20%;
The provisions of Specification 3.0.3 are not applicable.
: 2)       Deleted
: c. With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6 by greater than or equal to 20'F, prepare and submit a Special Report as required by ACTION b. above and within 4 hours either restore the area(s) to within the temperature limit(s) or declare the equipment in the affected area(s) inoperable.
: 3)       Verifying that the heaters dissipate 9.4 +1 kW when tested in accordance with ANSI N510-1980.
SURVEILLANCE REQUIREMENTS 4.7.14 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limits: a. At 'a"a F-... ' seve days when the alarm is OPERABLE, and;b. At le4 once per 12 hours when the alarm is inoperable.
: f.     After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1120 cfm +/- 20%; and
the frequency specified in the Surveillance Frequency Control Proqram MILLSTONE  
: g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1120 cfm + 20%.
-UNIT 3 3/4 7-32 Amendment No. 9-5, 4-00, 444-Mait M, 2306 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) 11 Inoperable Equipment Required AC I ION e. Two diesel generators e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.AND e.3 Following restoration of one diesel generator, restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement  
: h.     By performance of CRE unfiltered air inleakage testing in accordance with the CRE Habitability Program at a frequency in accordance with the CRE Habitability Program.
: b. above based on the initial loss of the remaining inoperable diesel generator.-r/-the frequency specified in the Surveillance Frequency Control Progqram[SURVEILLANCE RE(ENTS a. Detern imed OPERABL a pr7dy by verifying correct breaker alignr ents, indicated pow avilabilt,i and b. Dem .nsrtOPERABLE af ............  
* ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.
.dunn shutdown by trans erring (manually and automatically) unit power supply prom the normal circ fit to the alternate circuit.4.8.1.1.2 Each dfsel generator shall be demonstrated OPERABLE:*
MILLSTONE - UNIT 3                     3/4 7-17           Amendment No. 2, 23, 4-8-1, 20-, 2-20, 24-3
: a. Atb 1) Verifying the fuel level in the day tank, 2) Verifying the fuel level in the fuel storage tank, 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, 4) Verifying the lubricating oil inventory in storage, 5) Verifying the diesel starts from standby conditions and achieves generator voltage and frequency at 4160 +420 volts and 60 +0.8 Hz. The diesel generator shall be started for this test by using one of the following signals: a) Manual, or a.All planned starts for the purpose of these surveillances may be preceded by an engine prelube period.MILLSTONE
 
-UNIT 3 3/4 8-3a Amendment No. 4 s0, 64,b4--2,--94,24-0, m229 F.bruftry 2, 2001 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Program I b) Simulated loss-of-offsite power by itself, or c) Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d) An ESF Actuation test signal by itself.Verifying the generator is synchronized and gradually loaded in accordance with the manufacturer's recommendations between 4800-5000 kW* and operates with a load between 4800-5000 kW* for at least 60 minutes, and Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.6)7)b. At 1)a per 184 day by: V ing that the diesel generator starts from standby conditionS and attains generator voltage and freauencv of 4160 +/- 420 volts and 60+/-/0.8 Hz within 11 seconds after the start signal.2) Verifying the generator is synchronized to the associated emergency bus, loaded between 4800-5000 kW* in accordance with the manufacturer's recommendations, and operate with a load between 4800-5000 kW* for a least 60 minutes.T diesel generator shall be started for this test using one of the signals in S teillance Requirement 4.8.1.1.2.a.5.
M*rLc  29, 2007 PLANT SYSTEMS 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.9     Two independent Auxiliary Building Filter Systems shall be OPERABLE.
This test, if it is performed so it coincide w the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also s e to concurrently meet those requirements as well.Aop i -thanud after each operation of the diesel where the period o op ration was greater than or equal to 1 hour by checking for and removing ac mulated water from the day tank;At as* oncz nor3 by checking for and removing accumulated water from the fuel oil storage tanks;t/'I s f/C.d.e. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and: 1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has: a) An API Gravity of within 0.3 degrees at 607F, or a specific gravity of within 0.0016 at 60/607F, when compared to the supplier's certificate, or an absolute specific gravity at 60/607F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;* The operating band is meant as guidance to avoid routine overloading of the diesel.Momentary transients outside the load range shall not invalidate the test.MILLSTONE  
APPLICABILITY:         MODES 1,2, 3, and 4.
-UNIT 3 3/4 8-4 Amendment No. 4, 64, 41-2, 4-3-, 94-ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Proqram b) A kinematic viscosity at 40'C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity, SUS at 100°F of greater than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 125°F; and d) Water and sediment less than 0.05 percent by volume when tested in accordance with ASTM-D 1796-83.By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested ir accordance with ASTM-D975-81 except that: (1) the cetane index shall b determined in accordance with ASTM-D976 (this test is an appropriate approximation for cetane number as stated in ASTM-D975-81  
ACTION:
[Note E]), and (2) the analysis for sulfur may be performed in accordance with ASTM-D 1552-79, ASTM-D2622-82 or ASTM-D4294-83.
With one Auxiliary Building Filter System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. In addition, comply with the ACTION requirements of Specification 3.6.6.1.
: f. t by obtaining a sample of fuel oil in accordance with TM-D2276-78, and verifying that total particulate contamination is less thanmg/liter when checked in accordance with ASTM-D2276-78, Method A;g. As m , during shutdown, by: t c 1) DELETED 2) Verifying the generator capability to reject a load of greater than or equal to 595 kW while maintaining voltage at 4160 +/- 420 volts and frequency at3 Hz;3) Verifying the generator capability to reject a load of 4986 kW without tripping.
SURVEILLANCE REQUIREMENTS the frequency specified in the Surveillance Frequency Control Proqram 4.7.9     Each Au     ary Building Filter System shall be demonstrated OPERABLE:
The generator voltage shall not exceed 5000 volts during and 4784 volts following the load rejection;
: a. At         - p, r 31 day on a S B-,,6ERBD TEST BASIS by initiating, from the contr 1room, flow through the HEPA filters and charcoal adsorbers and verifying a sy em flow rate of 30,000 cfm +/-10% and that the system operates for at least 10 con nuous hours with the heaters operating;
: 4) Simulating a loss-of-offsite power by itself, and: a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts from standby conditions on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +/- 420 volts and 60 +/- 0.8 Hz during this test.Ir MILLSTONE  
: b. AtEa           p22- 4 rntnhs or following painting, fire, or chemical release in any     ,
-UNIT 3 3/4 8-5 Amendment No.4, 4-0, 64, 73, 400, 140,4-2,148, -+94 MaiIt 29, 2007 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
ventilation zone communicating with the system by:
: 8) Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 5335 kW;9) Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2)automatically energizing the emergency loads with offsite power;11) DELETED 12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within +/- 10% of its design interval; and 13) DELETED h. At '--t per 10 yzarS by starting both diesel generators simultaneously from ,.stal by conditions, during shutdown, and verifying that both diesel generators achie generator voltage and frequency at 4160 + 420 volts and 60 + 0.8 Hz in less tha or equal to 11 seconds; and i. At by draining each fuel oil storage tank, removing the ace ted ediment and cleaning the tank using a sodium hypochlorite solution.ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE  
: 1)     Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is 30,000 cfm I+10%;
-UNIT 3 3/4 8-7 Amendment No. 64, 49, 4-00, 4-4-2, 4-4, 4-94, 234-Febrdtry 2 9O4 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 2)     Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl MILLSTONE - UNIT 3                 3/4 7-20           Amendment No. 2, 84, +-2-, 4-84, 2404, 206,
: j. Atper 1 mt by verifying the diesel generator operates for at least 24 hrs. During the first 2 hours of this test, the diesel generator shall be loaded betwee 5400-5500kW*
                                                                                              --2-34--
and during the remaining 22 hours of this test, the diesel generato 7hall be loaded between 4800-50OOkW*.
 
The generator voltage and frequency all be 4160 +/- 420 volts and 60 +/- 0.8 Hz within 11 seconds after the start signal; t steady-state generator voltage and frequency shall be maintained within these li *ts during this test.** Within 5 minutes after completing this 24-hour test, perf Specification 4.8.1.1.2.a.5) excluding the requirement to start the diesel from tandby conditions.***
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86°F), a relative humidity of 70%, and a face velocity of 52 ft/min; and
: k. At least, froxXbac connecti by verifying that the fuel transfer pump transfers fuel to the day tank of each diesel via the installed cross-by v*fying that the following diesel generator jel gen ator starting: 1. At !4 locke 1) Engine overspeed, frequency specified in the veillance Frequency Control Program 2) Lube oil pressure low (2 of 3 logic), I'3) Generator differential, and 4) Emergency stop.* The operating band is meant as guidance to avoid routine overloading of the diesel.Momentary transients outside the load range shall not invalidate the test.** Diesel generator loadings may include gradual loading as recommended by the manufacturer.
: 3)       Verifying a system flow rate of 30,000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1980.
: c. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 52 ft/min; the frequency specified in the Surveillance Frequency Control Program
: d. A   ..-   .
: 1)       Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.8 inches Water Gauge while operating the system at a flow rate of 30,000 cfm +/-10%,
: 2)       Verifying that the system starts on a Safety Injection test signal, and
: 3)       Verifying that the heaters dissipate 180 +/-18 kW when tested in accordance with ANSI N510-1980.
: e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 30,000 cfm +/- 10%;
and
: f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 30,000 cfm +/-1 0%.
* ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.
MILLSTONE - UNIT 3                 3/4 7-21                   Amendment No. 2, 87, --2--, 4-84, 206-
 
Rme 24, 1997ý-
PLANT SYSTEMS 3/4.7.14 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.14       The temperature limit of each area shown in Table 3.7-6 shall not be exceeded.
APPLICABILITY:           Whenever the equipment in an affected area is required to be OPERABLE.
ACTION:
With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6:
: a.       By less than 20'F and for less than 8 hours, record the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s).
: b.     By less than 20'F and for greater than or equal to 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specification 3.0.3 are not applicable.
: c.     With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6 by greater than or equal to 20'F, prepare and submit a Special Report as required by ACTION b. above and within 4 hours either restore the area(s) to within the temperature limit(s) or declare the equipment in the affected area(s) inoperable.
SURVEILLANCE REQUIREMENTS 4.7.14       The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limits:
: a.     At 'a"a       F-...' seve days when the alarm is OPERABLE, and;
: b.     At le4 once per 12 hours when the alarm is inoperable.
the frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3                           3/4 7-32             Amendment No. *-5, 9-5, 4-00, 444-
 
Mait     M, 2306 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) 11   Inoperable Equipment                                       Required AC I ION
: e. Two diesel generators         e.2     Restore one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
AND e.3     Following restoration of one diesel generator, restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement b. above based on the initial loss of the remaining inoperable diesel generator.               -r
                              /-the frequency specified in the Surveillance Frequency Control Progqram[
SURVEILLANCE RE(ENTS
: a.       Detern imed OPERABL a 1at*c                    pr7dy by verifying correct breaker alignr ents, indicated pow         avilabilt,i and shutdown dunn prom af ............ unit. power supply                  by
: b. Dem erring .
nsrtOPERABLE (manually   and automatically)                               the normal trans circ fit to the alternate circuit.
demonstrated OPERABLE:*
4.8.1.1.2   Each dfsel generator shall be
: a.       Atb
: 1)       Verifying the fuel level in the day tank,
: 2)       Verifying the fuel level in the fuel storage tank, the storage
: 3)       Verifying the fuel transfer pump starts and transfers fuel from system to the day tank, inventory in storage,
: 4)       Verifying the lubricating oil m229 Verifying the diesel starts from standby conditions       and achieves generator
: 5)                                                                        following of theHz.
60 +0.8 and one                    signals:
The diesel and frequency at 4160this    +420  volts test by using generator shall be started for voltage a)     Manual, or prelube of these surveillances   may be preceded by an engine a.Allstarts for planned              the purpose period.
s0, Amendment No. 4 64,b4--2,--94,24-0, 3/4 8-3a MILLSTONE UNIT 3-
 
F.bruftry 2, 2001 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Program I b)     Simulated loss-of-offsite power by itself, or c)     Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d)     An ESF Actuation test signal by itself.
: 6)        Verifying the generator is synchronized and gradually loaded in accordance with the manufacturer's recommendations between 4800-5000 kW* and operates with a load between 4800-5000 kW* for at least 60 minutes, and
: 7)        Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
: b.     At   a         per   184 day by:
: 1)            V ing that the diesel generator starts from standby conditionS and             /
attains generator voltage and freauencv of 4160 +/- 420 volts and 60+/-             /
0.8 Hz within 11 seconds after the start signal.
: 2)       Verifying the generator is synchronized to the associated emergency bus, loaded between 4800-5000 kW* in accordance with the manufacturer's recommendations, and operate with a load between 4800-5000 kW* for a t least 60 minutes.
T diesel generator shall be started for this test using one of the signals in               'I S teillance Requirement 4.8.1.1.2.a.5. This test, if it is performed so it coincide s    /
w the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also s e to concurrently meet those requirements as well.
C.      Aop     i           -     thanud after each operation of the diesel where the period o f op ration was greater than or equal to 1 hour by checking for and removing ac mulated water from the day tank;
: d.      At as* oncz nor3               by checking for and removing accumulated water from the fuel oil storage tanks;
: e.     By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and:
: 1)       By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:
a)     An API Gravity of within 0.3 degrees at 607F, or a specific gravity of within 0.0016 at 60/607F, when compared to the supplier's certificate, or an absolute specific gravity at 60/607F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;
* The operating band is meant as guidance to avoid routine overloading of the diesel.
Momentary transients outside the load range shall not invalidate the test.
MILLSTONE - UNIT 3                           3/4 8-4           Amendment No. 4, 64, 41-2, 4-3-,     94-
 
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Proqram b)     A kinematic viscosity at 40'C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity, SUS at 100°F of greater than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the supplier's certification; c)     A flash point equal to or greater than 125°F; and d)     Water and sediment less than 0.05 percent by volume when testedt in accordance with ASTM-D 1796-83.
By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested ir accordance with ASTM-D975-81 except that: (1) the cetane index shall bc determined in accordance with ASTM-D976 (this test is an appropriate approximation for cetane number as stated in ASTM-D975-81 [Note E]),
and (2) the analysis for sulfur may be performed in accordance with ASTM-D 1552-79, ASTM-D2622-82 or ASTM-D4294-83.
: f. t                             by obtaining a sample of fuel oil in accordance with TM-D2276-78, and verifying that total particulate contamination is less than
* mg/liter when checked in accordance with ASTM-D2276-78, Method A;
: g.       As                 m       , during shutdown, by:
: 1)       DELETED                                                                         Ir
: 2)       Verifying the generator capability to reject a load of greater than or equal to 595 kW while maintaining voltage at 4160 +/- 420 volts and frequency at 60+/-* 3 Hz;
: 3)       Verifying the generator capability to reject a load of 4986 kW without tripping. The generator voltage shall not exceed 5000 volts during and 4784 volts following the load rejection;
: 4)       Simulating a loss-of-offsite power by itself, and:
a)     Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b)     Verifying the diesel starts from standby conditions on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +/- 420 volts and 60 +/- 0.8 Hz during this test.
MILLSTONE - UNIT 3                       3/4 8-5             Amendment No.4, 4-0, 64, 73, 400, 140,4-2,148,-+94
 
MaiIt 29, 2007 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: 8)     Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 5335 kW;
: 9)     Verifying the diesel generator's capability to:
a)       Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b)       Transfer its loads to the offsite power source, and c)       Be restored to its standby status.
: 10)   Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by:
(1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;
: 11)   DELETED
: 12)   Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within +/- 10% of its design interval; and
: 13)   DELETED
: h. At '--t       per 10 yzarS by starting both diesel generators simultaneously from     ,.
stal by conditions, during shutdown, and verifying that both diesel generators achie generator voltage and frequency at 4160 + 420 volts and 60 + 0.8 Hz in less tha or equal to 11 seconds; and
: i. At                           by draining each fuel oil storage tank, removing the ace       ted ediment and cleaning the tank using a sodium hypochlorite solution.
ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                         3/4 8-7           Amendment No. 64, 49, 4-00, 4-4-2, 4-4, 4-94, 234-
 
Febrdtry 2 9O4 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: j.       Atper               1 mt       by verifying the diesel generator operates for at least 24 hrs. During the first 2 hours of this test, the diesel generator shall be loaded betwee 5400-5500kW* and during the remaining 22 hours of this test, the diesel generato 7hall         be loaded between 4800-50OOkW*. The generator voltage and frequency all be 4160 +/- 420 volts and 60 +/- 0.8 Hz within 11 seconds after the start signal; t steady-state generator voltage and frequency shall be maintained within these li *ts during this test.** Within 5 minutes after completing this 24-hour test, perf     Specification 4.8.1.1.2.a.5) excluding the requirement to start the diesel from tandby conditions.***
: k.       At least,                     by verifying that the fuel transfer pump transfers fuel froxXbac                      to the day tank of each diesel via the installed cross-connecti
: 1.      At !4                          by v*fying that the following diesel generator locke                        jel gen ator starting:
: 1)       Engine overspeed, frequency specified in the veillance Frequency Control Program
: 2)       Lube oil pressure low (2 of 3 logic),   I'
: 3)       Generator differential, and
: 4)       Emergency stop.
* The operating band is meant as guidance to avoid routine overloading of the diesel.
Momentary transients outside the load range shall not invalidate the test.
** Diesel generator loadings may include gradual loading as recommended by the manufacturer.
If Surveillance Requirement 4.8.1.1.2.a.5) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated between 4800-5000 kW for 2 hours or until operating temperature has stabilized.
If Surveillance Requirement 4.8.1.1.2.a.5) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated between 4800-5000 kW for 2 hours or until operating temperature has stabilized.
MILLSTONE  
MILLSTONE - UNIT 3                           3/4 8-8                 Amendment No. 1-0, 64, 440, t-94  -
-UNIT 3 3/4 8-8 Amendment No. 1-0, 64, 440, t ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE: a. 125-volt Battery Bank 301A-l, and an associated full capacity charger, b. 125-volt Battery Bank 301A-2, and an associated full capacity charger, c. 125-volt Battery Bank 301 B-I and an associated full capacity charger, and d. 125-volt Battery Bank 301B-2 and an associated full capacity charger.APPLICABILITY:
 
MODES 1, 2, 3, and 4.ACTION: a. With either Battery Bank 301A-1 or 301B-l, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.b. With either Battery Bank 301A-2 or 301B-2 inoperable, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS f --Ithe frequency specified in the Surveillance Frequency Control Program 4.8.2.1 Each 12 -volt battery bank and charger shall be demonstrated OPERABLE: a. A --- t ----r 7 day by verifying that: 1) The parameters in Table 4.8-2a meet the Category A limits, and 2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.MILLSTONE  
ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1   As a minimum, the following D.C. electrical sources shall be OPERABLE:
-UNIT 3 3/4 8-11 Amendment No. -6+
: a. 125-volt Battery Bank 301A-l, and an associated full capacity charger,
: b. 125-volt Battery Bank 301A-2, and an associated full capacity charger,
: c. 125-volt Battery Bank 301 B-I and an associated full capacity charger, and
: d. 125-volt Battery Bank 301B-2 and an associated full capacity charger.
APPLICABILITY:         MODES 1, 2, 3, and 4.
ACTION:
: a. With either Battery Bank 301A-1 or 301B-l, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b. With either Battery Bank 301A-2 or 301B-2 inoperable, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS f   --Ithe frequency specified in the Surveillance Frequency Control Program 4.8.2.1   Each 12 -volt battery bank and charger shall be demonstrated OPERABLE:
: a. A   ---t - - --   r 7 day by verifying that:
: 1)       The parameters in Table 4.8-2a meet the Category A limits, and
: 2)       The total battery terminal voltage is greater than or equal to 129 volts on float charge.
MILLSTONE - UNIT 3                         3/4 8-11                           Amendment No. -6+
* 95197 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
* 95197 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: b. At least ..... per 92 da"- and within 7 days after a battery discharge with battery teinal voltage below 110 volts, or battery overcharge with battery terminal v itage above 150 volts, by verifying that:The parameters in Table 4.8-2a meet the Category B limits, 2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohm, and 3) The average electrolyte temperature of six connected cells is above 60'F.Ppei 18 inanth by verifying that: the frequency specified in the Surveillance Frequency Control Program The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,--4 3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohm, and) Each battery charger will supply at least the amperage indicated in Table 4.8-2b at greater than or equal to 132 volts for at least 24 hours.d. t least .... e per 18 mI th ., during shutdown, by verifying that the battery apacity is adequate to supply and maintain in OPERABLE status all of the actual o simulated emergency loads for the design duty cycle when the battery is s'uected to a battery service test;e. t 16 , during shutdown, by verifying that the battery apacity is at least 80% of the manufacturer's rating when subjected to a rformance discharge test. 6, ii .I 6-ionth intei-vai this performance d charge test may be performed in li of the battery service test required by S ecification 4.8.2.1d.;
: b. At least
and f. At pe, during shu own, by giving performance discharge tests of battery capacity to any battery th t shows signs of degradation or has reached 85% of the service life expected or the application.
                              .....       per 92 da"- and within 7 days after a battery discharge with battery teinal voltage below 110 volts, or battery overcharge with battery terminal v itage above 150 volts, by verifying that:
Degradation is indicated when the battery capacity drops ore than 10% of rated capacity from its average on previous performance tests, or s below 90% of the manufacturer's rating.MILLSTONE  
                          *)        The parameters in Table 4.8-2a meet the Category B limits,
-UNIT 3 Set 1 1 bA-4+ 004 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION APPLICABILITY:
: 2)         There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohm, and
MODES 1, 2, 3, and 4.ACTION: a. With one of the required trains of A.C. emergency busses not OPERABLE, restore the inoperable train to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.b. With one A.C. vital bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. vital bus within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and (2) reenergize the A.C.vital bus from its associated inverter connected to its associated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined OPERABLE in the specified manner kaM-oee per- days by verifying correct breaker alignment and indicated voltage on e es.Ithe frequency specified in the Surveillance Frequency Control Proqramr MILLSTONE  
: 3)         The average electrolyte temperature of six connected cells is above 60'F.
-UNIT 3 3/4 8-17 Amendment No.-64, 22-september 18, 2008 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION (Continued)
Ppei             18 inanth by verifying that:
: 4) Two 125 volt DC Busses consisting of: a) Bus #301B-1 energized from Battery Bank #301B-1, and b) Bus #301 B-2 energized from Battery Bank #301 B-2.APPLICABILITY:
the frequency                       The cells, cell plates, and battery racks show no visual indication of specified in the                    physical damage or abnormal deterioration, Surveillance
MODES 5 and 6.ACTION: With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity additions that could result in loss of required SDM or boron concentration, movement of recently irradiated fuel assemblies, crane operation with loads over the fuel storage pool, or operations 4" with a potential for draining the reactor vessel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible.SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner t least.f.ee pefr day-s by verifying correct breaker alignment and indicated voltae usses.Ithe frequency specified in the Surveillance Frequencv Control Program MILLSTONE  
                -- 4                The cell-to-cell and terminal connections are clean, tight, and coated with Frequency anticorrosion material, Control Program
-UNIT 3 3/4 8-18a Amendment No. 446, 230,443 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of the following reactivity conditions is met; either: a. A Keff of 0.95 or less, or b. A boron concentration of greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).Additionally, the CVCS valves of Specification 4.1.1.2.2 shall be closed and secured in position.APPLICABILITY:
: 3)         The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohm, and
MODE 6.*ACTION: a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and positive reactivity additions and initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to the limit specified in the COLR, whichever is the more restrictive.
                          )         Each battery charger will supply at least the amperage indicated in Table 4.8-2b at greater than or equal to 132 volts for at least 24 hours.
: b. With any of the CVCS valves of Specification 4.1.1.2.2 not closed** and secured in position, immediately close and secure the valves.SURVEILLANCE REQUIREMENTS 4.9.1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a. Removing or unbolting the reactor vessel head, and b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling cavity shall be determined by chemical analysis at a pE 7a hPur.4.9.1.1.3 The CVCS valves of Specification 4.1.1.2. s ýrified closed and locked at lem-oefee-pet-31 day .t Ithe frequency specified in the Surveillance Frequency Control Proqram-T='
: d. t least
* The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.** Except those opened under administrative control.MILLSTONE  
                              ....       e per 18 .mI th , during shutdown, by verifying that the battery apacity is adequate to supply and maintain in OPERABLE status all of the actual o simulated emergency loads for the design duty cycle when the battery is s'uected to a battery service test;
-UNIT 3 3/4 9-1 Amendment No. -0, 60, 99, 4-4-3, 2-0-3, 24-9, 240 F 20, 2002 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be greater than or equal to 800 ppm. 4-APPLICABILITY:
: e. t 16                           , during shutdown, by verifying that the battery apacity is at least 80% of the manufacturer's rating when subjected to a rformance discharge test. 6,ii         I. 6-ionth intei-vai this performance d charge test may be performed in li of the battery service test required by S ecification 4.8.2.1d.; and
Whenever fuel assemblies are in the spent fuel pool.ACTION: a. With the boron concentration less than 800 ppm, initiate action to bring the boron concentration in the fuel pool to at least 800 ppm within 72 hours, and b. With the boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool and loads over the spent fuel racks.SURVEILLANCE REQUIREMENTS 4.9.1.2 Verify that the boron concentration in the fuel pool is greater than or equal to 800 ppm evey" 7 dwgy.lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE  
: f. At                 pe,             during shu own, by giving performance discharge tests of battery capacity to any battery th t shows signs of degradation or has reached 85% of the service life expected or the application. Degradation is indicated when the battery capacity drops ore than 10% of rated capacity from its average on previous performance tests, or s below 90% of the manufacturer's rating.
-UNIT 3 3/4 9-1a Amendment No. 4-2, 4-58, 4-89, 203-REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 Two Source Range Neutron Flux Monitors shall be OPERABLE with continuous visual indication in the control room, and one with audible indication in the containment and control room.APPLICABILITY:
MILLSTONE - UNIT 3
MODE 6.ACTION: a. With one of the above required monitors inoperable immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.b. With both of the above required monitors inoperable determine the boron concentration of the Reactor Coolant System within 4 hours and at least once per 12 hours thereafter.
 
SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of: a. A CHANNEL CHECK and verification of audible counts  
Set 1 1 bA-4+ 004 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION APPLICABILITY: MODES 1, 2, 3, and 4.
: b. A CHANNEL CALIBRATION at Ithe frequency specified in the Surveillance Frequency Control Program* Neutron detectors are excluded from CHANNEL CALIBRATION.
ACTION:
MILLSTONE  
: a. With one of the required trains of A.C. emergency busses not OPERABLE, restore the inoperable train to OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
-UNIT 3 3/4 9-2 Amendment No. 44-7, 20-3, 430 Mare.h 17, 2004 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status: a. The equipment access hatch shall be either: 1. closed and held in place by a minimum of four bolts, or 2. open under administrative control
: b. With one A.C. vital bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. vital bus within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and (2) reenergize the A.C.
* and capable of being closed and held in place by a minimum of four bolts, b. A personnel access hatch shall be either: 1. closed by one personnel access hatch door, or 2. capable of being closed by an OPERABLE personnel access hatch door, under administrative control,*
vital bus from its associated inverter connected to its associated D.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
and /c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1. Closed by an isolation valve, blind flange, or manual valve, or 2. Be capable of being closed under administrative control.*APPLICABILITY:
: c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
During movement of fuel within the containment building.ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of fuel in the containment building.SURVEILLANCE REQUIREMENTS 4.9.4.a Verify each required containment penetrations is in the required status ole-enee per 4.9.4.b DELETED Ithe frequency specified in the Surveillance Frequency Control Program Administrative controls shall ensure that appropriate personnel are aware that the equipment access hatch penetration, personnel access hatch doors and/or other containment penetrations are open, and that a specific individual(s) is designated and available to close the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g. cables and hoses)that could prevent closure of the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations must be capable of being quickly removed.MILLSTONE  
SURVEILLANCE REQUIREMENTS 4.8.3.1   The specified busses shall be determined OPERABLE in the specified manner kaM-oee per- days by verifying correct breaker alignment and indicated voltage on e           es.
-UNIT 3 3/4 9-4 Amendment No. 203, 2+9-REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.*
Ithe frequency specified in the Surveillance Frequency Control Proqramr MILLSTONE - UNIT 3                         3/4 8-17                       Amendment No.-64, 22-
APPLICABILITY:
 
MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.ACTION: With no RHR loop OPERABLE or in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and suspend loading irradiated fuel assemblies in the core and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible.
september 18, 2008 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION (Continued)
Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at ...a...... H 12....f" t Ithe frequency specified in the Surveillance Frequency Control Program I The RHR loop may be removed from operation for up to 1 hour per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.I, MILLSTONE  
: 4)     Two 125 volt DC Busses consisting of:
-UNIT 3 3/4 9-8 Amendment No. 4-14, 23 REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.*
a)       Bus #301B-1 energized from Battery Bank #301B-1, and b)       Bus #301 B-2 energized from Battery Bank #301 B-2.
APPLICABILITY:
APPLICABILITY:         MODES 5 and 6.
MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet.ACTION: a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.b. With no RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and immediately initiate corrective action to return the required RHR loop to operation.
ACTION:
Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at lcatt zncz per 12 ho .Ithe frequency specified in the Surveillance Frequency Control Program* The RHR loop may be removed from operation for up to 1 hour per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron J, concentration less than required to meet the boron concentration of LCO 3.9.1.1.MILLSTONE  
With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity additions that could result in loss of required SDM or boron concentration, movement of recently irradiated fuel assemblies, crane operation with loads over the fuel storage pool, or operations     4" with a potential for draining the reactor vessel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible.
-UNIT 3 3/4 9-9 Amendment No. 4047, 2-3 Februay 20, 2002 REFUELING OPERATIONS 3/4.9.10 WATER LEVEL -REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.APPLICABILITY:
SURVEILLANCE REQUIREMENTS 4.8.3.2     The specified busses shall be determined energized in the required manner t least
During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6.ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel.SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required deptht least Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE  
.f.ee pefr day-s by verifying correct breaker alignment and indicated voltae             usses.
-UNIT 3 3/4 9-11 Amendment No. -2&3-Oeoe 25, 1990 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL -STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.APPLICABILITY:
Ithe frequency specified in the Surveillance Frequencv Control Program MILLSTONE - UNIT 3                           3/4 8-18a                 Amendment No. 446, 230,443
Whenever irradiated fuel assemblies are in the storage pool.ACTION: a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.b. The provisions of Specification 3.0.3 are not applicable.
 
SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at J.t .... &.... , days when irradiated fuel assemblies are in the fuel storage pool.Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE  
066*2UU 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.1       The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:
-UNIT 3 3/4 9-12 Amendment No. 55-3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).APPLICABILITY:
: a.       A Keff of 0.95 or less, or
MODE 2.ACTION: a. With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by 'Specification 3.1.1.1 is restored.SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at 4.10.1.2 Each full-le control rod not fully inserted shall be demonstrated capable of full insertion when tripped from ast the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to les an the limits of Specification 3.1.1.1.Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE  
: b.       A boron concentration of greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).
-UNIT 3 3/4 10-1 Amendment No. 4+3-Dceember 10, 2003 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.10.2.1.2 below.APPLICABILITY:
Additionally, the CVCS valves of Specification 4.1.1.2.2 shall be closed and secured in position.
MODE 1.ACTION: With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either: a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or b. Be in HOT STANDBY within 6 hours.SURVEILLANCE REQUIREMENTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least one@ per hou during PHYSICS TESTS.4.10.2.1.2 The Surveillance Req rements of the below listed specifications shall be performed at p-er 12 hurs during YSICS TESTS: a. Specifications 4.2.2. .2 and 4.2.2.1.3, and b. Spec ation 4.2.3. 1.b. Spei[the frequency specified in the Surveillance Frequency Control Program[MILLSTONE  
APPLICABILITY:           MODE 6.*
-UNIT 3 3/4 10-2 Amendment No. 444-SPECIAL TEST EXCETIlONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541 'F.APPLICABILITY:
ACTION:
MODE 2.ACTION: a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.b. With a Reactor Coolant System operating loop temperature (Tavg) less than 541 °F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at lcast once per heur during PHYSICS TESTS.4.10.3.2 Each Intermediate and P er Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST w in 12 hours prior to initiating PHYSICS TESTS.4.10.3. The Reactor Coolant System teierature (Tavg) shall be determined to be greater than or equal to 541'F at 1 uring PHYSICS TESTS.Fthe frequencv specified in the Surveillance Frequencv Control Proqramr MILLSTONE  
: a.       With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and positive reactivity additions and initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to the limit specified in the COLR, whichever is the more restrictive.
-UNIT 3 3/4 10-4 1annaty 3fi9866 SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided: a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.APPLICABILITY:
: b.       With any of the CVCS valves of Specification 4.1.1.2.2 not closed** and secured in position, immediately close and secure the valves.
During operation below the P-7 Interlock Setpoint.ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers.SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least ete per- heur during STARTUP and PHYSICS TESTS.4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating STARTUP and PHYSICS TESTS.Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE  
SURVEILLANCE REQUIREMENTS 4.9.1.1.1     The more restrictive of the above two reactivity conditions shall be determined prior to:
-UNIT 3 3/4 10-5 Sept-mber 30N "008 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
: a.     Removing or unbolting the reactor vessel head, and
: b.     Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling cavity shall be determined by chemical analysis at a             pE 7a hPur.
4.9.1.1.3     The CVCS valves of Specification 4.1.1.2. s           ýrified closed and locked at lem-oefee-pet-31 day .                                                                               t Ithe frequency specified in the Surveillance Frequency Control Proqram-T='
* The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
** Except those opened under administrative control.
MILLSTONE - UNIT 3                           3/4 9-1           Amendment No. -0, 60, 99, 4-4-3, 2-0-3, 24-9, 240
 
F         20, 2002
                                                                                        .bi*uly REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.2     The soluble boron concentration of the Spent Fuel Pool shall be greater than or equal to 800 ppm.                                                                               4-APPLICABILITY:
Whenever fuel assemblies are in the spent fuel pool.
ACTION:
: a.     With the boron concentration less than 800 ppm, initiate action to bring the boron concentration in the fuel pool to at least 800 ppm within 72 hours, and
: b.     With the boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool and loads over the spent fuel racks.
SURVEILLANCE REQUIREMENTS 4.9.1.2     Verify that the boron concentration in the fuel pool is greater than or equal to 800 ppm evey"7 dwgy.
lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                         3/4 9-1a             Amendment No. 4-2, 4-58, 4-89, 203-
 
REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2       Two Source Range Neutron Flux Monitors shall be OPERABLE with continuous visual indication in the control room, and one with audible indication in the containment and control room.
APPLICABILITY:           MODE 6.
ACTION:
: a.     With one of the above required monitors inoperable immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.
: b.     With both of the above required monitors inoperable determine the boron concentration of the Reactor Coolant System within 4 hours and at least once per 12 hours thereafter.
SURVEILLANCE REQUIREMENTS 4.9.2       Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:
: a.     A CHANNEL CHECK and verification of audible counts at-least-e*se-per
: b.     A CHANNEL CALIBRATION at Ithe frequency specified in the Surveillance Frequency Control Program
* Neutron detectors are excluded from CHANNEL CALIBRATION.
MILLSTONE - UNIT 3                         3/4 9-2                 Amendment No. 44-7, 20-3, 430
 
Mare.h 17, 2004 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4       The containment building penetrations shall be in the following status:
: a.     The equipment access hatch shall be either:
: 1. closed and held in place by a minimum of four bolts, or
: 2. open under administrative control
* and capable of being closed and held in place by a minimum of four bolts,
: b.     A personnel access hatch shall be either:
: 1. closed by one personnel access hatch door, or
: 2. capable of being closed by an OPERABLE personnel access hatch door, under administrative control,* and                                               /
: c.     Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
: 1. Closed by an isolation valve, blind flange, or manual valve, or
: 2.     Be capable of being closed under administrative control.*
APPLICABILITY:         During movement of fuel within the containment building.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of fuel in the containment building.
SURVEILLANCE REQUIREMENTS 4.9.4.a     Verify each required containment penetrations is in the required status         ole-enee per 4.9.4.b     DELETED Ithe frequency specified in the Surveillance Frequency Control Program Administrative controls shall ensure that appropriate personnel are aware that the equipment access hatch penetration, personnel access hatch doors and/or other containment penetrations are open, and that a specific individual(s) is designated and available to close the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g. cables and hoses) that could prevent closure of the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations must be capable of being quickly removed.
MILLSTONE - UNIT 3                         3/4 9-4                       Amendment No. 203, 2+9-
 
                                                                                          -O2*"ý,6 REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1     At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.*
APPLICABILITY:         MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.
ACTION:
With no RHR loop OPERABLE or in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and suspend loading irradiated fuel assemblies in the core and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access t
from the containment atmosphere to the outside atmosphere within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.8.1     At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at ...a...... 12....
H f"
Ithe frequency specified in the Surveillance Frequency Control Program I The RHR loop may be removed from operation for up to 1 hour per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.               I, MILLSTONE - UNIT 3                         3/4 9-8                     Amendment No. 4-14,23
 
REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2     Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.*
APPLICABILITY:           MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet.
ACTION:
: a.     With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
: b.     With no RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.8.2     At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at lcatt zncz per 12 ho .
Ithe frequency specified in the Surveillance Frequency Control Program
* The RHR loop may be removed from operation for up to 1 hour per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron     J, concentration less than required to meet the boron concentration of LCO 3.9.1.1.
MILLSTONE - UNIT 3                           3/4 9-9                     Amendment No. 4047, 2-3
 
Februay 20, 2002 REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10     At least 23 feet of water shall be maintained over the top of the reactor vessel flange.
APPLICABILITY:           During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6.
ACTION:
With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel.
SURVEILLANCE REQUIREMENTS 4.9.10     The water level shall be determined to be at least its minimum required deptht least Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                           3/4 9-11                         Amendment No. -2&3-
 
Oeoe 25, 1990 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11     At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY:           Whenever irradiated fuel assemblies are in the storage pool.
ACTION:
: a.     With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.
: b.     The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.11     The water level in the storage pool shall be determined to be at least its minimum required depth at J.t....     ,
                                &.... days when irradiated fuel assemblies are in the fuel storage pool.
Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                         3/4 9-12                             Amendment No. 55-
 
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1     The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).
APPLICABILITY:           MODE 2.
ACTION:
: a.     With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
: b.     With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at 4.10.1.2 Each full-le         control rod not fully inserted shall be demonstrated capable of full insertion when tripped from       ast the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to les an the limits of Specification 3.1.1.1.
Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                           3/4 10-1                           Amendment No. 4+3-
 
Dceember 10, 2003 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2.1     The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
: a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
: b.     The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.10.2.1.2 below.
APPLICABILITY:           MODE 1.
ACTION:
With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either:
: a.     Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or
: b.     Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least one@ per hou during PHYSICS TESTS.
4.10.2.1.2 The Surveillance Req rements of the below listed specifications shall be performed at           p-er 12 hurs during         YSICS TESTS:
: a.     Specifications 4.2.2. .2 and 4.2.2.1.3, and
: b. b. Spec       ation 4.2.3. 1.
Spei
[the frequency specified in the Surveillance Frequency Control Program[
MILLSTONE - UNIT 3                           3/4 10-2                           Amendment No. 444-
 
SPECIAL TEST EXCETIlONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3       The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:
: a.       The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
: b.       The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and
: c.       The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541 'F.
APPLICABILITY:           MODE 2.
ACTION:
: a.       With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
: b.       With a Reactor Coolant System operating loop temperature (Tavg) less than 541 °F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.10.3.1     The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at lcast once per heur during PHYSICS TESTS.
4.10.3.2     Each Intermediate and P er Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST w in 12 hours prior to initiating PHYSICS TESTS.
4.10.3.       The Reactor Coolant System teierature (Tavg) shall be determined to be greater than or equal to 541'F at 1                               uring PHYSICS TESTS.
Fthe frequencv specified in the Surveillance Frequencv Control Proqramr MILLSTONE - UNIT 3                           3/4 10-4
 
1annaty 3fi9866 SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4     The limitations of Specification 3.4.1.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided:
: a.     The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
: b.     The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
APPLICABILITY:         During operation below the P-7 Interlock Setpoint.
ACTION:
With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers.
SURVEILLANCE REQUIREMENTS 4.10.4.1   The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least ete per-heur during STARTUP and PHYSICS TESTS.
4.10.4.2   Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating STARTUP and PHYSICS TESTS.
Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3                         3/4 10-5
 
Sept-mber 30N "008 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
RG 1.183), which were considered in completing the vulnerability assessments, are documented in the UFSAR/current licensing basis. Compliance with these RGs is consistent with the current licensing basis as described in the UFSAR and other licensing basis documents.
RG 1.183), which were considered in completing the vulnerability assessments, are documented in the UFSAR/current licensing basis. Compliance with these RGs is consistent with the current licensing basis as described in the UFSAR and other licensing basis documents.
: d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVs, operating at the flow rate required by the Surveillance Requirements, at a Frequency of 48 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.e. The quantitative limits on unfiltered air inleakage into the CRE.These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph  
: d.     Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVs, operating at the flow rate required by the Surveillance Requirements, at a Frequency of 48 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.
: c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.
: e.     The quantitative limits on unfiltered air inleakage into the CRE.
Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.Insert 1 The provisions of Surveillance Requirement 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs  
These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
: c. and d., respectively.
Insert 1           The provisions of Surveillance Requirement 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c. and d., respectively.
6.8.5 Written procedures shall be established, implemented and maintained covering Section I.E, Radiological Environmental Monitoring, of the REMODCM.6.8.6 All procedures and procedure changes required for the Radiological Environmental Monitoring Program (REMP) of Specification 6.8.5 above shall be reviewed by an individual (other than the author) from the organization responsible for the REMP and approved by appropriate supervision.
6.8.5 Written procedures shall be established, implemented and maintained covering Section I.E, Radiological Environmental Monitoring, of the REMODCM.
6.8.6 All procedures and procedure changes required for the Radiological Environmental Monitoring Program (REMP) of Specification 6.8.5 above shall be reviewed by an individual (other than the author) from the organization responsible for the REMP and approved by appropriate supervision.
Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the organization responsible for the REMP, within 14 days of implementation.
Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the organization responsible for the REMP, within 14 days of implementation.
MILLSTONE
MILLSTONE - UNIT 3                          6-17e                              Amendment No. 245
-UNIT 3 6-17e Amendment No. 245 INSERTS FOR TECHNICAL SPECIFICATIONS MARKUPS INSERT 1 (for TS 6.8.4)i .Surveillance Frequency Control Progqram This program provides controls for surveillance frequencies.
 
The program shall ensure that surveillance requirements specified in the technical specification are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.a. The Surveillance Frequency Control Program shall contain a list of frequencies of those surveillance requirements for which the frequency is controlled by the program.b. Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI
INSERTS FOR TECHNICAL SPECIFICATIONS MARKUPS INSERT 1 (for TS 6.8.4)
: i. Surveillance Frequency Control Progqram This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specification are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
: a. The Surveillance Frequency Control Program shall contain a list of frequencies of those surveillance requirements for which the frequency is controlled by the program.
: b. Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
: c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the frequencies established in the Surveillance Frequency Control Program.
 
INSERTS FOR TECHNICAL SPECIFICATIONS BASES INSERT 2 The surveillance frequency is controlled under the Surveillance Frequency Control Program I
 
Serial No. 10-711 Docket No. 50
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Serial No. 10-711 Docket No. 50-423 Attachment 5 Page 2 of 2 2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
 
Response:
Serial No. 10-711 Docket No. 50-423 Attachment 5 Page 2 of 2
No.No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.
: 2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
In addition, the changes do not impose any new or different requirements.
Response: No.
The changes do not alter assumptions made in the safety analysis.
No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements.
The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.
: 3. Do the proposed changes involve a significant reduction in the margin of safety?Response:
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
No.The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies.
: 3. Do the proposed changes involve a significant reduction in the margin of safety?
Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Dominion will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based upon the reasoning presented above, Dominion concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.
Response: No.
Serial No. 10-711 Docket No. 50-423 ATTACHMENT 6 MARKED-UP TECHNICAL SPECIFICATIONS BASES CHANGES DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Insert 2 The surveillance frequency is controlled under the Surveillance Frequency Control Program.
The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Dominion will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
LDBCR ,5-MP3-,25 MarchJqeJ.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
7 , REACTIVITY CONTROL SYSTEMS lat frequency specified in the Surveillance BASES lFrequency Control Program MOVABLE CONTROL ASSEMBLIES (Continued)
Based upon the reasoning presented above, Dominion concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.
Control rod positions and OPERABILITY of the position indicators are required to be verified on a nemil h basis of on-ce per 12 hou's with more frequent verifications required if an automatic monitoring channel is inoperable.
 
These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
Serial No. 10-711 Docket No. 50-423 ATTACHMENT 6 MARKED-UP TECHNICAL SPECIFICATIONS BASES CHANGES DOMINION NUCLEAR CONNECTICUT, INC.
The Digital Rod Position Indication (DRPI) System is defined as follows:* Rod position indication as displayed on DRPI display panel (MB4), or* Rod position indication as displayed by the Plant Process Computer System.With the above definition, LCO 3.1.3.2, "ACTION a." is not applicable with either DRPI display panel or the plant process computer points OPERABLE.The plant process computer may be utilized to satisfy DRPI System requirements which meets LCO 3.1.3.2, in requiring diversity for determining digital rod position indication.
MILLSTONE POWER STATION UNIT 3
Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least @ncc cach 12 h, ur , except during the time when the rod position deviation monitor is inoperable, then o are the Demand Position Indication System and the DRPI System at least once each 4 hours. the frequency specified in the SurveillanceFrequency Control Program I The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours.Trelmical Spcfcto SR 4.1.3.2.1 detcrminc2s each digital fod position indicator tob O-ERL*__E by ýin 1 the Dtntiad Psition Indi.ation tm and the System agr-, wlithin 1-2 steps at least ecncc eaeh 12 hours, emeept d~rig the time whe the td positiondeiaion mntris inperable, then 1 conipaIe theL Demlanld Position 1 Indicatiu Systini and the ORP system -At ]ARA* non swch 4 hoursw.The Rod Deviation Monitor is generated ofnly &fra the DRPI panel at MB4. Therefore, w~henrd flieii peilfb 1 inl SURVeILLANCqet  
 
*EQUIREMENf~TS eveiy 4 frm D1uplicative paarps ILLSTIONE  
BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Insert 2 The surveillance frequency is controlled under the Surveillance Frequency Control Program.
-UNIT 3 B 3/4 1-4 Amendment No.i0, LDDC9R 65-MP3-025 M hr 1  , 26,6-REACTIVITY CONTROL SYSTEMS the frequency specified in the Surveillance Frequency Control Program BASES MOVABLE CONTROL ASSEMBLIES (Continued)d Additional surveillance is required to ensure the plant ocess computer indications are in agreement with those displayed on the DRPI. This a ditional SURVEILLANCE REQUIREMENT is as follows: Each rod position indication as displ ed by the plant process computer shall be determined to be OPERABLE by rifying the rod position indication as displayed on the DRPI display panel agrees wi te rod position indication as displayed by the plant process computer at least once per 1 2 hours.The rod position indication, as displayed by DRPI display panel (MB4), is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a refueling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position.
 
Additionally, the plant process computer provides the same type and level of accuracy as the DRPI System (MB4). The plant process computer does not provide any alarm or rod position deviation monitoring as does DRPI display panel (MB4).For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.
LDBCR ,5-MP3-,25 MarchJqeJ. 7 , 2006~~s'*
In the event the plant is unable to verify the rod(s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours for this verification.
REACTIVITY CONTROL SYSTEMS                         lat frequency specified in the Surveillance BASES                                             lFrequency Control Program MOVABLE CONTROL ASSEMBLIES (Continued)
For LCO 3.1.3.6 the control bank insertion limits are specified in the CORE OPERATING LIMITS REPORT (COLR). These insertion limits are the initial assumptions in safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions, assumptions of available SHUTDOWN MARGIN, and initial reactivity insertion rate.The applicable I&C calibration procedure (Reference 1.) being current indicates the associated circuitry is OPERABLE.There are conditions when the Lo-Lo and Lo alarms of the RIL Monitor are limited below the RIL specified in the COLR. The RIL Monitor remains OPERABLE because the lead control rod bank still has the Lo and Lo-Lo alarms greater than or equal to the RIL.MILLSTONE  
Control rod positions and OPERABILITY of the               position indicators are required to be verified on a nemil hbasis of on-ce per 12 hou's with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
-UNIT 3 B 3/4 1-5 Amendment No. 60, LBDCR No. 06 MP3 0 14 J22.1,-2O06-POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)
The Digital Rod Position Indication (DRPI) System is defined as follows:
(2) APLND (for base load operation).
* Rod position indication as displayed on DRPI display panel (MB4), or
Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.Each of these is measurable but will normally only be determined periodically as .peei.ie in Specicationc 1.2.2 and .2.3. : This periodic surveillance is sufficient to ensure that the limits are maintained provided:
* Rod position indication as displayed by the Plant Process Computer System.
in accordance with the Surveillance d Frequency Control Program./a. Control rods in a single group move together with no individual rod inse ion differing by more than +/-12 steps, indicated, from the group demand position;b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.FNAH will be maintained within its limits provided Conditions  
With the above definition, LCO 3.1.3.2, "ACTION a." is not applicable with either DRPI display panel or the plant process computer points OPERABLE.
: a. through d. above are maintained.
The plant process computer may be utilized to satisfy DRPI System requirements which meets LCO 3.1.3.2, in requiring diversity for determining digital rod position indication.
The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNAH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.The RCS total flow rate and FNAH are specified in the CORE OPERATING LIMITS REPORT (COLR) to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow rate, that is based on 10% steam generator tube plugging, is retained in the Technical Specifications.
Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least @ncc cach 12 h, ur , except during the time when the rod position deviation monitor is inoperable, then o are the Demand Position Indication System and the DRPI System at least once each 4 hours.           the frequency specified in the Surveillance
MILLSTONE  
* Frequency Control Program                               I The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours.
-UNIT 3 B 3/4 2-3 Amendment No. -50, 60, 24-7, LBDCR N11?4.. 08 MP3 014 O9eteber 21, 2008-POWER DISTRIBUTION LIMITS in accordance with the Surveillance BASES IFrequency Control Program 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RC FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continu d)Margin is maintained between the safety analysis limit DNBR and t design limit DNBR. This margin is more than sufficient to offset the effect of rod bow and an other DNB penalties that may occur. The remaining margin is available for plant design flex* ility.When an FQ measurement is taken, an allowance for both experimmntal error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a 11 core map taken with the incore detector flux mapping system and a 3% allowance is appr nriate for manufacturing tolerance.
Trelmical Spcfcto SR 4.1.3.2.1 detcrminc2s each digital fod position indicator tob O-ERL*__E by v*.. ýin 1the Dtntiad Psition Indi.ation Sy-*tm and the DP"* System agr-,
The heat flux hot channel factor, FQ(Z), is measured periodically sing the incore detector system.These measurements are generally taken with the core at or near steady state conditions.
wlithin 1-2 steps at least ecncc eaeh 12 hours, emeept d~rig the time whe the td positiondeiaion mntris inperable, then   1 conipaIe theL Demlanld Position 1 Indicatiu Systini and the ORP system -At]ARA* non swch 4 hoursw.
Using the measured three dimensional power distributions, it is possible to derive FQM(Z), a computed value of FQ(Z). However, because this value represents a steady state condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations.
The Rod Deviation Monitor is generated ofnly &fra         the DRPI panel at MB4. Therefore, w~henrd flieii peilfb1 inl SURVeILLANCqet *EQUIREMENf~TS eveiy 4 frm D1uplicative paarps         ILLSTIONE - UNIT 3                         B 3/4 1-4                                 Amendment No. i0,
To account for these possible variations, the steady state limit of FQ(Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst case transient conditions.
 
The W(Z) and W(Z)BL, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion.
LDDC9R 65-MP3-025 M a* hr1 , 26,6-REACTIVITY CONTROL SYSTEMS the frequency specified in the Surveillance Frequency Control Program BASES MOVABLE CONTROL ASSEMBLIES (Continued)d Additional surveillance is required to ensure the plant ocess computer indications are in agreement with those displayed on the DRPI. This a ditional SURVEILLANCE REQUIREMENT is as follows:
Evaluation of the steady state FQ(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.
Each rod position indication as displ ed by the plant process computer shall be determined to be OPERABLE by rifying the rod position indication as displayed on the DRPI display panel agrees wi te rod position indication as displayed by the plant process computer at least once per 12 hours.
When RCS flow rate and FNAH are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation.
The rod position indication, as displayed by DRPI display panel (MB4), is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a refueling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position. Additionally, the plant process computer provides the same type and level of accuracy as the DRPI System (MB4). The plant process computer does not provide any alarm or rod position deviation monitoring as does DRPI display panel (MB4).
Measurement errors for RCS total flow rate and for FNAH have been taken into account in determination of the design DNBR value.The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators.
For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. In the event the plant is unable to verify the rod(s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours for this verification.
To perform the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated at least once per 18 months. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.Any fouling which might bias the RCS flow rate measurement can be detected by monitoring and trending various plant performance parameters.
For LCO 3.1.3.6 the control bank insertion limits are specified in the CORE OPERATING LIMITS REPORT (COLR). These insertion limits are the initial assumptions in safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions, assumptions of available SHUTDOWN MARGIN, and initial reactivity insertion rate.
If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.MILLSTONE  
The applicable I&C calibration procedure (Reference 1.) being current indicates the associated circuitry is OPERABLE.
-UNIT 3 B 3/4 2-4 Amendment No. 2, 60, 4-40, 24-7, L=BDGR No. 04 MP3-01 524, 2005 POWER DISTRIBUTION LIMITS lin accordance with the Surveillance Frequency Control Program BASESW cT HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW TE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
There are conditions when the Lo-Lo and Lo alarms of the RIL Monitor are limited below the RIL specified in the COLR. The RIL Monitor remains OPERABLE because the lead control rod bank still has the Lo and Lo-Lo alarms greater than or equal to the RIL.
The 12-h periodic surveillance of indicated RCS flo is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation defined in Specifications 3.2.3.1.3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
MILLSTONE - UNIT 3                           B 3/4 1-5                                 Amendment No. 60,
Radial power distribution measurements are made during STARTUP testing and periodically during POWER OPERATION.
 
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in FQ is depleted.
LBDCR No. 06 MP3 0 14 J22.1,-2O06-POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)
A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
(2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.
The two sets of four symmetric thimbles is a unique set of eight detector locations.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
Each of these is measurable but will normally only be determined periodically as .peei.ie in Specicationc 1.2.2 and   .2.3.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient.
:   This periodic surveillance is sufficient to ensure that the limits are maintained provided:                 in accordance with the Surveillance d Frequency Control Program./
The indicated Tavg values MILLSTONE  
: a. Control rods in a single group move together with no individual rod inse ion differing by more than +/-12 steps, indicated, from the group demand position;
-UNIT 3 B 3/4 2-5 Amendment No. 2-2, -50, 60, 2-4-, Ae~knowledgcJd by NRC lemer dated 08/245/0 LBDCR t4-NMP3--62 Matreh 25, 2004-POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued) and the indicated pressurizer pressure values are specified in the CORE OPERATING LIMITS REPORT. The calculated values of the DNB related parameters will be an average of the indicated values for the OPERABLE channels.The -4 2 hour periodic surveillance of these parameters through instrument readout* sufficient to ensure that the parameters are restored within their limits following load changes a other expected transient operation.
: b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;
Measurement uncertainties have been accounted for in determin the parameter limits.k lin accordance with the Surveillance Frequency Control ProgramI MILLSTONE  
: c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
-UNIT 3 B 3/4 2-6 Amendment No. 2, 60, -bV lNRC kiki by 08/25/ttda OI3.I.
: d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
LBD30cl No. UT-1vr.3-UM Feb~adry.
FNAH will be maintained within its limits provided Conditions a. through d. above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
4 20....3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated action and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNAH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The RCS total flow rate and FNAH are specified in the CORE OPERATING LIMITS REPORT (COLR) to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow rate, that is based on 10% steam generator tube plugging, is retained in the Technical Specifications.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
MILLSTONE - UNIT 3                         B 3/4 2-3                       Amendment No. -50,60, 24-7,
The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
 
The periodic surveillance tests performed atrh mini..mum frgucn.ieie.
LBDCR N11?4.. 08 MP3 014 O9eteber 21, 2008-POWER DISTRIBUTION LIMITS in accordance with the Surveillance BASES                                             IFrequency Control Program 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RC FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continu d)
are sufficient to demonstrate this capability.
Margin is maintained between the safety analysis limit DNBR and t design limit DNBR. This margin is more than sufficient to offset the effect of rod bow and an other DNB penalties that may occur. The remaining margin is available for plant design flex* ility.
The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set for each functional unit. The Allowable Values (Nominal Trip Setpoints  
When an FQ measurement is taken, an allowance for both experimmntal error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a 11core map taken with the incore detector flux mapping system and a 3% allowance is appr nriate for manufacturing tolerance.
+/- the calibration tolerance) are considered the Limiting Safety System Settings as identified in 1OCFR50.36 and have been selected to mitigate the consequences of accidents.
The heat flux hot channel factor, FQ(Z), is measured periodically sing the incore detector system.
A Setpoint is considered to be consistent with the nominal value when the measured "as left" Setpoint is within the administratively controlled  
These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive FQM(Z), a computed value of FQ(Z). However, because this value represents a steady state condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations.
(+) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint).
To account for these possible variations, the steady state limit of FQ(Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst case transient conditions. The W(Z) and W(Z)BL, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state FQ(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.
Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.
When RCS flow rate and FNAH are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation. Measurement errors for RCS total flow rate and for FNAH have been taken into account in determination of the design DNBR value.
The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. To perform the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated at least once per 18 months. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.
Any fouling which might bias the RCS flow rate measurement can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.
MILLSTONE - UNIT 3                         B 3/4 2-4                 Amendment No. 2, 60, 4-40, 24-7,
 
L=BDGR No. 04 MP3-01 5 Feznl*y 24, 2005 POWER DISTRIBUTION LIMITS BASESW                  lin accordance with the SurveillancecTFrequency Control Program HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW                         TE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
The 12-h     periodic surveillance of indicated RCS flo is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation defined in Specifications 3.2.3.1.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during POWER OPERATION.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in FQ is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations.
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. The indicated Tavg values MILLSTONE - UNIT 3                         B 3/4 2-5                   Amendment No. 2-2,-50,60, 2-4-,
Ae~knowledgcJd by NRC lemer dated 08/245/0
 
LBDCR     t4-NMP3--62 Matreh 25, 2004-POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued) and the indicated pressurizer pressure values are specified in the CORE OPERATING LIMITS REPORT. The calculated values of the DNB related parameters will be an average of the                   k indicated values for the OPERABLE channels.
The -42 hour periodic surveillance of these parameters through instrument readout* sufficient to ensure that the parameters are restored within their limits following load changes a other expected transient operation. Measurement uncertainties have been accounted for in determin   the parameter limits.
lin accordance with the Surveillance Frequency Control ProgramI MILLSTONE - UNIT 3                         B 3/4 2-6                     Amendment No. 2, 60, 2-t-*,
                                                          - lNRC        by bV       kiki       08/25/ttda OI3.I.
 
LBD30cl No. UT-1vr.3-UM Feb~adry. 4 20....
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated action and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed atrh mini..mum frgucn.ieie. are sufficient to demonstrate this capability.
The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set for each functional unit. The Allowable Values (Nominal Trip Setpoints +/- the calibration tolerance) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to mitigate the consequences of accidents. A Setpoint is considered to be consistent with the nominal value when the measured "as left" Setpoint is within the administratively controlled (+) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.
Measurement and Test Equipment accuracy is administratively controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991.
Measurement and Test Equipment accuracy is administratively controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991.
OPERABILITY determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation.
OPERABILITY determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation.
The Allowable Value specified in Table 3.3-4 defines the limit beyond which a channel is inoperable.
The Allowable Value specified in Table 3.3-4 defines the limit beyond which a channel is inoperable. If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be OPERABLE.
If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be OPERABLE.MILLSTONE  
MILLSTONE - UNIT 3                           B 3/4 3-1                               Amendment No. 4-59, A Am.... _   ._Ii440P 1i Ah !-IF-* 1 AR094
-UNIT 3 B 3/4 3-1 Amendment No. 4-59, A Am.... _ ._Ii440P 1i Ah 1 AR094 LBDCR 1 o-fMP3-603 Febr3uary 213, 2010 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
 
The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels.
LBDCR 1o-fMP3-603 Febr3uary 213, 2010 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties.
The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels. Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation.
Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes.
The above Bases does not apply to the Control Building Inlet Ventilation radiation monitors ESF Table (Item 7E). For these radiation monitors the allowable values are essentially nominal values.
Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics.
Due to the uncertainties involved in radiological parameters, the methodologies of WCAP- 10991 were not applied. Actual trip setpoints will be reestablished below the allowable value based on calibration accuracies and good practices.
Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation.
The OPERABILITY requirements for Table 3.3-3, Functional Units 7.a, "Control Building Isolation, Manual Actuation," and 7.e, "Control Building Isolation, Control Building Inlet Ventilation Radiation," are defined by table notation "*". These functional units are required to be OPERABLE at all times during plant operation in MODES 1,2, 3, and 4. These functional units are also required to be OPERABLE during movement of recently irradiated fuel assemblies, as specified by table notation
The above Bases does not apply to the Control Building Inlet Ventilation radiation monitors ESF Table (Item 7E). For these radiation monitors the allowable values are essentially nominal values.Due to the uncertainties involved in radiological parameters, the methodologies of WCAP- 10991 were not applied. Actual trip setpoints will be reestablished below the allowable value based on calibration accuracies and good practices.
"*". The Control Building Isolation Manual Actuation and Control Building Inlet Ventilation Radiation are required to be OPERABLE during movement of recently irradiated fuel assemblies (i.e.,
The OPERABILITY requirements for Table 3.3-3, Functional Units 7.a, "Control Building Isolation, Manual Actuation," and 7.e, "Control Building Isolation, Control Building Inlet Ventilation Radiation," are defined by table notation "*". These functional units are required to be OPERABLE at all times during plant operation in MODES 1, 2, 3, and 4. These functional units are also required to be OPERABLE during movement of recently irradiated fuel assemblies, as specified by table notation"*". The Control Building Isolation Manual Actuation and Control Building Inlet Ventilation Radiation are required to be OPERABLE during movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 350 hours*). Table notation "*" 7 of Table 4.3-2 has the same applicability.
fuel that has occupied part of a critical reactor core within the previous 350 hours*). Table notation "*"   7 of Table 4.3-2 has the same applicability.
The verification of response time at the sp, eified provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analysis.
The verification of response time at the sp, eified frequenci*e provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analysis. No credit is taken in the analysis for those channels with response times indicated as not applicable (i.e., N.A.).             "     --     nsert Required ACTION 4. of Table 3.3-1 is modified by a Note to indicate that normal plant control operations that individually add limited positive reactivity (e.g., temperature or boron fluctuations associated with RCS inventory management or temperature control) are not precluded by this ACTION provided they are accounted for in the calculated SDM. The proposed change permits operations introducing positive reactivity additions but prohibits the temperature change or overall boron concentration from decreasing below that required to maintain the specified SDM or required boron concentration.
No credit is taken in the analysis for those channels with response times indicated as not applicable (i.e., N.A.). " -- nsert Required ACTION 4. of Table 3.3-1 is modified by a Note to indicate that normal plant control operations that individually add limited positive reactivity (e.g., temperature or boron fluctuations associated with RCS inventory management or temperature control) are not precluded by this ACTION provided they are accounted for in the calculated SDM. The proposed change permits operations introducing positive reactivity additions but prohibits the temperature change or overall boron concentration from decreasing below that required to maintain the specified SDM or required boron concentration.
During fuel assembly cleaning evolutions that involve the handling or cleaning of two fuel assemblies coincidentally, recently irradiated fuel is fuel that has occupied part of a critical reactor core within the previous 525 hours.                                               /
During fuel assembly cleaning evolutions that involve the handling or cleaning of two fuel assemblies coincidentally, recently irradiated fuel is fuel that has occupied part of a critical reactor core within the previous 525 hours. /MILLSTONE  
MILLSTONE - UNIT 3                             B 3/4 3-2   Amendment No. 3,-9-1-, 4-59, 4-74, 4-8-4,2 19-,230O
-UNIT 3 B 3/4 3-2 Amendment No. 3,-9-1-, 4-59, 4-74, 4-8-4,2 19-,230O Fb.... 24, 2Z,0.INSTRUMENTATION Sin accordance with the Surveillance BASES IFrequency Control Program.-I 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATIO;ý/and ENGINEER-ED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT N (Continued)
 
For slave relays, or any auxiliary relays in ESFAS circuits thatae of the type Potter & Brumfield MDR series relays, the SLAVE RELAY TEST is performed-, an fley (at least @nee e,,veryff 18 rnon,) prov f~li tq the Ireiability assesirnJl Citalia p, eslnt 1 ,d in we7-.3.7, "R.liabiiy  
Fb.... 24, 2Z,0.
..... .. of uf... and D 1 mi 1 field MDR .... relays," and WCAF-i.;5o, "Extiiunsiu of sf Siuzv,,,lanc , Test inte val." The r .liability assessments prormzefad as part cfthe afcr-ementioncd WCAP9 are relay speeifie and apply ofnly tc r[k alld DIgillld MDl z 1 1o 1day .aL ttato ns 1 llally a.l liaIuII, t-lh May havO tc b@ replacEd pcri@Edially inf accodfrdanc  
INSTRUMENTATION                                                       Sin accordance with the Surveillance BASES                                                                 IFrequency Control Program.-I 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATIO;ý/and ENGINEER-ED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT                                           N (Continued)
%ith th@ glidancc@
For slave relays, or any auxiliary relays in ESFAS circuits thatae of the type Potter & Brumfield MDR series relays, the SLAVE RELAY TEST is performed-, an                                   fley (at least @nee e,,veryff 18 rnon,)   prov       f~li tq elay* *ie*t          the Ireiability assesirnJl Citalia p, eslnt 1 ,d in we7-.3.7,           "R.liabiiy
givenif in IXCAPR 13879 for.Sf 1R reletys.REACTOR TRIP BREAKER This trip function applies to the reactor trip breakers (RTBs) exclusive of individual trip mechanisms.
                            .....           ..       ofuf... and D1 mi1 field MDR .... relays," and WCAF-i.;5o, "Extiiunsiu ofsf                    ,,.*,ay Siuzv,,,lanc Test  , inte val."   The r     .liability assessments prormzefad as part cfthe afcr-ementioncd WCAP9 are relay speeifie and apply ofnly tc r[k alld DIgillld MDl                 1 1o 1day .           ttato     1 llally       z    liaIuII, ns t-lh r*lays aL a.l May havO tc b@ replacEd pcri@Edially inf accodfrdanc %ith th@ glidancc@ givenif in                 IXCAPR 13879 for.
The LCO requires two OPERABLE trains of trip breakers.
Sf 1R reletys.
A trip breaker train consists of all trip breakers associated with a single RTS logic train that are racked in, closed, and capable of supplying power to the control rod drive (CRD) system. Thus, the train may consist of the main breaker, bypass breaker, or main breaker and bypass breaker, depending upon the system configuration.
REACTOR TRIP BREAKER This trip function applies to the reactor trip breakers (RTBs) exclusive of individual trip mechanisms. The LCO requires two OPERABLE trains of trip breakers. A trip breaker train consists of all trip breakers associated with a single RTS logic train that are racked in, closed, and capable of supplying power to the control rod drive (CRD) system. Thus, the train may consist of the main breaker, bypass breaker, or main breaker and bypass breaker, depending upon the system configuration. Two OPERABLE trains ensure no single random failure can disable the RTS trip capability.
Two OPERABLE trains ensure no single random failure can disable the RTS trip capability.
These trip functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip functions must be OPERABLE when the RTBs or associated bypass breakers are closed, and the CRD system is capable of rod withdrawal.
These trip functions must be OPERABLE in MODE 1 or 2 when the reactor is critical.
BYPASSED CHANNEL* - Technical Specifications 3.3.1 and 3.3.2 often allow the bypassing of instrument channels in the case of an inoperable instrument or for surveillance testing.
In MODE 3, 4, or 5, these RTS trip functions must be OPERABLE when the RTBs or associated bypass breakers are closed, and the CRD system is capable of rod withdrawal.
A BYPASSED CHANNEL shall be a channel which is:
BYPASSED CHANNEL* -Technical Specifications 3.3.1 and 3.3.2 often allow the bypassing of instrument channels in the case of an inoperable instrument or for surveillance testing.A BYPASSED CHANNEL shall be a channel which is: Required to be in its accident or tripped condition, but is not presently in its accident or tripped condition using a method described below; or Prevented from tripping.MILLSTONE  
Required to be in its accident or tripped condition, but is not presently in its accident or tripped condition using a method described below; or Prevented from tripping.
-UNIT 3 B 3/4 3-2b Amendment No. 2-4-9, Ak -- .l.d.. d b, NRC ltter-- dated 025,'05 INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR BASES (continued)
MILLSTONE - UNIT 3                                     B 3/4 3-2b                                 Amendment No. 2-4-9, Ak--. l.d.. d b, NRC ltter--dated 025,'05
Required ACTION b. is modified by a Note which permits plant temperature changes provided the temperature change is accounted for in the calculated SDM. Introduction of temperature changes, including temperature increases when a positive MTC exists, must be evaluated to ensure they do not result in a loss of required SDM.2. All dilution flowpaths are isolated and placed under administrative control (locked closed). This action provides redundant protection and defense in depth (safety overlap)to the SMMs. In this configuration, a boron dilution event (BDE) cannot occur. This is the basis for not having to analyze for BDE in MODE 6. Since the BDE cannot occur with the dilution flow paths isolated, the SMMs are not required to be OPERABLE as the event cannot occur and OPERABLE SMMs provide no benefit.3. Increase the SHUTDOWN MARGIN surveillance frequency from ery"24--hotrr to every 12 hours. This action in combination with the above, provi defense in depth and overlap to the loss of the SMMs. i Surveillance Requirements the frequency specified in the SSurveillance Frequency Control Program The SMMs are subject to an AGOe;veT " 92 dayR to ensure each train of SMM is fully operational.
 
