Text

Amends 178 & 159 to Licenses DPR-70 & DPR-75,respectively, Eliminating Defined Term Controlled Leakage,Removing Controlled Leakage Flow from Reactor Coolant Sys Operational Leakage LCO & Establishing New Seal Injection Flow LCO
	ML20094D404
Person / Time
Site:	Salem
Issue date:	10/30/1995
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Public Service Electric & Gas Co, Philadelphia Electric Co, Delmarva Power & Light Co, Atlantic City Electric Co
Shared Package
ML20094D408	List: ML20094D404; ML20094D420;
References
	DPR-70-A-178, DPR-75-A-159 NUDOCS 9511030359
	Download: ML20094D404 (28)
	v • d • e

. _. _ _. _.. _.. _. _ _. _ _ _ -

4 l

t pug i

g i

UNITED STATES

5 j

NUCLEAR REGULATORY COMMISSION I

WASHINGTON. D.C. 2008H001 49.....,o l

PUBLIC SERVICE ELECTRIC _& GAS COMPANY l

PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY i

f

-DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION. UNIT NO. I t

AMENDMENT TO FACILITY OPERATING LICENSE i

f Amendment No.178 License No. DPR-70 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated March 30, 1995, as supplemented August Is, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is' reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license.is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:

9511030359 951030 PDR ADOCK 05000272 P

PDR

4 1

i i

(2) Technical Snecifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 178, are hereby incorporeted in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMISSION F. Stolz, Dir or ject Directorate

-2 ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 30, 1995 h

b ATTACHMENT TO LICEREE AMENDMENT NO.178 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Revise Appendix A as follows:

Remove Paaes Insert Paaes I

VI VI 1-2 1-2 l

1 1-3 1-3 1-7 1-7 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 5-6b 3/4 5-6b B 3/4 4-4 B 3/4 4-4 8 3/4 5-la B 3/4 5-2 B 3/4 5-2 B 3/4 5-3

1@.EE DEFINITIONS SECTION R&QE 1.0 DEFINITIONS DEFINED TERMS 1-1 ACTION.

1-l' AXIAL FLUX DIFFERENCE 1-1 CHANNEL CALIBRATION 1-1 CHANNEL CHECK 1-1 CHANNEL FUNCTIONAL TEST 1-1

' CONTAINMENT INTEGRITY 1-2

. CORE ALTERATION 1-2 QOSE EQUIVALENT I-131-1-2

-E-AVERAGE DISINTEGRATION ENERGY 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 FREQUENCY NOTATION.

1-3 FULLY WITHDRAWN 1-3 GASEOUS RADWASTE TREATMENT SYSTEM 1-3 IDENTIFIED LEAKAGE 1-3 MEMBER (S) OF THE PUBLIC 1-4

'OFFSITE DOSE CALCULATION MANUAL (ODCM) 1-4 OPERABLE - OPERABILITY.

1-4 OPERATIONAL MODE.

1-4 PHYSICS TESTS 1-5 PRESSURE BOUNDARY LEAKAGE 1-5 PROCESS CONTROL PROGRAM (PCP) 1-5 PURGE-PURGING.

1-5 QUADRANT POWER TILT RATIO 1-5 RATED THERMAL POWER 15 REACTOR TRIP SYSTEM RESPONSE TIME 1-6 REPORTABLE EVENT.

1-6 SHUTDOWN MARGIN 1-6 SITE BOUNDARY 1-6 SOLIDIFICATION.

1-6

' SOURCE CHECK.

1-6 STAGGERED TEST BASIS.

1-6 THERMAL POWER 1-7

. UNIDENTIFIED LEAKAGE.

1-7

-UNRESTRICTED AREA 1-7 VENTILATION EXHAUST TREATMENT SYSTEM.

1-7 VENTING 1-7 SALEM - UNIT 1 I

Amendment No.178 l

b.

