Text

Amends 49 & 12 to Licenses NPF-39 & NPF-85,respectively, Revising Tech Specs to Conform to NRC Positions on Inservice Insps & Monitoring of Unidentified Leakage in Generic Ltr 88-01 Re IGSCC in Austenitic Stainless Steel Piping
	ML20070K156
Person / Time
Site:	Limerick
Issue date:	03/05/1991
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20070K159	List: ML20070K156; ML20070K165;
References
	GL-88-01, NUDOCS 9103180327
	Download: ML20070K156 (29)
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e UNITED STATES a

NUCLEAR REGULATORY COMMISSION i

\\..... /l wAssiwoTow, p. c. rossa PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. NPF-39 1.

The Nuclear Regulatory Comission (the Comission) has found that A.

The appif :ation for amendment by Philadelphia Electric Company (the lice.1see) dated December 21, 1990, 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 It t

B.

The f acility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission:

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and suety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; 0.

The issuence of this amendment will not be inimical to the comon defense and security or to the health and safety of the publict and E.

The issuance of this amendment is in accordance with 10 CFR Pa been satisfied.of the Comission's regulations and all applicable requirements have 2.

Accordingly, the license is amended by changes to the of Facility Operating License No NPF-39 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 49, are hereby incorporated into this license.

Philadelphia Electric Comoany shall operate the facility i

in accordance with the Technical Specifications and the Environmental 1

Protection Plan.

9103180327 910305 ADOCK0500g2 pa

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This license amendment is effective 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

[L, i

Welter R. Butler, Director Project Directorate I-2 Division of Reactor Projects - 1/II

Attachment:

Changes to the Technical Specifications Date of Issuance: March 5. 1991 l

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This license amendment is effective 30 days f rom the date of issuance.

FOR THE flVCLEAR REGULATORY COMMISSION

/S/

Walter R. Dutler, Director Project Directorate 1-2 Division of Reactor Projects. 1/11

Attachment:

Changes to the Technical Specifications

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M ATTACHMENT TO LICENSE AMENDMENT N0. 49 TACILITY OPERATING LICENSE NO. NPF-39

%CKET N0. 50-352 Replace the following pages of the Appendix A Technical 5pecifications with the attached pages.

contain vertical lines indicating the area of change.The revised pages a Overleaf pages are provided to maintain document completeness.*

Remove Insert 3/4 0-3 3/4 0-3 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 B 3/4 0-5 B 3/4 0-6 B 3/4 0-5 B 3/4 0 6*

B 3/t 4-3 B 3/4 4-4 8 3/4 4-3 B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 B 3/4 4-5*

B 3/4 4-6 l

APPLICABILITY I

SURVEILLANCE REQUIREMENTS (Continued)

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing intervice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days i

Monthly At least once per 31 days Quarterly or every 3 months At lear,t once per 92 days Semiannually or every 6 months At least once per 184 days i

Every 9 months At least once per 276 days Year?v or annually At least once per 366 days The provisions of Specification 4.0.2 are applicable to the above c.

required frequencies for performing inservice inspection and testing activities.

d.

Performance of the'above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.-

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to e.

supersede the requirements of any Technical Specification.

f.

The Inservite Inspection (ISI) Program for riping identified in NRC.

Generic letter 88-01 shall be performed in accordance with the staff positions on schedule, methods and personnel, and sample expansion included in the Generic Letter. Details for implementation of.these requirements are included as augmented inspection requirements in the ISI Program.

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LIMERICK - UNIT 1 3/4 0-3 Amendment No. II.,49

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REACTOR COOLANT SYSTEM OPERAfl0NAL LEAKAGE LIMITING CADITION FOR OPERATION h"-

3.4.3.2 Reactor coolant system leakage shall be limitrd to a.

No PRESSURE BOUNDARY LEAKAGE.

b.

5 gpm UNIDENTIFIED LEAKAGE.

c.

30 gpm total leakage, d.

25 gpm total leakage averaged over any 24-hour period, 1 gpm leakage at a reactor coolant system pressure of 950 e.

110 psig from any reactor coolant system pressure isnlation valve specified in Table 3.4,3.2-1.

f.