This test shalle de verification that the SMMs are set per the CORE OPERATING LIMITS ORT. c--Finsert-2-1
INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR BASES (continued)
[ANALOG CHANNEL OPERATIONAL TEST MILLSTONE  
Required ACTION b. is modified by a Note which permits plant temperature changes provided the temperature change is accounted for in the calculated SDM. Introduction of temperature changes, including temperature increases when a positive MTC exists, must be evaluated to ensure they do not result in a loss of required SDM.
-UNIT 3 B 3/4 3-9 Amendment No. 4-64, 2-30  
: 2.       All dilution flowpaths are isolated and placed under administrative control (locked closed). This action provides redundant protection and defense in depth (safety overlap) to the SMMs. In this configuration, a boron dilution event (BDE) cannot occur. This is the basis for not having to analyze for BDE in MODE 6. Since the BDE cannot occur with the dilution flow paths isolated, the SMMs are not required to be OPERABLE as the event cannot occur and OPERABLE SMMs provide no benefit.
¥ "ILT_ I i_,i.Jj.l._,il f1s .tl--dt , ii', ".J t Felbruar-" 21, 2005 3/4.4 REACTOR COOLANT SYSTEM BASES The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4.The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.The loop stop valves have motor operators.
: 3.       Increase the SHUTDOWN MARGIN surveillance frequency from ery"24--hotrr to every 12 hours. This action in combination with the above, provi defense in depth and overlap to the loss of the SMMs.                                   i Surveillance Requirements                             the frequency specified in the SSurveillance Frequency Control Program The SMMs are subject to an AGOe;veT         "92 dayR to ensure each train of SMM is fully operational. This test shalle de verification that the SMMs are set per the CORE OPERATING LIMITS               ORT. c--Finsert-2-1
If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated.
[ANALOG CHANNEL OPERATIONAL TEST MILLSTONE - UNIT 3                           B 3/4 3-9                       Amendment No. 4-64, 2-30
Although operating procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators.
 
The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task.Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion.
                                                                            ¥ T*T*.f*T*    "ILT_   J*,l    I KT*"*) a**l L"*
The time period provided in ACTION 3.4.1.5.b allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours and MODE 5 within 30 hours. The allowed ACTION times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.Surveillance Requirement 4.4.1.5 is performed at st ... 3e 31 das to ensure that the RCS loop stop valves are open, with power removed from the loop stop valve operators.
i_,i.Jj.l._,il f1s   .     tl--dt ii',,    t
The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every-l2-houm that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency  
                                                                                                                          ".J Felbruar-" 21, 2005 3/4.4 REACTOR COOLANT SYSTEM BASES The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4.
-ensures that the required flow is available ia bacd cb all nzrirg 4-lb Amendment No. 60, 40, 99, 44-5, 20, 2-2-4, A.ck 1 1 wikdgd by NRC lctte 1 dated 08,'25/05  
The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.
-96ý2O6-3/4.4 REACTOR COOLANT SYSTEM BASES (continued)
The loop stop valves have motor operators. If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated. Although operating procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators. The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task.
For the isolated loop being restored, the power to both loop stop valves has been restored Surveillance 4.4.1.6.2 indicates that the reactor shall be determined subcritical by at least the amount required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours of opening the cold leg or hot leg stop valve.The SHUTDOWN MARGIN requirement in Specification 3.1.1.1.2 is specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops filled. Specification 3.1.1.1.2 cannot be used to determine the required SHUTDOWN MARGIN for MODE 5 loops isolated condition.
Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The time period provided in ACTION 3.4.1.5.b allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours and MODE 5 within 30 hours. The allowed ACTION times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
Specification 3.1.1.2 requires the SHUTDOWN MARGIN to be greater than or equal to the limits specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops not filled provided CVCS is aligned to preclude boron dilution.
Surveillance Requirement 4.4.1.5 is performed at st .           .. 31 das to ensure that the 3e RCS loop stop valves are open, with power removed from the loop stop valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every-l2-houm that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency -             ensures that the required flow is available ia bacdcball            nzrirg 4-lb Amendment No. 60, 40, 99, 44-5, 4-9*, 20, 2-2-4, A.ck 1 1wikdgd by NRC lctte1 dated 08,'25/05
This specification is for loops not filled and therefore is applicable to an all loops isolated condition.
 
Specification 3.9.1.1 requires Keff of 0.95 or less, or a boron concentration of greater than or equal to the limit specified in the COLR in MODE 6.Specification 3.1.1.1.2 or 3.1.1.2 for MODE 5, both require boron concentration to be determined at I.Pas, .a2 ,. SR 4.1.1.1.2.1 .b.2 and 4.1.1.2.1.b.l satisfy the requirements of Speci 1 .1.1.1.2 and 3.1.1.2 respectfully.
                                                                                              -96ý2O6-3/4.4 REACTOR COOLANT SYSTEM BASES (continued)
Specification 3.9.1.1 for MODE 6 requires boron concentration to be d at m.ist , ta.h 72 hur-. S.R. 4.9.1.1.2 satisfy the requirements of Specification 3.9.1.1. [the frequency specified in the Surveillance Frequency Control Program Per Specifications 3.4.1.2, ACTION c.; 3.4.1.3, ACTION c.; 3.4.1.4.1, ACTION b.; and 3.4.1.4.2, ACTION b., suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1.2 is required to assure continued safe operation.
For the isolated loop being restored, the power to both loop stop valves has been restored Surveillance 4.4.1.6.2 indicates that the reactor shall be determined subcritical by at least the amount required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours of opening the cold leg or hot leg stop valve.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The SHUTDOWN MARGIN requirement in Specification 3.1.1.1.2 is specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops filled. Specification 3.1.1.1.2 cannot be used to determine the required SHUTDOWN MARGIN for MODE 5 loops isolated condition.
Specification 3.1.1.2 requires the SHUTDOWN MARGIN to be greater than or equal to the limits specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops not filled provided CVCS is aligned to preclude boron dilution. This specification is for loops not filled and therefore is applicable to an all loops isolated condition.
Specification 3.9.1.1 requires Keff of 0.95 or less, or a boron concentration of greater than or equal to the limit specified in the COLR in MODE 6.
Specification 3.1.1.1.2 or 3.1.1.2 for MODE 5, both require boron concentration to be determined at I.Pas, .     a2           ,. SR 5Rt2**
4.1.1.1.2.1 .b.2 and 4.1.1.2.1.b.l satisfy the requirements of Speci 1           .1.1.1.2 and 3.1.1.2 respectfully. Specification 3.9.1.1 for MODE 6 requires boron concentration to be             d at       m.ist
                                                              ,     ta.h 72 hur-. S.R. 4.9.1.1.2 satisfy the requirements of Specification 3.9.1.1.                             [the frequency specified in the Surveillance Frequency Control Program Per Specifications 3.4.1.2, ACTION c.; 3.4.1.3, ACTION c.; 3.4.1.4.1, ACTION b.; and 3.4.1.4.2, ACTION b., suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1.2 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.


==References:==
==References:==
: 1. Letter NEU-94-623, dated July 13, 1994; Mixing Evaluation for Boron Dilution Accident in Modes 4 and 5, Westinghouse HR-59782.2. Memo No. MP3-E-93-821, dated October 7, 1993.MILLSTONE  
: 1. Letter NEU-94-623, dated July 13, 1994; Mixing Evaluation for Boron Dilution Accident in Modes 4 and 5, Westinghouse HR-59782.
-UNIT 3 B 3/4 4-1If Amendment No. 2-4-7, 230 REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (continued)
: 2.     Memo No. MP3-E-93-821, dated October 7, 1993.
Insert 2 The t2"rrmr periodic surveillances quire that pressurizer level be maintained at programmed level within +/- 6% of full sc .The surveillance is performed by observing the indicated level. The 12 het ~itfterfvval hats ccri she by operotng practiee to be 3i~ffieien regelfrly ttsses level for nfy deyiattion:_arA to eznqsimrec that the. apprpriattc level existsi inth." During transitory conditions, i.e., power changes, the operators will maintain programmed level, and deviations greater than 6% will be corrected within 2 hours. Two hours has been selected for pressurizer level restoration after a transient to avoid an unnecessary downpower with pressurizer level outside the operating band. Normally, alarms are also available for early detection of abnormal level indications.
MILLSTONE - UNIT 3                           B 3/4 4-1If                           Amendment No. 2-4-7, 230
Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure.A minimum required available capacity of pressurizer heaters ensures that the RCS pressure can be maintained.
 
The capability to maintain and control system pressure is important for maintaining subcooled conditions in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of the reactor coolant. Unless adequate heater capacity is available, the hot high-pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system. Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to a loss of single-phase natural circulation and decreased capability to remove core decay heat.The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity of at least 175 kW. The heaters are capable of being powered from either the offsite power source or the emergency power supply. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation.
REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (continued)                           Insert 2 The t2"rrmr periodic surveillances quire that pressurizer level be maintained at programmed level within +/- 6% of full sc . The surveillance is performed by observing the indicated level. The 12 het ~itfterfvval hats ccri she by operotng practiee to be 3i~ffieien ttsses regelfrly ."      leveltransitory During    for nfy deyiattion:_arA to eznqsimrec that the.apprpriattc level existsi inth conditions, i.e., power changes, the operators will maintain programmed level, and deviations greater than 6% will be corrected within 2 hours. Two hours has been selected for pressurizer level restoration after a transient to avoid an unnecessary downpower with pressurizer level outside the operating band. Normally, alarms are also available for early detection of abnormal level indications.
By maintaining the pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. The requirement for two groups of pressurizer heaters, each having a capacity of 175 kW, is met by verifying the capacity of the pressurizer heater groups A and B. Since the pressurizer heater groups A and B are supplied from the emergency 480V electrical buses, there is reasonable assurance that these heaters can be energized during a loss of offsite power to maintain natural circulation at HOT STANDBY. Providing an emergency (Class 1 E) power source for the required pressurizer heaters meets the requirement of NUREG-0737, "A Clarification of TMI Action Plan Requirements," II.E.3.1, "Emergency Power Requirements for Pressurizer Heaters." If one required group of pressurizer heaters is inoperable, restoration is required within 72 hours. The Completion Time of 72 hours is reasonable considering that a demand caused by loss of offsite power would be unlikely in this time period. Pressure control may be maintained during this time using normal station powered heaters.MODE 3 The requirement for the pressurizer to be OPERABLE, with a level less than or equal to 89%, ensures that a steam bubble exists. The 89% level preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure control for MODE 3 and to ensure a bubble is present in the pressurizer.
Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure.
Initial pressurizer level is not significant for those events analyzed for MODE 3 in Chapter 15 of the FSAR.MILLSTONE  
A minimum required available capacity of pressurizer heaters ensures that the RCS pressure can be maintained. The capability to maintain and control system pressure is important for maintaining subcooled conditions in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of the reactor coolant. Unless adequate heater capacity is available, the hot high-pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system. Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to a loss of single-phase natural circulation and decreased capability to remove core decay heat.
-UNIT 3 B 3/4 4-2a Amendment No. 460,-249 LtiilcR fu. 04-MVIF3-Mi5 ftbiuamy 24, 205 REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (cont'd.)
The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity of at least 175 kW. The heaters are capable of being powered from either the offsite power source or the emergency power supply. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. The requirement for two groups of pressurizer heaters, each having a capacity of 175 kW, is met by verifying the capacity of the pressurizer heater groups A and B. Since the pressurizer heater groups A and B are supplied from the emergency 480V electrical buses, there is reasonable assurance that these heaters can be energized during a loss of offsite power to maintain natural circulation at HOT STANDBY. Providing an emergency (Class 1E) power source for the required pressurizer heaters meets the requirement of NUREG-0737, "A Clarification of TMI Action Plan Requirements," II.E.3.1, "Emergency Power Requirements for Pressurizer Heaters."
Insert 2 The 12 he periodic surveillance requires that during MODE 3 op ation, pressurizer level is maintained below the nominal upper limit to provide a minimu /ace for a steam bubble. The surveillance is performed by observing the indicated level. R.c 12 h-Ea- interval has bee showvn b, pr~n sppfiatu ieea;. to be 9"ffieie..t to reg..lal ases level. fer an~y devatin d to.ns... that a steam b.ub-ble cxis in t*h p......izr Alarms are also available for early detection of abnormal level indications.
If one required group of pressurizer heaters is inoperable, restoration is required within 72 hours. The Completion Time of 72 hours is reasonable considering that a demand caused by loss of offsite power would be unlikely in this time period. Pressure control may be maintained during this time using normal station powered heaters.
The basis for the pressurizer heater requirements is identical to MODES 1 and 2.3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
MODE 3 The requirement for the pressurizer to be OPERABLE, with a level less than or equal to 89%, ensures that a steam bubble exists. The 89% level preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure control for MODE 3 and to ensure a bubble is present in the pressurizer. Initial pressurizer level is not significant for those events analyzed for MODE 3 in Chapter 15 of the FSAR.
Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is met.ACTION statements a, b, and c distinguishes the inoperability of the power operated relief valves ," (PORV). Specifically, a PORV may be designated inoperable but it may be able to automatically and manually open and close and therefore, able to perform its function.
MILLSTONE - UNIT 3                             B 3/4 4-2a                         Amendment No. 460,-249
PORV inoperability may be due to seat leakage which does not prevent automatic or manual use and does not create the possibility for a small-break LOCA. For these reasons, the block valve may be closed but the action requires power to be maintained to the valve. This allows quick access to the PORV for pressure control. On the other hand if a PORV is inoperable and not capable of being automatically and manually cycled, it must be either restored or isolated by closing the associated block valve and removing power.Note: PORV position indication does not affect the ability of the PORV to perform any of its safety functions.
 
Therefore, the failure of PORV position indication does not cause the PORV to be inoperable.
LtiilcR fu. 04-MVIF3-Mi5 ftbiuamy 24, 205 REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (cont'd.)                                                             Insert 2 The 12 he periodic surveillance requires that during MODE 3 op ation, pressurizer level is maintained below the nominal upper limit to provide a minimu /ace for a steam bubble. The surveillance is performed by observing the indicated level. R.c 12-1-h-Ea- interval has bee showvn b, sppfiatu pr~n       ieea;. to be 9"ffieie..t to reg..lal ases level. fer an~y devatin d to
However, failed position indication of these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3.
.ns... that a steam b.ub-ble cxis in t*h p......izr Alarms are also available for early detection of abnormal level indications.
Automatic operation of the PORVs is created to allow more time for operators to terminate an Inadvertent ECCS Actuation at Power. The PORVs and associated piping have been demonstrated to be qualified for water relief. Operation of the PORVs will prevent water relief from the pressurizer safety valves for which qualification for water relief has not been demonstrated.
The basis for the pressurizer heater requirements is identical to MODES 1 and 2.
If the PORVs are capable of automatic operation but have been declared inoperable, closure of the PORV block valve is acceptable since the Emergency Operating Procedures provide guidance to assure that the PORVs would be available to mitigate the event.OPERABILITY and setpoint controls for the safety grade PORV opening logic are maintained in ,4 the Technical Requirements Manual.MILLSTONE  
3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is met.
-UNIT 3 B 3/4 4-2b Amendment No. -160, 4-6-Ackniowuedged by NRC tiutit daitd G8/2j/05
ACTION statements a, b, and c distinguishes the inoperability of the power operated relief valves     ,"
_,iiV.ei fTito. tity-Mi titi-.)i yi May " 2"006 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)
(PORV). Specifically, a PORV may be designated inoperable but it may be able to automatically and manually open and close and therefore, able to perform its function. PORV inoperability may be due to seat leakage which does not prevent automatic or manual use and does not create the possibility for a small-break LOCA. For these reasons, the block valve may be closed but the action requires power to be maintained to the valve. This allows quick access to the PORV for pressure control. On the other hand if a PORV is inoperable and not capable of being automatically and manually cycled, it must be either restored or isolated by closing the associated block valve and removing power.
An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in RCS LCO 3.4.6. 1, "Leakage Detection Systems." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.Thu 72 huui F-1 qUlmal -l~~iabik iiitet v a! to trend LEAKAG E and regiirzs th--impet!anc of early leakage detcltieU ii the prevntion of  
Note: PORV position indication does not affect the ability of the PORV to perform any of its safety functions. Therefore, the failure of PORV position indication does not cause the PORV to be inoperable. However, failed position indication of these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3.
.4.4.6.2.1.e nsert 2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated.
Automatic operation of the PORVs is created to allow more time for operators to terminate an Inadvertent ECCS Actuation at Power. The PORVs and associated piping have been demonstrated to be qualified for water relief. Operation of the PORVs will prevent water relief from the pressurizer safety valves for which qualification for water relief has not been demonstrated. If the PORVs are capable of automatic operation but have been declared inoperable, closure of the PORV block valve is acceptable since the Emergency Operating Procedures provide guidance to assure that the PORVs would be available to mitigate the event.
The 150 gallons per day limit is measured at room temperature as described in Reference  
OPERABILITY and setpoint controls for the safety grade PORV opening logic are maintained in           ,4 the Technical Requirements Manual.
: 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG If it is not practical to assign the LEAKAGE to an individual SQ all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the surveillance is not required to be performed until 12 hours after establishment of steady state operation.
MILLSTONE - UNIT 3                         B 3/4 4-2b                         Amendment No. -160,4                                                               Ackniowuedged by NRC tiutit daitd G8/2j/05
For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. nsert 27 Tlhe Sauxvvttlneu of 72 houuis is a izvauiiubk juL 1 v tn t 1 i 1 d iniuay to sereodtay LEAKAGE end meeg~izes the impCoAflc of early leakage detzztion in the preyentfiet of-acidenits-The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5).4.4.6.2.2 The Surveillance Requirements for RCS pressure isolation valves provide assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.MILLSTONE  
 
-UNIT 3 B 3/4 4-4g-Amettdment+ý.
_,iiV.ei fTito. tity-Mi     titi-.)iyi May       " 2"006 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)
r LBDCR Nm. m8-MF3-0mi 3 Marlch 1s, 208 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in RCS LCO 3.4.6. 1, "Leakage Detection Systems."
ACTIONS (Continued) e.If the required action and completion time of ACTION d. is not met, the reactor must be brought to HOT STANDBY (MODE 3) within 6 hours and COLD SHUTDOWN (MODE 5) within 36 hours. The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS 4.4.8.1 Ithe frequency specified in the Surveillance Frequency Control Program Surveillance Requirement 4.4.8.1 requires performing aamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at F-Very 7 day .This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken.This Surveillance Requirement provides an indication of any increase in the noble gas specific activity.Trending the results of this Surveillance Requirement allows proper remedial action to be taken before reaching the LCO limit under normal operating cond.thons.
Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.
ofmve between 2n7d6y--aft~ern a pow er- chane lo p5%rTbabwithi ea 1 ou erio i f e stbihdbcue h oievl Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sampe due to masking from radioisotopes with similar decay energies, such as F- 18 and 1- 134, it is acce ble to include the minimum detectable activity for Kr-85 in the Surveillance Requirement 4.4.8. alculation.
Thu 72 huui F-1 qUlmal     -             l~~iabik iiitet v a! to trend LEAKAG E and regiirzs       th--
If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-nt is not detected, it should be assumed to be present at the minimum detectable activity.
impet!anc     of early leakage detcltieU         ii the prevntion
Insert 2]A Note modifies the Surveillance Requirement to allow entry into and tl'on in MODE 4, MODE 3, and MODE 2 prior to performing the Surveillance R
                                                              *iLdentl*of         .
This allows the Surveillance Requirement to be performed in thoe O prior to entering MODE 1.4.4.8.2 .../This Surveillance Reqie efre oenueidn pcfc activity remains within the LCO limit during n operation.
4.4.6.2.1.e                                                                                           nsert 2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated.
and following fast power changes when iodine spiking is m ore apt to occur. .1,IN .1 4 day kfiAeqienicy is adeqtUate tv ft~c, C Seno hi. tilt iodiiie actliV t 1,3 .V l,....erin ;no ble g..ii... ar;, arivt is. m-esa-iave weryD -7 .&ys. The frequency of between 2 and 6 hours after a power change _> 15% RTP within a I hour period is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.MILLSTONE  
The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG If it is not practical to assign the LEAKAGE to an individual SQ all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
-UNIT 3 B 3/4 4-6b Amendment No. .
The Surveillance is modified by a Note which states that the surveillance is not required to be performed until 12 hours after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.                                                       nsert 27 Tlhe Sauxvvttlneu F*equenicy of 72 houuis is a izvauiiubk juL 1v tn t1i 1 d pu*        iniuay to sereodtay LEAKAGE end meeg~izes the impCoAflc of early leakage detzztion in the preyentfiet of
-acidenits- The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5).
4.4.6.2.2 The Surveillance Requirements for RCS pressure isolation valves provide assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.
MILLSTONE - UNIT 3                                   B 3/4 4-4g                               -Amettdment+ý.               r
 
LBDCR Nm. m8-MF3-0mi 3 Marlch 1s, 208 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
ACTIONS (Continued) e.
If the required action and completion time of ACTION d. is not met, the reactor must be brought to HOT STANDBY (MODE 3) within 6 hours and COLD SHUTDOWN (MODE 5) within 36 hours. The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS 4.4.8.1               Ithe frequency specified in the Surveillance Frequency Control Program Surveillance Requirement 4.4.8.1 requires performing aamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at                     F-Very 7 day . This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken.
This Surveillance Requirement provides an indication of any increase in the noble gas specific activity.
Trending the results of this Surveillance Requirement allows proper remedial action to be taken before reaching the LCO limit under normal operating cond.thons.                       ofmve between 2n7d6y--
Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sampedue to masking from radioisotopes with similar decay energies, such as F- 18 and 1-134, it is acce ble to include the minimum detectable activity for Kr-85 in the Surveillance Requirement 4.4.8. alculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-nt is not detected, it should be assumed to be present at the minimum detectable activity.                           Insert 2]
A Note modifies the Surveillance Requirement to allow entry into and                         tl'on in MODE 4, MODE aft~ern    3, and a pow chane MODE er-        2 prior to performing   the lo p5%rTbabwithi ea1 ou erioSurveillance f    ei stbihdbcue h allows R q*,ahment.      This     oievlthe Surveillance Requirement to be performed in thoe O                         prior to entering MODE 1.
4.4.8.2                                                                         .       ..                           /
This Surveillance Reqie                     efre         oenueidn pcfc activity remains within the LCO limit during n               operation. and following fast power changes when iodine spiking is more apt to occur. .1,IN .1     4 day kfiAeqienicy is adeqtUate tv ft~c,C *,idchSenohi. tilt iodiiie actliV t 1,3
                                                                                                                .V l,
.... ;noerin      ble g..ii...
arivt ar;,     is. m-esa-iave weryD -7.&ys. The frequency of between 2 and 6 hours after a power change _>15% RTP within a I hour period is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.
MILLSTONE - UNIT 3                                 B 3/4 4-6b                     Amendment No.                       .
 
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
The AOT in MODE 4 considers the facts that only one of the relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. The RCS must be depressurized and a vent must be established within the following 12 hours if the required relief valve is not restored to OPERABLE within the required AOT of 7 days.d.The consequences of operational events that will overpressure the RCS are more severe at lower temperatures (Ref. 8). Thus, with one of the two required relief valves inoperable in MODE 5 or in MODE 6 with the head on, the AOT to restore two valves to OPERABLE status is 24 hours.The AOT represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE relief valve to protect against overpressure events. The RCS must be depressurized and a vent must be established within the following 12 hours if the required relief valve is not restored to OPERABLE within the required AOT of 24 hours.e.The RCS must be depressurized and a vent must be established within 12 hours when both A'required Cold Overpressure Protection relief valves are inoperable.
The AOT in MODE 4 considers the facts that only one of the relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. The RCS must be depressurized and a vent must be established within the following 12 hours if the required relief valve is not restored to OPERABLE within the required AOT of 7 days.
The vent must be sized > 2.0 square inches to ensure that the flow capacity is greater than that required for the worst case cold overpressure transient reasonable during the applicable MODES.This action is needed to protect the RCPB from a low temperature overpressure event and a possible non-ductile failure of the reactor vessel.The time required to place the plant in this Condition is based on the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
d.
SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Performance of an ANALOG CHANNEL OPERATIONAL TEST is required within 31 days prior to entering a condition in which the PORV is required to be OPERABLE a 3--dys on each required PORV to verify and, as necessary, adjust its lift setpoint.
The consequences of operational events that will overpressure the RCS are more severe at lower temperatures (Ref. 8). Thus, with one of the two required relief valves inoperable in MODE 5 or in MODE 6 with the head on, the AOT to restore two valves to OPERABLE status is 24 hours.
Th ALOG CHANNEL OPERATIONAL TEST will verify the setpoint in accordanc ith the nominal values given in Figures 3.4-4a and 3.4-4b. PORV actuation could de ssurize the RCS;therefore, valve operation is not required.MILLSTONE  
The AOT represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE relief valve to protect against overpressure events. The RCS must be depressurized and a vent must be established within the following 12 hours if the required relief valve is not restored to OPERABLE within the required AOT of 24 hours.
-UNIT 3 B 3/4 4-24 Amendment No. 5, +9-lat the frequency specified in the Surveillance Frequency Control Program thereafter tlttct flu. fJ7-M?1P3-W juiue 19, 2007-REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (Performance
e.
]BRATION (on each required PORV ac tion channel is required 0-,- niont, to adjust the channel so that it responds and the lve opens within the required range and accuracy to a known input. .7 The PORV block valve must be verified open and ,'OPPS must be verified armed ty-72-lbn to provide a flow path and a cold overpressure prol ction actuation circuit for each required PORV to perform its function when required.
The RCS must be depressurized and a vent must be established within 12 hours when both               A' required Cold Overpressure Protection relief valves are inoperable.
The alve is remotely verified open in the main control room. This Surveillance is performed if cr dit is being taken for the PORV to satisfy the LCO.'Insert 2 The block valve is a remotely controlled, mot ope ted valve. The power to the valve operator is not required to be removed, and the ma 1 operat r is not required to be locked in the open position.
The vent must be sized > 2.0 square inches to ensure that the flow capacity is greater than that required for the worst case cold overpressure transient reasonable during the applicable MODES.
Thus, the block valve can be sed in the ent the PORV develops excessive leakage or does not close (sticks open) after ieving an ove essure transient.
This action is needed to protect the RCPB from a low temperature overpressure event and a possible non-ductile failure of the reactor vessel.
I,,1 I IIIUL I%'%iulA1%y a %,JJ3%J a , M III VIV I~ YV UI%'L 11 a~~i1 a CLI V%1 %,%ill Vi a v UAUY t It? ~lt t~tJ~lEAI II LlA. t'VI~ttJII~fJ~h, O~t'I tO VLAV jfJO ~l il~S~ttILtIl tit. VtI~t LIt' ~I 4.4.9.3.2 Each required RHR suction relief valve shall be demonstra d OPERABLE by verifying the RHR suction valves, 3RHS*MV8701A and 3RHS*M8701C, are pen when suction relief valve 3RHS*RV8708A is being used to meet the LCO and by ver ing the RHR suction valves, 3RHS*MV8702B and 3RHS*MV8702C, are open when suc ion relief valve 3RHS*RV8708B is being used to meet the LCO. Each required RIIR suction reli f valve shall also be demonstrated OPERABLE by testing it in accordance with 4.0.5. This Su illance is only required to be performed if the RHR suction relief valve is being used to me this LCO., -Periodi7caly The RHR suction valves are 4erified to be open every-12 -hourm. "fivei1mer-m-a e are *,y ew e or er ft rn " orm ve efintra 4 sue tis vol ve stan- " I] f" To~~~~1~~~~
The time required to place the plant in this Condition is based on the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
----zesreAtr int 0th eentrcl room that Yeritv the IU-IR 9tuteiett yalvc Demft z.The ASME Code for Operation and Maintenance of Nuclear Power Plants, (Reference 9), test per 4.0.5 verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.4" MILLSTONE
SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Performance of an ANALOG CHANNEL OPERATIONAL TEST is required within 31 days prior to entering a condition in which the PORV is required to be OPERABLE a enve*y 3--dys on each required PORV to verify and, as necessary, adjust its lift setpoint. Th       ALOG CHANNEL OPERATIONAL TEST will verify the setpoint in accordanc                 ith the nominal values given in Figures 3.4-4a and 3.4-4b. PORV actuation could de ssurize the RCS; therefore, valve operation is not required.
-UNIT 3 B 3/4 4-25 Amendment No. 4-1, 9, 2-06,
MILLSTONE - UNIT 3                         B 3/4 4-24                     Amendment No. 5, +9-lat the frequency specified in the Surveillance Frequency Control Program thereafter