1HQ33 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Eagg 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS 3/4 5-1 1

3/4.5.2 ECCS SUBSYSTEMS - T, a 3 50

F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T, < 3 50

F 3/4 5-6 3/4.5.4 SEAL INJECTION FLOW.

3/4 5-6b l

3/4.5.5 REFUELING WATER STORAGE TANK 3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.

3/4 6-1 Containment Leakage.

3/4 6-2 Containment Air Locks 3/4 6-5 Internal Pressure.

3/4 6-6 Air Temperature 3/4 6-7 Containment Structural Integrity 3/4 6-8 Containment Ventilation System 3/4 6-Sa 1

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6-9 Spray Additive System.

3/4 6-10 Containment Cooling System 3/4 6-11 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-12 3/4,6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers 3/4 6-18 Electric Hydrogen Recombiners.

3/4 6-19 SALEM UNIT 1 VI Amendment No.178 l

DEFINITIONS i

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either a.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or b.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of specification 3.6.3.1.

1.7.2 All equipment hatches are closed and seeled, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE.

1.8 NOT USED 1

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component i

within the reactor pressure vessel with the vessel head removed and fuel in l

the vessel.

Suspension of CORE ALTERATION shall not preclude completion of f

movement of a component to a safe conservative position.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries i

per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM - UNIT 1 1-2 Amendment No.178 l

h DEFINITIONS-thyroid dose conversion factors used for this calculation shall be those

' listed in Table III of TID-14844 " Calculation of Distance Factors for Power and Test Reactor Sites."

' b - AVERAGE DISINTEGRATION _ ENERGY 1.11 5 shall be the average (weighted in proportion to the concentration of,

. each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolanc.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when th2 monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1

1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive.

FULLY WITHDRAWN will be specified in the current reload

analysis, GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collectiig primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release tv the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except Reactor Coolant Pump Seal Water Injection) into l

closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No.178 l

~. __.

b 8

DEFINITIONS b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LFAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage (except Reactor Coolant Pump Seal Water Injection) which is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

4 VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release, to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.37 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required du' ring VENTING.

Vent, used in system names, does not imply a VENTING process.

SALEM - UNIT 1 1-7 Amendment No.178 l

.~

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, d.

10 GPM IDENTIFIED LEAKAGE from the reactor Coolant System.

APPLICABILITY: MODES 1, 2,

3 and 4 ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY l

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by; a.

Monitoring the c.

cainment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

Monitoring the containment sump inventory at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i

SALEM - UNIT 1 3/4 4-15 Amendment No.178 l

. _.... ~ _.. _.. - -.. _..........-

64 t

REACTOR COOLANT SYSTEM l

SURVEILLANCE REQUIREMENTS (Continued) c.

NOT USED l

l l-d.

Performance of a Reactor Coolant System water inventory balance at l

'least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The water inventory balance shall be performed with the plant at steady state conditions. The provisions of specification 4.0.4 are not applicable for entry-l into Mode 4, and

{

l e.

Monitoring the reactor head flange leakoff system at least once l

per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

i l

l l

4

. SALEM - UNIT 1 3/4 4 16 Amendment No.178 l

f

~.

l I

I i

I l

EMERGENCY CORE COOLING SYSTEMS SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION I

l 3.5.4 Reactor coolant pump seal injection flow shall be s40 gpm with centrifugal charging pump discharge header pressure a2430 psig and che charging flow control valve full open.

APPLICABILITY: MODES 1, 2, and 3 ACTION:

With seal injection flow not within the limit, adjust manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure a2430 psig and the charging flow control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4 At least once per 31 days, verify manual seal injection throttle valves are adjusted to give a flow within the limit with centrifugal charging pump discharge header pressure =2430 psig, and the charging flow control valve full

)

open.

The provisions of Specification 4.0.4 are not applicable for entry into Mode 3.