2 gpm increase in VH10ENTIFIED LEAKAGE over a 24-hour period.

l APPLICABILITY: OPERATIONAL C0HDITIONS 1, 2, and 3.

ACTION:

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTOOWN within 12 a.

hours and in COLD SHUTOOWN within the axt 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With any reactor coolant system leakage greater than the limits in b, c and/or d above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTOOWN within the fo lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , With any reactor coolant system pressure isolation va' ve leakage greater c.

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 one other closed manual, deactivated automatic, or check

valves, or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN withi I

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

d.

With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , With any reactor coolant system leakage greater than the limit in f above, e.

identify the source of leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.

LIMERICK - UNIT 1 3/4 4-9 Amendment No. 28,49

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

Monitoring the primary containment atmospheric gaseous radioactivity at a.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (not a means of quantifying leakage),

b.

Monitoring the drywell floor drain sump and drywell equipment drain tank flow rate at least once per eight (8) hours.

l c.

Monitoring the drywell unit coolers condensate flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , d.

Monitoring the primary containment pressure at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (notameansofquantifyingleakage),

Monitoring the reactor vessel head flange leak detection system at e.

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and f.

Monitoring the primary containment temperature at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (not a means of quantifying leakage).

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3. ?-l shall be demonstrated OPERABLE by leak testing pursuant to Specificatto,,.v.5 and verifying the leakage of each valve to be within the specified limit:

a.

At least once per 18 months, and b.

Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The provisions of Specification 4.0.4 are not applicaoie for entry into OPERATIONAL CONDITION 3.

i 4.4.3.2.3 The high/ low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints set less than the allowable values in Table 3.4.3.2-1 hu pa-Formance of a:

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a.

b.

CHANNEL CALIBRATION at least once per 18 months.

LIMERICK - UNIT 1 3/4 4-10 Amendment No. 33, 49

3/4.0 APPLICABILITY 1ASES If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with ACTION requirements, e.g.,

Specification 3.0.3, a 24-hour allowance is proviced to permit a celay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed.

The purpose of this allowance is to permit the completion of a surveillance before a shutdown would be required to correly with ACTION requirements or before other remedial measures would be required that may preclude the completion of a surveillance.

The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in completing the required surveillance.

This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of CONDITION changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0.4 is allowed.

If a surveillance is not completed within the 24-hour allowance, the time limits of the ACTION requirements are applicable at that time.

When a surveillance is performed within the 24-hour allowance and the Surveillance Requirements are not met, the time limits of the ACTION requirements are applicable at the time that the surveillance is terminated.

Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply.

However, the Surveillance Requirements have to be met to demonstrate that inoperable equipment has been restored to OPERABLE status.

Specification 4.0.4 establishes the requirement that all applicable surveillances must P met before entry into an OPERATIONAL CONDITION or other condition of operation specified in the Applicability statement.

The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into an OPERATIONAL CONDITION or other specified condition for which these systems and components ense e safe operation of the facility.

This provision applies to changes in OPERATIONAL CONDITIONS or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during in'tial plant startup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower CONDITION of operation.

Specification 4.0.5 establishes the requirement that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a.

These requirements apply except when relief has been provided in writing by the Commission. Additionally, the Inser'tice Inspection Program conforms to the NRC staff positions identified in NRC G ueric Letter 88-01, "NRC Position on IGSCC in BWR Austinetic Stainless Steel Piping," as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990.

LIMERICK - UNIT 1 8 3/4 0-5 Amendment No, II. 49

3/4.0 APPLICABILITY BASES This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda.

This clarifica-tion is provided to ensure consistency in surveillance intervals throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and test'ing activities.

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and acplicable Addenda.

The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL CONDITION or other specified condition takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows pumps and valves to be tested up to one week after return to normal operation.

The Technical Specification definition of OPERABLE does not allow a grace period before a component, which is not capable of performing its specified function, is declared inoperable and takes prece-dance over the ASME Boiler and Pressure Vessel Code provision that allows a valve to be incapable of performin before being declared inoperable. g its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LIMERICK - UNIT 1 B 3/4 0-6 Amendment No. A4, 49 g

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,REA'CTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 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 with the recomendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

In conformance with Regulatory Guide 1.45, the channel calibration tests will verify the ability to detect a1gpmleakinlegsthan1hourandanatmosphericgaseousradioactivitysystem sensitivity of 10" J)C/cc.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action. The limit of 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period and the monitoring of drywell floor drain sump and drywell equipment drain tank flow rate at least once every eight (8) hours conforms with HRC staff positions specified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," as revised by NRC Safety Evaluation dated March 6, 1990.