JI. V I --viL Rine,, 19, 2007 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) 4.4.9.3.3 periodically The RCS vent of> 2.0 square inches is proven OPERABLE by verifying its open condition.
tlttct        flu. fJ7-M?1P3-W juiue 19, 2007-REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (
it. "_imuI c 2.ho a~ vent vatII YdVe that. cann~t be, locked ope./ll other. wont path. ýA remfoved Pressurizer satety valve tits this category.This passive vent arrangement must only be open to be OPERABLE.
Performance                                              ]BRATION (on each required PORV ac                                    tion channel is required                   0-,- niont, the required range and accuracy to a known to adjust the channel so that it responds and the input.                                     .7               lve opens within The PORV block valve must be verified open and                                     ,'OPPS must be verified armed                            ty-72-lbn to provide a flow path and a cold overpressure prol                                ction actuation circuit for each required PORV to perform its function when required. The                                     alve is remotely verified open in the main control room. This Surveillance is performed if cr                                dit is being taken for the PORV to satisfy the LCO.'Insert                                                                                        2 The block valve is a remotely controlled, mot ope ted valve. The power to the valve operator is not required to be removed, and the ma                              1operat r is not required to be locked in the open position. Thus, the block valve can be                             sed in the ent the PORV develops excessive leakage or does not close (sticks open) after ieving an ove essure transient.
This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LO. < Insert 2 4.4.9.3.4 and 4.4.9.3.5 To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all SIH pumps and all but one centrifugal charging pump are verified incapable of injecting into the RCS.The SIH pumps and charging pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control.Alternate methods of control may be employed using at least two independent means to prevent an injection into the RCS. This may be accomplished through any of the following methods: 1) placing the pump in pull to lock (PTL) and pulling its UC fuses, 2) placing the pump in pull to lock (PTL) and closingthe pump discharge valve(s) to the injection line, 3) closing the pump discharge valve(s) to the injection line and either removing power from the valve operator(s) or locking manual valves closed, and 4) closing the valve(s) from the injection source and either removing power from the valve operator(s) or locking manual valves closed.An SIH pump may be energized for testing or for filling the Accumulators provided it is incapable of injecting into the RCS.Tie Fiegueiiey of ,2 ii eu.,,io,., ,uiidiin, uL..,, X1 .,..ii .and alarms .........bl. to the REFERENCES n sert2 1. ASME Boiler and Pressure Vessel Code, Section XI, Appendix G "Fracture Toughness for Protection Against Failure," 1995 Edition.2. ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," dated February 26, 1999.3. Generic Letter 88-11 4. ASME, Boiler and Pressure Vessel Code, Section III 5. FSAR, Chapter 15 6. 1OCFR50, Section 50.46 7. 10CFR50, Appendix K 8. Generic Letter 90-06 9. ASME Code for Operation and Maintenance of Nuclear Power Plants MILLSTONE
I,,1 I IIIUL    I%'%iulA1%y           a %,JJ3%J            a ,     M    III VIV YV    I~ UI%'L      a~~i111    a   CLI V%1 %,%ill Vi    a v UAUY t It? ~lt t~tJ~lEAI II LlA. t'VI~ttJII~fJ~h,         O~t'I  tO  VLAV    jfJO  ~l  il~S~ttILtIl  ~I    tit. VtI~t     LIt' 4.4.9.3.2 Each required RHR suction relief valve shall be demonstra d OPERABLE by verifying the RHR suction valves, 3RHS*MV8701A and 3RHS*M8701C, are pen when suction relief valve 3RHS*RV8708A is being used to meet the LCO and by ver ing the RHR suction valves, 3RHS*MV8702B and 3RHS*MV8702C, are open when suc ion relief valve 3RHS*RV8708B is being used to meet the LCO. Each required RIIR suction reli f valve shall also be demonstrated OPERABLE by testing it in accordance with 4.0.5. This Su illance is only required to be performed if the RHR suction relief valve is being used to me this LCO.
-UNIT 3 B 3/4 4-26 Amendment No. +54, +W, LBDCKt flu. 06-MF3-02 Atigt~s 10, 2006 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continue periodically flush upon he xchanger return to service and procedural compliance is relied upon to ensure ehas is not present within the heat exchanger u-tubes.Surveil e Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed -di if the containment has been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure.
                                                    ,  -Periodi7caly The RHR suction valves are 4erified to be open every-12 -hourm. "fivei1mer-m-a e are *,y ew e or er ft rn " orm ve efintra 4 sue tis vol ve stan- "
< Insert 2 Surveillance Requirement 4.5.2.d.2 addresses periodic inspectiog; etcontainment sump to ensure that it is unrestricted and stays in proper operating condition.
~~~~1~~~~            ----
the me th rcgucny is based ont the need to pzrform +his Zor cilaoc e~ic tL -filtion, that !Fy~ duMu h'., 4n the need to- ha-vo -.ec to e hoc Ti. his f~requerny is 9 Iffi-eiont.
I] f"To zesreAtr int 0th eentrcl room that Yeritv the IU-IR 9tuteiett yalvc                                                  Demft z.
t"M -~c~ -or-admM d 4 t.aiid L~uIufI1IId by upc~nianig cpie~icnv.
The ASME Code for Operation and Maintenance of Nuclear Power Plants, (Reference 9), test per                                                              4" 4.0.5 verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.
The Emergency Core Cooling System (ECCS) has several piping cross connection points for use during the post-LOCA recirculation phase of operation.
MILLSTONE - UNIT 3                                                 B 3/4 4-25                                  Amendment No. 4-1, 9, 2-06,
These cross-connection points allow the Recirculation Spray System (RSS) to supply water from the containment sump to the safety injection and charging pumps. The RSS has the capability to supply both Train A and B safety injection pumps and both Train A and B charging pumps. Operator action is required to position valves to establish flow from the containment sump through the RSS subsystems to the safety injection and charging pumps since the valves are not automatically repositioned.
The quarterly stroke testing (Technical Specification 4.0.5) of the ECC/RSS recirculation flowpath valves discussed below will not result in subsystem inoperability (except due to other equipment manipulations to support valve testing) since these valves are manually aligned in accordance with the Emergency Operating Procedures (EOPs) to establish the recirculation flowpaths.
It is expected the valves will be returned to the normal pre-test position following termination of the surveillance testing in response to the accident.
Failure to restore any valve to the normal pre-test position will be indicated to the Control Room Operators when the ESF status panels are checked, as directed by the EOPs. The EOPs direct the Control Room Operators to check the ESF status panels early in the event to ensure proper equipment alignment.
Sufficient time before the recirculation flowpath is required is expected to be available for operator action to position any valves that have not been restored to the pretest position, including localmanual valve operation.
Even if the valves are not restored to the pre-test position, sufficient capability will remain to meet ECCS post-LOCA recirculation requirements.
As a result, stroke testing of the ECCS recirculation valves discussed below will not result in a loss of system independence or redundancy, and both ECCS subsystems will remain OPERABLE.When performing the quarterly stroke test of 3SIH*MV8923A, the control switch for safety injection pump 3SIH*PIA is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*P1A in ull-to-lock, the Train A ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a., applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*PIA defeated to allow stroke testing of 3SIH*MV8923A.
In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath.
Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated).
During the injection phase the redundant subsystem (Train B) is frilly functional, as is a significant portion of the Train A subsystem.
During the recirculation phase, the Train A RSS subsystem can supply water from the containment sump to the Train A MILLSTONE
-UNIT 3 B 3/4 5-2b Amendment No. 400, 44-7, 4-5-4 EMERGENCY CORE COOLING SYSTEMS BASES (Continued)
APPLICABILITY In MODES 1, 2, 3, and 4, a design basis accident (DBA) could lead to a fission product release to containment that leaks to the secondary containment boundary.
The large break LOCA, on which this system's design is based, is a full-power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and reactor coolant system pressure decrease.
With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.In MODES 5 and 6, the probability and consequence of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the SLCRS is not required to be OPERABLE.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 7-day Completion Time is based on the low probability of a DBA occurring during this period. The Completion Time is adequate to restore the volume of TSP to within the technical specification limits.If the TSP cannot be restored within limits within the 7-day Completion Time, the plant must be brought to a MODE in which the LCO does not apply. The specified Completign Times for reaching MODES 3 and 4 are those used throughout the technical specifications; they were chosen to allow reaching the specified conditions from full power in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.5.5 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 ...tu,,y ,,fonce per 24 ..,th- is t-qtli- d to determince visually that a mitnimum of 971 4 ubie feet is ccntoaincd in the T-SP Storage 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 > 7.0.haskets is only feasiblew during outages. Operating
.ep Ience -h-019- Aohc-Wn Hi Sfif Frequeney tteepkbl e~de to the inrni the. Itm of TH"P1 placed in the conitainmentii bailding.MILLSTONE
-UNIT 3 B 3/4 5-5 Amendment No. 44-S, 206 T "T' fT"1']~ XT- AnC A Xr'"q CONTAINMENT SYSTEMS BASES The design of the Containment RSS is sufficiently independent so that an active failure in the recirculation spray mode, cold leg recirculation mode, or hot leg recirculation mode of the ECCS has no effect on its ability to perform its engineered safety function.
In other words, the failure in one subsystem does not affect the capability of the other subsystem to perform its designated safety function of assuring adequate core cooling in the event of a design basis LOCA. As long as one subsystem is OPERABLE, with one pump capable of assuring core cooling and the other pump capable of removing heat from containment, the RSS system meets its design requirements.
The LCO 3.6.2.2. ACTION applies when any of the RSS pumps, heat exchangers, or associated components are declared inoperable.
All four RSS pumps are required to be OPERABLE to meet the requirements of this LCO 3.6.2.2. During the injection phase of a Loss Of Coolant Accident all four RSS pumps would inject into containment to perform their containment heat removal function.
The minimum requirement for the RSS to adequately perform this function is to have at least one subsystem available.
Meeting the requirements of LCO 3.6.2.2. ensures the minimum RSS requirements are satisfied.
Surveillance Requirement 4.6.2.2.c requires that at lAt once per 24 months, verification is made that on a CDA test signal, each RSS pump starts automatically after receipt of an RWST Low-Low level signal. Tlt;~ 24 1 1101 1 Al fiCU1C is baSCd M! th Me tO PV, ifbri this stnvelhmee ttL th+e c.,onid.itions that apply du"ing a plant ot.agc and for .npl.... d ,.. nSief. t if the......' .. ..was pe.-rformed with the raetr-r At p.w... Gp..t.ing
.
has that these eomponznfts pass the suffelte Me la 1 1 pc etfou lned at thl 1 24 1 1 1 0 1 1 d 1 fIcgueIuVy.
The.efe..
the freueeywa eeneluded to be. neeta from a relability stanpin~Xt.
Thisi change hasn ad.'erce im0pact on plant saf.. y. &#xfd;'-' 2 Surveillance Requirements 4.6.2.1 .d and 4.6.2.2.e require verification that each spray nozzle is unobstructed following maintenance that could cause nozzle blockage.
Normal plant operation and maintenance activities are not expected to trigger performance of these surveillance requirements.
However, activities, such as an inadvertent spray actuation that causes fluid flow through the nozzles, a major configuration change, or a loss of foreign material control when working within the respective system boundary may require surveillance performance.
An evaluation, based on the specific situation, will determine the appropriate test method (e.g., visual inspection, air or smoke flow test) to verify no nozzle obstruction.
MILLSTONE
-UNIT 3 B 3/4 6-2a Amendment No.
LBDCR Ne. 04 NIP3 015-FebUaty 24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive matenal to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for these isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. FSAR Table 6.2-65 lists all containment isolation valves. The addition or deletion of any containment isolation valve shall be made in accordance with Section 50.59 of 10CFR50 and approved by the committee(s) as described in the QAP Topical Report.For the purposes of meeting this LCO, the safety function of the containment isolation valves is to shut within the time limits assumed in the accident analyses.
As long as the valves can shut within the time limits assumed in the accident analyses, the valves are OPERABLE.
Where the valveposition indication does not affect the operation of the valve, the indication is not required for valve OPERABILITY under this LCO. Position indication for containment isolation 4-valves is covered by Technical Specification 6.8.4.e., Accident Monitoring Instrumentation.
Failed position indication on these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3.
Maintaining the valves OPERABLE, when position indication fails, facilitates troubleshooting and correction of the failure, allowing the indication to be restored "as soon as practicable." With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated.
The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and deactivated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration.
If the containment isolation valve on a closed system becomes inoperable, the remaining barrier is a closed system since a closed system is an acceptable alternative to an automatic valve.However, actions must still be taken to meet Technical Specification ACTION 3.6.3.d and the valve, not normally considered as a containment isolation valve, and closest to the containment wall should be put into the closed position.
No leak testing of the alternate valve is necessary to satisfy the ACTION statement.
Placing the manual valve in the closed position sufficiently deactivates the penetration for Technical Specification compliance.
Closed system isolation valves applicable to Technical Specification ACTION 3.6.3.d are included in FSAR Table 6.2-65, and are the isolation valves for those penetrations credited as General Design Criteria 57. The specified time (i.e., 72 hours) of Technical Specification
/ACTION 3.6.3.d is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3 and 4. In the event the affected penetration is , isolated in accordance with 3.6.3.d, the affected penetration flow path must be verified to be isolated on a periodic basis, (Surveillance Requirement 4.6.1.1 .a). This is necessary to assure leak tightness of containment and that containment penetrations requiring isolation following an accidentare isolated.'], d,_3,yy,?:yy _ 33'gy, " ,5,,U MILLSTONE
-UNIT 3 B 3/4 6-3 Amendment No. 2-8, 63&#xfd;, 442-, 24-6, Autiuwtudged by NRC tuttei dated ft/25165 Insert 2 LBDCR o5-. .3-25 Mlch 7, 200(7.CONTAINMENT SYSTEMS BASES 3/4.6.6.1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM (Continued)
Surveillance Requirements jInsert 2 -a Cumulative operation of the SLCRS with heaters operating for at least 10 continuous hours'1-day puriud is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.The 31 day fequency was dc;e0cped in consideration of the ktewn reliabilit, of fan metefs and cnr1.This test is performed one STAGGERD T6 TBSI pet 31-days.b. c. e. and f These surveillances verify that the required SLCRS filter testing is performed in accordance with Regulatory Guide 1.52, Revision 2. ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters.
The surveillances include testing HEPA filter performance, charcoal adsorber efficiency, system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
The 720 hours of operation requirement originates from Regulatory Guide 1.52, Revision 2, March 1978, Table 2, Note "c", which states that "Testing should be performed (1) initially, (2) at least once per 18 months thereafter for systems maintained in a standby status or after 720 hours of system operations, and (3) following painting, fire, or chemical release in any ventilation zone communicating with the system." This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits, as well as providing trend data. The 720 hour figure is an arbitrary number which is equivalent to a 30 day period. This criteria is directed to filter systems that are normally in operation and also provide emergency air cleaning functions in the event of a Design Basis Accident.
The applicable filter units are not normally in operation and the sample canisters are typically removed due to the 18 month criteria.d Fnsdi /L r -2 The_ utomatic strtup ensures that each SLCRS train responds pro erly. .heznc pe- 21 mzntith fi bl thl I.urIv a --lLI EI~ ti.od lll t i app 1 J 1 .LL. .-1 1 lti y ldw ll a plant ouItage, and the, potential for. an unplanned-trnisien ifle wh u~ilncvas perforrnd wi;th the reactor It power The surveillance verifies that the SLCRS starts on a SIS test signal. It also includes the automatic functions to isolate the other ventilation systems that are not part of the safety-related postaccident operating configuration and to start up and to align the ventilation systems that flow through the secondary containment to the accident condition.
MILLSTONE  
-UNIT 3 B 3/4 6-6 Amendment No. 8-7, 4-2-3, 4-84,-206,  


24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.6.2 SECONDARY CONTAINMENT (continued)
JLUJL)JL'*.i  JI. V I -- viL J-U*JJ7 Rine,, 19, 2007 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) 4.4.9.3.3                                                                          periodically The RCS vent of> 2.0 square inches is proven OPERABLE by verifying its open condition.
In MODES 5 and 6, the probability and consequences of a DBA are low due to the RCS temperature and pressure limitation in these MODES. Therefore, Secondary Containment is not required in MODES 5 and 6.ACTIONS In the event Secondary Containment OPERABILITY is not maintained, Secondary Containment OPERABILITY must be restored within 24 hours. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a DBA occurring during this time period. Therefore, it is considered that there exists no loss of safety function while in the ACTION Statement.
it.      "_imuI    c    2.ho    f*gbi a~ vent vatII YdVe that. cann~t  be, locked    ope./ll other. rag*_'ve wont path.&#xfd;A remfoved Pressurizer satety valve tits this category.
Inoperability of the Secondary Containment does not make the SLCRS fans and filters inoperable.
This passive vent arrangement must only be open to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LO. <                        Insert 2 4.4.9.3.4 and 4.4.9.3.5 To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all SIH pumps and all but one centrifugal charging pump are verified incapable of injecting into the RCS.
Therefore, while in this ACTION Statement solely due to inoperability of the Secondary Containment, the conditions and required ACTIONS associated with Specification 3.6.6.1 (i.e., Supplementary Leak Collection and Release System) are not required to be entered.If the Secondary Containment OPERABILITY cannot be restored to OPERABLE status within the required completion time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within the following 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full-power conditions in an orderly manner and without challenging plant systems.Surveillance Requirements 4.6.6.2.1 Maintaining Secondary Containment OPERABILITY requires maintaining each door in each access opening in a closed position except when the access opening is being used for normal entry and exit. The normal time allowed for passage of equipment and personnel through each access opening at a time is defined as no more than 5 minutes. The access opening shall not be blocked open. During this time, it is not considered necessary to enter the ACTION statement.
The SIH pumps and charging pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control.
A 5-minute time is considered acceptable since the access opening can be quickly closed without special provisions and the probability of occurrence of a DBA concurrent with equipment and/or personnel transit time of 5 minutes is low.T 3 1 dJay frequenc~y for this~ suvifmc is bae on e uJgmiet and ii--e~~~i~~ in view o~qac , f the other indi;at;i 1 s of ne~~plX stettw thett aI.e etv ailable. to the perator.MII LSTONE -UNIT 3 B 3/4 6-8 Amendment No. 8-7, 4-26, ILnsert 2] gedd by NRC kitua datd OW25ffiO .U 110. Uq-Mvt'-:9 i it Nvme 10, 2005 PLANT SYSTEMS BASES AUXILIARY FEEDWATER SYSTEM (Continued)
Alternate methods of control may be employed using at least two independent means to prevent an injection into the RCS. This may be accomplished through any of the following methods:
If all three AFW pumps are inoperable in MODE 1, 2, or 3, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with non safety related equipment.
: 1) placing the pump in pull to lock (PTL) and pulling its UC fuses, 2) placing the pump in pull to lock (PTL) and closingthe pump discharge valve(s) to the injection line, 3) closing the pump discharge valve(s) to the injection line and either removing power from the valve operator(s) or locking manual valves closed, and 4) closing the valve(s) from the injection source and either removing power from the valve operator(s) or locking manual valves closed.
In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW pump to OPERABLE status. Required ACTION e. is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW pump is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.
An SIH pump may be energized for testing or for filling the Accumulators provided it is incapable of injecting into the RCS.
SR 4.7.1.2. la. verifies the correct alignment for manual, power operated, and automatic valves in the auxiliary feedwater water and steam supply flow paths to provide assurance that the proper flow paths exist for auxiliary feedwater operation.
Tie Fiegueiiey of ,2 .uu.* ii                  ,uiidiin, uL..,,                  .,..ii and. alarms X1 eu.,,io,.,            .........      bl. to the REFERENCES                                                                                                    n      sert2
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.
: 1.      ASME Boiler and Pressure Vessel Code, Section XI, Appendix G "Fracture Toughness for Protection Against Failure," 1995 Edition.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulations; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.  
: 2.      ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," dated February 26, 1999.
'T ,, 31t d 4 y u I i, a,,. on engineerin jUdgmenAt, ig G91n9i8tent With th@ procedura cz bgvming ,t-v oper~atiet, an onruret correct -Valve positietts.
: 3.      Generic Letter 88-11
Isr The SR is modified by a Note that states one or more auxiliary feedwater pumps may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the auxiliary feedwater mode of operation, provided it is not otherwise inoperable.
: 4.      ASME, Boiler and Pressure Vessel Code, Section III
This exception to pump OPERABILITY allows the pump(s) and associated valves to be out of their normal standby alignment and temporarily incapable of automatic initiation without declaring the pump(s)inoperable.
: 5.      FSAR, Chapter 15
Since auxiliary feedwater may be used during STARTUP, SHUTDOWN, HOT STANDBY operations, and HOT SHUTDOWN operations for steam generator level control, and these manual operations are an accepted function of the auxiliary feedwater system, OPERABILITY (i.e., the intended safety function) continues to be maintained.
: 6.     10CFR50, Section 50.46
MILLSTONE
: 7.     10CFR50, Appendix K
-UNIT 3 B 3/4 7-2c I
: 8.     Generic Letter 90-06
~1 LBDCR 037-WMP3-633 ne- 25, 200-7-PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section.Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature.
: 9.      ASME Code for Operation and Maintenance of Nuclear Power Plants MILLSTONE - UNIT 3                                B 3/4 4-26                                   Amendment No. +54,+W,
Ti=24 hour frqtnyi based n op..at~in e.Ap.AimiceA tintekd to heniuing Uf flic Paidin Iii ...o d- udthe MvODES. This surveillance requirement verifies that the avera water temperature of the UHS is less than or equal to 75&deg;F. Insert 2 Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70'F during the applicable MODES. The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70'F. The frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. Additionally, the system provides temperature control for the control room envelope (CRE) during normal and post-accident operations.
 
The control room emergency ventilation system is comprised of the CRE emergency air filtration system and a temperature control system.The control room emergency air filtration system consists of two redundant systems that recirculate and filter the air in the CRE and a CRE boundary that limits the inleakage of unfiltered air. Each control room emergency air filtration system consists of a moisture separator, electric 4 heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system.The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions.
LBDCKt flu. 06-MF3-02 Atigt~s 10, 2006 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continue                    periodically flush upon he xchanger return to service and procedural compliance is relied upon to ensureehas      is not present within the heat exchanger u-tubes.
This area encompasses the control room, and other non-critical areas including adjacent support offices, MILLSTONE
Surveil
-UNIT 3 B 3/4 7-10 Amendment No. 149,4-36,444, 244, LBDCR G7-IIMP3-G33 PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
* e Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed    -di                if the containment has been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure. <                                                                          Insert 2 Surveillance Requirement 4.5.2.d.2 addresses periodic inspectiog; etcontainment sump to ensure that it is unrestricted and stays in proper operating condition. the 24* me th rcgucny is based ont the need to pzrform +his        cilaoc      ZortLe~ic -filtion, that !Fy~ duMu  h'.,            4n the need  to-ha-vo        to -.ec hoc  e   Ti.
ACTIONS (Continued) their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.Immediate action(s), in accordance with the LCO ACTION Statements, means that the required action should be pursued without delay and in a controlled manner.During movement of recently irradiated fuel assemblies
aiid i* L~uIufI1IId by upc~nianig cpie~icnv. his f~requerny is 9 Iffi-eiont. t"M      -~c~
: d. With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the emergency mode or suspend the movement of fuel. Initiating and L /maintaining operation of the OPERABLE train in the emergency mode ensures: (i) OPERABILITY of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected.e. With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'd' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving the movement of recently irradiated fuel assemblies.
                                                                                        -or-admM  d 4 t.
This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position.SURVEILLANCE REQUIREMENTS 4.7.7.a Insert 2 The CRE environment should be chec eriodically to ens that the CRE temperature control system is functioning properly.
The Emergency Core Cooling System (ECCS) has several piping cross connection points for use during the post-LOCA recirculation phase of operation. These cross-connection points allow the Recirculation Spray System (RSS) to supply water from the containment sump to the safety injection and charging pumps. The RSS has the capability to supply both Train A and B safety injection pumps and both Train A and B charging pumps. Operator action is required to position valves to establish flow from the containment sump through the RSS subsystems to the safety injection and charging pumps since the valves are not automatically repositioned. The quarterly stroke testing (Technical Specification 4.0.5) of the ECC/RSS recirculation flowpath valves discussed below will not result in subsystem inoperability (except due to other equipment manipulations to support valve testing) since these valves are manually aligned in accordance with the Emergency Operating Procedures (EOPs) to establish the recirculation flowpaths. It is expected the valves will be returned to the normal pre-test position following termination of the surveillance testing in response to the accident. Failure to restore any valve to the normal pre-test position will be indicated to the Control Room Operators when the ESF status panels are checked, as directed by the EOPs. The EOPs direct the Control Room Operators to check the ESF status panels early in the event to ensure proper equipment alignment. Sufficient time before the recirculation flowpath is required is expected to be available for operator action to position any valves that have not been restored to the pretest position, including localmanual valve operation. Even if the valves are not restored to the pre-test position, sufficient capability will remain to meet ECCS post-LOCA recirculation requirements. As a result, stroke testing of the ECCS recirculation valves discussed below will not result in a loss of system independence or redundancy, and both ECCS subsystems will remain OPERABLE.
I, .4 E .ten., n' f i V1 , ,, , , att 95 &deg;F at ie ast ,n i , per 12 houi s is su iffi eie t. It is not necessary to c le the C R E ventilation chillers.
When performing the quarterly stroke test of 3SIH*MV8923A, the control switch for safety injection pump 3SIH*PIA is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*P1A in ull-to-lock, the Train A ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a., applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*PIA defeated to allow stroke testing of 3SIH*MV8923A. In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated). During the injection phase the redundant subsystem (Train B) is frilly functional, as is a significant portion of the Train A subsystem. During the recirculation phase, the Train A RSS subsystem can supply water from the containment sump to the Train A MILLSTONE - UNIT 3                            B 3/4 5-2b                        Amendment No. 400, 44-7, 4-5-4
The CRE is manned during operations covered by the techni specifications.
 
Typically, temperature aberrations will be readily apparent.'o 4.7.7.b Standby systems should be checked periodically to ensure that they function properl .As-th env 'ir amont and NORM.l eo erating eenditiea f en this system ar. .t. .too severe. festitt the t -_in3 MILLSTONE
EMERGENCY CORE COOLING SYSTEMS BASES (Continued)
-UNIT 3 B 3/4 7-13b Amendment No. -t LBDCR .5-.P3.025 MArch 7--'4, 00&PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
APPLICABILITY In MODES 1, 2, 3, and 4, a design basis accident (DBA) could lead to a fission product release to containment that leaks to the secondary containment boundary. The large break LOCA, on which this system's design is based, is a full-power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and reactor coolant system pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.
SURVEILLANCE REQUIREMENTS (Continued) once ever; 31 days on a STAGGERED TEST BASIS ...... an adequate check of s-system. This surveillance requirement verifies a system flow rate of 1,120 cfm +/- 20%.Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized.
In MODES 5 and 6, the probability and consequence of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the SLCRS is not required to be OPERABLE.
These operations are sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters due to the humidity in the ambient air.4.7.7.c las specified in the Surveillance Frequency Control Program 'The performance of the control room emergency filtration systems should be chec ced periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, 4inimum flow rate, and the physical properties of the activated charcoal.
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 frequency is at4ee*.efiee pe,; 4moits or following painting, fire, or chemical release in any ventilation zone k communicating with the system.ANSI N510-1980 will be used as a procedural guide for surveillance testing.Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
The 7-day Completion Time is based on the low probability of a DBA occurring during this period. The Completion Time is adequate to restore the volume of TSP to within the technical specification limits.
4.7.7.c. 1 This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 1,120 cfm +/- 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in the regulatory guide.4.7.7.c.2 This surveillance requires that a representative carbon sample be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters.
If the TSP cannot be restored within limits within the 7-day Completion Time, the plant must be brought to a MODE in which the LCO does not apply. The specified Completign Times for reaching MODES 3 and 4 are those used throughout the technical specifications; they were chosen to allow reaching the specified conditions from full power in an orderly manner and without challenging plant systems.
The laboratory analysis is required to be performed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.MILLSTONE  
SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.5.5 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 . tu,,y    .. ,,fonce per    .. 24    ,th- is t-qtli- d to determince visually that a mitnimum of 9714 ubie feet is ccntoaincd in the T-SP Storage Baskets.
-UNIT 3 B 3/4 7-14 Amendment No. 4-36, 4-84, 206, LBDCRf 07&#xfd;-MP3-633 June 25, 200, PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
This requirement ensures that there is an adequate volume of TSP to adjust the pH of the post LOCA sump solution to a value > 7.0.
haskets is only feasiblew during outages. Operating    ep . Ience -h-019-Aohc-Wn Hi Sfif Frequeney tteepkbl e~de to the inrni          the. Itm of TH"P1 placed in the conitainmentii bailding.
MILLSTONE - UNIT 3                            B 3/4 5-5                        Amendment No. 44-S, 206
 
T "T' fT"1']~ XT- AnC A Xr'"q AA*{*'
CONTAINMENT SYSTEMS BASES The design of the Containment RSS is sufficiently independent so that an active failure in the recirculation spray mode, cold leg recirculation mode, or hot leg recirculation mode of the ECCS has no effect on its ability to perform its engineered safety function. In other words, the failure in one subsystem does not affect the capability of the other subsystem to perform its designated safety function of assuring adequate core cooling in the event of a design basis LOCA. As long as one subsystem is OPERABLE, with one pump capable of assuring core cooling and the other pump capable of removing heat from containment, the RSS system meets its design requirements.
The LCO 3.6.2.2. ACTION applies when any of the RSS pumps, heat exchangers, or associated components are declared inoperable. All four RSS pumps are required to be OPERABLE to meet the requirements of this LCO 3.6.2.2. During the injection phase of a Loss Of Coolant Accident all four RSS pumps would inject into containment to perform their containment heat removal function. The minimum requirement for the RSS to adequately perform this function is to have at least one subsystem available. Meeting the requirements of LCO 3.6.2.2. ensures the minimum RSS requirements are satisfied.
Surveillance Requirement 4.6.2.2.c requires that at lAt once per 24 months, verification is made that on a CDA test signal, each RSS pump starts automatically after receipt of an RWST Low-Low level signal. Tlt;~ 24 111011Al fiCU1C is baSCd M!th Me tO PV,ifbri this stnvelhmee ttL th+e c.,onid.itions that apply du"ing a plant ot.agc and p*...ti.l for  .npl....,.. d nSief. t if the was  pe.-rformed with the raetr-r At p.w... Gp..t.ing  .     *Vcience*has shon** that these eomponznfts pass the suffelte    la 11 Me  lnedpcetfou          1 24 111011d1 fIcgueIuVy. The.efe..
at thl                                        the freueeywa eeneluded to be. neeta                from a relability stanpin~Xt. Thisi change hasn ad.'erce im0pact on plant saf..y.                  &#xfd;'-'2 Surveillance Requirements 4.6.2.1 .d and 4.6.2.2.e require verification that each spray nozzle is unobstructed following maintenance that could cause nozzle blockage. Normal plant operation and maintenance activities are not expected to trigger performance of these surveillance requirements. However, activities, such as an inadvertent spray actuation that causes fluid flow through the nozzles, a major configuration change, or a loss of foreign material control when working within the respective system boundary may require surveillance performance. An evaluation, based on the specific situation, will determine the appropriate test method (e.g., visual inspection, air or smoke flow test) to verify no nozzle obstruction.
MILLSTONE - UNIT 3                                B 3/4 6-2a                                        Amendment No.
 
LBDCR Ne. 04 NIP3 015-FebUaty 24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive matenal to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for these isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. FSAR Table 6.2-65 lists all containment isolation valves. The addition or deletion of any containment isolation valve shall be made in accordance with Section 50.59 of 10CFR50 and approved by the committee(s) as described in the QAP Topical Report.
For the purposes of meeting this LCO, the safety function of the containment isolation valves is to shut within the time limits assumed in the accident analyses. As long as the valves can shut within the time limits assumed in the accident analyses, the valves are OPERABLE. Where the valveposition indication does not affect the operation of the valve, the indication is not required for valve OPERABILITY under this LCO. Position indication for containment isolation              4-valves is covered by Technical Specification 6.8.4.e., Accident Monitoring Instrumentation.
Failed position indication on these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3. Maintaining the valves OPERABLE, when position indication fails, facilitates troubleshooting and correction of the failure, allowing the indication to be restored "as soon as practicable."
With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and deactivated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration.
If the containment isolation valve on a closed system becomes inoperable, the remaining barrier is a closed system since a closed system is an acceptable alternative to an automatic valve.
However, actions must still be taken to meet Technical Specification ACTION 3.6.3.d and the valve, not normally considered as a containment isolation valve, and closest to the containment wall should be put into the closed position. No leak testing of the alternate valve is necessary to satisfy the ACTION statement. Placing the manual valve in the closed position sufficiently deactivates the penetration for Technical Specification compliance.
Closed system isolation valves applicable to Technical Specification ACTION 3.6.3.d are included in FSAR Table 6.2-65, and are the isolation valves for those penetrations credited as General Design Criteria 57. The specified time (i.e., 72 hours) of Technical Specification              /
ACTION 3.6.3.d is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3 and 4. In the event the affected penetration is                ,
isolated in accordance with 3.6.3.d, the affected penetration flow path must be verified to be isolated on a periodic basis, (Surveillance Requirement 4.6.1.1 .a). This is necessary to assure leak tightness accidentareof containment isolated.'], and  that containment d,_3,yy,?:yy          penetrations requiring L**,""            _ 33'gy,isolation
                                                                            "      following
                                                                                    ,5,,U an MILLSTONE - UNIT 3                            B 3/4 6-3                Amendment No. 2-8, 63&#xfd;, 442-, 24-6, Autiuwtudged by NRC tuttei dated ft/25165 Insert 2
 
LBDCR o5-. 3-25  .
Mlch 7, 200(7.
CONTAINMENT SYSTEMS BASES 3/4.6.6.1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM (Continued)
Surveillance Requirements                                                                    jInsert 2 -
a Cumulative operation of the SLCRS with heaters operating for at least 10 continuous hours' 1-day puriud is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
The 31 day fequency was dc;e0cped in consideration of the ktewn reliabilit, of fan metefs and cnr1.This test is performed one STAGGERD T6 TBSI                      pet 31-days.
: b. c. e. and f These surveillances verify that the required SLCRS filter testing is performed in accordance with Regulatory Guide 1.52, Revision 2. ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The surveillances include testing HEPA filter performance, charcoal adsorber efficiency, system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.
Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
The 720 hours of operation requirement originates from Regulatory Guide 1.52, Revision 2, March 1978, Table 2, Note "c", which states that "Testing should be performed (1) initially, (2) at least once per 18 months thereafter for systems maintained in a standby status or after 720 hours of system operations, and (3) following painting, fire, or chemical release in any ventilation zone communicating with the system."
This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits, as well as providing trend data. The 720 hour figure is an arbitrary number which is equivalent to a 30 day period. This criteria is directed to filter systems that are normally in operation and also provide emergency air cleaning functions in the event of a Design Basis Accident. The applicable filter units are not normally in operation and the sample canisters are typically removed due to the 18 month criteria.
d                                                                              /L Fnsdir -2 The_utomatic strtup ensures that each SLCRS train responds pro erly. .heznc pe- 21 mzntith fi            bl I.urIv            )*    llltthl i            JLL.