This exemption is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant i

System pressure stabilizes at 2235 1 20 psig.

l l

l i

i l

i i

1 SALEM - UNIT 1 3/4 5-6b Amendment No.178 l

~._ -

~

e d

REACTOR COOLANT SYSTg BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) 1 The total steam generator tube leakage limit of 1 GPM for all steam generators (but not more than 500 gpd for any steam generator) ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System Leakage or failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the' Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action'.

1 SALEM - UNIT 1 B 3/4 4-4 Amendment No.178 l

EMERGENCY CORE COOLING SYSTEMS 1

BASES

~

ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The surveillance requirements, which are provided to ensure.the OPERABILITY of each component, ensure that, at a minimum, the assumptions used in the safety analysis are met and that subsystem OPERABILITY is maintained.

i The safety analyses make assumptions with respect to:

1) both the maximum and minimum total system resistance, and 2) both the maximum and minimum branch injection line resistance. These resistances, in conjunction with the ranges of potential pump performance, are used to calculate the maximum and minir.um ECCS flow assumed in the safety analyses.

The maximum and minimum flow surveillance requirements in conjunction with the maximum and minimum pump performance curves ensures that the assumptions of total system resistance and the distribution of that system resistance among the various paths are met.

The maximum total pump flow surveillance requirements ensure the pump runout limits of 560 gpm for the centrifugal charging pumps and 675 gpm for the safety injection pumps are not exceeded.

The surveillance requirement for the maximum difference between the maximum and minimum individual injection line flows ensure that the minimum individual injection line resistance assumed for the spilling line following a LOCA is met.

3/4.5.4 SEAL INJECTION FLOW The Reactor Coolant Pump IRCP) seal injection flow restriction limits the amount of ECCS flow that could be diverted from the injection path following an ECCS actuation. This limit is based on safety analysis assumptions, since RCP seal injection flow is not isolated during Safety Injection (SI).

The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance. Line pressure and flow must be known to establish the i

proper line resistance. Flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure, and that the centrifugal charging pump discharge pressure is greater than or equal to 2430 psig.

Charging pump header pressure is used instead of RCS pressure, since it is

)

more representative of flow diversion during an accident.

The additional LCO modifier, charging flow control valve full open, is required since the valve is designed to fail open.

With the LCO specified discharge pressure and control valve position, a flow limit is established.

This flow limit is used in the accident analysis.

A provision has been added to exempt surveillance requirement 4.0.4 for entry into MODE 3, since the surveillance cannot be performed in a lower mode.

The exemption is permitted for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has r;abilized within 2 20 psig of normal operating pressure.

The RCS pressure requirement produces the conditions necessary to correctly set the manual throttle valves. The exemption is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure timely surveillance completion once the necessary conditions are established.

SALEM - UNIT 1 B 3/4 5-2 Amendment No.178

1 I

a b.

EMERGENCY CORE COOLING SYSTEMS 4

-BASES 4

3/4.5.5 REFUELING WATER STORAGE TANK l

.The OPERABILITY of the RWST as part of the ECCS ensures that a d

sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

i 1

The' limits on RWST minimum volume and boron concentration ensure that:

4 (1) sufficient water is available within containment to permit r,ecirculation 1

cooling flow to the core, (2) the. reactor will remain suberitical in the. cold condition following a small LOCA assuming complete mixing of the RWST, RCS, and ECCS water volumes with all control rods inserted except the most reactive control assembly (ARI-1), and (3) the reactor will remain subcritical in the l

cold condition following a large break LOCA (break flow area > 3.0 sq. f t.)

assuming complete mixing of.the RWST, RCS, and ECCS water and other sources of water that may eventually reside in the sump following a LOCA with all control rods assumed to be out (ARO).

The limits on contained water volume and boron concentration also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within.

containment after a LOCA.

This pH band minimizes the evolution cf iodine and 3

minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes-an allowance for water not usable because of tank discharge line location or other physi. cal characteristics.

SALEM - UNIT 1 B 3/4 5-3 Amendment No.178 l

jeneu p

UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enana annt

\\...../

PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER ANO LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.159 License No. DPR-75 1.