The ACTION requirement for the 2 gpm increase in UNIDENTIFIED LEAKAGE limit ensures that such leakage is identified or a plant shutdown is initiated to allow further investigation and corrective action. Once identified, reactor operation may continue dependent upon the impact on total leakage.

The ACTION requirements for pressure isolation valves (P!Vs) are used in conjunction with the system specifications for which PlVs are listed in Table 3.4.3.2-1 and with primary containment isolation valve requirements to ensure that plant operation is appropriately limited.

The Surveillance Requirements for the 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.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during -

these periods.

L1MERICK - UNIT 1 8 3/4 4-3 Amendment No. H, 49

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3/4.4.4 CHEMISTRY (Continued)

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Conductivity measurements are required on a continuous basis since changes in l

this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must i

also be within their acceptable limits. With the cor4uctivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding l

the limits.

The surveillance requirements provide adequate assurance that-concentrations t

in excess of the limits will be detected in sufficient time to take corrective

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3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure i

outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on t

specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific s i te. parameters, such as SITE BOUNDARY location and meteorological conditions, r

were not considered in this evaluation, t

The ACTION statement pt.rmitting POWER OPERATION to continue for limited time v

periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT l-131, but less than or equal to 4 microcuries per gram DOSE EQUIVALENT I-131, accommodates possible iodine spikino phenomenon which may l

occur following changes in the THERMAL POWER. Operation wIth specific activity levels exceeding 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram DOSE EQUIVALENT l-131 must be restricted since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY fol hwing a postulated steam line rupture.

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Closing the main steam line isolatic valves prevents the. release of activity i

to the environs should a steam line rupture occur outside containment. The 3

surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take l

l corrective action.

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3/4.4.6 PRESSURE / TEMPERATURE LIMITS i

All components in the reactor coolant system are designed to withstand the j

effects of cyclic loads due to system temperature and pressure changes. These j

cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum i

J specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

1 LIMERICK - UNIT 1 B 3/4 4-4 Amendment No. 20. 40,49 i

i REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code See'. ion !!!,

Appendix G.

The curves are based on the RTNOT and stress intenatW factor information for the reactor vessel components.

Fracture toughness limits and the basis for compliance are more f ully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, " Fracture Toughness."

The reactor vessel materials have been tested to determine their initial RT The results of these tests are shown in Table B 3/4.4.6 1.

NOT.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, NDT.

based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figuro B 3/4.4.6-1 and the recommenda-tions of Regulator Vessel Materials."y Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor The pressure / temperature limit curve, Figurt 3.4.6.1+1, curves A, B and C, are based on the non beltline, discontinuity areas of the RPV which do not receive significant neutron fluence and the RTHDis will, therefore, not shift.

areas of the RPV until 12 EFPY when the beltline material's RTThese limit NDT *III 'hIIt due to neutron fluence and the beltline curves will intersect the non-beltline discontinuity curves.

Figure 3,4,6.1-1, but are included on FSAR Figure 5.3 4The non-limiting be The actual shift in RTNDT of the vessel material will be estabitsbed period-ically during operation by removing and evaluating, in accordance with 10 CFR Part 50, Appendix H, irradiated reactor vessel flux wire and Charpy specimens installed near the inside wall of the reactor vessel in the core area.

Since the l

neutron spectra at the flux wires, Charpy specimens and vessel inside radius are essentially identical, the irradiated Charpy specimens can be used with confi-dence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4,6.1-1 shall be adjusted, as required, on the Guide 1.99, Revision 2. basis of the ftux wire and Charpy specimen data and recomm This would include showing the beltline (versus non-beltline discontinuity) limits when they become bounding.