                                                                  . EI~
a 1 1- -lLI  - .11    ldw ti.od th*tulf  y app lti  ll a plant ouItage, and the,potential for.an unplanned-trnisien ifle u~ilncvas        wh perforrnd wi;th the reactor It power The surveillance verifies that the SLCRS starts on a SIS test signal. It also includes the automatic functions to isolate the other ventilation systems that are not part of the safety-related postaccident operating configuration and to start up and to align the ventilation systems that flow through the secondary containment to the accident condition.
MILLSTONE - UNIT 3                          B 3/4 6-6                Amendment No. 8-7, 4-2-3, 4-84,-206,
 
FL*tI-yIJ  24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.6.2 SECONDARY CONTAINMENT (continued)
In MODES 5 and 6, the probability and consequences of a DBA are low due to the RCS temperature and pressure limitation in these MODES. Therefore, Secondary Containment is not required in MODES 5 and 6.
ACTIONS In the event Secondary Containment OPERABILITY is not maintained, Secondary Containment OPERABILITY must be restored within 24 hours. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a DBA occurring during this time period. Therefore, it is considered that there exists no loss of safety function while in the ACTION Statement.
Inoperability of the Secondary Containment does not make the SLCRS fans and filters inoperable. Therefore, while in this ACTION Statement solely due to inoperability of the Secondary Containment, the conditions and required ACTIONS associated with Specification 3.6.6.1 (i.e., Supplementary Leak Collection and Release System) are not required to be entered.
If the Secondary Containment OPERABILITY cannot be restored to OPERABLE status within the required completion time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within the following 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full-power conditions in an orderly manner and without challenging plant systems.
Surveillance Requirements 4.6.6.2.1 Maintaining Secondary Containment OPERABILITY requires maintaining each door in each access opening in a closed position except when the access opening is being used for normal entry and exit. The normal time allowed for passage of equipment and personnel through each access opening at a time is defined as no more than 5 minutes. The access opening shall not be blocked open. During this time, it is not considered necessary to enter the ACTION statement. A 5-minute time is considered acceptable since the access opening can be quickly closed without special provisions and the probability of occurrence of a DBA concurrent with equipment and/or personnel transit time of 5 minutes is low.
T 3 1 dJay frequenc~y for this~    suvifmc is bae on e                    uJgmiet    and ii-
    -e~~~i~~          in, view o~qacf the other indi;at;i 1 s of ne~~plX      stettw thett aI.e etv ailable. to the perator.
MII LSTONE - UNIT 3                                B 3/4 6-8                          Amendment No. 8-7, 4-26, ILnsert 2]          -FTtI*~*(                              A-,*iuw gedd by NRC kitua datd OW25ffi  O    .U
 
urL)*t 110. Uq-Mvt'-:9    i it Nvme 10, 2005 PLANT SYSTEMS BASES AUXILIARY FEEDWATER SYSTEM (Continued)
If all three AFW pumps are inoperable in MODE 1, 2, or 3, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with non safety related equipment. In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW pump to OPERABLE status. Required ACTION e. is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW pump is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.
SR 4.7.1.2. la. verifies the correct alignment for manual, power operated, and automatic valves in the auxiliary feedwater water and steam supply flow paths to provide assurance that the proper flow paths exist for auxiliary feedwater operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulations; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. 'T 31t ,, d4 y f*iutCiiy u    I i, a,,. on engineerin jUdgmenAt, ig G91n9i8tent With th@ procedura cz        bgvming        ,t-v  oper~atiet, an onruret correct -Valve positietts.                                  Isr The SR is modified by a Note that states one or more auxiliary feedwater pumps may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the auxiliary feedwater mode of operation, provided it is not otherwise inoperable. This exception to pump OPERABILITY allows the pump(s) and associated valves to be out of their normal standby alignment and temporarily incapable of automatic initiation without declaring the pump(s) inoperable. Since auxiliary feedwater may be used during STARTUP, SHUTDOWN, HOT STANDBY operations, and HOT SHUTDOWN operations for steam generator level control, and these manual operations are an accepted function of the auxiliary feedwater system, OPERABILITY (i.e., the intended safety function) continues to be maintained.
MILLSTONE - UNIT 3                            B 3/4 7-2c                                                    I
 
                                                                                                              ~
LBDCR 037-WMP3-633 ne-25, 200 PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section.
Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature. Ti=
24 hour frqtnyi          based n op..at~in e.Ap.AimiceA tintekd to heniuing Uf  flic Paidin Iii    . .. o d-      udthe applI~*tib  MvODES. This surveillance requirement verifies that the avera water temperature of the UHS is less than or equal to 75&deg;F.                          Insert 2 Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70'F during the applicable MODES. The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70'F. The frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. Additionally, the system provides temperature control for the control room envelope (CRE) during normal and post-accident operations.
The control room emergency ventilation system is comprised of the CRE emergency air filtration system and a temperature control system.
The control room emergency air filtration system consists of two redundant systems that recirculate and filter the air in the CRE and a CRE boundary that limits the inleakage of unfiltered air. Each control room emergency air filtration system consists of a moisture separator, electric            4 heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system.
The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and other non-critical areas including adjacent support offices, MILLSTONE - UNIT 3                          B 3/4 7-10              Amendment No. 149,4-36,444, 244,
 
LBDCR G7-IIMP3-G33 PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
ACTIONS (Continued) their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.
Immediate action(s), in accordance with the LCO ACTION Statements, means that the required action should be pursued without delay and in a controlled manner.
During movement of recently irradiated fuel assemblies
: d.      With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the emergency mode or suspend the movement of fuel. Initiating and                    L /
maintaining operation of the OPERABLE train in the emergency mode ensures:
(i) OPERABILITY of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected.
: e.      With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'd' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving the movement of recently irradiated fuel assemblies. This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position.
SURVEILLANCE REQUIREMENTS 4.7.7.a                                                              Insert 2 The CRE environment should be chec                  eriodically to ens        that the CRE temperature control system is functioning properly.            I, .              4 E. ten.,                n'f i ,V1 ,, , ,
95 at&deg;Fie        ast i,nper 12              iffi t. It is not necessary to c le the C R E ventilation
                              , houi s is su eie Typically, temperature aberrations will be readilycovered chillers. The    CRE    is manned  during operations      apparent.
                                                                                'o by the techni      specifications. att 4.7.7.b Standby systems should be checked periodically to ensure that they function properl . As-th env'ir amont and NORM.l eo erating eenditiea f en this system ar. .t. too      . severe. festitt the t  -_in3 MILLSTONE - UNIT 3                              B 3/4 7-13b                                    Amendment No.    -t
 
LBDCR .5-.P3.025 MArch 7--'4,00&
PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
SURVEILLANCE REQUIREMENTS (Continued) once ever; 31 days on a STAGGERED TEST BASIS ......               an adequate check of s
-system. This surveillance requirement verifies a system flow rate of 1,120 cfm +/- 20%.
Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized. These operations are sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters due to the humidity in the ambient air.
4.7.7.c       specified in the Surveillance Frequency Control Program las                                                                  '
The performance of the control room emergency filtration systems should be chec ced periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, 4inimum flow rate, and the physical properties of the activated charcoal. The frequency is at4ee*.efiee pe,; 4moits or following painting, fire, or chemical release in any ventilation zone               k communicating with the system.
ANSI N510-1980 will be used as a procedural guide for surveillance testing.
Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
4.7.7.c. 1 This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 1,120 cfm +/- 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in the regulatory guide.
4.7.7.c.2 This surveillance requires that a representative carbon sample be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters. The laboratory analysis is required to be performed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.
MILLSTONE - UNIT 3                     B 3/4 7-14                   Amendment No. 4-36, 4-84, 206,
 
LBDCRf 07&#xfd;-MP3-633 June 25, 200, PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)
SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm +/- 20%, during system operation when testing in accordance with ANSI N510-1980.
SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm +/- 20%, during system operation when testing in accordance with ANSI N510-1980.
4.7.7.d After 720 hours of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters.
4.7.7.d After 720 hours of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters.
The laboratory analysis is required to be performed within 31 days after removal of the sample.ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.The maximum surveillance interval is 900 hours, per Surveillance Requirement 4.0.2. The 720 hours of operation requirement originates from Nuclear Regulatory Guide 1.52, Table 2, Note C. This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits as well as providing trending data.4.7.7.e. I This surveillance verifies that the pressure drop across the combined HEPA filters and charcoal adsorbers banks at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm +/- 20%. ThC fieqUMIY iS at !a 01,,, M Z&#xfd; ,,,__. __s.4.7.7.e.2 Deleted. 2 4.7.7.e.3 This surveillance verifies that the heaters can dissipate 1 kW at 480V when tested in accordance with ANSI N510-1980.
The laboratory analysis is required to be performed within 31 days after removal of the sample.
Thc .........
ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.
is a. feas un, c p 24 s. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.MILLSTONE  
The maximum surveillance interval is 900 hours, per Surveillance Requirement 4.0.2. The 720 hours of operation requirement originates from Nuclear Regulatory Guide 1.52, Table 2, Note C. This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits as well as providing trending data.
-UNIT 3 B 3/4 7-15 Amendment No. 4-36,4-8, 4-84,2403,2-06, Februar, 20, 2002 PLANT SYSTEMS BASES 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM Insert2 The 0 BILITY of the Auxiliary Bui ng Filter System, and associated filters and fans, ensures that r *oactive materials leaki rom the equipment within the charging pump, I1 component cooling wa ump and hea changer areas following a LOCA are filtered prior to reaching the environment.
4.7.7.e. I This surveillance verifies that the pressure drop across the combined HEPA filters and charcoal adsorbers banks at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm +/- 20%. ThC fieqUMIY iS at !a 01,,,         M Z&#xfd; ,,,__.
perati f the system with the heaters operating for at least 10 continuous hoursi R I I id*. p is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. e operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses.
__s.
ANSI N510-1980 will be used as a procedural guide for surveillance testing. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters.
4.7.7.e.2 Deleted.                                                                 2 4.7.7.e.3 This surveillance verifies that the heaters can dissipate       1 kW at 480V when tested in accordance with ANSI N510-1980. Thc ......... is a. feas un,c p 24 in*u          s. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.
The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System is required to be available to support the Auxiliary Building Filter System and the Supplementary Leak Collection and Release System (SLCRS). The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System consists of two redundant trains, each capable of providing 100% of the required flow. Each train has a two position, "Off' and "Auto," remote control switch. With the remote control switches for each train in the "Auto" position, the system is capable of automatically transferring operation to the redundant train in the event of a low flow condition in the operating train. The associated fans do not receive any safety related automatic start signals (e.g. Safety Injection Signal).Placing the remote control switch for a Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train in the "Off" position to start the redundant train or to perform post maintenance testing to verify availability of the redundant train will not affect the availability of that train, provided appropriate administrative controls have been established to ensure the remote control switch is immediately returned to the "Auto" position after the completion of the specified activities or in response to plant conditions.
MILLSTONE - UNIT 3                       B 3/4 7-15     Amendment No. 4-36,4-8, 4-84,2403,2-06,
These administrative controls include the use of an approved procedure and a designated individual at the control switch for the respective Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train who can rapidly respond to instructions from procedures, or control room personnel, based on plant conditions.
 
MILLSTONE  
Februar, 20, 2002 PLANT SYSTEMS BASES Insert2 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM The 0         BILITY of the Auxiliary Bui ng Filter System, and associated filters and fans, ensures that r *oactive materials leaki       rom the equipment within the charging pump,         I1 component cooling wa         ump and hea changer areas following a LOCA are filtered prior to reaching the environment. perati           f the system with the heaters operating for at least 10 continuous hoursi RI I id*. p           is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. e operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.
-UNIT 3 B 3/4 7-23 Amendment No. 8-7, 4-1-9, 4-36, 4-84, 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION Technical Specification 3.8.1.1 .b. 1 requires each of the diesel generator day tanks contain a minimum volume of 278 gallons. Technical Specification 3.8.1.2.b.1 requires a minimum volume of 278 gallons be contained in the required diesel generator day tank. This capacity ensures that a minimum usable volume of 189 gallons is available.
The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System is required to be available to support the Auxiliary Building Filter System and the Supplementary Leak Collection and Release System (SLCRS). The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System consists of two redundant trains, each capable of providing 100% of the required flow. Each train has a two position, "Off' and "Auto," remote control switch. With the remote control switches for each train in the "Auto" position, the system is capable of automatically transferring operation to the redundant train in the event of a low flow condition in the operating train. The associated fans do not receive any safety related automatic start signals (e.g. Safety Injection Signal).
This volume permits operation of the diesel generators for approximately 27 minutes with the diesel generators loaded to the 2,000 hour rating of 5335 kw. Each diesel generator has two independent fuel oil transfer pumps. The shutoff level of each fuel oil transfer pump provides for approximately 60 minutes of diesel generator operation at the 2000 hour rating. The pumps start at day tank levels to ensure the minimum level is maintained.
Placing the remote control switch for a Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train in the "Off" position to start the redundant train or to perform post maintenance testing to verify availability of the redundant train will not affect the availability of that train, provided appropriate administrative controls have been established to ensure the remote control switch is immediately returned to the "Auto" position after the completion of the specified activities or in response to plant conditions. These administrative controls include the use of an approved procedure and a designated individual at the control switch for the respective Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train who can rapidly respond to instructions from procedures, or control room personnel, based on plant conditions.
The loss of the two redundant pumps would cause day tank level to drop below the minimum value.Technical Specification 3.8. 1.1 .b.2 requires a minimum volume of 32,760 gallons be contained in each of the diesel generator's fuel storage systems. Technical Specification 3.8.1.2.b.2 requires a minimum volume of 32,760 gallons be contained in the required diesel generator's fuel storage system. This capacity ensures that a minimum usable volume (29,180 gallons) is available to permit operation of each of the diesel generators for approximately three days with the diesel generators loaded to the 2,000 hour rating of 5335 kW. The ability to cross-tie the diesel generator fuel oil supply tanks ensures that one diesel generator may operate up to approximately six days. Additional fuel oil can be supplied to the site within twenty-four hours after contacting a fuel oil supplier.Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation.
MILLSTONE - UNIT 3                           B 3/4 7-23           Amendment No. 8-7, 4-1-9, 4-36, 4-84,
Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum SDM or refueling boron concentration.
 
This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION Technical Specification 3.8.1.1 .b. 1 requires each of the diesel generator day tanks contain a minimum volume of 278 gallons. Technical Specification 3.8.1.2.b.1 requires a minimum volume of 278 gallons be contained in the required diesel generator day tank. This capacity ensures that a minimum usable volume of 189 gallons is available. This volume permits operation of the diesel generators for approximately 27 minutes with the diesel generators loaded to the 2,000 hour rating of 5335 kw. Each diesel generator has two independent fuel oil transfer pumps. The shutoff level of each fuel oil transfer pump provides for approximately 60 minutes of diesel generator operation at the 2000 hour rating. The pumps start at day tank levels to ensure the minimum level is maintained. The loss of the two redundant pumps would cause day tank level to drop below the minimum value.
Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.Suspension of these activities does not preclude completion of actions to establish a safe conservative condition.
Technical Specification 3.8. 1.1 .b.2 requires a minimum volume of 32,760 gallons be contained in each of the diesel generator's fuel storage systems. Technical Specification 3.8.1.2.b.2 requires a minimum volume of 32,760 gallons be contained in the required diesel generator's fuel storage system. This capacity ensures that a minimum usable volume (29,180 gallons) is available to permit operation of each of the diesel generators for approximately three days with the diesel generators loaded to the 2,000 hour rating of 5335 kW. The ability to cross-tie the diesel generator fuel oil supply tanks ensures that one diesel generator may operate up to approximately six days. Additional fuel oil can be supplied to the site within twenty-four hours after contacting a fuel oil supplier.
These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power source and distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the unit safety systems.Surveillance Requirements 4.8.1.1.2.a.6,(mc-th!-A and ".1.1,2.b.2"' (onee mer.' f-8&deg;+/-:&#xfd;-'...
Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.
and 4.8.1.1.2.j R1 ... me th ....The Surveillances 4.8.1.1.2.a.6 and 4.8.1.1.2.b.2 verify that the diesel generators are capable of synch-ronizing with the offsite electrical system and loaded to greater than or equal to continuous rating of the machine. A minimum time of 60 minutes is required to stabilize engine temperatures, while MILLSTONE  
Suspension of these activities does not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power source and distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the unit safety systems.
-UNIT 3 B 3/4 8-1 c Amendment No. 9-7, 4-42, 4-3-7, 4-94, 2-0-0, "23 LBDCRf 08-MvP3 -02 1 SeptemberL 1 9, 2008 3/4.8 ELECTRICAL POWER SYSTEMS BASES minimizing the time that the diesel generator is connected to the offsite source. Surveillance Requirement 4.8.1.1.2.j requires demonstration onco per- 18 mnths that the diesel generator can start and run continuously at full load capability for an interval of not less than 24 hours, > 2 hours of which are at a load equivalent to 110% of the continuous duty rating and the remainder of the time at a load equivalent to the continuous duty rating of the diesel generator.
Surveillance Requirements 4.8.1.1.2.a.6,(mc-th!-A and ".1.1,2.b.2"'(onee mer.'f-8&deg;+/-:&#xfd;-'...     and 4.8.1.1.2.j R1 me... th ....
The load band is provided to avoid routine overloading of the diesel generator.
The Surveillances 4.8.1.1.2.a.6 and 4.8.1.1.2.b.2 verify that the diesel generators are capable of synch-ronizing with the offsite electrical system and loaded to greater than or equal to continuous rating of the machine. A minimum time of 60 minutes is required to stabilize engine temperatures, while MILLSTONE - UNIT 3                           B 3/4 8-1 c     Amendment No. 9-7, 4-42, 4-3-7, 4-94, 2-0-0, "23
Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain diesel generator OPERABILITY.
 
The load band specified accounts for instrumentation inaccuracies, operational control capabilities, and human factor characteristics.
LBDCRf 08-MvP3 -02 1 SeptemberL 1 9, 2008 3/4.8 ELECTRICAL POWER SYSTEMS BASES minimizing the time that the diesel generator is connected to the offsite source. Surveillance Requirement 4.8.1.1.2.j requires demonstration onco per- 18 mnths that the diesel generator can start and run continuously at full load capability for an interval of not less than 24 hours, > 2 hours of which are at a load equivalent to 110% of the continuous duty rating and the remainder of the time at a load equivalent to the continuous duty rating of the diesel generator. The load band is provided to avoid routine overloading of the diesel generator. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain diesel generator OPERABILITY. The load band specified accounts for instrumentation inaccuracies, operational control capabilities, and human factor characteristics. The note (*)
The note (*)acknowledges that a momentary transient outside the load range shall not invalidate the test.Surveillance Requirements 4.8.1.1.2.a.5jh~jy 4 ......  
acknowledges that a momentary transient outside the load range shall not invalidate the test.
(,O,,,, per ..... Dy s, 4.8.1.1.2.gP.4.b fi8 fthff 4.8.1.1.2.g.5 hgnM.Ts)ad481.1.2.g-.6.b , ......Several diesel generator surveillance requirements specify that the Lemergency diesel generator, are started from a standby condition.
Surveillance Requirements 4.8.1.1.2.a.5jh~jy...... *,48.1.1.2.b.1 4                    (,O,,,, per..... Dy s, 4.8.1.1.2.gP.4.b fi8 fthff       4.8.1.1.2.g.5     hgnM.Ts)ad481.1.2.g-.6.b                   , ......
Standby conditions for a diesel generator means the diese engine coolant and lubricating oil are being circulated and temperatures are maintained within design ranges. Design ranges for standby temperatures are greater than or equal to the low temperature alarm setpoints and less than or equal to the standby "keep-warm" he Insert 2 temperatures for each respective sub-system.
Several diesel generator surveillance requirements specify that the Lemergency diesel generator, are started from a standby condition. Standby conditions for a diesel generator means the diese engine coolant and lubricating oil are being circulated and temperatures are maintained within design ranges. Design ranges for standby temperatures are greater than or equal to the low temperature alarm setpoints and less than or equal to the standby "keep-warm" he                                   Insert 2 temperatures for each respective sub-system.
Surveillance Requirement 4.8.1.1.2.i (4 1o8--Mte The)/The existing "standby condition" stipulation contained in specification 4.8.1.1.2.a.5 is su erseded when performing the hot restart demonstration required by 4.8.1.1.2.j.
Surveillance Requirement 4.8.1.1.2.i (41o8--Mte     The)/
The existing "standby condition" stipulation contained in specification 4.8.1.1.2.a.5 is su erseded when performing the hot restart demonstration required by 4.8.1.1.2.j.
Any time the OPERABILITY of a diesel generator has been affected by repair, mai enance, or replacement activity, or by modification that could affect its interdependency, post aintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
Any time the OPERABILITY of a diesel generator has been affected by repair, mai enance, or replacement activity, or by modification that could affect its interdependency, post aintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.
MILLSTONE  
MILLSTONE - UNIT 3                           B 3/4 8-1d               Amendment No. 9-7, 2, 1-+3,     1-94, 2-1-0,
-UNIT 3 B 3/4 8-1d Amendment No. 9-7, 2, 1-+3, 1-94, 2-1-0, LDDCR 6? -MP3 033 Jiu, 25, 2007.ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued)
 
The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1975 & 1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Sections 5 and 6 of IEEE Std 450-1980 replaced Sections 4 and 5 of IEEE Std 450-1975, otherwise the balance of IEEE Std 450-1975 applies.,"C-&#xfd;lnsert Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity.Table 4.8-2a specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity.
LDDCR 6?-MP3 033 Jiu, 25, 2007.
The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2a is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.If the required power sources or distribution systems are not OPERABLE in MODES 5 and 6, operations involving CORE ALTERATIONS, positive reactivity changes, movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the MILLSTONE  
ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued)
-UNIT 3 B 3/4 8-2 Amendment No.
The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1975 & 1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Sections 5 and 6 of IEEE Std 450-1980 replaced Sections 4 and 5 of IEEE Std 450-1975, otherwise the balance of IEEE Std 450-1975 applies.,"C-&#xfd;lnsert Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity.
Table 4.8-2a specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2a is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.
If the required power sources or distribution systems are not OPERABLE in MODES 5 and 6, operations involving CORE ALTERATIONS, positive reactivity changes, movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the MILLSTONE - UNIT 3                           B 3/4 8-2                                     Amendment No.
 
3/4.9 REFUELING OPERATIONS BASES 3/4.9.8.1 HIGH WATER LEVEL (continued)
3/4.9 REFUELING OPERATIONS BASES 3/4.9.8.1 HIGH WATER LEVEL (continued)
ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operations, except as permitted in the Note to the LCO.If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations.
ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operations, except as permitted in the Note to the LCO.
Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation.
If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum refueling boron concentration.
If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.
This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition.
If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition.
If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 and the refueling water level > 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.
If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements.
If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.
With the unit in MODE 6 and the refueling water level > 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.
The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.
If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Surveillance Requirement This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant.
Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.Surveillance Requirement This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant.The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core .... T fqt..n.y of 12 h....is suffieiefft, eensider-ing the fiew, tcmpzratz, pump eentrzl, and alarm indietions availabk1 to MILSON -prae UnI 3h Bete 3/4r fo4 Amendmentg No. ]4M, syst, 23 MILLSTONE  
The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core ....T fqt..n.y of 12 h....
-UNIT 3 B 3/4 9-4 Amendment No. +07,24-9,2 3/4.9 REFUELING OPERATIONS BASES The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.Surveillance Requirement This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant.The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The FiPquic~y of 12 hluus is eon the Howl c ,clir, t, e trol. and ttlam i tions avail ble to the 1 tm, I .I -T- -.Al A 111VII LIN, 13YaL%,111 ill L11%, rRWLITB B R M Tonom. tEHD MILLSTONE  
is suffieiefft, eensider-ing the fiew, tcmpzratz, pump eentrzl, and alarm indietions availabk1 to MILSON               UnI 3h
-UNIT 3 B 3/4 9-7 Amendment No. 409, 24-9,-2-3-}}
                  -prae             fo4    Bete 3/4r         syst,      AmendmentgNo. ]4M,       23 MILLSTONE - UNIT 3                         B 3/4 9-4                 Amendment No. +07,24-9,2 3/4.9 REFUELING OPERATIONS BASES The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.
Surveillance Requirement This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant.
The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The FiPquic~y of 12 hluus is sufficiu*n, eon e*iJngi the Howlt,c *. ,clir,,Itxtl*me trol. and ttlam i nJ*i tions avail ble to the1    a*ee tm, I . I   -T-     -
                                        . Al BR M rRWLITB tEHD A 111VII             LIN, 13YaL%,111 ill L11%,
Tonom.
MILLSTONE - UNIT 3                                   B 3/4 9-7       Amendment No. 409, 24-9,-2-3-}}

Latest revision as of 20:10, 10 March 2020

License Amendment Request to Relocate TS Surveillance Frequencies to Licensee Controlled Program in Accordance with TSTF-425, Revision 3, Attachment 3 Through Attachment 6, Marked-up TS Pages and TS Bases Pages
ML11193A224
Person / Time
Site: Millstone Dominion icon.png
Issue date: 07/05/2011
From:
Dominion Nuclear Connecticut
To:
Office of Nuclear Reactor Regulation
References
10-711
Download: ML11193A224 (92)
v • d • e


Text

fbi- ul iy 28, 2600 PLANT SYSTEMS AUXILIARY FEED WATER SYSTEM

'4/

LIMITING CONDITION FOR OPERATION ACTION: (Continued)

A'

.'r Inoperable Equipment Required ACTION Required

+

e. Three auxiliary feedwater e. I pumps in MODE 1, 2, or 3.

- - - ----- NOTE -

LCO 3.0.3 and all other LCO required ACTIONS requiring MODE changes are suspended until one AFW pump is restored to OPERABLE status. /

/

Immediately initiate ACTION to restore one auxiliary feedwater pump to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At 1--st ofte r 31 days by: specified in the Surveillance Frequency Control Progqram
  • 7--1the frequency

- . .- . .- . .-N kU I t -. . . .. . . - . . . .

Auxiliary fjedwater pumps may be considered OPERABLE during alignment and operation l/r steam generator level control, if they are capable of being manually realgnedo the auxiliary feedwater mode of operation.

raligne? .. . . ..

Verifyiot each auxiliary feedwater . power manual, . operated,

. .. and automatic

. . valve

. in/

each wter flow path and in each required steam supply flow path to the steam turbin drriven auxiliary feedwater pump, that is not locked, sealed, or otherwise sec d in position, is in the correct positin. .....

b. Spciiaton, Specification 40, cays -n go I 4.0.5,P2 bby:

t4

, tested pursuant to

1) Verifying that on recirculation flow each motor-driven pump develops a total head of greater than or equal to 3385 feet;
2) Verifying that on recirculation flow the steam turbine-driven pump develops a total head of greater than or equal to 3780 feet when the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

MILLSTONE - UNIT 3 3/4 7-5 Amendment No. 96, 400, 4-2-7, 4-39,

-206,-435-

Febmuary 28, 2007 PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

c. At p by verifying that each auxiliary feedwater pump starts as igned automatically upon receipt of an Auxiliary Feedwater Actuation test signsna. For the steam turbine-driven auxiliary feedwater pump, the provisions of Specifi tion 4.0.4 are not applicable for entry into MODE 3.

4.7.1.2.2 An auxiliary fe dwater flow path to each steam generator shall be demonstrated OPERABLE following each OLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying flow to ach steam generator.

Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3 3/4 7-5a Amendment No. 244- 4-

September 11, 1997 PLANT SYSTEMS DEMINERALIZED WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The demineralized water storage tank (DWST) shall be OPERABLE with a water volume of at least 334,000 gallons.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the DWST inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

a. Restore the DWST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or
b. Demonstrate the OPERABILITY of the condensate storage tank (CST) as a backup supply to the auxiliary feedwater pumps and restore the DWST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST shall be demonstrated OPERABLE at last pe 12 hu,, by verifying the water volume is within its limits when the tank is the sup y source for the auxiliary feedwater pumps.

4.7.1.3.2 The CST shall be demonstrated OPERABLE least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the combined volume of both the DWST and CST is at least 384,000 gallons of water whenever the CST and DWST are the supply source for he auxiliary feedwater pumps.

Ithe frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 3 3/4 7-6 Amendment No. +50-

Jftnutftry 31,986 PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microCurie/gram DOSE EQUIVALENT 1-131.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

With the specific activity of the Secondary Coolant System greater than 0.1 microCurie/gram DOSE EQUIVALENT 1-131, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by ........ o t, sampling a ono ifFa, 4. -.

determining the Gross Radioactivity and DOSE EQUIVALENT 1-131 Concentration at the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 7-7

JaT. .a .y 31 , 1 98 6 Delete Table 4.7-1 TABLE 4.7-1 N iLEU IN A K Y kL IUU L A IN 1 L)1I V1 l*

,l IJVA I L 11V 11 SAMPLE AND ANALYSTS PR ' MRAM TWY-P MEASUREMENT SAMPL D ANALYSIS AN DNALYSIS FRE NCY

1. Gross Radioacti At least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Determination
2. Isotopic Analysis for DO a) Once per 31 days, when-EQUIVALENT 1-131 ever the gross radio-Concentration activity determination indicates concentrations "a

"* ,,eater ctta bthan wavl l e l im10% of the i t fo r ad io i n s .

~ever b) Oncethe pergross 6 nths, when-r~a-.-

.activity determ inatio indicates concentrations less than or equal to 10%

of the allowable limitj for radioiodines.

MILLSTONE - UNIT 3 3/4 7-8

1O79 99 PLANT SYSTEMS STEAM GENERATOR ATMOSPHERIC RELIEF BYPASS LINES LIMITING CONDITION FOR OPERATION 3.7.1.6 Each steam generator atmospheric relief bypass valve (SGARBV) line shall be OPERABLE, with the associated main steam atmospheric relief isolation (block) valve in the open position.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS

a. With one required SGARBV line inoperable, restore required SGARBV line to OPERABLE status within 7 days or be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. LCO 3.0.4 is not applicable.
b. With two or more required SGARBV lines inoperable, restore all but one required SGARBV line to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 4 without reliance upon steam generator for heat removal within the next 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.6.1 Verify one complete cycle of each SGARBV , o fi,-tths.

4.7.1.6.2 Verify one complete cycle of each mai eam atmospheric relief isolation (block) valve ... 18 ~n.*

lat the frequency specified in the Surveillance Frequency Control Program I MILLSTONE - UNIT 3 3/4 7-9a Amendment No. 14ý

PLANT SYSTEMS 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent reactor plant component cooling water safety loops shall be OPERABLE.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

With only one reactor plant component cooling water safety loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.3 At least two reactor plant component cooling water safety loops shall be demonstrated OPERABLE:

a. At by verifying that each valve (manual, power-operated, or autobtic) servicing safety-related equipment that is not locked, sealed, or I_

otherwis _ ' in position is in its correct position; and

. __ cured

b. At *. --- by verifying that:
1) Eac tornatic e actuates to its correct position on its associated Engineere fety F e actuation signal, and
2) Each Component Coo . r System pump starts automatically on an SIS test signal.

Ithe frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3 3/4 7-11 Amendment No. 4-2-4, -M&6-

ly 24.-, -VV2-PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE.

APPLICABILITY: MODES 1,2,3, and 4.

ACTION:

With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.4 At least two service water loops shall be demonstrated OPERABLE:

a. At by verifying that each valve (manual, power-operated, or aut tic) servicing safety-related equipment that is not locked, sealed, or otherwi secured in position is in its correct position; and
b. At`- -- ,, r 24 mo-tts by verifying that:
1) Ea utoma valve servicing safety-related equipment actuates to its correct tion its associated Engineered Safety Feature actuation signal, and
2) Each Service Water Sys pump starts automatically on an SIS test signal.

Ithe frequency specified in the Surveillance Frequency Control Proqram I MILLSTONE - UNIT 3 3/4 7-12 Amendment No. 4-2-4, 2e6

August 29, 1995 PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with an average water temperature j, of less than or equal to 75°F.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

If the UHS temperature is above 75'F, monitor the UHS temperature once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the UHS temperature does not drop below 75'F during this period, place the plant in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. During this period, if the UHS temperature increases above 77°F, place the plant in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.5 The UHS shall be determined OPERABLE:

a. A+ 1-f per 24 hoofs by verifying the average water temperature to be within lim
b. At least oncer 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by verifying the average water temperature to be within limits w the erage water temperature exceeds 70 0 F.

Ithe freauencv swecified in the Surveillance Freauencv Control Proaram MILLSTONE - UNIT 3 3/4 7-13 Amendment No. ++9-

S~pt@Mb@r 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION /

ACTION: (Continued)

e. With both Control Room Emergency Air Filtration Systems inoperable, or with the OPERABLE Control Room Emergency Air Filtration System required to be in the emergency mode by ACTION d. not capable of being powered by an OPERABLE emergency power source, or with one or more Control Room Emergency Air Filtration System Trains inoperable due to an inoperable CRE boundary, immediately suspend the movement of recently irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.7.7 Each Control Room Emergency Air Filtration System shall be demonstrated OPERABLE: the frequency specified in the Surveillance Frequency Control Proqram ]

a. At m by verifying that the control room air temperature is less than ual to 95'F;
b. At pen by initiating, from the contr room, flow through the HEPA filters and charcoal adsorbers and verifying a sys m flow rate of 1,120 cfm +/- 20% and that the system operates for at least 10 cont uous hours with the heaters operating;
c. At - or following painting, fire, or chemical release in any ventilation zone communicating with the system by:
1) Verifying that the system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the system flow rate is 1,120 cfm +/- 20%;
2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 54 ft/min; and
3) Verifying a system flow rate of 1,120 cfm +/- 20% during system operation when tested in accordance with ANSI N510-1980.

MILLSTONE - UNIT 3 3/4 7-16 Amendment No. 2, --23, 44, 4-84, 203, 206, 23-7, 243

Septem.be 18, 2008 PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5%

when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86 0 F),

and a relative humidity of 70%, and a face velocity of 54 ft/min.

,T Ethe frequency specified in the Surveillance Frequency Control Proqram

e. At' least .. pr"24 .mnth by:
1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.75 inches Water Gauge while operating the system at a flow rate of 1,120 cfm + 20%;
2) Deleted
3) Verifying that the heaters dissipate 9.4 +1 kW when tested in accordance with ANSI N510-1980.
f. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 1120 cfm +/- 20%; and
g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 1120 cfm + 20%.
h. By performance of CRE unfiltered air inleakage testing in accordance with the CRE Habitability Program at a frequency in accordance with the CRE Habitability Program.
  • ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.

MILLSTONE - UNIT 3 3/4 7-17 Amendment No. 2, 23, 4-8-1, 20-, 2-20, 24-3

M*rLc 29, 2007 PLANT SYSTEMS 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.9 Two independent Auxiliary Building Filter Systems shall be OPERABLE.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