The Nuclear Regulatory Commission (the Comission or the NRC) has found that:

A.

The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated March 30, 1995, as supplemented August 18, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The. facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and j

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of j

the Comission's regulations and all applicable requirements have been i

satisfied.

2.

Accordingly, the license is amanded by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby amended to read as follows:

i

. (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No.159, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION q

n F. Stolz, Dire

)r roject Directorat

-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: October 30, 1995 l

l j

J

ATTACHMENT TO LICENSE AMENDMENT NO.159 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET N0. 50-311 Revise Appendix A as follows:

Remove Paaes Insert'Paaes I

I VI VI 1-2 1-2 1-3 1-3 1-7 1-7 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 5-8a 3/4 5-8a B 3/4 4-4 B 3/4 4-4 8 3/4 5-2 8 3/4 5-2 B 3/4 5-3 8 3/4 5-3

INDEX DEFINITIONS SECTION Rhg8 1.0 DEFINITIONS DEFINED TERMS 1-1 ACTION.

1-1 AXIAL FLUX DIFFERENCE 1-1 CHANNEL CALIBRATION 1-1 CHANNEL CHECK 1-1 CHANNEL FUNCTIONAL TEST 1-1 CONTAINMENT INTEGRITY 1-2 CORE ALTERATION 1-2 DOSE EQUIVALENT I-131 1-2 E-AVERAGE DISINTEGRATION ENERGY 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 FREQUENCY NOTATION 1-3 FULLY WITHDRAWN 1-3 GASEOUS RADWASTE TREATMENT SYSTEM 1-3 IDENTIFIED LEAKAGE 1-3 MEMBER (S) OF THE PUBLIC 1-4 OFFSITE DO3E CALCULATION MANUAL (ODCM) 1-4 OPERABLE - OPERABILITY.

1-4 OPERATIONAL MODE 1-4 PHYSICS TESTS 1-5 PRESSURE BOUNDARY LEAKAGE 1-5 PROCESS CONTROL PROGRAM (PCP) 1-5 PURGE-PURGING 1-5 QUADRANT POWER TILT RATIO 1-5 RATED THERMAL POWER 1-5 REACTOR TRIP SYSTEM RESPONSE TIME 1-6 REPORTABLE EVENT.

1-6 SHUTDOWN MARGIN 1-6 SITE BOUNDARY 1-6 SOLIDIFICATION.

1-6 SOURCE CHECK.

1-6 STAGGERED TEST BASIS 1-6 THERMAL POWER 1-7 UNIDENTIFIED LEAKAGE 1-7 UNRESTRICTED AREA 1-7 VENTILATION EXHAUST TREATMENT SYSTEM.

1-7 VENTING 1-7 SALEM - UNIT 2 I

Amendment No.159 l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAgg 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T.,, a 3 50*F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T.,, < 3 50 *F 3/4 5-7 3/4.5.4 SEAL INJECTION FLOW 3/4 5-8a l

3/4.5.5 REFUELING WATER STORAGE TANK 3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.

3/4 6-1 Containment Leakage.

3/4 6-2 Containment Air Locks.

3/4 6-4 Internal Pressure 3/4 6-6 Air Temperature.

3/4 6-7 Containment Structural Integrity 3/4 6-8 Containment Ventilation System 3/4 6-9 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System 3/4 6-10 Spray Additive System.

3/4 6-11 Containment Cooling System 3/4 6-12 3/4.6.3 CONTAINMENT ISOLATION VALVES 3/4 6-14 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers 3/4 6-21 Electric Hydrogen Recombiners.

3/4 6-22 SALEM - UNIT 2 VI Amendment No.159 l

. _ _ ~.. _. -.

i.

' DEFINITIONS i

j CONTAINMENT INTEGRITY 1.7.

CONTAINMENT INTEGRITY shall exist when:

j 1.7.1 All penetrations required to be closed during accident conditions are either:

j a.