The pressure temperature limit lines shown in Figures 3.4.6.1 1, curves C, and A, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are pro-vided in Table 4.4.6.1.3-1 Appendix H to 10 CFR Part 50.to assure compliance with the requirements of LIMERICK - UNIT 1 8 3/4 4-5 Amendment No. 36 l

APR 2 01990 J

i REACTOR COOLANT SYSTEM l

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j 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPE RABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The 3

minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that i

the structural integrity of these components will be maintained at an acceptable 1

level throughout the life of the plants Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Sectica XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55e(g)(6)(1). Additionally, the Inservice inspection Program conforms to the NRC staff positions identified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping " as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990, 1

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability _for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation, LIMERICK - UNIT 1 B 3/4 4-6 Amendment No. 49

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UMTED STATES NUCLEAR REGULATORY COMMISSION i,

f w Aswiwaro w. o. c. r w n s.,.....j PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 12 License No. NPF-85 1.

The Nuclear Regulatory Comission (the Comission) has found that A.

The application for amendment by Philadelphia Electric Company (the licensee) dated December 21, 1990, complies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter is B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission:

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 Comission's regulations:

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 E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised throuch Amendment No. 12, are hereby incorporated into this license.

Philadelphia Electric Company shall operate the f acility l

in accordance with the Technical Specifications and the Environmental Protection Plan.

7 3.

This license amendnient it, eff ective 30 days from the date of issuance.

FOR THE HUCLEAR REGULATORY COMMISSION l

/S/

Walter R. Butler, Director Project Directorate 1-2 Division of P,eactor Projects. 1/11

Attachment:

Changes to the Technical Specifications Date of issuance: March 5, 1991 1l(I).

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e 3.

This license amendment is effective 30 days from the date of issuance.

FOR THE NUCLEAR REGULAf0RY Com!$$10N k.

Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects - I/!!

Attachment Changes to the Technical Specifications Date of Issuance: March 5, 1991 i

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I ATTACHMENT TO LICENSE AMENDMENT N0.12 FACILITY OPERA 11NG LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following sages of the Appendix A Technical Specifications with i

the attached pages.

contain vertical lines indicating the area of change. Tie revised pages are id Overleaf pages are provided to maintain document completeness.*

Remove insert xix xix xx xx*

3/4 0-3 3/4 0-3

~

3/4 4-9 3/4 4-10 3/4 4-9 3/4 4-10 J

B 3/4 0-5 B 3/4 0-6 8 3/4 0-5 8 3/4 0-6*

B 3/4 4-3 B 3/4 4-3

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B 3/4 4-3a B 3/4 4-4 B 3/4 4-3a B 3/4 4-4*

B 3/4 4-5 B 3/4 4-6 B 3/4 4-5*

8 3/4 4-6

t4%

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INDEX 9

BASES SECTION PAGE i

INSTRUMENTATION (Continued)

Seismic Monitoring Instrumentation.......................... B 3/4 3-4 (Deleted).................................................. B 3/4 3-4 Remote Shutdown System Instrumentation and tcntrols......... B 3/4 3-5 Accident Monitoring Instrumentation......................... B 3/4 3-5 Source Range Monitors....................................... B 3/4 3-5 Traversing In-Core Probe System.............................

B 3/4 3-5 Chlorine and Toxic Sas Detection Systems.................... B 3/4 3-6 Fire Detection Instrumentation.............................. B 3/4 3-6 Loose-Part Detection System.................................

B 3/4 3-6 (Deleted).................................................. B 3/4 3-6 Of fgas Moni toring Instrumentation........................... B 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTIONSYSTEM......................... B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................. B 3/4-3-7 Bases Figure B 3/4.3-1 Reacte: Vessel Water Leve1........................... B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REC IRCULATION SYSTEM....................................

3/4.4.2 SAFETY / RELIEF VALVES........................................ B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................... B 3/4 4-3 Operational Leakage......................................... B 3/4 4-3 3/4.4.4 CHEMISTRY................................................... B 3/4 4-3a i

LIMERICK - UNIT 2 xix Amendment No. fl.12 y

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INDEX l

BASES SECTION PAGE l

l REACTOR COOLANT SYSTEM (Continued) l 3/4.4.5 SPECIFIC ACTIVITY.......................................

B 3/4 4-4 j

3/4.4.6 PRESSURE / TEMPERATURE LINITS.............................

B 3/4 4-4 l

Bases Table B 3/4.4.6-1 Reactor Yessel Toughness.................