With one Auxiliary Building Filter System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. In addition, comply with the ACTION requirements of Specification 3.6.6.1.

SURVEILLANCE REQUIREMENTS the frequency specified in the Surveillance Frequency Control Proqram 4.7.9 Each Au ary Building Filter System shall be demonstrated OPERABLE:

a. At - p, r 31 day on a S B-,,6ERBD TEST BASIS by initiating, from the contr 1room, flow through the HEPA filters and charcoal adsorbers and verifying a sy em flow rate of 30,000 cfm +/-10% and that the system operates for at least 10 con nuous hours with the heaters operating;
b. AtEa p22- 4 rntnhs or following painting, fire, or chemical release in any ,

ventilation zone communicating with the system by:

1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is 30,000 cfm I+10%;
2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl MILLSTONE - UNIT 3 3/4 7-20 Amendment No. 2, 84, +-2-, 4-84, 2404, 206,

--2-34--

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86°F), a relative humidity of 70%, and a face velocity of 52 ft/min; and

3) Verifying a system flow rate of 30,000 cfm +/- 10% during system operation when tested in accordance with ANSI N510-1980.
c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,* shows the methyl iodide penetration less than or equal to 2.5% when tested in accordance with ASTM D3803-89 at a temperature of 30'C (86'F), a relative humidity of 70%, and a face velocity of 52 ft/min; the frequency specified in the Surveillance Frequency Control Program
d. A ..- .
1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.8 inches Water Gauge while operating the system at a flow rate of 30,000 cfm +/-10%,
2) Verifying that the system starts on a Safety Injection test signal, and
3) Verifying that the heaters dissipate 180 +/-18 kW when tested in accordance with ANSI N510-1980.
e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at a flow rate of 30,000 cfm +/- 10%;

and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 30,000 cfm +/-1 0%.
  • ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.

MILLSTONE - UNIT 3 3/4 7-21 Amendment No. 2, 87, --2--, 4-84, 206-

Rme 24, 1997ý-

PLANT SYSTEMS 3/4.7.14 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.14 The temperature limit of each area shown in Table 3.7-6 shall not be exceeded.

APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE.

ACTION:

With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6:

a. By less than 20'F and for less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, record the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s).
b. By less than 20'F and for greater than or equal to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area(s) exceeded the limit(s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specification 3.0.3 are not applicable.
c. With one or more areas exceeding the temperature limit(s) shown in Table 3.7-6 by greater than or equal to 20'F, prepare and submit a Special Report as required by ACTION b. above and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area(s) to within the temperature limit(s) or declare the equipment in the affected area(s) inoperable.

SURVEILLANCE REQUIREMENTS 4.7.14 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limits:

a. At 'a"a F-...' seve days when the alarm is OPERABLE, and;
b. At le4 once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the alarm is inoperable.

the frequency specified in the Surveillance Frequency Control Proqram MILLSTONE - UNIT 3 3/4 7-32 Amendment No. *-5, 9-5, 4-00, 444-

Mait M, 2306 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (continued) 11 Inoperable Equipment Required AC I ION

e. Two diesel generators e.2 Restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

AND e.3 Following restoration of one diesel generator, restore remaining inoperable diesel generator to OPERABLE status following the time requirements of ACTION Statement b. above based on the initial loss of the remaining inoperable diesel generator. -r

/-the frequency specified in the Surveillance Frequency Control Progqram[

SURVEILLANCE RE(ENTS

a. Detern imed OPERABL a 1at*c pr7dy by verifying correct breaker alignr ents, indicated pow avilabilt,i and shutdown dunn prom af ............ unit. power supply by
b. Dem erring .

nsrtOPERABLE (manually and automatically) the normal trans circ fit to the alternate circuit.

demonstrated OPERABLE:*

4.8.1.1.2 Each dfsel generator shall be

a. Atb
1) Verifying the fuel level in the day tank,
2) Verifying the fuel level in the fuel storage tank, the storage
3) Verifying the fuel transfer pump starts and transfers fuel from system to the day tank, inventory in storage,
4) Verifying the lubricating oil m229 Verifying the diesel starts from standby conditions and achieves generator
5) following of theHz.

60 +0.8 and one signals:

The diesel and frequency at 4160this +420 volts test by using generator shall be started for voltage a) Manual, or prelube of these surveillances may be preceded by an engine a.Allstarts for planned the purpose period.

s0, Amendment No. 4 64,b4--2,--94,24-0, 3/4 8-3a MILLSTONE UNIT 3-

F.bruftry 2, 2001 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Program I b) Simulated loss-of-offsite power by itself, or c) Simulated loss-of-offsite power in conjunction with an ESF Actuation test signal, or d) An ESF Actuation test signal by itself.

6) Verifying the generator is synchronized and gradually loaded in accordance with the manufacturer's recommendations between 4800-5000 kW* and operates with a load between 4800-5000 kW* for at least 60 minutes, and
7) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
b. At a per 184 day by:
1) V ing that the diesel generator starts from standby conditionS and /

attains generator voltage and freauencv of 4160 +/- 420 volts and 60+/- /

0.8 Hz within 11 seconds after the start signal.

2) Verifying the generator is synchronized to the associated emergency bus, loaded between 4800-5000 kW* in accordance with the manufacturer's recommendations, and operate with a load between 4800-5000 kW* for a t least 60 minutes.

T diesel generator shall be started for this test using one of the signals in 'I S teillance Requirement 4.8.1.1.2.a.5. This test, if it is performed so it coincide s /

w the testing required by Surveillance Requirement 4.8.1.1.2.a.5, may also s e to concurrently meet those requirements as well.

C. Aop i - thanud after each operation of the diesel where the period o f op ration was greater than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by checking for and removing ac mulated water from the day tank;

d. At as* oncz nor3 by checking for and removing accumulated water from the fuel oil storage tanks;
e. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to storage tanks and:
1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degrees at 607F, or a specific gravity of within 0.0016 at 60/607F, when compared to the supplier's certificate, or an absolute specific gravity at 60/607F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;

  • The operating band is meant as guidance to avoid routine overloading of the diesel.

Momentary transients outside the load range shall not invalidate the test.

MILLSTONE - UNIT 3 3/4 8-4 Amendment No. 4, 64, 41-2, 4-3-, 94-

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) the frequency specified in the Surveillance Frequency Control Proqram b) A kinematic viscosity at 40'C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity, SUS at 100°F of greater than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the supplier's certification; c) A flash point equal to or greater than 125°F; and d) Water and sediment less than 0.05 percent by volume when testedt in accordance with ASTM-D 1796-83.

By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested ir accordance with ASTM-D975-81 except that: (1) the cetane index shall bc determined in accordance with ASTM-D976 (this test is an appropriate approximation for cetane number as stated in ASTM-D975-81 [Note E]),

and (2) the analysis for sulfur may be performed in accordance with ASTM-D 1552-79, ASTM-D2622-82 or ASTM-D4294-83.

f. t by obtaining a sample of fuel oil in accordance with TM-D2276-78, and verifying that total particulate contamination is less than
  • mg/liter when checked in accordance with ASTM-D2276-78, Method A;
g. As m , during shutdown, by:
1) DELETED Ir
2) Verifying the generator capability to reject a load of greater than or equal to 595 kW while maintaining voltage at 4160 +/- 420 volts and frequency at 60+/-* 3 Hz;
3) Verifying the generator capability to reject a load of 4986 kW without tripping. The generator voltage shall not exceed 5000 volts during and 4784 volts following the load rejection;
4) Simulating a loss-of-offsite power by itself, and:

a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts from standby conditions on the auto-start signal, energizes the emergency busses with permanently connected loads within 11 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +/- 420 volts and 60 +/- 0.8 Hz during this test.

MILLSTONE - UNIT 3 3/4 8-5 Amendment No.4, 4-0, 64, 73, 400, 140,4-2,148,-+94

MaiIt 29, 2007 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

8) Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 5335 kW;
9) Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.

10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by:

(1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;

11) DELETED
12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each load block within +/- 10% of its design interval; and
13) DELETED
h. At '--t per 10 yzarS by starting both diesel generators simultaneously from ,.

stal by conditions, during shutdown, and verifying that both diesel generators achie generator voltage and frequency at 4160 + 420 volts and 60 + 0.8 Hz in less tha or equal to 11 seconds; and

i. At by draining each fuel oil storage tank, removing the ace ted ediment and cleaning the tank using a sodium hypochlorite solution.

ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 8-7 Amendment No. 64, 49, 4-00, 4-4-2, 4-4, 4-94, 234-

Febrdtry 2 9O4 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

j. Atper 1 mt by verifying the diesel generator operates for at least 24 hrs. During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded betwee 5400-5500kW* and during the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generato 7hall be loaded between 4800-50OOkW*. The generator voltage and frequency all be 4160 +/- 420 volts and 60 +/- 0.8 Hz within 11 seconds after the start signal; t steady-state generator voltage and frequency shall be maintained within these li *ts during this test.** Within 5 minutes after completing this 24-hour test, perf Specification 4.8.1.1.2.a.5) excluding the requirement to start the diesel from tandby conditions.***
k. At least, by verifying that the fuel transfer pump transfers fuel froxXbac to the day tank of each diesel via the installed cross-connecti
1. At !4 by v*fying that the following diesel generator locke jel gen ator starting:
1) Engine overspeed, frequency specified in the veillance Frequency Control Program
2) Lube oil pressure low (2 of 3 logic), I'
3) Generator differential, and
4) Emergency stop.
  • The operating band is meant as guidance to avoid routine overloading of the diesel.

Momentary transients outside the load range shall not invalidate the test.

    • Diesel generator loadings may include gradual loading as recommended by the manufacturer.

If Surveillance Requirement 4.8.1.1.2.a.5) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated between 4800-5000 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until operating temperature has stabilized.

MILLSTONE - UNIT 3 3/4 8-8 Amendment No. 1-0, 64, 440, t-94 -

ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:

a. 125-volt Battery Bank 301A-l, and an associated full capacity charger,
b. 125-volt Battery Bank 301A-2, and an associated full capacity charger,
c. 125-volt Battery Bank 301 B-I and an associated full capacity charger, and
d. 125-volt Battery Bank 301B-2 and an associated full capacity charger.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With either Battery Bank 301A-1 or 301B-l, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With either Battery Bank 301A-2 or 301B-2 inoperable, and/or one of the required full capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS f --Ithe frequency specified in the Surveillance Frequency Control Program 4.8.2.1 Each 12 -volt battery bank and charger shall be demonstrated OPERABLE:

a. A ---t - - -- r 7 day by verifying that:
1) The parameters in Table 4.8-2a meet the Category A limits, and
2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.

MILLSTONE - UNIT 3 3/4 8-11 Amendment No. -6+

  • 95197 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
b. At least

..... per 92 da"- and within 7 days after a battery discharge with battery teinal voltage below 110 volts, or battery overcharge with battery terminal v itage above 150 volts, by verifying that:

  • ) The parameters in Table 4.8-2a meet the Category B limits,
2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10-6 ohm, and
3) The average electrolyte temperature of six connected cells is above 60'F.

Ppei 18 inanth by verifying that:

the frequency The cells, cell plates, and battery racks show no visual indication of specified in the physical damage or abnormal deterioration, Surveillance

-- 4 The cell-to-cell and terminal connections are clean, tight, and coated with Frequency anticorrosion material, Control Program

3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10-6 ohm, and

) Each battery charger will supply at least the amperage indicated in Table 4.8-2b at greater than or equal to 132 volts for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. t least

.... e per 18 .mI th , during shutdown, by verifying that the battery apacity is adequate to supply and maintain in OPERABLE status all of the actual o simulated emergency loads for the design duty cycle when the battery is s'uected to a battery service test;

e. t 16 , during shutdown, by verifying that the battery apacity is at least 80% of the manufacturer's rating when subjected to a rformance discharge test. 6,ii I. 6-ionth intei-vai this performance d charge test may be performed in li of the battery service test required by S ecification 4.8.2.1d.; and
f. At pe, during shu own, by giving performance discharge tests of battery capacity to any battery th t shows signs of degradation or has reached 85% of the service life expected or the application. Degradation is indicated when the battery capacity drops ore than 10% of rated capacity from its average on previous performance tests, or s below 90% of the manufacturer's rating.

MILLSTONE - UNIT 3

Set 1 1 bA-4+ 004 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one of the required trains of A.C. emergency busses not OPERABLE, restore the inoperable train to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With one A.C. vital bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. bus: (1) reenergize the A.C. vital bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; and (2) reenergize the A.C.

vital bus from its associated inverter connected to its associated D.C. bus within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined OPERABLE in the specified manner kaM-oee per- days by verifying correct breaker alignment and indicated voltage on e es.

Ithe frequency specified in the Surveillance Frequency Control Proqramr MILLSTONE - UNIT 3 3/4 8-17 Amendment No.-64, 22-

september 18, 2008 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION (Continued)

4) Two 125 volt DC Busses consisting of:

a) Bus #301B-1 energized from Battery Bank #301B-1, and b) Bus #301 B-2 energized from Battery Bank #301 B-2.

APPLICABILITY: MODES 5 and 6.

ACTION:

With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity additions that could result in loss of required SDM or boron concentration, movement of recently irradiated fuel assemblies, crane operation with loads over the fuel storage pool, or operations 4" with a potential for draining the reactor vessel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible.

SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner t least

.f.ee pefr day-s by verifying correct breaker alignment and indicated voltae usses.

Ithe frequency specified in the Surveillance Frequencv Control Program MILLSTONE - UNIT 3 3/4 8-18a Amendment No. 446, 230,443

066*2UU 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:

a. A Keff of 0.95 or less, or
b. A boron concentration of greater than or equal to the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

Additionally, the CVCS valves of Specification 4.1.1.2.2 shall be closed and secured in position.

APPLICABILITY: MODE 6.*

ACTION:

a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and positive reactivity additions and initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until Keff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to the limit specified in the COLR, whichever is the more restrictive.
b. With any of the CVCS valves of Specification 4.1.1.2.2 not closed** and secured in position, immediately close and secure the valves.

SURVEILLANCE REQUIREMENTS 4.9.1.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and
b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.1.2 The boron concentration of the Reactor Coolant System and the refueling cavity shall be determined by chemical analysis at a pE 7a hPur.

4.9.1.1.3 The CVCS valves of Specification 4.1.1.2. s ýrified closed and locked at lem-oefee-pet-31 day . t Ithe frequency specified in the Surveillance Frequency Control Proqram-T='

  • The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
    • Except those opened under administrative control.

MILLSTONE - UNIT 3 3/4 9-1 Amendment No. -0, 60, 99, 4-4-3, 2-0-3, 24-9, 240

F 20, 2002

.bi*uly REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1.2 The soluble boron concentration of the Spent Fuel Pool shall be greater than or equal to 800 ppm. 4-APPLICABILITY:

Whenever fuel assemblies are in the spent fuel pool.

ACTION:

a. With the boron concentration less than 800 ppm, initiate action to bring the boron concentration in the fuel pool to at least 800 ppm within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
b. With the boron concentration less than 800 ppm, suspend the movement of all fuel assemblies within the spent fuel pool and loads over the spent fuel racks.

SURVEILLANCE REQUIREMENTS 4.9.1.2 Verify that the boron concentration in the fuel pool is greater than or equal to 800 ppm evey"7 dwgy.

lat the frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 9-1a Amendment No. 4-2, 4-58, 4-89, 203-

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 Two Source Range Neutron Flux Monitors shall be OPERABLE with continuous visual indication in the control room, and one with audible indication in the containment and control room.

APPLICABILITY: MODE 6.

ACTION:

a. With one of the above required monitors inoperable immediately suspend all operations involving CORE ALTERATIONS and operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1.
b. With both of the above required monitors inoperable determine the boron concentration of the Reactor Coolant System within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK and verification of audible counts at-least-e*se-per
b. A CHANNEL CALIBRATION at Ithe frequency specified in the Surveillance Frequency Control Program
  • Neutron detectors are excluded from CHANNEL CALIBRATION.

MILLSTONE - UNIT 3 3/4 9-2 Amendment No. 44-7, 20-3, 430

Mare.h 17, 2004 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment access hatch shall be either:
1. closed and held in place by a minimum of four bolts, or
2. open under administrative control
  • and capable of being closed and held in place by a minimum of four bolts,
b. A personnel access hatch shall be either:
1. closed by one personnel access hatch door, or
2. capable of being closed by an OPERABLE personnel access hatch door, under administrative control,* and /
c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
1. Closed by an isolation valve, blind flange, or manual valve, or
2. Be capable of being closed under administrative control.*

APPLICABILITY: During movement of fuel within the containment building.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of fuel in the containment building.

SURVEILLANCE REQUIREMENTS 4.9.4.a Verify each required containment penetrations is in the required status ole-enee per 4.9.4.b DELETED Ithe frequency specified in the Surveillance Frequency Control Program Administrative controls shall ensure that appropriate personnel are aware that the equipment access hatch penetration, personnel access hatch doors and/or other containment penetrations are open, and that a specific individual(s) is designated and available to close the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g. cables and hoses) that could prevent closure of the equipment access hatch penetration, a personnel access hatch door and/or other containment penetrations must be capable of being quickly removed.

MILLSTONE - UNIT 3 3/4 9-4 Amendment No. 203, 2+9-

-O2*"ý,6 REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.*

APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.

ACTION:

With no RHR loop OPERABLE or in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and suspend loading irradiated fuel assemblies in the core and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access t

from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at ...a...... 12....

H f"

Ithe frequency specified in the Surveillance Frequency Control Program I The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1. I, MILLSTONE - UNIT 3 3/4 9-8 Amendment No. 4-14,23

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.*

APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet.

ACTION:

a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.
b. With no RHR loop in operation, suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.1 and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at lcatt zncz per 12 ho .

Ithe frequency specified in the Surveillance Frequency Control Program

  • The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period, provided no operations are permitted that could cause introduction of coolant into the RCS with boron J, concentration less than required to meet the boron concentration of LCO 3.9.1.1.

MILLSTONE - UNIT 3 3/4 9-9 Amendment No. 4047, 2-3

Februay 20, 2002 REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.

APPLICABILITY: During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required deptht least Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 9-11 Amendment No. -2&3-

Oeoe 25, 1990 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at J.t.... ,

&.... days when irradiated fuel assemblies are in the fuel storage pool.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 9-12 Amendment No. 55-

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 33 gpm of a solution containing greater than or equal to 6600 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at 4.10.1.2 Each full-le control rod not fully inserted shall be demonstrated capable of full insertion when tripped from ast the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to les an the limits of Specification 3.1.1.1.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 10-1 Amendment No. 4+3-

Dceember 10, 2003 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2.1 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
b. The limits of Specifications 3.2.2.1 and 3.2.3.1 are maintained and determined at the frequencies specified in Specification 4.10.2.1.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specification 3.2.2.1 or 3.2.3.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1.1, and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2.1 and 3.2.3.1, or
b. Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least one@ per hou during PHYSICS TESTS.

4.10.2.1.2 The Surveillance Req rements of the below listed specifications shall be performed at p-er 12 hurs during YSICS TESTS:

a. Specifications 4.2.2. .2 and 4.2.2.1.3, and
b. b. Spec ation 4.2.3. 1.

Spei

[the frequency specified in the Surveillance Frequency Control Program[

MILLSTONE - UNIT 3 3/4 10-2 Amendment No. 444-

SPECIAL TEST EXCETIlONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and
c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 541 'F.

APPLICABILITY: MODE 2.

ACTION:

a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
b. With a Reactor Coolant System operating loop temperature (Tavg) less than 541 °F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at lcast once per heur during PHYSICS TESTS.

4.10.3.2 Each Intermediate and P er Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST w in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating PHYSICS TESTS.

4.10.3. The Reactor Coolant System teierature (Tavg) shall be determined to be greater than or equal to 541'F at 1 uring PHYSICS TESTS.

Fthe frequencv specified in the Surveillance Frequencv Control Proqramr MILLSTONE - UNIT 3 3/4 10-4

1annaty 3fi9866 SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of STARTUP and PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and
b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least ete per-heur during STARTUP and PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating STARTUP and PHYSICS TESTS.

Ithe frequency specified in the Surveillance Frequency Control Program MILLSTONE - UNIT 3 3/4 10-5

Sept-mber 30N "008 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

RG 1.183), which were considered in completing the vulnerability assessments, are documented in the UFSAR/current licensing basis. Compliance with these RGs is consistent with the current licensing basis as described in the UFSAR and other licensing basis documents.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVs, operating at the flow rate required by the Surveillance Requirements, at a Frequency of 48 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.
e. The quantitative limits on unfiltered air inleakage into the CRE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

Insert 1 The provisions of Surveillance Requirement 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c. and d., respectively.

6.8.5 Written procedures shall be established, implemented and maintained covering Section I.E, Radiological Environmental Monitoring, of the REMODCM.

6.8.6 All procedures and procedure changes required for the Radiological Environmental Monitoring Program (REMP) of Specification 6.8.5 above shall be reviewed by an individual (other than the author) from the organization responsible for the REMP and approved by appropriate supervision.

Temporary changes may be made provided the intent of the original procedure is not altered and the change is documented and reviewed by an individual (other than the author) from the organization responsible for the REMP, within 14 days of implementation.

MILLSTONE - UNIT 3 6-17e Amendment No. 245

INSERTS FOR TECHNICAL SPECIFICATIONS MARKUPS INSERT 1 (for TS 6.8.4)

i. Surveillance Frequency Control Progqram This program provides controls for surveillance frequencies. The program shall ensure that surveillance requirements specified in the technical specification are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
a. The Surveillance Frequency Control Program shall contain a list of frequencies of those surveillance requirements for which the frequency is controlled by the program.
b. Changes to the frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the frequencies established in the Surveillance Frequency Control Program.

INSERTS FOR TECHNICAL SPECIFICATIONS BASES INSERT 2 The surveillance frequency is controlled under the Surveillance Frequency Control Program I

Serial No.10-711 Docket No. 50-423 ATTACHMENT 4 CROSS-REFERENCES NUREG-1431 TO MPS3 TS SURVEILLANCE FREQUENCIES REMOVED DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 1 of 12 CROSS-REFERENCE NUREG-1431 TS SURVEILLANCE REQUIREMENT FREQUENCIES TO MILLSTONE UNIT 3 TS SURVEILLANCE REQUIREMENT FREQUENCIES REMOVED Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Shutdown margin Verify SDM in Modes 2 w/keff < 1 ,3 4, and 5 SR 3.1.1.1 ---

Verify SDM in Modes l and 2 --- 4.1.1.1.1 Verify SDM in Modes 3, 4, and 5 Loops Filled --- 4.1.1.1.2.1 .b Verify Valve Position 3-CHS v305 4.1.1.1.2.2 Verify SDM in Cold Shutdown Loops Not Filled --- 4.1.1.2.1.b Verify Valve Positions --- 4.1.1.2.2 Core Reactivity Verify Reactivity _1% SR 3.1.2.1 4.1.1.1.2 Rod Group Alignment Verify Rod Position within Alignment SR 3.1.4.1 4.1.3.1.1 Verify Rod Movement SR 3.1.4.2 4.1.3.1.2 Verify Rod Drop Times --- 4.1.3.4.c Shutdown Bank Insertion Limits Verify Insertion Limits SR 3.1.5.1 4.1.3.5.b Control Bank Insertion limit Verify Limits within COLR SR 3.1.6.2 4.1.3.6 Verify Control Bank Rod Sequence and Overlap SR 3.1.6.3 PositionIndication System Verify DigitalRod Position Operable DRPI vs. Demand 4.1.3.2.1 Position Indication System Verify DigitalRod Position Operable DRPI vs. Demand 4.1.3.2.2 Position Indication System Agree When Exercised Physics Test Exceptions Verify RCS Loop Temperature SR 3.1.8.2 4.10.3.3 Verify Thermal Power <5% SR 3.1.8.3 4.10.3.1 Verify Thermal Power < 85% --- 4.10.2.1.1 Verify SDM SR 3.1.8.4 4.10.1.1 Perform Specs 4.2.2.1.2, 4.2.2.1.3, and 4.2.3.1.2 --- 4.10.2.1.2 Determine Thermal Power < P-7 --- 4.10.4.1 FQ(Z) Limits - RAOC Verify FQ(Z) limits - measured SR 3.2.1.1 4.2.2.1.2.d(2)

Verify FQ(Z) limits Base Load Operations - Measured --- 4.2.2.1.4.d(2)

Verify FQ (Z) limits SR 3.2.1.2 Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 2 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

N FAH Limits N

Verify FAH (Z) limits SR 3.2.2.1 4.2.3.1.2.b AFD Limits - RAOC Verify AFD Within Limit SR 3.2.3.1 4.2.1.1.1.a Base Loaded Operations- Determine by Measurement the --- 4.2.1.1.3 AFD for Each Operable Excore Base Loaded Operations- Updated Target AFD --- 4.2.1.1.4 QPTR Verify QPTR by calculation SR 3.2.4.1 4.2.4.1.a Verify QPTR w/ incore detectors SR 3.2.4.2 4.2.4.2 RPS Instrumentation Table 4.3-1 Perform Channel Check SR 3.3.1.1 Channel Check Column Perform Calorimetric - actual power adjust if > 2% SR 3.3.1.2 Table 4.3-1 Functional Unit (FU) 2 Compare and Adjust NIS to Incore > 3% SR 3.3.1.3 Table 4.3-1, FU 2 Perform TADOT Rx Trip Breakers SR 3.3.1.4 Table 4.3-1 FUs, 18 & 21 Perform Actuation Logic Test SR 3.3.1.5 Table 4.3-1 FU, Unit 19 Calibrate NIS to Incore SR 3.3.1.6 Table 4.3-1, FU 2 Table 4.3-1 Perform COT - 184 days SR 3.3.1.7 Analog Channel Operational Test Column Perform COT (MPS3 - Quarterly Frequency) SR 3.3.1.8 Table 4.3-1 FU 6 Perform TADOT SR 3.3.1.9 Table 4.3-1 TADOT column Perform Channel Calibration w/time constants SR 3.3.1.10 ---

Perform Channel Calibration w/o neutron detectors SR 3.3.1.11 Table 4.3-1, FUs 2, 3, 5, &6 Perform Channel Calibration w/ RTDs SR 3.3.1.12 ---

Perform COT - 18 months SR 3.3.1.13 Table 4.3-1, FU 17 Perform TADOT - 18 months SR 3.3.1.14 Table 4.3-1, FUs 1 &21 Verify Response Time SR 3.3.1.16 4.3.1.2 ESFAS Instrumentation Table 4.3-2, SR 3.3.2.1 c an l e column Perform C hannel C heck Channel Check Column Perform Actuation Logic Test - 92 days SR 3.3.2.2 Table 4.3-2, FU 10 Table 4.3-2, FL~s SR 3.3.2.3 a .3 -2,4 a 6b Table Perform Actuation Logic Test - 31 days 1.b,2.b,3.a.2,3.b.2,4.b,5.a&b,6.b,7.c Perform Master Relay Test SR 3.3.2.4 Table 4.3-2 Master Relay Test Column Perform Frequency)COT - 184 days (MPS3- Quarterly SR 3.3.2.5 Table Operational Analog Channel 4.3-2 Test Column Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 3 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Perform Slave Relay Test - 92 days SR 3.3.2.6 Table 4.3-2 Slave Relay Test Column Perform TADOT - 92 days SR 3.3.2.7 Table 4.3-2, FUs 1.a, 2.a, 3.a.1, Perform TADOT - 18 months SR 3.3.2.8 3.b.1, 4.d.1, 4.d.2, 5.c, 6.a, 7.a, 7.b, and 9.c Perform Channel Calibration SR 3.3.2.9 Table 4.3-2 Channel Calibration Column Verify Time Response SR 3.3.2.10 4.3.2.2 Radiation Monitoring Instrumentation MPS3 Perform Check, Calibrate,and Analog COT 4.3.3.1 PAM Instrumentation PAM Channel Check SR 3.3.3.1 4.3.3.6.1 PAM Channel Calibration SR 3.3.3.2 4.3.3.6.1 Remote Shutdown System Perform Channel Check SR 3.3.4.1 4.3.3.5.1 Verify Control and Transfer Switch Function SR 3.3,4.2 4.3.3.5.2 Perform Channel Calibration SR 3.3.4.3 4.3.3.5.1 Perform TADOT of Reactor Trip Breaker SR 3.3.4.4 Shutdown Margin Monitor - MPS3 Perform Analog COT --- 4.3.5.a Verify Monitor Count Rate --- 4.3.5.b LOP EDG Start Instrumentation Perform Channel Check SR 3.3.5.1 ---

Perform TADOT SR 3.3.5.2 Table 4.3-2, FU 8 Perform Channel Calibration SR 3.3.5.3 Table 4.3-2, FU 8 Perform Response Time Testing ----- 4.3.2.2 Containment Purge and Vent Isolation Perform Channel Check SR 3.3.6.1 Perform Actuation Logic Test - 31 days SR 3.3.6.2 Perform Master Relay Test - 3 days SR 3.3.6.3 ---

Perform Actuation - 92 days SR 3.3.6.4 ---

Perform Master Relay Test -92 days SR 3.3.6.5 ---

Perform COT SR 3.3.6.6 ---

Perform Slave Relay Test SR 3.3.6.7 ---

Perform TADOT SR 3.3.6.8 4.6.3.2.c Perform Channel Calibration SR 3.3.6.9 ---

CREFS (Control Building Isolation)

Perform Channel Check SR 3.3.7.1 Table 4.3-2, FUs 7.d & e Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 4 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Perform COT SR 3.3.7.2 Table 4.3-2, FUs 7.d & e Perform Actuation Logic Test - 31 days SR 3.3.7.3 Table 4.3-2, FU 7.c Perform Master Relay Test - 31 days SR 3.3.7.4 Table 4.3-2, FU 7.c Perform Actuation Logic Test - 92 days SR 3.3.7.5 ---

Perform Master Relay Test - 92 days SR 3.3.7.6 ---

Perform Slave Relay Test SR 3.3.7.7 Table 4.3-2 ,FU 7.c Perform TADOT SR 3.3.7.8 Table 4.3-2, FUs 7.a & b Perform Channel Calibration SR 3.3.7.9 Table 4.3-2, FUs 7.a & b FBACS Actuation Instrumentation Perform Channel Check SR 3.3.8.1 Note 1 Perform COT SR 3.3.8.2 Note 1 Perform Actuation Logic Test SR 3.3.8.3 Note 1 Perform TADOT SR 3.3.8.4 Note 1 Perform Channel Calibration SR 3.3.8.5 Note 1 BDPS (Shutdown Monitor)

Perform Channel Check SR 3.3.9.1 ---

Perform COT SR 3.3.9.2 ---

Perform Channel Calibration SR 3.3.9.3 ---

RCS Press Temp & Flow Limits Verify Pressurizer Pressure SR 3.4.1.1 4.2.5 Verify RCS Average Temperature SR 3.4.1.2 4.2.5 Verify RCS Total Flow SR 3.4.1.3 4.2.3.1.3.b Verify RCS Total Flow w/ Heat Balance SR 3.4.1.4 ---

Calibrate RCS Total Flow Indicators --- 4.2.3.1.4 RCS Minimum Temp for Criticality Verify RSC Average Temperature in Each Loop SR 3.4.2.1 ---

RCS Temperature, Pressure, Verify Limits SR 3.4.3.1 4.2.5 Loop Operation - Modes 1 and 2 Verify Each Loop Operating SR 3.4.4.1 4.4.1.1 Loop Operation - Mode 3 Verify Required Loops Operating SR 3.4.5.1 4.4.1.2.3 Verify Steam Generator Water Level > 17% SR 3.4.5.2 4.4.1.2.2 Verify Breaker Alignment and Power Available SR 3.4.5.3 4.4.1.2.1 Loop Operation - Mode 4 Verify Loop Operation - RHR or RCS SR 3.4.6.1 4.4.1.3.3 Verify Steam Generator Water Level > 17% SR 3.4.6.2 4.4.1.3.2 Verify Breaker Alignment and Power Available SR 3.4.6.3 4.4.1.3.1 Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 5 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Loop Operation - Mode 5 -Loops Filled Verify RHR Loop Operating SR 3.4.7.1 4.4.1.4.1.2 Verify Steam Generator water Level > 17% SR 3.4.7.2 4.4.1.4.1.1 Verify Breaker Alignment and Power Available RHR SR 3.4.7.3 4.4.1.4.1.3 Pumps Verify Loop Operation - Mode 5 -Loops - Not Filled Verify RHR Loop Operating SR 3.4.8.1 4.4.1.4.2.2 Verify Breaker Alignment and Power Available RHR Pumps SR 3.4.8.2 4.4.1.4.2.1 Pressurizer (Modes I and 2/Mode 3)

Verify Water Level SR 3.4.9.1 4.4.3.1.1/4.4.3.2.1 Verify Heater Capacity of Required Groups SR 3.4.9.2 4.4.3.1.2/4.4.3.2.2 Verify Heater banks can be Powered from Emergency SR 3.4.9.4 Power Supply Pressurizer PORVS Cycle each Block Valve SR 3.4.11.1 4.4.4.2 Cycle each PORV SR 3.4.11.2 4.4.4.1.b Cycle each SOV Valve and Check Valve on the Air SR 3.4.11.3 Accumulators in PORV Control Systems Verify PORVs and Block Valves can be Powered SR 3.4.11.4 ---

from Emergency Power Sources Perform Channel Calibration --- 4.4.4.1.a Perform ACOT on PORV High PressurizePressure --- 4.4.4.1 .c Verify High PressureAuto Open is Enabled --- 4.4.4.1 .d LTOP Systems Verify only one HPI pump is capable of injecting into SR 3.4.12.1 4.4.9.3.4 the RCS.

Verify a maximum of one charging pump is capable of SR 3.4.12.2 4.4.9.3.5 injecting into the RCS.

Verify each accumulator is isolated. SR 3.4.12.3 ---

Verify each RHR Suction Valve is open for each Relief SR 3.4.12.4 4.4.9.3.2.a Valve Verify required RCS vent [2.07] square inches open SR 3.4.12.5 4.4.9.3.3 Verify PORV block valve is open for each required SR 3.4.12.6 4.4.9.3.1.c PORV.

Verify PORV COPPSArmed --- 4.4.9.3.1.c Verify RHR Suction Isolation Valve is Locked Open with Operator Power Removed for Required RHR SR 3.4.12.7 ---

Suction Relief Valve.

Perform COT on each Required PORV SR 3.4.12.8 4.4.9.3.1.a Perform Channel Calibration on each Required PORV SR 3.4.12.9 4.4.9.3.1.b Channel Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 6 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Operational Leakage Verify RCS Operational Leakage SR 3.4.13.1 4.4.6.2.1.d Verify SG Leakage < 150 gpd SR 3.4.13.2 4.4.6.2.1.e Monitor Reactor Head Flange Leakoff System --- 4.4.6.2.1 .f Measure Leakage to RCP Seals --- 4.4.6.2.1.c RCS PIVs Verify leakage from each is < 0.5 gpm SR 3.4.14.1 4.4.6.2.2.a Verify RHR Autoclosure Interlock Prevents Opening SR 3.4.14.2 4.5.2. d.1 Verify RHR Autoclosure Interlock Auto Close SR 3.4.14.3 RCS Leakage Detection Instrumentation Perform Channel Check - Particulate Rad Monitor SR 3.4.15.1 4.4.6.1.a Perform COT - Particulate Rad Monitor SR 3.4.15.2 4.4.6.1.a Perform Channel Calibration Sump Monitor SR 3.4.15.3 4.4.6.1.b Perform Channel Calibrationcontainment atmosphere SR 3.4.15.4 4.4.6.1 a radioactivitymonitor.

Perform Channel Calibration containment air cooler. SR 3.4.15.5 ---

RCS Specific Activity Verify RCS gross specific activity SR 3.4.16.1 ---

Verify reactor coolant Dose Equivalent 1-131 SR 3.4.16.2 4.4.8.2 Determine E Bar SR 3.4.16.3 ---

Verify Xe-133 --- 4.4.8.1 RCS Loop Isolation Valves Verify Open and Power Remove from Isolation Valves SR 3.4.17.1 4.4.1.5 RCS Loops Test Exceptions Verify power < P-7 SR 3.4.19.1 4.10.4.1 Accumulators Verify Accumulator isolation valve open SR 3.5.1.1 4.5.1.a.2)

Verify borated Water Volume SR 3.5.1.2 4.5.1.a. 1)

Verify N2 Pressure SR 3.5.1.3 4.5.1.a.1)

Verify Boron Concentration SR 3.5.1.4 4.5.1. b Verify Power removed from isolation valve SR 3.5.1.5 4.5.1.c ECCS - Operating Verify Valve Lineup SR 3.5.2.1 4.5.2.a Verify Valve Position SR 3.5.2.2 4.5.2.b.2)

Verify Piping Sufficiently Full SR 3.5.2.3 4.5.2. b. 1)

Verify Automatic Valve Actuation SR 3.5.2.5 4.5.2.e.1)

Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 7 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Verify Automatic Pump Start SR 3.5.2.6 4.5.2.e.2)

Verify RHR Pump Stop Automatically - LoLo RWST --- 4.5.2. e. 3)

Verify Throttle Valve Position SR 3.5.2.7 4.5.2.g.2)

Inspection Sump Components SR 3.5.2.8 4.5.2.d.2)

Verify Auto Interlock Prevents Valve Opening --- 4.5.2.d.1)

RWST Verify Water Temperature SR 3.5.4.1 4.5.4.b & 4.6.2.1.a.2)

Verify Water Volume SR 3.5.4.2 4.5.4.a.1)

Verify Boron Concentration SR 3.5.4.3 4.5.4.a.2)

Seal Injection Flow Verify Throttle Valve Position SR 3.5.5.1 ---

pH Tri-sodium PhosphateStorage Baskets Verify Minimum volume of TSP --- 4.5.5 BIT Verify Water Temperature SR 3.5.6.1 Note 1 Verify Water Volume SR 3.5.6.2 Note 1 Verify Water Boron Concentration SR 3.5.6.3 Note 1 Containment Air Locks Verify Interlock Operation SR 3.6.2.2 4.6.1.3.c Containment Isolation Valves Verify 42" Purge Valves Sealed Closed SR 3.6.3.1 / 4.6.1.7.1 Verify 8" Purge Valves Closed SR 3.6.3.2 4.6.1.7.1 Verify Valves Outside Containment in Correct Position SR 3.6.3.3 4.6.1.1.a Verify Isolation Time of Valves SR 3.6.3.5 ---

Cycle Weight/Spring Loaded Check Valves SR 3.6.3.6 ---

Perform Leak Rate Test of Purge Valves SR 3.6.3.7 ---

Verify Automatic Valves Actuate to Correct Position SR 3.6.3.8 4.6.3.2.a, b & c Cycle Non Testable Weight/Spring Loaded Check Valves SR 3.6.3.9 ---

Verify Purge Valves Blocked SR 3.6.3.10 ---

Containment Pressure Verify Pressure SR 3.6.4.1 4.6.1.4 Containment Air Temperature Verify Average Air Temperature SR 3.6.5.1 4.6.1.5 Spray Systems Verify Valve Position SR 3.6.6D. 1 4.6.2.1.a. 1)

Verify Valve Actuation SR 3.6.6D.3 4.6.2.1.c.1)

Verify Pump Start on Auto Signal SR 3.6.6D.4 4.6.2.1.c.2)

Verify Nozzle are not Obstructed SR 3.6.6D.5 Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 8 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Recirculation Spray Verify Casing Cooling Temperature SR 3.6.6E.1 ---

Verifying Casing Cooling Volume SR 3.6.6E.2 ---

Verify Casing Cooling Boron Concentration SR 3.6.6E.33---

Verify Valve Position SR 3.6.6E.4 4.6.2.2.a Verify Actuation of Pumps and Valves SR 3.6.6E.6 4.6.2.2.c & d Verify Nozzle are not Obstructed SR 3.6.6E.7 ---

Spray Additive System Verify Valve position SR 3.6.7.1 Note 1 Verify Tank Volume SR 3.6.7.2 Note 1 Verify Tank Solution Concentration SR 3.6.7.3 Note 1 Actuate Each Flow Path Valve SR 3.6.7.4 Note 1 Verify Spray Additive Flow Rate SR 3.6.7.5 Note 1 Iodine Cleanup System Operate train with heaters SR 3.6.11.1 Note 1 Verify train Actuation SR 3.6.11.3 Note 1 Verify Filter Bypass Operation SR 3.6.11.4 Note 1 Steam Jet Air Ejectors Verify Air Ejectoroutside ContainmentIsolation Valve Closed --- 4.6.5.1.1 Supplementary Leak Collection and Release System MPS3 Verify Manual Train Actuation & Operate Heaters SR 3.6.13.1 4.6.6.1.a Verify FilterPenetrationand Bypass Leakage --- 4.6.6.1 .b.1)

Verify FilterPressureDrop --- 4.6.6.1.d.1)

Verify Acutation on Safety Injection Signal SR 3.6.13.3 4.66.61. d.2)

Verify HeaterCapacity --- 4.6.6.1 d.3)

Verify Damper Open SR 3.6.13.4 Verify Each train Flow SR 3.6.13.5 4.6.6.1.a./4.66.61.b.3)

Secondary Containment (MPS3)

Verify Doors in Access Opening are Closed --- 4.6.6.2.1 Verify Each SLCR System Produce Negative Pressure --- 4.6.6.2.2 Main Steam Isolation Valves Actuate Valves SR 3.7.2.2 ---

MFIVs and MFRVs Actuate Valves SR 3.7.3.2 ---

Atmospheric Dump Valves -

Cycle Dump Valves SR 3.7.4.1 4.7.1.6.1 Cycle Block Valves SR 3.7.4.2 4.7.1.6.2 Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 9 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

AFW Verify Valve Position SR 3.7.5.1 4.7.1.2.1.a Verify Auto Valve Actuation SR 3.7.5.3 ---

Verify Pump Auto Actuation SR 3.7.5.4 4.7.1.2.1 .c Verify Pump Head --- 4.7.1.2.1.b.1 & 2 Condensate Storage Tank Verify Volume of DWST SR 3.7.6.1 4.7.1.3.1 Component Cooling Verify Valve Position SR 3.7.7.1 4.7.3.a Verify Valve Actuation SR 3.7.7.2 4.7.3.b.1)

Verify Pump Actuation SR 3.7.7.3 4.7.3.b.2)

Service Water Verify Valve Position SR 3.7.8.1 4.7.4.a Verify Valve Actuation SR 3.7.8.2 4.7.4.b.1)

Verify Pump Actuation SR 3.7.8.3 4.7.4.b.2)

Ultimate Heat Sink Verify Water Level SR 3.7.9.1 ---

Verify Water Temperature SR 3.7.9.2 4.7.5.a Operate Cooling Tower SR 3.7.9.3 Note 1 Verify Fan Actuation SR 3.7.9.4 Note 1 CR Emergency Ventilation Verify Control Room Air Temperature --- 4.7.7.a Verify Manual Train Actuation Operate Heaters SR 3.7.10.1 4.7.7. b Verify Train Actuation Actual or Simulated Signal SR 3.7.10.3 ---

Verify Envelope Pressurization SR 3.7.10.4 ---

Verify FilterPenetrationand Bypass Leakage --- 4.7.7.c.1)

Verify System Flow Rate --- 4.7.7.b/4.7.7.c.3)