Capable of being closed by an OPERABLE containment automatic 1

isolation valve system, or i

H Closed by manual valves', blind flanges, or deactivated b.

4

~

automatic valves secured in.their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.

i 1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakagefrates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g.,

welds, bellows or 0-rings) is OPERABLE.

1.8 NOT USED l-CORE ALTERATION 19 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DOSE EQUIV) LENT I.131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM - UNIT 2 1-2 Amendment No.159 l

i I

I DEFINITIONS l

thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 " Calculation of Distance Factors for Power and Test Reactor Sites."

b - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of tl.e total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive.

FULLY WITHDRAWN will be established by the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREAINENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the prima.ry system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except Reactor Coolant Pump Seal Water Injection) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 2 1-3 Amendment No.159 I

i

.\\

l i

o l

I DEFINITIONS b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

l THERMAL POWER j

i l

1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

8 UNIDENTIFIED LEAKAGE 1.34' UNIDENTIFIED LEAKAGE shall be all leakage (except Reactor Coolant Pump l

l Seal Water Injection) which is not IDENTIFIED LEAKAGE.

l l

UNRESTRICTED AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, l

access to which is not controlled by the licensee for purposes of protection

[

of individuals from exposure to radiation and radioactive materials, or any l

area within the SITE BOUNDARY used for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.36 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radiciodine and radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous ~ exhaust stream prior to the release to the environment (such a system is not considered to have any effect on l

noble gas effluents).

Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

l W ING 1.37 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING pror.ess.

SALEM - UNIT 2 1-7 Amendment No.159 l

-.. ~.. -.

'e j

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURS. BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator, d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant-System, and I

e.

NOT USED f.

1 GPM leakage at a Reactor Coolant System pressure of 2230 120 psig, from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2,

3 and 4 ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN during within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by; a.

Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Monitoring the containment sump inventory at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 2 3/4 4-17 Amendment No.159

_. _. ~. _... -- -_. _. _ ~

i REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l c.

NOT USED l

l d.

Performance of a Reactor Coolant System water inventory balance at j

least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The watet inventory balance shall be performed with the plant at steady state conditions. The provisions of specification 4.0.4 are not applicable for entry into Mode 4, and Monitoring the reactor head flange leakoff system at least once per e.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.7.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to Spegification 4.0.5, l

except that in lieu of any leakage testing required by' Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within j

its limit:

a.

At least once per 18 months.

b.

Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, c.

Prior to returning the valve to service following maintenance repair or replacement work on the valve.

d.

For the Residual Heat Removal and Safety Injection Systems hot and cold leg injection valves and accumulator valves listed in Table 3.4-1 the testing will be done within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve.

For all other systems testing will be done once per refueling.

The provisions of specification 4.0.4 are not applicable for entry into EMODE 3 or 4.

SALEM - UNIT 2 3/4 4-18 Amendment No.159 l

o t

d EMERGENCY CORE COOLING SYSTEMS SEAL INJECTION PLOW LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be s40 gpm with centrifugal charging pump discharge header pressure a2430 psig and the charging flow control valve full open.

APPLICABILITY: MODES 1, 2, and 3 ACTION:

With seal injection flow not within the limit, adjust manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure a2430 psig and the charging flow control valve full open j

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURV3ILLANCE REQUIREMENTS 4.5.4 At least once per 31 days, verify manual seal injection throttle valves are adjusted to give a flow within the limit with centrifugal charging pump discharge header pressure =2430 psig, and the charging flow control valve full open.

I The provisions of Specification 4.0.4 are not applicable for entry into l

Mode 3.

This exemption is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at 2235 2 20 psig.

SALEM - UNIT 2 3/4 5-Ba Amendment No.159 l

e d

REACTOR COOLANT SYSTEM

. BASES 3/4.4.6 STEAM GENERATORS (Continued)

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be evaluated for reportability to the Commission pursuant to the applicable sections of 10CFR50.72 and 10CFR50.73.