B 3/4 4-7 Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) At 1/4 T As A Function of Service Life......................

B 3/4 4-8 l

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................

B 3/4 4-6 3/4.4.8 STRUCTURAL INTEGRITY....................................

B 3/4 4-6 t

3/4.4.9 RESIDUAL HEAT REM 0 VAL...................................

B 3/4 4-6 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN............

B 3/4 5-1 3/4.5.3 SUPPRESSION CHAMBER................................

B 3/4 5-2 i

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAllMENT Primary Containment Integrity......................

B 3/4 6-1 Primary Containment Leakage........................

B 3/4 6-1 Primary Containment Air Lock.......................

B 3/4 6-1 i

MSIV Leakage Control System........................

B 3/4 6-1 Primary Containment Structural Integrity...........

B 3/4 6-2 Orywell and Suppression Chamber Internal Pressure........................................

B 3/4 6-2 Drywel l Ave rage Ai r Temperature....................

B 3/4 6-2 Drywell and Suppression Cha:nber Purge System.......

B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................

B 3/4'6-3 LIMERICK - UNIT 2 xn

[PPLICABILITY s

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SURVEILLANCE REQUIREMENTS (Continued)

ASME Boiler and Pressure vessel Required frequencies l

Code and applicable Addenda for-performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days i

Yearly or annually At least once per 366 days The provisions of Specification 4.0.2.are applicable to the above c.

required frequencies for performing inservice inspection and testing activities.

d.

Performance of the above inservice Inspection and testing activities shall be in addition to other specified Surveillance Requirements.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to e.

supersede the requirements of any Technical Specification.

f.

The Inservice Inspection (ISI) Program for piping-identified in NRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods and personnel, and sample expansion included in the Generic Letter. Details for implementation of these requirements are included as augmented inspection requirements in the'!SI-Program.

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LIMERICK - UNIT 2 3/4 0-3 Anendment No.12

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REACTOR COOLANT SYSTEM i

I OPERATIONA.L LEAKAGE j

LIMITING COW 0! TION FOR OPERATION 3.4.3.2 Reactor coo? ant system leakage shall be limited to:

4 4.

No PRESSURE B0bHDARY LEAKAGE.

I b.

5 gpm UNIDENTIFIED LEAKAGr..

c.

30 gpm total leakage.

4 d.

25 gpm total leakage averaged over any 24-hour period.

1 gpm leakage at a reactor coolant system pressure of 950 +10 psig from any e.

+

reactor coolant system pressure isolation valve specified Tn Table 3.4.3.2-1.

f.

2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With any PRESSURE BOUNDARY LEAXAGE, be in at least HOT SHUTDOWN within 12 a.

hours and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With any reactor coolant system leakage greater than the limits in_b. c and/or d above, reduce the leakage rate to within the-limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and i' COLD SHUTO within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , With any reactor coolant system pressure isolation valve leakage greater c.

than the above limit, isolete 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 one i

other closed manual, deactivated automatic, or check

valves, c: 'be-in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN w following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i t

d.

With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperoble i

monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore the; t

inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next'12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ano in COLD SHilT00WN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With any reactor coolant system leakage greater than the limit in f.above.

e.

identify the source'of leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in 4t.least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUT 00Wd within the fo D owing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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Which have been verified nor to exceed the allowable leakage limit at the last refueling outage or after thi. last time the valve was disturbed, whichever is more recent.

LIMERICK - UNIT 2 3/4 4-9 Aff.eadment No.12 a.

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I

_ REACTOR COOLANT SYSTEM 1

SURVEILLANCE REQUIREMENTS l

l 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within l

each of the above limits by:

Monitoring the primary containment atmospheric gaseous radioactivity at 6.

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (not a means of quantifying leakage),

b, Monitoring the drywell floor drain sump and drywell eyuipment drain tank flow rate at least once per eight (8) hours, l

Monitoring the drywell unit coolers u noensate flow rate at least once c.

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , d.

Monitoring the primary containment pressure at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (not a means of quantifying leakage),

Monitoring the reactor vessel head flange leak detection system at e.

least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and f.

Monitoring the primary containment temperature at lent once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (notameansofquantifyingleakage).

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak tr. sting pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:

a.