Verify FilterPressureDrop --- 4.7.7.e. 1)

Verify Heater Capacity --- 4.7.7.e.3)

CR Air Condition System Verify Train Capacity SR 3.7.11.1 ---

ECCS PREACS (MPS3 Auxiliary Building Filter System)

Verify Manual Train Actuation and Operate Heaters SR 3.7.12.1 4.7.9.a Verify Train Actuation Actual or Simulated Signal SR 3.7.12.3 4.7.9.d.2)

Verify FilterPenetrationand Bypass Leakage --- 4.7.9.b. 1)

Verify System Flow Rate --- 4.7.9.a/4.7.9.b.3)

Verify Envelope Negative Pressure SR 3.7.12.4 Verify Pressure Drop Across HEPA and Adsorbers --- 4.7.9.d. 1)

Note 1 - This system is not included in the MPS3 design or TS.

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Serial No.10-711 Docket No. 50-423 Attachment 4 Page 10 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Verify Bypass Damper Closure SR 3.7.12.5 ---

Verify HeaterCapacity --- 4.7.9.d.3)

Fuel Building Air Cleanup Operate Heaters SR 3.7.13.1 ---

Verify Automatic Train Actuation SR 3 7.13.3 ---

Verify Envelope Negative Pressure SR 3.7.13.4 ---

Verify Bypass Damper Closure SR 3.7.13.5 ---

Penetration Room Air Cleanup System -

Operate Heaters SR 3,7.14.1 Note 1 Verify Automatic Train Actuation SR 3.7.14.3 Note 1 Verify Envelope Negative Pressurization SR 3.7.14.4 Note 1 Verify Bypass Damper Closure SR 3.7.14.5 Note 1 Fuel Storage Pool Water Level Verify Water Level SR 3.7.15.1 4.9.11 Fuel Storage Pool Boron Verify Boron Concentration SR 3.7.16.1 4.9.1.2 Secondary Specific Activity Verify Secondary Activity SR 3.7.18.1 4.7.1.4 Area Temperature Monitoring (MPS3)

Verify Temperature is Within Limits --- 4.7.14.a AC Sources -Operating Verify Breaker Alignment Offsite Circuits SR 3.8.1.1 4.8.1.1.1 .a Verify EDG Starts - Achieves Voltage & Frequency SR 3.8.1.2 4.8.1.1.2.a.5)

Synchronize and Load for > 60 minutes SR 3.8.1.3 4.8.1.1.2.a.6) &4.8.1.1.2.b.2)

Verify Day Tank Level SR 3.8.1.4 4.8.1.1.2.a.1)

Remove Accumulate Water for Day Tank SR 3.8.1.5 4.8.1.1.2.c Verify Operation of Transfer Pump SR 3.8.1.6 4.8.1.1.2.a.3)

Verify EDG Starts - Achieves Voltage & Frequency in SR 3.8.1.7 4.8.1.1.2.b.1) 10 seconds Verify Auto and Manual Transfer of AC power Sources SR 3.8.1.8 4.8 11.1 b

- Offsite Sources Verify the EDG alignment for standby power --- 4.8.1.1.2.a.7)

Verify Largest Load Rejection SR 3.8.1.9 48.1.1.2.g.2)

Verify EDG Does Not Trip with Load Rejection SR 3.8.1.10 4.8.1.1.2.g.3)

Verify De-energize, Load Shed and Re energize SR 3.8.111 4.8.11.2.g.4)a) & b)

Emergency Bus with Loss of Offsite Power Verify EDG Start on ESF Signal SR 3.8.1.12 4.8.1.1.2.g.5)

Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 11 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Verify EDG Noncritical Trips are Bypassed SR 3.8.1.13 4.8.1.1.2.g.6)c)

Verify the Lockout FeaturesPrevent EDG Starting --- 4.8.1.1.2.1 Run EDG for 24 Hours SR 3.8.1.14 4.8.1.1.2.j Verify EDG Starts Post Operation - Achieves Voltage & SR 3.8.1.15 Frequency Verify EDG Synchronizes w/Offsite Power and SR 3.8.1.16 4811.2.g.9) a), b) &c)

Transfers Load Verify ESF Signal overrides Test Mode of EDG SR 3.8.1.17 4.8.1.1.2. g.10)

Verify Load Sequencers are with Design Tolerance SR 3.8.1.18 4.8.1.1.2. g. 12)

Verify EDG Start on Loss of Offsite Power with ESF SR 3.8.1.19 4.8.1.1.2.g.6)a) & b)

Verify When Started Simultaneously from Standby SR 3.8.1.20 4.8.1.1.2.h Each EDGs Reach Rated Voltage and Frequency Verify auto-connected loads are < 5335kW --- 4.8.1.1.2.g.8)

Diesel FO and Starting Air Verify FO Storage Tank Volume SR 3.8.3.1 4.8.1.1.2.a.2)

Verify Lube Oil Inventory SR 3.8.3.2 4.8.1.1.2.a.4)

Verify EDG Air Start Receive Pressure SR 3.8.3.4 ---

Check and Remove Accumulate Water from FO Tanks SR 3.8.3.5 4.8.1.1.2.d Verify Total Particulate <10mg/liter --- 4.8.1.1.2.f Verify Operation of FO Transfer Pumps via the Installed 4.8.11.2.k Cross-ConnectLines Clean FO Storage Tanks --- 4.8.1.1.2.i DC Sources Operating Verify Battery Terminal Voltage SR 3.8.4.1 4.8.2.1.a.2)

Verify Station Battery Chargers Capable of Supplying SR 3.8.4.2 4.8.2.1 c.4)

[x]Amp for [y]Hours Verify Battery Capacity SR 3.8.4.3 4.8.2.1.d Verify No Visable Corrosion at Terminal or Connectors --- 4.8.2. 1 b.2) and resistance > xx ohms Verify No Visual Indication of Physical Damage --- 4.8.2.1.c. 1)

Cell-to-Cell And Terminal Connections Clean & Tight --- 4.8.2.1.c.2)

Verify Cell-to-Cell Resistance is < xx ohms --- 4.8.2.1 .c.3)

Battery Parameters Verify each Battery Float Current is < [2] amps. SR 3.8.6.1 4.8.2.1.a.1 & b.1)

Verify each Battery Pilot Cell Voltage is >[2.07] V SR 3.8.6.2 4.8.2.1.a.1 & b.1)

Verify each Battery Cell Electrolyte Level is > to SR 3.8.6.3 4.8.2.1.a.1 & b.i)

Minimum Design Limits.

Verify each Battery Pilot Cell Temperature > to SR 3.8.6.4 4.8.2.1.b.3)

Minimum Design Limits.

Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 Attachment 4 Page 12 of 12 Technical Specification Section Title/ TSTF 425 MPS3 Surveillance Description*

Verify Each Battery Connected Cell Voltage is>[2.07] V. SR 3.8.6.5 4.8.2.1.b.1)

Verify Station and EDG Battery Capacity - >80% After SR 3.8.6.6 4.8.2.1.e Performance Test Complete Performance Discharge Test if Signs of --- 4.8.2.1 .f Degradation or after reaching 85% of Service Life Inverters - Operating Verify Correct Inverter Voltage & Alignment to Required SR 3.8.7.1 AC Vital Buses.

Inverters - Shutdown Verify Correct Inverter Voltage & Alignment to Required SR 3.8.8.1 AC Vital Buses.

Distribution System - Operating Verify Correct Breaker Alignments & Voltage to AC/DC SR 3.8.9.1 4.8.3.1 and AC Vital Bus Electrical Distribution Subsystems.

Distribution System - Shutdown Verify Correct Breaker Alignments & Voltage to AC/DC SR 3.8.10.1 4.8.3.2 and AC Vital Bus Electrical Distribution Subsystems.

Boron Concentration Verify Boron Concentration is Within COLR Limit SR 3.9.1.1 4.9.1.1.2 Primary Grade Water Source Isolation Valves Verify Each Valve that Isolates Unborated Water SR 3.9.2.1 4.9.1.1.3 Sources is Secured in the Closed Position Nuclear Instrumentation Perform Channel Check SR 3.9.3.1 4.9.2.a Perform Channel Calibration SR 3.9.3.2 4.9.2.b Containment Penetrations Verify each Required Containment Penetration is in the SR 3.9.4.1 4.9.4.a Required Status.

Verify Each Required Containment Purge and Exhaust Valve Actuates to the Isolation Position on an Actuated SR 3.9.4.2 or Simulated Actuation Signal.

RHR and Coolant Circulation - High Water Level Verify One Loop is in Operation and Circulating Reactor SR 3.9.5.1 4.9.8.1 Coolant at a Flow Rate of > [2800] gpm.

RHR and Coolant Circulation - Low Water Level Verify One Loop is in Operation and Circulating Reactor SR 3.9.6.1 4.9.8.2 Coolant at a flow rate of > [2800] gpm.

Verify Correct Breaker Alignment and Indicated Power SR 3.9.6.2 Available to Required RHR Pump Not in Operation.

Refueling Cavity Water Level Verify Refueling Cavity Water Level is >23 ft Above The SR 3.9.7.1 4.9.10 Top of Reactor Vessel Flange.

Note 1 - This system is not included in the MPS3 design or TS.

--- Surveillance not included in ITS or MPS TSs Italicized text denotes MPS3-specific surveillances

Serial No.10-711 Docket No. 50-423 ATTACHMENT 5 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.10-711 Docket No. 50-423 Attachment 5 Page 1 of 2 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request:

This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for Westinghouse Pressurized Water Reactors (NUREG-1431), to allow relocation of specific technical specification (TS) surveillance frequencies to a licensee controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642), "Relocate Surveillance Frequencies to Licensee Control

- RITSTF Initiative 5b" and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).

The proposed changes are consistent with NRC-approved Industry/TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b." The proposed changes relocate surveillance frequencies to a licensee controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456). In addition, administrative/editorial deviations of the TSTF-425 inserts and the existing TS wording are being proposed to fit the custom TS format.

Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91 (a), the Dominion analysis of the issue of no significant hazards consideration is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.

Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.

The systems and components required by the TSs for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Serial No.10-711 Docket No. 50-423 Attachment 5 Page 2 of 2

2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements.

The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

3. Do the proposed changes involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Dominion will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the reasoning presented above, Dominion concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.

Serial No.10-711 Docket No. 50-423 ATTACHMENT 6 MARKED-UP TECHNICAL SPECIFICATIONS BASES CHANGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Insert 2 The surveillance frequency is controlled under the Surveillance Frequency Control Program.

LDBCR ,5-MP3-,25 MarchJqeJ. 7 , 2006~~s'*

REACTIVITY CONTROL SYSTEMS lat frequency specified in the Surveillance BASES lFrequency Control Program MOVABLE CONTROL ASSEMBLIES (Continued)

Control rod positions and OPERABILITY of the position indicators are required to be verified on a nemil hbasis of on-ce per 12 hou's with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

The Digital Rod Position Indication (DRPI) System is defined as follows:

  • Rod position indication as displayed on DRPI display panel (MB4), or
  • Rod position indication as displayed by the Plant Process Computer System.

With the above definition, LCO 3.1.3.2, "ACTION a." is not applicable with either DRPI display panel or the plant process computer points OPERABLE.

The plant process computer may be utilized to satisfy DRPI System requirements which meets LCO 3.1.3.2, in requiring diversity for determining digital rod position indication.

Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least @ncc cach 12 h, ur , except during the time when the rod position deviation monitor is inoperable, then o are the Demand Position Indication System and the DRPI System at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. the frequency specified in the Surveillance

  • Frequency Control Program I The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Trelmical Spcfcto SR 4.1.3.2.1 detcrminc2s each digital fod position indicator tob O-ERL*__E by v*.. ýin 1the Dtntiad Psition Indi.ation Sy-*tm and the DP"* System agr-,

wlithin 1-2 steps at least ecncc eaeh 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, emeept d~rig the time whe the td positiondeiaion mntris inperable, then 1 conipaIe theL Demlanld Position 1 Indicatiu Systini and the ORP system -At]ARA* non swch 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sw.

The Rod Deviation Monitor is generated ofnly &fra the DRPI panel at MB4. Therefore, w~henrd flieii peilfb1 inl SURVeILLANCqet *EQUIREMENf~TS eveiy 4 frm D1uplicative paarps ILLSTIONE - UNIT 3 B 3/4 1-4 Amendment No. i0,

LDDC9R 65-MP3-025 M a* hr1 , 26,6-REACTIVITY CONTROL SYSTEMS the frequency specified in the Surveillance Frequency Control Program BASES MOVABLE CONTROL ASSEMBLIES (Continued)d Additional surveillance is required to ensure the plant ocess computer indications are in agreement with those displayed on the DRPI. This a ditional SURVEILLANCE REQUIREMENT is as follows:

Each rod position indication as displ ed by the plant process computer shall be determined to be OPERABLE by rifying the rod position indication as displayed on the DRPI display panel agrees wi te rod position indication as displayed by the plant process computer at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The rod position indication, as displayed by DRPI display panel (MB4), is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a refueling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position. Additionally, the plant process computer provides the same type and level of accuracy as the DRPI System (MB4). The plant process computer does not provide any alarm or rod position deviation monitoring as does DRPI display panel (MB4).

For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent upon the plant to verify the trippability of the inoperable control rod(s). Trippability is defined in Attachment C to a letter dated December 21, 1984, from E. P. Rahe (Westinghouse) to C. 0. Thomas (NRC). This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism. In the event the plant is unable to verify the rod(s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a. Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for this verification.

For LCO 3.1.3.6 the control bank insertion limits are specified in the CORE OPERATING LIMITS REPORT (COLR). These insertion limits are the initial assumptions in safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions, assumptions of available SHUTDOWN MARGIN, and initial reactivity insertion rate.

The applicable I&C calibration procedure (Reference 1.) being current indicates the associated circuitry is OPERABLE.

There are conditions when the Lo-Lo and Lo alarms of the RIL Monitor are limited below the RIL specified in the COLR. The RIL Monitor remains OPERABLE because the lead control rod bank still has the Lo and Lo-Lo alarms greater than or equal to the RIL.

MILLSTONE - UNIT 3 B 3/4 1-5 Amendment No. 60,

LBDCR No. 06 MP3 0 14 J22.1,-2O06-POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

(2) APLND (for base load operation). Penalty deviation minutes for base load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as .peei.ie in Specicationc 1.2.2 and .2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided: in accordance with the Surveillance d Frequency Control Program./
a. Control rods in a single group move together with no individual rod inse ion differing by more than +/-12 steps, indicated, from the group demand position;
b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;
c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FNAH will be maintained within its limits provided Conditions a. through d. above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

The FNAH as calculated in Specification 3.2.3.1 is used in the various accident analyses where FNAH influences parameters other than DNBR, e.g., peak clad temperature, and thus is the maximum "as measured" value allowed.

The RCS total flow rate and FNAH are specified in the CORE OPERATING LIMITS REPORT (COLR) to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow rate, that is based on 10% steam generator tube plugging, is retained in the Technical Specifications.

MILLSTONE - UNIT 3 B 3/4 2-3 Amendment No. -50,60, 24-7,

LBDCR N11?4.. 08 MP3 014 O9eteber 21, 2008-POWER DISTRIBUTION LIMITS in accordance with the Surveillance BASES IFrequency Control Program 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RC FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continu d)

Margin is maintained between the safety analysis limit DNBR and t design limit DNBR. This margin is more than sufficient to offset the effect of rod bow and an other DNB penalties that may occur. The remaining margin is available for plant design flex* ility.

When an FQ measurement is taken, an allowance for both experimmntal error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a 11core map taken with the incore detector flux mapping system and a 3% allowance is appr nriate for manufacturing tolerance.

The heat flux hot channel factor, FQ(Z), is measured periodically sing the incore detector system.

These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive FQM(Z), a computed value of FQ(Z). However, because this value represents a steady state condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations.

To account for these possible variations, the steady state limit of FQ(Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(Z) or W(Z)BL, that accounts for the calculated worst case transient conditions. The W(Z) and W(Z)BL, factors described above for normal operation are specified in the COLR per Specification 6.9.1.6. Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state FQ(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

When RCS flow rate and FNAH are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation. Measurement errors for RCS total flow rate and for FNAH have been taken into account in determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. To perform the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated at least once per 18 months. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Any fouling which might bias the RCS flow rate measurement can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. 2, 60, 4-40, 24-7,

L=BDGR No. 04 MP3-01 5 Feznl*y 24, 2005 POWER DISTRIBUTION LIMITS BASESW lin accordance with the SurveillancecTFrequency Control Program HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW TE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The 12-h periodic surveillance of indicated RCS flo is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation defined in Specifications 3.2.3.1.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during POWER OPERATION.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in FQ is depleted. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. The indicated Tavg values MILLSTONE - UNIT 3 B 3/4 2-5 Amendment No. 2-2,-50,60, 2-4-,

Ae~knowledgcJd by NRC lemer dated 08/245/0

LBDCR t4-NMP3--62 Matreh 25, 2004-POWER DISTRIBUTION LIMITS BASES DNB PARAMETERS (Continued) and the indicated pressurizer pressure values are specified in the CORE OPERATING LIMITS REPORT. The calculated values of the DNB related parameters will be an average of the k indicated values for the OPERABLE channels.

The -42 hour periodic surveillance of these parameters through instrument readout* sufficient to ensure that the parameters are restored within their limits following load changes a other expected transient operation. Measurement uncertainties have been accounted for in determin the parameter limits.

lin accordance with the Surveillance Frequency Control ProgramI MILLSTONE - UNIT 3 B 3/4 2-6 Amendment No. 2, 60, 2-t-*,

- lNRC by bV kiki 08/25/ttda OI3.I.

LBD30cl No. UT-1vr.3-UM Feb~adry. 4 20....

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated action and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed atrh mini..mum frgucn.ieie. are sufficient to demonstrate this capability.

The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set for each functional unit. The Allowable Values (Nominal Trip Setpoints +/- the calibration tolerance) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to mitigate the consequences of accidents. A Setpoint is considered to be consistent with the nominal value when the measured "as left" Setpoint is within the administratively controlled (+) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.

Measurement and Test Equipment accuracy is administratively controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991.

OPERABILITY determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation.

The Allowable Value specified in Table 3.3-4 defines the limit beyond which a channel is inoperable. If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be OPERABLE.

MILLSTONE - UNIT 3 B 3/4 3-1 Amendment No. 4-59, A Am.... _ ._Ii440P 1i Ah !-IF-* 1 AR094

LBDCR 1o-fMP3-603 Febr3uary 213, 2010 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels. Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation.

The above Bases does not apply to the Control Building Inlet Ventilation radiation monitors ESF Table (Item 7E). For these radiation monitors the allowable values are essentially nominal values.

Due to the uncertainties involved in radiological parameters, the methodologies of WCAP- 10991 were not applied. Actual trip setpoints will be reestablished below the allowable value based on calibration accuracies and good practices.

The OPERABILITY requirements for Table 3.3-3, Functional Units 7.a, "Control Building Isolation, Manual Actuation," and 7.e, "Control Building Isolation, Control Building Inlet Ventilation Radiation," are defined by table notation "*". These functional units are required to be OPERABLE at all times during plant operation in MODES 1,2, 3, and 4. These functional units are also required to be OPERABLE during movement of recently irradiated fuel assemblies, as specified by table notation

"*". The Control Building Isolation Manual Actuation and Control Building Inlet Ventilation Radiation are required to be OPERABLE during movement of recently irradiated fuel assemblies (i.e.,

fuel that has occupied part of a critical reactor core within the previous 350 hour0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />s*). Table notation "*" 7 of Table 4.3-2 has the same applicability.

The verification of response time at the sp, eified frequenci*e provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analysis. No credit is taken in the analysis for those channels with response times indicated as not applicable (i.e., N.A.). " -- nsert Required ACTION 4. of Table 3.3-1 is modified by a Note to indicate that normal plant control operations that individually add limited positive reactivity (e.g., temperature or boron fluctuations associated with RCS inventory management or temperature control) are not precluded by this ACTION provided they are accounted for in the calculated SDM. The proposed change permits operations introducing positive reactivity additions but prohibits the temperature change or overall boron concentration from decreasing below that required to maintain the specified SDM or required boron concentration.

During fuel assembly cleaning evolutions that involve the handling or cleaning of two fuel assemblies coincidentally, recently irradiated fuel is fuel that has occupied part of a critical reactor core within the previous 525 hours0.00608 days <br />0.146 hours <br />8.680556e-4 weeks <br />1.997625e-4 months <br />. /

MILLSTONE - UNIT 3 B 3/4 3-2 Amendment No. 3,-9-1-, 4-59, 4-74, 4-8-4,2 19-,230O

Fb.... 24, 2Z,0.

INSTRUMENTATION Sin accordance with the Surveillance BASES IFrequency Control Program.-I 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATIO;ý/and ENGINEER-ED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT N (Continued)

For slave relays, or any auxiliary relays in ESFAS circuits thatae of the type Potter & Brumfield MDR series relays, the SLAVE RELAY TEST is performed-, an fley (at least @nee e,,veryff 18 rnon,) prov f~li tq elay* *ie*t the Ireiability assesirnJl Citalia p, eslnt 1 ,d in we7-.3.7, "R.liabiiy

..... .. ofuf... and D1 mi1 field MDR .... relays," and WCAF-i.;5o, "Extiiunsiu ofsf ,,.*,ay Siuzv,,,lanc Test , inte val." The r .liability assessments prormzefad as part cfthe afcr-ementioncd WCAP9 are relay speeifie and apply ofnly tc r[k alld DIgillld MDl 1 1o 1day . ttato 1 llally z liaIuII, ns t-lh r*lays aL a.l May havO tc b@ replacEd pcri@Edially inf accodfrdanc %ith th@ glidancc@ givenif in IXCAPR 13879 for.

Sf 1R reletys.

REACTOR TRIP BREAKER This trip function applies to the reactor trip breakers (RTBs) exclusive of individual trip mechanisms. The LCO requires two OPERABLE trains of trip breakers. A trip breaker train consists of all trip breakers associated with a single RTS logic train that are racked in, closed, and capable of supplying power to the control rod drive (CRD) system. Thus, the train may consist of the main breaker, bypass breaker, or main breaker and bypass breaker, depending upon the system configuration. Two OPERABLE trains ensure no single random failure can disable the RTS trip capability.

These trip functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip functions must be OPERABLE when the RTBs or associated bypass breakers are closed, and the CRD system is capable of rod withdrawal.

BYPASSED CHANNEL* - Technical Specifications 3.3.1 and 3.3.2 often allow the bypassing of instrument channels in the case of an inoperable instrument or for surveillance testing.

A BYPASSED CHANNEL shall be a channel which is:

Required to be in its accident or tripped condition, but is not presently in its accident or tripped condition using a method described below; or Prevented from tripping.

MILLSTONE - UNIT 3 B 3/4 3-2b Amendment No. 2-4-9, Ak--. l.d.. d b, NRC ltter--dated 025,'05

INSTRUMENTATION 3/4.3.5 SHUTDOWN MARGIN MONITOR BASES (continued)

Required ACTION b. is modified by a Note which permits plant temperature changes provided the temperature change is accounted for in the calculated SDM. Introduction of temperature changes, including temperature increases when a positive MTC exists, must be evaluated to ensure they do not result in a loss of required SDM.

2. All dilution flowpaths are isolated and placed under administrative control (locked closed). This action provides redundant protection and defense in depth (safety overlap) to the SMMs. In this configuration, a boron dilution event (BDE) cannot occur. This is the basis for not having to analyze for BDE in MODE 6. Since the BDE cannot occur with the dilution flow paths isolated, the SMMs are not required to be OPERABLE as the event cannot occur and OPERABLE SMMs provide no benefit.
3. Increase the SHUTDOWN MARGIN surveillance frequency from ery"24--hotrr to every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This action in combination with the above, provi defense in depth and overlap to the loss of the SMMs. i Surveillance Requirements the frequency specified in the SSurveillance Frequency Control Program The SMMs are subject to an AGOe;veT "92 dayR to ensure each train of SMM is fully operational. This test shalle de verification that the SMMs are set per the CORE OPERATING LIMITS ORT. c--Finsert-2-1

[ANALOG CHANNEL OPERATIONAL TEST MILLSTONE - UNIT 3 B 3/4 3-9 Amendment No. 4-64, 2-30

¥ T*T*.f*T* "ILT_ J*,l I KT*"*) a**l L"*

i_,i.Jj.l._,il f1s . tl--dt ii',, t

".J Felbruar-" 21, 2005 3/4.4 REACTOR COOLANT SYSTEM BASES The safety analyses performed for the reactor at power assume that all reactor coolant loops are initially in operation and the loop stop valves are open. This LCO places controls on the loop stop valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3 and 4.

The inadvertent closure of a loop stop valve when the Reactor Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow. If the reactor is at rated power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and a RCS loop is in operation removing decay heat, closure of the loop stop valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.

The loop stop valves have motor operators. If power is inadvertently restored to one or more loop stop valve operators, the potential exists for accidental closure of the affected loop stop valve(s) and the partial loss of forced reactor coolant flow. With power applied to a valve operator, only the interlocks prevent the valve from being operated. Although operating procedures and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop stop valve operators. The time period of 30 minutes to remove power from the loop stop valve operators is sufficient considering the complexity of the task.

Should a loop stop valve be closed in MODES 1 through 4, the affected valve must be maintained closed and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.1.6, "Reactor Coolant System Isolated Loop Startup." Opening the closed loop stop valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The time period provided in ACTION 3.4.1.5.b allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed ACTION times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirement 4.4.1.5 is performed at st . .. 31 das to ensure that the 3e RCS loop stop valves are open, with power removed from the loop stop valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since Surveillance Requirement 4.4.1.1 requires verification every-l2-houm that all loops are operating and circulating reactor coolant, thereby ensuring that the loop stop valves are open. The frequency - ensures that the required flow is available ia bacdcball nzrirg 4-lb Amendment No. 60, 40, 99, 44-5, 4-9*, 20, 2-2-4, A.ck 1 1wikdgd by NRC lctte1 dated 08,'25/05

-96ý2O6-3/4.4 REACTOR COOLANT SYSTEM BASES (continued)

For the isolated loop being restored, the power to both loop stop valves has been restored Surveillance 4.4.1.6.2 indicates that the reactor shall be determined subcritical by at least the amount required by Specifications 3.1.1.1.2 or 3.1.1.2 for MODE 5 or Specification 3.9.1.1 for MODE 6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of opening the cold leg or hot leg stop valve.

The SHUTDOWN MARGIN requirement in Specification 3.1.1.1.2 is specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops filled. Specification 3.1.1.1.2 cannot be used to determine the required SHUTDOWN MARGIN for MODE 5 loops isolated condition.

Specification 3.1.1.2 requires the SHUTDOWN MARGIN to be greater than or equal to the limits specified in the CORE OPERATING LIMITS REPORT for MODE 5 with RCS loops not filled provided CVCS is aligned to preclude boron dilution. This specification is for loops not filled and therefore is applicable to an all loops isolated condition.

Specification 3.9.1.1 requires Keff of 0.95 or less, or a boron concentration of greater than or equal to the limit specified in the COLR in MODE 6.

Specification 3.1.1.1.2 or 3.1.1.2 for MODE 5, both require boron concentration to be determined at I.Pas, . a2 ,. SR 5Rt2**

4.1.1.1.2.1 .b.2 and 4.1.1.2.1.b.l satisfy the requirements of Speci 1 .1.1.1.2 and 3.1.1.2 respectfully. Specification 3.9.1.1 for MODE 6 requires boron concentration to be d at m.ist

, ta.h 72 hur-. S.R. 4.9.1.1.2 satisfy the requirements of Specification 3.9.1.1. [the frequency specified in the Surveillance Frequency Control Program Per Specifications 3.4.1.2, ACTION c.; 3.4.1.3, ACTION c.; 3.4.1.4.1, ACTION b.; and 3.4.1.4.2, ACTION b., suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1.2 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

References:

1. Letter NEU-94-623, dated July 13, 1994; Mixing Evaluation for Boron Dilution Accident in Modes 4 and 5, Westinghouse HR-59782.
2. Memo No. MP3-E-93-821, dated October 7, 1993.

MILLSTONE - UNIT 3 B 3/4 4-1If Amendment No. 2-4-7, 230

REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (continued) Insert 2 The t2"rrmr periodic surveillances quire that pressurizer level be maintained at programmed level within +/- 6% of full sc . The surveillance is performed by observing the indicated level. The 12 het ~itfterfvval hats ccri she by operotng practiee to be 3i~ffieien ttsses regelfrly ." leveltransitory During for nfy deyiattion:_arA to eznqsimrec that the.apprpriattc level existsi inth conditions, i.e., power changes, the operators will maintain programmed level, and deviations greater than 6% will be corrected within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Two hours has been selected for pressurizer level restoration after a transient to avoid an unnecessary downpower with pressurizer level outside the operating band. Normally, alarms are also available for early detection of abnormal level indications.

Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure.

A minimum required available capacity of pressurizer heaters ensures that the RCS pressure can be maintained. The capability to maintain and control system pressure is important for maintaining subcooled conditions in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of the reactor coolant. Unless adequate heater capacity is available, the hot high-pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system. Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to a loss of single-phase natural circulation and decreased capability to remove core decay heat.

The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity of at least 175 kW. The heaters are capable of being powered from either the offsite power source or the emergency power supply. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. The requirement for two groups of pressurizer heaters, each having a capacity of 175 kW, is met by verifying the capacity of the pressurizer heater groups A and B. Since the pressurizer heater groups A and B are supplied from the emergency 480V electrical buses, there is reasonable assurance that these heaters can be energized during a loss of offsite power to maintain natural circulation at HOT STANDBY. Providing an emergency (Class 1E) power source for the required pressurizer heaters meets the requirement of NUREG-0737, "A Clarification of TMI Action Plan Requirements," II.E.3.1, "Emergency Power Requirements for Pressurizer Heaters."

If one required group of pressurizer heaters is inoperable, restoration is required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. 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 reasonable considering that a demand caused by loss of offsite power would be unlikely in this time period. Pressure control may be maintained during this time using normal station powered heaters.

MODE 3 The requirement for the pressurizer to be OPERABLE, with a level less than or equal to 89%, ensures that a steam bubble exists. The 89% level preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure control for MODE 3 and to ensure a bubble is present in the pressurizer. Initial pressurizer level is not significant for those events analyzed for MODE 3 in Chapter 15 of the FSAR.

MILLSTONE - UNIT 3 B 3/4 4-2a Amendment No. 460,-249

LtiilcR fu. 04-MVIF3-Mi5 ftbiuamy 24, 205 REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER (cont'd.) Insert 2 The 12 he periodic surveillance requires that during MODE 3 op ation, pressurizer level is maintained below the nominal upper limit to provide a minimu /ace for a steam bubble. The surveillance is performed by observing the indicated level. R.c 12-1-h-Ea- interval has bee showvn b, sppfiatu pr~n ieea;. to be 9"ffieie..t to reg..lal ases level. fer an~y devatin d to

.ns... that a steam b.ub-ble cxis in t*h p......izr Alarms are also available for early detection of abnormal level indications.

The basis for the pressurizer heater requirements is identical to MODES 1 and 2.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is met.

ACTION statements a, b, and c distinguishes the inoperability of the power operated relief valves ,"

(PORV). Specifically, a PORV may be designated inoperable but it may be able to automatically and manually open and close and therefore, able to perform its function. PORV inoperability may be due to seat leakage which does not prevent automatic or manual use and does not create the possibility for a small-break LOCA. For these reasons, the block valve may be closed but the action requires power to be maintained to the valve. This allows quick access to the PORV for pressure control. On the other hand if a PORV is inoperable and not capable of being automatically and manually cycled, it must be either restored or isolated by closing the associated block valve and removing power.

Note: PORV position indication does not affect the ability of the PORV to perform any of its safety functions. Therefore, the failure of PORV position indication does not cause the PORV to be inoperable. However, failed position indication of these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3.

Automatic operation of the PORVs is created to allow more time for operators to terminate an Inadvertent ECCS Actuation at Power. The PORVs and associated piping have been demonstrated to be qualified for water relief. Operation of the PORVs will prevent water relief from the pressurizer safety valves for which qualification for water relief has not been demonstrated. If the PORVs are capable of automatic operation but have been declared inoperable, closure of the PORV block valve is acceptable since the Emergency Operating Procedures provide guidance to assure that the PORVs would be available to mitigate the event.

OPERABILITY and setpoint controls for the safety grade PORV opening logic are maintained in ,4 the Technical Requirements Manual.

MILLSTONE - UNIT 3 B 3/4 4-2b Amendment No. -160,4 Ackniowuedged by NRC tiutit daitd G8/2j/05

_,iiV.ei fTito. tity-Mi titi-.)iyi May " 2"006 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in RCS LCO 3.4.6. 1, "Leakage Detection Systems."

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

Thu 72 huui F-1 qUlmal - l~~iabik iiitet v a! to trend LEAKAG E and regiirzs th--

impet!anc of early leakage detcltieU ii the prevntion

  • iLdentl*of .

4.4.6.2.1.e nsert 2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated.

The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG If it is not practical to assign the LEAKAGE to an individual SQ all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. nsert 27 Tlhe Sauxvvttlneu F*equenicy of 72 houuis is a izvauiiubk juL 1v tn t1i 1 d pu* iniuay to sereodtay LEAKAGE end meeg~izes the impCoAflc of early leakage detzztion in the preyentfiet of

-acidenits- The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5).

4.4.6.2.2 The Surveillance Requirements for RCS pressure isolation valves provide assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

MILLSTONE - UNIT 3 B 3/4 4-4g -Amettdment+ý. r

LBDCR Nm. m8-MF3-0mi 3 Marlch 1s, 208 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

ACTIONS (Continued) e.

If the required action and completion time of ACTION d. is not met, the reactor must be brought to HOT STANDBY (MODE 3) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN (MODE 5) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS 4.4.8.1 Ithe frequency specified in the Surveillance Frequency Control Program Surveillance Requirement 4.4.8.1 requires performing aamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at F-Very 7 day . This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken.

This Surveillance Requirement provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance Requirement allows proper remedial action to be taken before reaching the LCO limit under normal operating cond.thons. ofmve between 2n7d6y--

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sampedue to masking from radioisotopes with similar decay energies, such as F- 18 and 1-134, it is acce ble to include the minimum detectable activity for Kr-85 in the Surveillance Requirement 4.4.8. alculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-nt is not detected, it should be assumed to be present at the minimum detectable activity. Insert 2]

A Note modifies the Surveillance Requirement to allow entry into and tl'on in MODE 4, MODE aft~ern 3, and a pow chane MODE er- 2 prior to performing the lo p5%rTbabwithi ea1 ou erioSurveillance f ei stbihdbcue h allows R q*,ahment. This oievlthe Surveillance Requirement to be performed in thoe O prior to entering MODE 1.

4.4.8.2 . .. /

This Surveillance Reqie efre oenueidn pcfc activity remains within the LCO limit during n operation. and following fast power changes when iodine spiking is more apt to occur. .1,IN .1 4 day kfiAeqienicy is adeqtUate tv ft~c,C *,idchSenohi. tilt iodiiie actliV t 1,3

.V l,

.... ;noerin ble g..ii...

arivt ar;, is. m-esa-iave weryD -7.&ys. The frequency of between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power change _>15% RTP within a I hour period is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

MILLSTONE - UNIT 3 B 3/4 4-6b Amendment No. .

REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)

The AOT in MODE 4 considers the facts that only one of the relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. The RCS must be depressurized and a vent must be established within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the required relief valve is not restored to OPERABLE within the required AOT of 7 days.

d.

The consequences of operational events that will overpressure the RCS are more severe at lower temperatures (Ref. 8). Thus, with one of the two required relief valves inoperable in MODE 5 or in MODE 6 with the head on, the AOT to restore two valves to OPERABLE status is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The AOT represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE relief valve to protect against overpressure events. The RCS must be depressurized and a vent must be established within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the required relief valve is not restored to OPERABLE within the required AOT of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e.

The RCS must be depressurized and a vent must be established within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when both A' required Cold Overpressure Protection relief valves are inoperable.

The vent must be sized > 2.0 square inches to ensure that the flow capacity is greater than that required for the worst case cold overpressure transient reasonable during the applicable MODES.

This action is needed to protect the RCPB from a low temperature overpressure event and a possible non-ductile failure of the reactor vessel.

The time required to place the plant in this Condition is based on the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Performance of an ANALOG CHANNEL OPERATIONAL TEST is required within 31 days prior to entering a condition in which the PORV is required to be OPERABLE a enve*y 3--dys on each required PORV to verify and, as necessary, adjust its lift setpoint. Th ALOG CHANNEL OPERATIONAL TEST will verify the setpoint in accordanc ith the nominal values given in Figures 3.4-4a and 3.4-4b. PORV actuation could de ssurize the RCS; therefore, valve operation is not required.

MILLSTONE - UNIT 3 B 3/4 4-24 Amendment No. 5, +9-lat the frequency specified in the Surveillance Frequency Control Program thereafter

tlttct flu. fJ7-M?1P3-W juiue 19, 2007-REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (

Performance ]BRATION (on each required PORV ac tion channel is required 0-,- niont, the required range and accuracy to a known to adjust the channel so that it responds and the input. .7 lve opens within The PORV block valve must be verified open and ,'OPPS must be verified armed ty-72-lbn to provide a flow path and a cold overpressure prol ction actuation circuit for each required PORV to perform its function when required. The alve is remotely verified open in the main control room. This Surveillance is performed if cr dit is being taken for the PORV to satisfy the LCO.'Insert 2 The block valve is a remotely controlled, mot ope ted valve. The power to the valve operator is not required to be removed, and the ma 1operat r is not required to be locked in the open position. Thus, the block valve can be sed in the ent the PORV develops excessive leakage or does not close (sticks open) after ieving an ove essure transient.

I,,1 I IIIUL I%'%iulA1%y a %,JJ3%J a , M III VIV YV I~ UI%'L a~~i111 a CLI V%1 %,%ill Vi a v UAUY t It? ~lt t~tJ~lEAI II LlA. t'VI~ttJII~fJ~h, O~t'I tO VLAV jfJO ~l il~S~ttILtIl ~I tit. VtI~t LIt' 4.4.9.3.2 Each required RHR suction relief valve shall be demonstra d OPERABLE by verifying the RHR suction valves, 3RHS*MV8701A and 3RHS*M8701C, are pen when suction relief valve 3RHS*RV8708A is being used to meet the LCO and by ver ing the RHR suction valves, 3RHS*MV8702B and 3RHS*MV8702C, are open when suc ion relief valve 3RHS*RV8708B is being used to meet the LCO. Each required RIIR suction reli f valve shall also be demonstrated OPERABLE by testing it in accordance with 4.0.5. This Su illance is only required to be performed if the RHR suction relief valve is being used to me this LCO.

, -Periodi7caly The RHR suction valves are 4erified to be open every-12 -hourm. "fivei1mer-m-a e are *,y ew e or er ft rn " orm ve efintra 4 sue tis vol ve stan- "

