ML20101S887

Amends 173 & 150 to Licenses DPR-53 & DPR-69,respectively, Revising TS to Provide Conditions Under Which SG Insp Intervals May Be Extended to 30-months,per GL 91-04

ML20101S887
Person / Time
Site:	Calvert Cliffs
Issue date:	07/13/1992
From:	Capra R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20101S889	List: ML20101S887 ML20101S894
References
GL-91-04, GL-91-4, NUDOCS 9207200127
Download: ML20101S887 (27)

Text

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'o UNITED STATES i

NUCLEAR REGULATORY COMMISSION g... + /[

WA5HINGTON, D.C. 20066 f

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BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 i

CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NU. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.173 License No. DPR-53 1.

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

A.

The application for. amendment by Baltimore Gas and Electric ~ Company (the licereee) dated March 25, 1992, as supplemented'on May 28 1992, complies with the standards and requirements of the. Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and c

regulations set forth in-10 CFR Chapter I; B.

The facility will o)erate in conformity with the application, the provisions of tie 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 complian:e with the Commission's regulations; =

0.

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 "sndment 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 Techn, cal Specifications as. indicated in the attachment to thisilicense. amendment,'

and_ paragraph 2.C.(2) of Facility __ Operating License ~No. DPR-53 is hereby amended to: read as follows::

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Technical Specification; The Technical Specifications contained in Appendices A and 8, as revised throagh Amendment No.173, are hereby incorporated in the license.

The licensee shal' operate the facility in accordance with the Technical Specifications.

3.

This license amendment is affactive at of the date of its issuance and shall be implemented within 30 odys.

FOR THE NUCLEAR REGULATORY COMMISSION W M c? f W Robert A. Capra, Director Project Directorate I-l Division of Reactor Projects - I/11 Office of Nucioar Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 13, 1992 i

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p %,g UNITED STATES 3,

i NUCLEAR REGULATORY COMMISSION I

WASHINGTON, o C. 20666 os,...../

I BALTIMORE GAS AND ELECTRIC COMPAN'Y DOCKET NO. 50-318 CALVERT CLIKS NUCLEAR POWER PLANT. UNIT NO. 2 I

l AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.150 License.No. DPR l 1.

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

A.

The application for amendment by Baltimore. Gas and Electric Company I

(the licensee) dated March 25, 1992, as supplemented on May 28, 1992, 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; q

B.

The facility will o)erate in conforn.ity._with the application, the provisions of t1e 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 Commission's regulations; D.

The issuance of this amendment will not be inimical to the common 1

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 l

51 of the Commission'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. DPR-69 is hereby amended to. read as follows:

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, (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.150, 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 the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Wh Robert A. Capra, Director Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 13,1992 l

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e e-e-m,4&4 ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT,NO.173 FACILITY OPERATING LICENSE NO. DPR-53 AMENDMENT NO.150 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 j

l Revise Appendix A as follows:

Remove Paaes Insert Paaes 3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 i

3/4 4-12*

3/4 4-12*

3/4 4-13*

3/4 4-13*

3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 B3/4 4-3 B3/4 4-3 83/4 4-4 B3/4 4-4 B3/4 4-5*

B3/4 4-5*

B3/4 4-6*

B3/4 4-6*

B3/4 4-7*

B3/4 4-7*

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e 3/4.4 kEACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.

Tubes in those treas where experience has indicated potential problems.

c.

The second and third inservice inspections may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

1 The results of each sample inspection shall be classified into one of the following three categories:

. _at_egon Inspection Results e

C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1%

of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note:

In all inspections, previously degraded tubes must exhibit significant (> 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Inspection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a.

The first inservice inspection shall be perfomed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be perfomed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If at least 20 percent of the tubes were inspected and the resu!ts were in the C-1 Category or if. at least 40 pt.rcent of the tubes were inspected and were in the C-2 Category during the previous inspection, the next inspection may be extended up to a maximum of 30 months in order to correspond with the next refueling l

CALVERT CLIFFS - UNIT 1 3/4 4-10 Amendment No. 173

3/4.4 REACTORC0OLANTSYSTEM SURVEILLANCE REQUIREMENTS (Continued) outoge if the results of the two previous inspections were not in the C-3 Category. However, if the results of either of the previous two inspections were in the C-2 Category, a1 engineering assessment shall be perfonned before operation beyond 24 months and ' hall provide assurance that all tubes will retain adequate structural margins against burst throughout normal operating, transient, and accident conditions until the end of the fuel cycle or 30 months, whichever occurs first.