3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.7.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent witn the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.7.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM.

This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

I The surveillance requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

SALEM - UNIT 2 B 3/4 4-4 Amendment No.159 l

.- -.... -. - ~.. -.. -. -

a d.

EMERGENCY CORE COOLING SYSTEMS

. BASES ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350*F,~one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity. condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one safety injection pump or one centrifugal charging pump to be OPERABLE'and the Surveillance requirement to verify'all safety injection pumps except the allowed OPERABLE safety injection pump to be inoperable below 312*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single POPS relief

valve, e

The surveillance requirements, which are provided to ensure the OPERABILITY of each component, ensure that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. The safety analyses make the assumptions with respect to:

1) both the maximum and minimum total system resistance, and 2) both the maximum and minimum branch injection'line resistance. These resistances, in conjunction with the ranges of potential pump performance, are used to calculate the maximum and minimum BCCS flow assumed in the. safety analyses.

The maximum and minimum flow surveillance requirements in conjunction with the maximum and minimum pump performance curves" ensures that the assumptions of total system resistance ^and the distribution of that system resistance among the various paths are met.

The maximum total pump flow surveillance requirements. ensure the pump runout' limits of 560 gpm for the centrifugal charging pumps and 675 gpm for the safety injection pumps are not exceeded.

The surveillance requirement for the maximum difference between the maximum and minimum individual injection line flows ensure that the minimum individual injection line resistance assumed for the spilling line following a LOCA is met.

3/4.5.4 SEAL INJECTION FLOW The Reactor Coolant Pump (RCP) seal injection flow restriction limits the amount of ECCS flow that would be diverted from the injection path following an ECCS actuation. This limit is based on safety analysis assumptions, since RCP seal injection flow is not isolated during Safety Injection (SI).

The LCO is not strictly a flow limit, but rather a flow limit based on a flow line resistance.

Line pressure and flow must be known to establish the proper line resistance.

Flow line resistance is determined by assuming that the RCS pressure is at normal operating pressure, and that the centrifugal charging pump discharge pressure is greater than or equal to 2430 peig.

Charging pump header pressure is used instead of RCS pressure, since it is more representative of flow diversion during an accident. The additional LCO SALEM - UNIT 2 B 3/4 5-2 Amendment No.159

O o

i I

l EMERGENCY CORE COOLING SYSTEMS BASES modifier, charging flow control valve full open, is required since the valve is designed to fail open. With the LCO specified discharge pressure and control valve position, a flow limit is established. This flow limit is used in the accident analysis.

A provision has been added to exempt surveillance requirement 4.0.4 for entry into MODE 3, since the surveillance cannot be performed in a lower mode.

L The exemption is permitted for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS pressure has stabilized within 2 20 psig of normal operating pressure. The RCS pressure requirement produces the conditions necessary to correctly set the manual l

throttle valves. The exemption is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure timely l

surveillance completion once the necessary conditions are established.

3/4.S.5 REFUELING WATER STORAGE TANK l

The OPERABILITY of the RWST as a part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits on RWST minimum volume and boron concentrations ensure thats l

(1) sufficient water is available within containment to permit recirculation cooling flow to the core, (2) the reactor will remain suberitical in the cold condition following a small LOCA assuming complete mixing of the RWST, RCS, l

and ECCS water volumes with all control rods inserted except the most reactive control assembly (ARI-1), and (3) the reactor will remain subcritical in the cold condition following a large break LOCA (break flow area > 3.0 sq. f t.)

assuming complete mixing of the RWST, RCS, and ECCS water and other sources of l

water that may eventually reside in the sump following a LOCA with all control I

rods assumed to be out (ARO). The limits on contained water volume and boron l

concentration also ensure a pH value of between 8.5 and 11.0 for the solution h

recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

SALEM - UNIT :

B 3/4 5-3 Amendment No.159 l

\\

-

ML20094D404

Text

Attachment:

Attachment:

Navigation menu

Search