~~~~1~~~~ ----

I] f"To zesreAtr int 0th eentrcl room that Yeritv the IU-IR 9tuteiett yalvc Demft z.

The ASME Code for Operation and Maintenance of Nuclear Power Plants, (Reference 9), test per 4" 4.0.5 verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.

MILLSTONE - UNIT 3 B 3/4 4-25 Amendment No. 4-1, 9, 2-06,

JLUJL)JL'*.i JI. V I -- viL J-U*JJ7 Rine,, 19, 2007 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) 4.4.9.3.3 periodically The RCS vent of> 2.0 square inches is proven OPERABLE by verifying its open condition.

it. "_imuI c 2.ho f*gbi a~ vent vatII YdVe that. cann~t be, locked ope./ll other. rag*_'ve wont path.ýA remfoved Pressurizer satety valve tits this category.

This passive vent arrangement must only be open to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LO. < Insert 2 4.4.9.3.4 and 4.4.9.3.5 To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all SIH pumps and all but one centrifugal charging pump are verified incapable of injecting into the RCS.

The SIH pumps and charging pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control.

Alternate methods of control may be employed using at least two independent means to prevent an injection into the RCS. This may be accomplished through any of the following methods:

1) placing the pump in pull to lock (PTL) and pulling its UC fuses, 2) placing the pump in pull to lock (PTL) and closingthe pump discharge valve(s) to the injection line, 3) closing the pump discharge valve(s) to the injection line and either removing power from the valve operator(s) or locking manual valves closed, and 4) closing the valve(s) from the injection source and either removing power from the valve operator(s) or locking manual valves closed.

An SIH pump may be energized for testing or for filling the Accumulators provided it is incapable of injecting into the RCS.

Tie Fiegueiiey of ,2 .uu.* ii ,uiidiin, uL..,, .,..ii and. alarms X1 eu.,,io,., ......... bl. to the REFERENCES n sert2

1. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G "Fracture Toughness for Protection Against Failure," 1995 Edition.
2. ASME Section XI, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," dated February 26, 1999.
3. Generic Letter 88-11
4. ASME, Boiler and Pressure Vessel Code,Section III
5. FSAR, Chapter 15
6. 10CFR50, Section 50.46
7. 10CFR50, Appendix K
8. Generic Letter 90-06
9. ASME Code for Operation and Maintenance of Nuclear Power Plants MILLSTONE - UNIT 3 B 3/4 4-26 Amendment No. +54,+W,

LBDCKt flu. 06-MF3-02 Atigt~s 10, 2006 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continue periodically flush upon he xchanger return to service and procedural compliance is relied upon to ensureehas is not present within the heat exchanger u-tubes.

Surveil

  • e Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed -di if the containment has been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure. < Insert 2 Surveillance Requirement 4.5.2.d.2 addresses periodic inspectiog; etcontainment sump to ensure that it is unrestricted and stays in proper operating condition. the 24* me th rcgucny is based ont the need to pzrform +his cilaoc ZortLe~ic -filtion, that !Fy~ duMu h'., 4n the need to-ha-vo to -.ec hoc e Ti.

aiid i* L~uIufI1IId by upc~nianig cpie~icnv. his f~requerny is 9 Iffi-eiont. t"M -~c~

-or-admM d 4 t.

The Emergency Core Cooling System (ECCS) has several piping cross connection points for use during the post-LOCA recirculation phase of operation. These cross-connection points allow the Recirculation Spray System (RSS) to supply water from the containment sump to the safety injection and charging pumps. The RSS has the capability to supply both Train A and B safety injection pumps and both Train A and B charging pumps. Operator action is required to position valves to establish flow from the containment sump through the RSS subsystems to the safety injection and charging pumps since the valves are not automatically repositioned. The quarterly stroke testing (Technical Specification 4.0.5) of the ECC/RSS recirculation flowpath valves discussed below will not result in subsystem inoperability (except due to other equipment manipulations to support valve testing) since these valves are manually aligned in accordance with the Emergency Operating Procedures (EOPs) to establish the recirculation flowpaths. It is expected the valves will be returned to the normal pre-test position following termination of the surveillance testing in response to the accident. Failure to restore any valve to the normal pre-test position will be indicated to the Control Room Operators when the ESF status panels are checked, as directed by the EOPs. The EOPs direct the Control Room Operators to check the ESF status panels early in the event to ensure proper equipment alignment. Sufficient time before the recirculation flowpath is required is expected to be available for operator action to position any valves that have not been restored to the pretest position, including localmanual valve operation. Even if the valves are not restored to the pre-test position, sufficient capability will remain to meet ECCS post-LOCA recirculation requirements. As a result, stroke testing of the ECCS recirculation valves discussed below will not result in a loss of system independence or redundancy, and both ECCS subsystems will remain OPERABLE.

When performing the quarterly stroke test of 3SIH*MV8923A, the control switch for safety injection pump 3SIH*PIA is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*P1A in ull-to-lock, the Train A ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a., applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*PIA defeated to allow stroke testing of 3SIH*MV8923A. In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated). During the injection phase the redundant subsystem (Train B) is frilly functional, as is a significant portion of the Train A subsystem. During the recirculation phase, the Train A RSS subsystem can supply water from the containment sump to the Train A MILLSTONE - UNIT 3 B 3/4 5-2b Amendment No. 400, 44-7, 4-5-4

EMERGENCY CORE COOLING SYSTEMS BASES (Continued)

APPLICABILITY In MODES 1, 2, 3, and 4, a design basis accident (DBA) could lead to a fission product release to containment that leaks to the secondary containment boundary. The large break LOCA, on which this system's design is based, is a full-power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and reactor coolant system pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.

In MODES 5 and 6, the probability and consequence of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the SLCRS is not required to be OPERABLE.

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 7-day Completion Time is based on the low probability of a DBA occurring during this period. The Completion Time is adequate to restore the volume of TSP to within the technical specification limits.

If the TSP cannot be restored within limits within the 7-day Completion Time, the plant must be brought to a MODE in which the LCO does not apply. The specified Completign Times for reaching MODES 3 and 4 are those used throughout the technical specifications; they were chosen to allow reaching the specified conditions from full power in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS Surveillance Requirement 4.5.5 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 . tu,,y .. ,,fonce per .. 24 ,th- is t-qtli- d to determince visually that a mitnimum of 9714 ubie feet is ccntoaincd in the T-SP Storage 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 > 7.0.

haskets is only feasiblew during outages. Operating ep . Ience -h-019-Aohc-Wn Hi Sfif Frequeney tteepkbl e~de to the inrni the. Itm of TH"P1 placed in the conitainmentii bailding.

MILLSTONE - UNIT 3 B 3/4 5-5 Amendment No. 44-S, 206

T "T' fT"1']~ XT- AnC A Xr'"q AA*{*'

CONTAINMENT SYSTEMS BASES The design of the Containment RSS is sufficiently independent so that an active failure in the recirculation spray mode, cold leg recirculation mode, or hot leg recirculation mode of the ECCS has no effect on its ability to perform its engineered safety function. In other words, the failure in one subsystem does not affect the capability of the other subsystem to perform its designated safety function of assuring adequate core cooling in the event of a design basis LOCA. As long as one subsystem is OPERABLE, with one pump capable of assuring core cooling and the other pump capable of removing heat from containment, the RSS system meets its design requirements.

The LCO 3.6.2.2. ACTION applies when any of the RSS pumps, heat exchangers, or associated components are declared inoperable. All four RSS pumps are required to be OPERABLE to meet the requirements of this LCO 3.6.2.2. During the injection phase of a Loss Of Coolant Accident all four RSS pumps would inject into containment to perform their containment heat removal function. The minimum requirement for the RSS to adequately perform this function is to have at least one subsystem available. Meeting the requirements of LCO 3.6.2.2. ensures the minimum RSS requirements are satisfied.

Surveillance Requirement 4.6.2.2.c requires that at lAt once per 24 months, verification is made that on a CDA test signal, each RSS pump starts automatically after receipt of an RWST Low-Low level signal. Tlt;~ 24 111011Al fiCU1C is baSCd M!th Me tO PV,ifbri this stnvelhmee ttL th+e c.,onid.itions that apply du"ing a plant ot.agc and p*...ti.l for .npl....,.. d nSief. t if the was pe.-rformed with the raetr-r At p.w... Gp..t.ing . *Vcience*has shon** that these eomponznfts pass the suffelte la 11 Me lnedpcetfou 1 24 111011d1 fIcgueIuVy. The.efe..

at thl the freueeywa eeneluded to be. neeta from a relability stanpin~Xt. Thisi change hasn ad.'erce im0pact on plant saf..y. ý'-'2 Surveillance Requirements 4.6.2.1 .d and 4.6.2.2.e require verification that each spray nozzle is unobstructed following maintenance that could cause nozzle blockage. Normal plant operation and maintenance activities are not expected to trigger performance of these surveillance requirements. However, activities, such as an inadvertent spray actuation that causes fluid flow through the nozzles, a major configuration change, or a loss of foreign material control when working within the respective system boundary may require surveillance performance. An evaluation, based on the specific situation, will determine the appropriate test method (e.g., visual inspection, air or smoke flow test) to verify no nozzle obstruction.

MILLSTONE - UNIT 3 B 3/4 6-2a Amendment No.

LBDCR Ne. 04 NIP3 015-FebUaty 24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive matenal to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for these isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. FSAR Table 6.2-65 lists all containment isolation valves. The addition or deletion of any containment isolation valve shall be made in accordance with Section 50.59 of 10CFR50 and approved by the committee(s) as described in the QAP Topical Report.

For the purposes of meeting this LCO, the safety function of the containment isolation valves is to shut within the time limits assumed in the accident analyses. As long as the valves can shut within the time limits assumed in the accident analyses, the valves are OPERABLE. Where the valveposition indication does not affect the operation of the valve, the indication is not required for valve OPERABILITY under this LCO. Position indication for containment isolation 4-valves is covered by Technical Specification 6.8.4.e., Accident Monitoring Instrumentation.

Failed position indication on these valves must be restored "as soon as practicable" as required by Technical Specification 6.8.4.e.3. Maintaining the valves OPERABLE, when position indication fails, facilitates troubleshooting and correction of the failure, allowing the indication to be restored "as soon as practicable."

With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and deactivated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration.

If the containment isolation valve on a closed system becomes inoperable, the remaining barrier is a closed system since a closed system is an acceptable alternative to an automatic valve.

However, actions must still be taken to meet Technical Specification ACTION 3.6.3.d and the valve, not normally considered as a containment isolation valve, and closest to the containment wall should be put into the closed position. No leak testing of the alternate valve is necessary to satisfy the ACTION statement. Placing the manual valve in the closed position sufficiently deactivates the penetration for Technical Specification compliance.

Closed system isolation valves applicable to Technical Specification ACTION 3.6.3.d are included in FSAR Table 6.2-65, and are the isolation valves for those penetrations credited as General Design Criteria 57. The specified time (i.e., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) of Technical Specification /

ACTION 3.6.3.d is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3 and 4. In the event the affected penetration is ,

isolated in accordance with 3.6.3.d, the affected penetration flow path must be verified to be isolated on a periodic basis, (Surveillance Requirement 4.6.1.1 .a). This is necessary to assure leak tightness accidentareof containment isolated.'], and that containment d,_3,yy,?:yy penetrations requiring L**,"" _ 33'gy,isolation

" following

,5,,U an MILLSTONE - UNIT 3 B 3/4 6-3 Amendment No. 2-8, 63ý, 442-, 24-6, Autiuwtudged by NRC tuttei dated ft/25165 Insert 2

LBDCR o5-. 3-25 .

Mlch 7, 200(7.

CONTAINMENT SYSTEMS BASES 3/4.6.6.1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM (Continued)

Surveillance Requirements jInsert 2 -

a Cumulative operation of the SLCRS with heaters operating for at least 10 continuous hours' 1-day puriud is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

The 31 day fequency was dc;e0cped in consideration of the ktewn reliabilit, of fan metefs and cnr1.This test is performed one STAGGERD T6 TBSI pet 31-days.

b. c. e. and f These surveillances verify that the required SLCRS filter testing is performed in accordance with Regulatory Guide 1.52, Revision 2. ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The surveillances include testing HEPA filter performance, charcoal adsorber efficiency, system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.

Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.

The 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation requirement originates from Regulatory Guide 1.52, Revision 2, March 1978, Table 2, Note "c", which states that "Testing should be performed (1) initially, (2) at least once per 18 months thereafter for systems maintained in a standby status or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operations, and (3) following painting, fire, or chemical release in any ventilation zone communicating with the system."

This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits, as well as providing trend data. The 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> figure is an arbitrary number which is equivalent to a 30 day period. This criteria is directed to filter systems that are normally in operation and also provide emergency air cleaning functions in the event of a Design Basis Accident. The applicable filter units are not normally in operation and the sample canisters are typically removed due to the 18 month criteria.

d /L Fnsdir -2 The_utomatic strtup ensures that each SLCRS train responds pro erly. .heznc pe- 21 mzntith fi bl I.urIv )* llltthl i JLL.

. EI~

a 1 1- -lLI - .11 ldw ti.od th*tulf y app lti ll a plant ouItage, and the,potential for.an unplanned-trnisien ifle u~ilncvas wh perforrnd wi;th the reactor It power The surveillance verifies that the SLCRS starts on a SIS test signal. It also includes the automatic functions to isolate the other ventilation systems that are not part of the safety-related postaccident operating configuration and to start up and to align the ventilation systems that flow through the secondary containment to the accident condition.

MILLSTONE - UNIT 3 B 3/4 6-6 Amendment No. 8-7, 4-2-3, 4-84,-206,

FL*tI-yIJ 24, 2005 CONTAINMENT SYSTEMS BASES 3/4.6.6.2 SECONDARY CONTAINMENT (continued)

In MODES 5 and 6, the probability and consequences of a DBA are low due to the RCS temperature and pressure limitation in these MODES. Therefore, Secondary Containment is not required in MODES 5 and 6.

ACTIONS In the event Secondary Containment OPERABILITY is not maintained, Secondary Containment OPERABILITY must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a DBA occurring during this time period. Therefore, it is considered that there exists no loss of safety function while in the ACTION Statement.

Inoperability of the Secondary Containment does not make the SLCRS fans and filters inoperable. Therefore, while in this ACTION Statement solely due to inoperability of the Secondary Containment, the conditions and required ACTIONS associated with Specification 3.6.6.1 (i.e., Supplementary Leak Collection and Release System) are not required to be entered.

If the Secondary Containment OPERABILITY cannot be restored to OPERABLE status within the required completion time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full-power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirements 4.6.6.2.1 Maintaining Secondary Containment OPERABILITY requires maintaining each door in each access opening in a closed position except when the access opening is being used for normal entry and exit. The normal time allowed for passage of equipment and personnel through each access opening at a time is defined as no more than 5 minutes. The access opening shall not be blocked open. During this time, it is not considered necessary to enter the ACTION statement. A 5-minute time is considered acceptable since the access opening can be quickly closed without special provisions and the probability of occurrence of a DBA concurrent with equipment and/or personnel transit time of 5 minutes is low.

T 3 1 dJay frequenc~y for this~ suvifmc is bae on e uJgmiet and ii-

-e~~~i~~ in, view o~qacf the other indi;at;i 1 s of ne~~plX stettw thett aI.e etv ailable. to the perator.

MII LSTONE - UNIT 3 B 3/4 6-8 Amendment No. 8-7, 4-26, ILnsert 2] -FTtI*~*( A-,*iuw gedd by NRC kitua datd OW25ffi O .U

urL)*t 110. Uq-Mvt'-:9 i it Nvme 10, 2005 PLANT SYSTEMS BASES AUXILIARY FEEDWATER SYSTEM (Continued)

If all three AFW pumps are inoperable in MODE 1, 2, or 3, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with non safety related equipment. In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW pump to OPERABLE status. Required ACTION e. is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW pump is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

SR 4.7.1.2. la. verifies the correct alignment for manual, power operated, and automatic valves in the auxiliary feedwater water and steam supply flow paths to provide assurance that the proper flow paths exist for auxiliary feedwater operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulations; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. 'T 31t ,, d4 y f*iutCiiy u I i, a,,. on engineerin jUdgmenAt, ig G91n9i8tent With th@ procedura cz bgvming ,t-v oper~atiet, an onruret correct -Valve positietts. Isr The SR is modified by a Note that states one or more auxiliary feedwater pumps may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the auxiliary feedwater mode of operation, provided it is not otherwise inoperable. This exception to pump OPERABILITY allows the pump(s) and associated valves to be out of their normal standby alignment and temporarily incapable of automatic initiation without declaring the pump(s) inoperable. Since auxiliary feedwater may be used during STARTUP, SHUTDOWN, HOT STANDBY operations, and HOT SHUTDOWN operations for steam generator level control, and these manual operations are an accepted function of the auxiliary feedwater system, OPERABILITY (i.e., the intended safety function) continues to be maintained.

MILLSTONE - UNIT 3 B 3/4 7-2c I

~

LBDCR 037-WMP3-633 ne-25, 200 PLANT SYSTEMS BASES SURVEILLANCE REQUIREMENTS For the surveillance requirements, the UHS temperature is measured at the locations described in the LCO write-up provided in this section.

Surveillance Requirement 4.7.5.a verifies that the UHS is capable of providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature. Ti=

24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frqtnyi based n op..at~in e.Ap.AimiceA tintekd to heniuing Uf flic Paidin Iii . .. o d- udthe applI~*tib MvODES. This surveillance requirement verifies that the avera water temperature of the UHS is less than or equal to 75°F. Insert 2 Surveillance Requirement 4.7.5.b requires that the UHS temperature be monitored on an increased frequency whenever the UHS temperature is greater than 70'F during the applicable MODES. The intent of this Surveillance Requirement is to increase the awareness of plant personnel regarding UHS temperature trends above 70'F. The frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.

3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM BACKGROUND The control room emergency ventilation system provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke. Additionally, the system provides temperature control for the control room envelope (CRE) during normal and post-accident operations.

The control room emergency ventilation system is comprised of the CRE emergency air filtration system and a temperature control system.

The control room emergency air filtration system consists of two redundant systems that recirculate and filter the air in the CRE and a CRE boundary that limits the inleakage of unfiltered air. Each control room emergency air filtration system consists of a moisture separator, electric 4 heater, prefilter, upstream high efficiency particulate air (HEPA) filter, charcoal adsorber, downstream HEPA filter, and fan. Additionally, ductwork, valves or dampers, and instrumentation form part of the system.

The CRE is the area within the confines of the CRE boundary that contains the spaces that control room occupants inhabit to control the unit during normal and accident conditions. This area encompasses the control room, and other non-critical areas including adjacent support offices, MILLSTONE - UNIT 3 B 3/4 7-10 Amendment No. 149,4-36,444, 244,

LBDCR G7-IIMP3-G33 PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

ACTIONS (Continued) their ability to control the reactor and maintain it in a safe shutdown condition in the event of a DBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

Immediate action(s), in accordance with the LCO ACTION Statements, means that the required action should be pursued without delay and in a controlled manner.

During movement of recently irradiated fuel assemblies

d. With one control room emergency air filtration system inoperable, action must be taken to restore the inoperable system to an OPERABLE status within 7 days. After 7 days, either initiate and maintain operation of the remaining OPERABLE control room emergency air filtration system in the emergency mode or suspend the movement of fuel. Initiating and L /

maintaining operation of the OPERABLE train in the emergency mode ensures:

(i) OPERABILITY of the train will not be compromised by a failure of the automatic actuation logic; and (ii) active failures will be readily detected.

e. With both control room emergency air filtration systems inoperable, or with the train required by ACTION 'd' not capable of being powered by an OPERABLE emergency power source, actions must be taken to suspend all operations involving the movement of recently irradiated fuel assemblies. This action places the unit in a condition that minimizes risk. This action does not preclude the movement of fuel to a safe position.

SURVEILLANCE REQUIREMENTS 4.7.7.a Insert 2 The CRE environment should be chec eriodically to ens that the CRE temperature control system is functioning properly. I, . 4 E. ten., n'f i ,V1 ,, , ,

95 at°Fie ast i,nper 12 iffi t. It is not necessary to c le the C R E ventilation

, houi s is su eie Typically, temperature aberrations will be readilycovered chillers. The CRE is manned during operations apparent.

'o by the techni specifications. att 4.7.7.b Standby systems should be checked periodically to ensure that they function properl . As-th env'ir amont and NORM.l eo erating eenditiea f en this system ar. .t. too . severe. festitt the t -_in3 MILLSTONE - UNIT 3 B 3/4 7-13b Amendment No. -t

LBDCR .5-.P3.025 MArch 7--'4,00&

PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued) once ever; 31 days on a STAGGERED TEST BASIS ...... an adequate check of s

-system. This surveillance requirement verifies a system flow rate of 1,120 cfm +/- 20%.

Additionally, the system is required to operate for at least 10 continuous hours with the heaters energized. These operations are sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters due to the humidity in the ambient air.

4.7.7.c specified in the Surveillance Frequency Control Program las '

The performance of the control room emergency filtration systems should be chec ced periodically by verifying the HEPA filter efficiency, charcoal adsorber efficiency, 4inimum flow rate, and the physical properties of the activated charcoal. The frequency is at4ee*.efiee pe,; 4moits or following painting, fire, or chemical release in any ventilation zone k communicating with the system.

ANSI N510-1980 will be used as a procedural guide for surveillance testing.

Any time the OPERABILITY of a HEPA filter or charcoal adsorber housing has been affected by repair, maintenance, modification, or replacement activity, post maintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.

4.7.7.c. 1 This surveillance verifies that the system satisfies the in-place penetration and bypass leakage testing acceptance criterion of less than 0.05% in accordance with Regulatory Position C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, while operating the system at a flow rate of 1,120 cfm +/- 20%. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in the regulatory guide.

4.7.7.c.2 This surveillance requires that a representative carbon sample be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978 and that a laboratory analysis verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters. The laboratory analysis is required to be performed within 31 days after removal of the sample. ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

MILLSTONE - UNIT 3 B 3/4 7-14 Amendment No. 4-36, 4-84, 206,

LBDCRf 07ý-MP3-633 June 25, 200, PLANT SYSTEMS BASES 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM (Continued)

SURVEILLANCE REQUIREMENTS (Continued) 4.7.7.c.3 This surveillance verifies that a system flow rate of 1,120 cfm +/- 20%, during system operation when testing in accordance with ANSI N510-1980.

4.7.7.d After 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, a representative carbon sample must be obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, and a laboratory analysis must verify that the representative carbon sample meets the laboratory testing criteria of ASTM D3803-89 and Millstone Unit 3 specific parameters.

The laboratory analysis is required to be performed within 31 days after removal of the sample.

ANSI N510-1980 is used in lieu of ANSI N510-1975 referenced in Revision 2 of Regulatory Guide 1.52.

The maximum surveillance interval is 900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, per Surveillance Requirement 4.0.2. The 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation requirement originates from Nuclear Regulatory Guide 1.52, Table 2, Note C. This testing ensures that the charcoal adsorbency capacity has not degraded below acceptable limits as well as providing trending data.

4.7.7.e. I This surveillance verifies that the pressure drop across the combined HEPA filters and charcoal adsorbers banks at less than 6.75 inches water gauge when the system is operated at a flow rate of 1,120 cfm +/- 20%. ThC fieqUMIY iS at !a 01,,, M Zý ,,,__.

__s.

4.7.7.e.2 Deleted. 2 4.7.7.e.3 This surveillance verifies that the heaters can dissipate 1 kW at 480V when tested in accordance with ANSI N510-1980. Thc ......... is a. feas un,c p 24 in*u s. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.

MILLSTONE - UNIT 3 B 3/4 7-15 Amendment No. 4-36,4-8, 4-84,2403,2-06,

Februar, 20, 2002 PLANT SYSTEMS BASES Insert2 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM The 0 BILITY of the Auxiliary Bui ng Filter System, and associated filters and fans, ensures that r *oactive materials leaki rom the equipment within the charging pump, I1 component cooling wa ump and hea changer areas following a LOCA are filtered prior to reaching the environment. perati f the system with the heaters operating for at least 10 continuous hoursi RI I id*. p is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. e operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. Laboratory testing of methyl iodide penetration shall be performed in accordance with ASTM D3803-89 and Millstone Unit 3 specific parameters. The heater kW measured must be corrected to its nameplate rating. Variations in system voltage can lead to measurements of kW which cannot be compared to the nameplate rating because the output kW is proportional to the square of the voltage.

The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System is required to be available to support the Auxiliary Building Filter System and the Supplementary Leak Collection and Release System (SLCRS). The Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation System consists of two redundant trains, each capable of providing 100% of the required flow. Each train has a two position, "Off' and "Auto," remote control switch. With the remote control switches for each train in the "Auto" position, the system is capable of automatically transferring operation to the redundant train in the event of a low flow condition in the operating train. The associated fans do not receive any safety related automatic start signals (e.g. Safety Injection Signal).

Placing the remote control switch for a Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train in the "Off" position to start the redundant train or to perform post maintenance testing to verify availability of the redundant train will not affect the availability of that train, provided appropriate administrative controls have been established to ensure the remote control switch is immediately returned to the "Auto" position after the completion of the specified activities or in response to plant conditions. These administrative controls include the use of an approved procedure and a designated individual at the control switch for the respective Charging Pump/Reactor Plant Component Cooling Water Pump Ventilation Train who can rapidly respond to instructions from procedures, or control room personnel, based on plant conditions.

MILLSTONE - UNIT 3 B 3/4 7-23 Amendment No. 8-7, 4-1-9, 4-36, 4-84,

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION Technical Specification 3.8.1.1 .b. 1 requires each of the diesel generator day tanks contain a minimum volume of 278 gallons. Technical Specification 3.8.1.2.b.1 requires a minimum volume of 278 gallons be contained in the required diesel generator day tank. This capacity ensures that a minimum usable volume of 189 gallons is available. This volume permits operation of the diesel generators for approximately 27 minutes with the diesel generators loaded to the 2,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> rating of 5335 kw. Each diesel generator has two independent fuel oil transfer pumps. The shutoff level of each fuel oil transfer pump provides for approximately 60 minutes of diesel generator operation at the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating. The pumps start at day tank levels to ensure the minimum level is maintained. The loss of the two redundant pumps would cause day tank level to drop below the minimum value.

Technical Specification 3.8. 1.1 .b.2 requires a minimum volume of 32,760 gallons be contained in each of the diesel generator's fuel storage systems. Technical Specification 3.8.1.2.b.2 requires a minimum volume of 32,760 gallons be contained in the required diesel generator's fuel storage system. This capacity ensures that a minimum usable volume (29,180 gallons) is available to permit operation of each of the diesel generators for approximately three days with the diesel generators loaded to the 2,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> rating of 5335 kW. The ability to cross-tie the diesel generator fuel oil supply tanks ensures that one diesel generator may operate up to approximately six days. Additional fuel oil can be supplied to the site within twenty-four hours after contacting a fuel oil supplier.

Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.

Suspension of these activities does not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power source and distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the unit safety systems.

Surveillance Requirements 4.8.1.1.2.a.6,(mc-th!-A and ".1.1,2.b.2"'(onee mer.'f-8°+/-:ý-'... and 4.8.1.1.2.j R1 me... th ....

The Surveillances 4.8.1.1.2.a.6 and 4.8.1.1.2.b.2 verify that the diesel generators are capable of synch-ronizing with the offsite electrical system and loaded to greater than or equal to continuous rating of the machine. A minimum time of 60 minutes is required to stabilize engine temperatures, while MILLSTONE - UNIT 3 B 3/4 8-1 c Amendment No. 9-7, 4-42, 4-3-7, 4-94, 2-0-0, "23

LBDCRf 08-MvP3 -02 1 SeptemberL 1 9, 2008 3/4.8 ELECTRICAL POWER SYSTEMS BASES minimizing the time that the diesel generator is connected to the offsite source. Surveillance Requirement 4.8.1.1.2.j requires demonstration onco per- 18 mnths that the diesel generator can start and run continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of which are at a load equivalent to 110% of the continuous duty rating and the remainder of the time at a load equivalent to the continuous duty rating of the diesel generator. The load band is provided to avoid routine overloading of the diesel generator. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain diesel generator OPERABILITY. The load band specified accounts for instrumentation inaccuracies, operational control capabilities, and human factor characteristics. The note (*)

acknowledges that a momentary transient outside the load range shall not invalidate the test.

Surveillance Requirements 4.8.1.1.2.a.5jh~jy...... *,48.1.1.2.b.1 4 (,O,,,, per..... Dy s, 4.8.1.1.2.gP.4.b fi8 fthff 4.8.1.1.2.g.5 hgnM.Ts)ad481.1.2.g-.6.b , ......

Several diesel generator surveillance requirements specify that the Lemergency diesel generator, are started from a standby condition. Standby conditions for a diesel generator means the diese engine coolant and lubricating oil are being circulated and temperatures are maintained within design ranges. Design ranges for standby temperatures are greater than or equal to the low temperature alarm setpoints and less than or equal to the standby "keep-warm" he Insert 2 temperatures for each respective sub-system.

Surveillance Requirement 4.8.1.1.2.i (41o8--Mte The)/

The existing "standby condition" stipulation contained in specification 4.8.1.1.2.a.5 is su erseded when performing the hot restart demonstration required by 4.8.1.1.2.j.

Any time the OPERABILITY of a diesel generator has been affected by repair, mai enance, or replacement activity, or by modification that could affect its interdependency, post aintenance testing in accordance with SR 4.0.1 is required to demonstrate OPERABILITY.

MILLSTONE - UNIT 3 B 3/4 8-1d Amendment No. 9-7, 2, 1-+3, 1-94, 2-1-0,

LDDCR 6?-MP3 033 Jiu, 25, 2007.

ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued)

The Surveillance Requirement for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1975 & 1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Sections 5 and 6 of IEEE Std 450-1980 replaced Sections 4 and 5 of IEEE Std 450-1975, otherwise the balance of IEEE Std 450-1975 applies.,"C-ýlnsert Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity.

Table 4.8-2a specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.

Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2a is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

If the required power sources or distribution systems are not OPERABLE in MODES 5 and 6, operations involving CORE ALTERATIONS, positive reactivity changes, movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the MILLSTONE - UNIT 3 B 3/4 8-2 Amendment No.

3/4.9 REFUELING OPERATIONS BASES 3/4.9.8.1 HIGH WATER LEVEL (continued)

ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operations, except as permitted in the Note to the LCO.

If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.

A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition.

If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 and the refueling water level > 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.

If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on the low probability of the coolant boiling in that time.

Surveillance Requirement This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant.

The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core ....T fqt..n.y of 12 h....

is suffieiefft, eensider-ing the fiew, tcmpzratz, pump eentrzl, and alarm indietions availabk1 to MILSON UnI 3h

-prae fo4 Bete 3/4r syst, AmendmentgNo. ]4M, 23 MILLSTONE - UNIT 3 B 3/4 9-4 Amendment No. +07,24-9,2 3/4.9 REFUELING OPERATIONS BASES The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on the low probability of the coolant boiling in that time.

Surveillance Requirement This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant.

The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The FiPquic~y of 12 hluus is sufficiu*n, eon e*iJngi the Howlt,c *. ,clir,,Itxtl*me trol. and ttlam i nJ*i tions avail ble to the1 a*ee tm, I . I -T- -

. Al BR M rRWLITB tEHD A 111VII LIN, 13YaL%,111 ill L11%,

Tonom.

MILLSTONE - UNIT 3 B 3/4 9-7 Amendment No. 409, 24-9,-2-3-

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