If two consecutive inspections following service under AYT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two co'isecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, theinspection interval may be extended to a maximum of once per 40 months.

b.

If the inservice inspection of a steam generator conducted in.

accordance with Table 4.4-2 at 40-month intervals fall in u.tegory C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequant inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 30 or 40 e nths, as applicable.

c.

Additional, unscheduled inservice inspections shall be perfonned on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1.

Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, 2.

A seismic or: Jrrence greater than the Operating Basis Earthauake, 3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.

A main steam line or feedwater line break, d.

The provisions of Specification 4.0.2 do not apply for extending the frequency for perfonning inservice inspections as specified in Specifications 4.4.5.3.a and b.

CALVERT CLIFFS - UNIT 1 3/4 4-11 Amendment No. 173

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3/4.4 REACTM_ CME 511 TIM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.

As used in this Specifications 1.

Imperfectio_r,1 means an exception to the dimensions, finish or -

contour of a tub 9 from that required by fabrication drawings or specificatit.ns.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be i

considered as imperfections.

2.

Dearadation means a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a tube.

3.

Iscaraded Tu_bg means.a tube contalning imperfections t 20% of j

the. nominal wall thickness caused by degradatioit.

4.

% Dearadation means the percentage of the tube wall thickness affected or removed by degradation.

5.

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

Any tube which does not pemit the passage of the eddy-current inspection probe shall-be deemed a defective tube.

6.

Pluaaina Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next ins)ection and is equal to 40% of the nominal tube wall thiccness.

7.

Unserviceable describes.the condition of a tube if it leaks or contains a defect large enough to affect its structural-integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or-a steam line or feedwater line-break as specifled in 4.4.5.3.c above.

8.

Tube Lnspection means an inspection of'the steam generator tube from th? point of entry (hot _ leg side) completely around the U-bend to'the. top support of the cold-leg.

.b.

The steam generator shall be detemined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table-4.4-2.

CALVERT CLIFFS - UNIT.1 3/4 4-12 Amendment No. 169,-173

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i 3/4.4 REACTOR COOLANT $YSTEM i

SURVEILLANCE REQUIREMENTS (Continued) 4 4.4.5.5 Reports a.

Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 15 days pursuant to Specification 6.9.2.

b.

The complete results of the steam generator tube Inservice inspection shall be included in the Annual Operatin the period in which this inspection was completec (g Report for pursuant to-i Specification 6.9.1.5.b).

This report shall include -

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indica'. ion of an imperfection.

3.

Identification of tubes plugged.

c.

Results of steam unerator tube inspections which fall-into i

Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-prior to resumption of

-plant operation.-

The written followup of this report shall provide a description of-investigations conducted to detennine l

cause of the tube degradation and corrective measures taken to prevent recurrence and shall be submitted within the next 30 days pursuant to Specification 6.9.2.

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CALVERT CLIFFS - UNIT 1 3/4.4 Amendment No. 169, 173

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3/4.4 RFACTOR COOLANT SYSTEM 3//.4.6 REACTOR COOLANT SYSTEM LEAKAGE Reactor Coolant System Leakace LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE B0UNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through all steam generators and 100 ga11cns-per-day through any one steam generator, and 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 H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next b hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

a.

Either:

1.

Monitoring the containment atmosphere particulate or gaseous radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,'or 2.

With the gaseous and partit.ulate monitors inoperable, conducting the containment atmosphere grab sam)1e analysis in accordance with the ACTION requirements of Ter.inical Specification 3.4.6.1.

CALVERT CLIFFS - UNIT 1 3/4 4-18 Amendment No. 173

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3/4.4 REACTOR C00 M SYSTEM SURVEILLANCE REQUIREMENTS (Cor,tinued) b.

Monitoring the containment sump discharge frequency at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, when the Containment Sump Level Alam System is OPERA 8LE, c.