At least once per 18 months, and Prior to returning the valve to service following maintenance, repair b.

or replacement work on the valve which could affect its leakage rate.

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

i 4

4.4.3.2.3 The high/ low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints set less than the allowable values in Table 3.4.3.2-1 by performance of a:

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a.

b.

CHANNEL CALIBRATION at least once per 18 months.

l

. LIMERICK - UNIT 2

_3/4 4-10 Amendment No. 12

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

l APPLICAdILITY 0'SES A

j If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply witn ACTION requirements, e.g.,

Specification 3.0.3. a 24-hour allowance is provided to perm lt a delay in implementing the ACTION requirements. This provides an adequate time limit to ccmplete Surveillance Requirements that have not been performed.

The purpose of this allowance is to permit the comp'etion of a surveillance before a shutdown would be required to comply with ACTION requirements or before other remedial measures would be required that may preclude the completion of a surveillance.

The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in completing the rM uired l

surveillance.

This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of CONDITION changes imposed by ACTION requirements and for completing Surveillance Requirements that are apolicable when an exception to the requirements of Specification 4.0.4 is allowed.

If a surveillance is not completed within 24-hour allowance, the time limits of the ACTION requirements are applicable at that time.

When a surveillance is performed within the 24-hour allowance and the Surveillance Reovirements are not met, the time limits of the ACTION requirements are apolicable at the time that the surveillance is terminated.

i Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that 6pply.

However, the Surveillance Requirements have to be met tc demonstrate that inoperable equipment has been restored to OPERABLE status.

Specification 4.0.4 establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL CONDITION or other condition of operation specified in the Applicability statement.

The purpose of this specification is to ensure that system and comporant OPERABILITY requirements or parameter limits are met before entry into an-OPERATIONAL CONDITION or other specified condition for which these systems and components ensure safe operation of the facility.

This provision applies to changes in OPERATIONAL CONDITIONS or other specified conditions associated with plant shutdown as well as startup.

j Under the provisions of this_ specification, the applicable Surveillance i

Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial _ plant startup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not ap;1y because this would delay placing the facility in a lower CONDITION of operation.

Specification 4.0.5 establishes the requirement that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Coou and Addenda as i

required by 10 CFR 50.55a. These requirements apply except when relief has been provided in writing by the-Cornission. Additionally, the Inservice Inspection P m gram conforms to the NRC staff positions identified in NRC Generic letter 88-01, "NRC Position on IGSCC in BWR Austinetic Stainless Steel Piping," as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990, i

LIMERICK - UNIT 2 8 3/4 0-5 Amendment No. 12 i

APPLICABILITY r

BASES This specification includes a clarification of the frequencies for performing the inst. v'ce inspection and testing activities required by Stetion XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda.

This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Srscifications and to ree:ve any ambiguities relative to the frequencies for perterning the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and pressure Vessel Code and applicable Addenda.

The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL CONDITION or other specified condition takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows one week after retu.rn to normal operation. pumps and valves to be tested up to The Technical Specification definition of OPERABLE does not allow a grace period before a component, which is not capable of performing its specified function, is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision that allows a valve to be incapable of performin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable. g its specified function for up to i

LIMERICK - UNIT 2 B 3/4 0-6 L

l

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REACTOR COOLANT SYSTEM

'l BASES 4

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGEDETECTIONSYSTQS, monitor and detect leakage t,om the reactor coolant pressure boun Coolant Pressure Boundary leakage Detection Systems," "sy 1973. sys In conformance with Regulatory Guide 1.45, the channel calibration tests will verify the ability to detect a1gpmleakinlegsthanihourandanatmosphericgaseousradioactivitysystem sensttivity of 10 pC/cc.

3/4.4.3.2 OPERATIONAL LEAKAGE predicted and experimentally observed behavior of cracks in pipes.T The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.

obtained from experiments suggests that for leakage somewhat greater than thatThe evidence specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

leakage rates exceed the values specified or the leakage is located and known to beHo PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to hilow further inv and correcthe action.

The li. nit of 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period and the monitoring of drywell floor drain sump and drywell equipment drain tank flow rate at least once every eight (8) hours conforms with NRC staff positions specified in NRC Generic letter 88-01, "NRC Position on ISSCC in BWR Austenitic Stainless Steel Piping," as revised by NRC Safety Evaluatic,1 dated March 6, 1990.