Detemining Reactor Coolant System leakage at least once per l

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> d"-ing steady state operation atid at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when required by ACTION 3.4.6.1.b. except when operating in the shutdown coolir.g mode, and F

d.

Monitoring-the reactor vessel head closure seal Leakage Detection System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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3/4.4 REACTOR COOLANT SYSTEJ BASES bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDB7.

3/4.4.5 STEAD GENERATORS The Surveillance Requiremen'.; for inspection of the steam generator tubes ensure that the structursl integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

An engineering assessment of steam generator tube integrity will confirm -

that no undue risk is associated with plant operation beyond 24 months of the previous steam generator tube inspection.

To provide this confinnation, the assessment would demonstrate that all tubes will retain adeauate structural margins against burst during all nonnal operating, transient, and accident conditions until the end of the fuel cycle. This evaluation would include the following elements:

1.

An assessment of the flaws found during the previous inspections.

2.

An assessment of the structural margins relative to the criteria of Regulatory Guide 1.121. " Bases for Plugging Degraded PWR Steam Generator Tubes," that can be expected before the enu of the fuel cycle or 30 months, whichever comes first.

3.

An update of the assessment model, as appropriate, based on comparison of the predicted results of the steam generator tube l

integrity assessment with actual inspection results from previous inspections.

l l

The plant is-expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in i

I negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result '

stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Primary Coolant System and the Secondary Coolant System (primary-to-secondary leakage = 1 gallon )er minute, total).

Cracks having a primary-to-secondary leakage less than t11s limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. 'Jperating plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator CALVERT CLIFFS - UNIT 1 B 3/4 4-3' Amenbment No. 173 9

3/4.4 REACTOR C00LANT SYSTEM BASES blowdown.

Lcakege in excess of this limit will require niant shutdowr, and an unschedu.ed inspection, during which the leaking tut.

will be located and plugged.

l Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develo) in service, it will be found during scheduled inservice steam generator tu)e examinations.

Plugging will be required for all tubes with imperfections exceeding the i

plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating p? ants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube j

wall thickness.

Whenever the results of any steam generator tubing inservice inspection l

fall into Category C-3, these results vill be promptly reported to the Comission pursuant to Specificat.icns 6.9.2 prior the resumption ofplant operation.

Such cases will be considered by the Comission on a case-by-case basis and may result in a recuirement for analysis, laboratory examinations, tests, additional ectiy-current inspection, and revision of the Technical Specificatiens, if necessary.

1 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 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 dettetion systems are consistent with the recomendations of Regulatory Guide 1.45, ' Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

3/4.4.6.2 Reactor Coolant System Leakage, L

Industry experience has shown that while a limited amount of leakage is l

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

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amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIFD LEAKAGE by the Leakage Detection Systems.

The total steam generator tube leakage limit of 1 GPM for all steam generators 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 tule rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

The 100 gallon per day leakage limit per steam generator ensures that steam-generator tube integrity is maintained in accordance wit" the recomendations of Generic letter 91-04.

CALVERT CLIFFS - UNIT 1 B 3/4 4-4 Amendment No. 173 P

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3/4.4 REACTOR COOLANT SYSTEM BASES PRESSURE BOUNDARY LEAKAGE of any maynitude 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.CXHjSTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolar.t System leakage or failure due to stress corrosion.

Mair.taining the chemistry within the Steady State Limits provides adequate corrosica protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects ofexceeding 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 significent effect on the structural integrity ef the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Trensient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

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

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that l

the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leckage rate of 1.0 gpm and a concurrent loss of offsite electrical power.

The values for the limits on specific u tivity represent interim limits based upon a parametric evaluation by the i

N (C of typical site locations. These values are conservative in that d

.pecific site parameters of the Calvert Cliffs site, such as SITE B3UNDARY b"

location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site.

This reevaluation may result in higher limits.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4.8-1, accomodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity l

levels exceeding 1.0 pCi/ gram DOSE EQUIVALENT I-131 tat within the limits i

shown on Figure 3.4.8-1 must be restricted to no more than 10 percent of the unit's yearly operating time since the activity levels allowed by CALVERT CLIFFS - UNIT 1 B 3/4 4-5 Amendment No. 48 173

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3/4.4 REACTOR COOLANT SYSTEM BASES Figure ?.4.8-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the SITE B0UNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T,,,be rupture since the saturation pressure of the primary coolant to < 500'F prevents the release of activity should a steam generator tu is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessivespecific activity levels in the primary coolant will be detected in sufficies Hme to take corrective action.

Information obtained on 1odine spiking w

. be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be pemissible if justified by the data obtained.

3/4.4.9 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 nomal load transients, reactor trips, and STARTUP and shutdown operation.

The various categories of load cycles used for-design purposes are provided in Section 4.1.1 of the UFSAR.

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.

Operation within the appropriate heatup and cooldown curves assures the integrity of th! reactor vessel against fracture induced by combinative themal and preuure stresses. As the vessel is subjected to increasing fluence, the tourness of the limiting material continues to decline, and ever more restrictive Pressure / Temperature limits must ba observed. The current limits, figures 3.4.9-1 and 3.4.9-2, are for a peak neutron fluence which corresponds to approximately 22 Effective,3.25x10 'N/cmFull Power to the inner surface of-the reactor vessel of <

r (E > 1 MeV,

l operation.

The reactor vessel materials have been tested to detemine their initial I

RT 1; the results of these tests are shown in Section 4.1.5 of the UFSAR.

l Reactor operation and resultart fast neutron (E > 1 McV) irradiation will I

cause an increase in the RTi The actual shift in RT l

material will be establishedt. periodically during opera, of the vesseltion by re evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimer.s are provided in "'1AR Table 4-13 and are approved by the NRC prior to implementation'in compliance with the requirements _of 10 CFR Part 50. Appendix H, The shift in the material fracture toughness, as represented by RTut, is calculated using Regulatory Guide 1.99, Revision 2.

For a fluence of 3.25x10"N/cm, the adjusted reference temperature (ART) value at the 1/4 T 2

position is 253.PF. At the 3/4 'T positio.1 the ART value is 193.8'F.

-CALE RT CLIFFS - UNIT 1 B 3/4 4-6 Amendment No. M, 173

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3/4.4 REACTOR COOLANT SYSTEM BASES Thes ?alues are used with procedures developed in the ASME Boiler and Pressure. Vessel Code, Section 111 A)pendix G to calculate heatup and cooldown limits in accordance with tie requirements of 10 CFR Part 50, Appendix G.

To develop composite pressure-temperature limits for the heatup transient, the isothermal, 1/4 I heatup, and 3/4 T heatup pressure-temperature limits are cocpared for a given themal rate.

Inen the most restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.

To develop a composite pressure-temperature limit for the cooldown event, the isothemal pressure-temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit associated with a cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.

l'oth 10 CFR Part 50, Appendix G and ASME, Code Appendix G require the development of pressure-temperature limits which are applicable to inservice hydrostatic tests. The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure (1.1 times nomal operating pressure) and locating the corresponding-temperature. This curve is shown for a fluence of

5. 3.25x10"N/cm' on Figures 3.4.9-1 and 3.4.9-2.

l Similarly,10 CFR Part 50 specifies that core critical limits be t

established based on material considerations. This limit is shown on the

(

heatup curve, Figure 3.4.9-1.

Note that this limit does not consider the core reactivity safety analyses that actually control the temperature at which the core can be brought critical.

The Lowest Service Temperature is the minimum allowable temperature at l

pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia).

This temperature is defined as equal to the most limiting RTm for the balance of the Reactor Coolant System components plus 100 F, per Article N3 2332 of Section III of tne ASME Boiler and Pressure Vesse Code.

The horizontal line between the minimum boltup temperature and the Lowest Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure.

The change in the line at 150 F on the cooldown curve is due to a cessation of RCP flow induced pressure deviation, since no RCPs are pemitted to operate during a cooldown below 150 F.

CALVERT CLIFFS - UNIT 1 8 3/4 4-7 Amendment No. MA,173

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2.

Tubes in those areas where experience has indicated potential problems.

c.