ACTION requirement for the 2 gpm increase in UNIDENTIFIEC LEAKAGE limit ensures t The such leakage is identified or a plant shutdown is initiated to allow further investigation and corrective action.

dependent upon the impact on total leakage.Once identified, reactor operation may continue The ACTION requirements for pressure isolation valves (PIVs) are used in L

7 and with primary containment isolation valve requirements to en operation is appropriately limited.

The Surveillance Requirements for the RCS pressure isolation valves provide added and consequent intersystem LOCA.of valve integrity thereby reducing the probability of gros assuranca IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

LIMERICK - UNIT 2 B 3/4 4-3 Amendment No.12

.. s REACTOR COOLANT SYSTEM i

BASES

\\

l 3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. -Chlorido limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 3.2 ppm limit on chlorides is permitted during POWER OPE A;J10H.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so-a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

Conductivity measurements are required-on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits._ With the conductivity meter inoperable, additicnel samples must be analyzed to ensure that the chloric 4: are not exceeding the limits.

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

action, f

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l LIMERICK - UNIT 2 B 3/4 4-3a Amendment No.12 l

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4 REACTOR COOLANT SYSTEM BASES l

3/4.4.5 $PECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the i

limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These velure are conservative s

in that specific site parameters, such as SITE BOUNDARY location and meteoro-logical conditions, wure not considered in this evaluation, The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4 micro-curies per gram 00SE EQUIVALENT I-131, accommodates r,cssible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 0.2 microcuric por gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram DOSE EQ"IVALENT I-131 must be restricted since these activity levels increase the 2-hour thyroid dose at the SITE B0UNDARY following a postulated steam line rupture, i

Closing the main steam line isolation valtes prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective actinn.

3/a.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

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L1MERICK - UNIT 2 B 3/4 4-4

.m R'EACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) l The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements cf 10 CFR 50 Appendix G_and ASME Code Section III, Appendix G.

The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components.

the basis for compliance are more fully discussed in FSAR Chapter 5, Para-Fracture graph 5.3.1.5, " Fracture Toughness."

The reactor vessel materials have been-tested to determine their initial RT The results of these tests are shown in Table B 3/4.4.6-1.

HDT.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, NDT.

based upon the fluence, nickel content and copper content of the material 1

in question, can be predicted using Bases Fi Vessel Materials."y Guide 1.99, Revision 2, gure.iation Embrittlement of ReactorB 3 tions of Regulator Rad curves A', B' and c', includes an assumed shift in RTThe pressure / temperature li NDT for the conditions at 8 EFPY, The actual shift in RTNDT of the vessel material will be established period-ically during operation by removing and evaluating, in accordance with 10 CFR Part 50, Appendix H, irradiated reactor vessel flux wire and charpy specimens-installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the charpy specimens and vessel inside radius are essentially identical, the irradiated charpy specimens can be used with con-fidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the flux wire and charpy specimen data and recommendations of, Regulatory Guide 1.99, Revision 2.

m The C, and C' pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves

, and A and A', for reactor criticality.and for inservice leak and hydrostatic testing _have been provided to assure compliance with the minimum and for inservice leak and hydrostatic testing. temperature requirements The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are pro-vided in Table 4.4.6.1.3-1 Appendix H to 10 CFR Part 50,to assure compliance with the requirements of 1

{l LIMERICK - UNIT 2 B 3/4 4-5

~

i REACTOR COOLANT SYSTEM

(

I BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provideo on each of the main steam lines to minimize the potential leakage paths ' rom the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single f ailure considerations reesi'e that two valves be OPERABLE.

The surveillance requirements are based on t w operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these compenents will be maintained at-an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit insevice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

The inservice inspection program for ASME Code Class 1, 2, and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g)-except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in HRC Generic 1.etter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," as approved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990.

3/4.4.9 RES. ;AL HEAT REMOVAL A single shutdown cooling mode loop providet sufficient heat removal capability for removing core decay _ heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in osieretion, l

LIMERICK - UNIT 2 B 3/4 4-6 Amendment No. 12

ML20070K156

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