The second and third inservice inspections may be less than a full tube ins)ection by concentrating (selecting at laast 50% of the tubes to 3e inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

CategoII Inspection Results q

C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

i i

C-3 More than 10% of the total tubes inspected l

are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (> 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Inspection Freauer.cies - The above required inservice inspections l

I of steam generator tubes shall be perfomed at the following frequencies:

a.

The first inservice inspection shall-be perfomed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequant inservice inspections shall be perfomed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If at least 20 percent of the tubes wer)e ins)ected and the results were in the C-1 Category (See Note or < f at least 40 percent of the tubes were inspected and were in the C-2 Category during the previous inspection, the next inspection may be extended up to a maximum of 30 months in order to correspond with the next refueling outage if the results of the two previous inspections

• NOTE:

For Cycle 9 ar. inspection of 15% of the steam generator tubes with inspection results in the C-1 Category shall be acceptable to extend the next inspection up to 30 months to coincide with the next refueling outage.

CALVERT CLIFFS - UNIT 2 3/4 4-10 Amendment No. 150.

e e

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) were not in the C-3 Category.

However, if the results of either of the previous two inspections were in the C-2 Category, an engineering assessment shall be performed before operatton beyond 24 months and shall provide assurance that all tubes will retain adequate structural margins against burst throughout nomal operating, transient, and accident conditions until the end of the fuel cycle or 30 months, whichever occurs first.

If two consecutive ins)ections following service under AVT conditions, not including tie preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b.

If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection-frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 30 months or 40 months, as applicable.

c.

Additional, unscheduled inservice inspections shall be perfomed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:-

1.

Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, 2.

A seismic occurrence greater than the Operating Basis Earthquake, 3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.

A main steam line or feedwater line break.

d.

The provisions of Specificatin 4.0.2 do not apply for extending i

the frequency for perfoming inservice inspections as specified in Specifications 4.4.5.3.a and b.

l CALYERT CLIFFS - UNIT 2 3/4 4-11 Amendment No. 150

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Cor _'tued) 4.4.5.4 Acceptance Criteria a.

As used in this Specification:

1.

Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

Dearadation means a service-induced cracking,

wastage, wear f

or general corrosion occurring on either inside or outside of a tube.

3.

Deoraded Tube means a tube containing imperfections > 20% of the nominal wall thickness caused by degradation.

4.

% Degradatign means the percentage of the tube wall thickness affected or removed by degradation.

I 5.

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6.

Pluaaina Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable or'or to the next iiispection and is equal to 40% of the nominal tube wall thickness.

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c. abovn. -

8.

Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg, b.

The stea:.; generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

CALVERT CLIFFS - UNIT 2 3/4 4-12 Amendment No. 449', 150

3/4.4 R MCTOR COOLANT TYSIEM

[

SURVEILLANCE REQUIREMENTS (Continued) f 4.4.5.5 Reports a.

Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days pursuant to Specification 6.9.2.

b.

The complete results of the steam generator tube inservice inspection shall be included in the Annual Operatin theperiodinwhichthisins)ectionwascompleted(gReportfor pursuant to Specification 6.9.1.5.b).

Tais report shall include:

1.

Number and extent of tubes inspected.-

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.

Identification of tubes plugged.

c.

Results of steam generator tube inspections which fall into Category C-3 require verbal notification of'the NRC Regional Administrator by telephone within-24 hours prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the' tube degradation and corrective measures taken to prLvent recurrence and shall be submitted within the next 30 days pursuant to Specification 6.9.2.

i L

l 1

CALVERT CLIFFS - UNIT 2 3'/4 4-13 Anendment No.L199,:150 m*

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3/4.4 REACTOR C0OLANT SYSTEN 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Reactor Coolant System Leakace LINITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited tot a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, 1 GPM total primary-to-secondar generators and 100 gallons-per y leakage through all steam c.

day through any one steam generator, and d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLI_CABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above _ limits, excluding PRESSURE B0UNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

4 a.

Either:

1.

Monitoring the containment atmosphere particulate or gaseous radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or 2.

With the gaseous and particulate monitors inoperable, conducting the containment atmosphere grab sample analysis in accordance with the ACTION requirements of Technical Specification 3.4.6.1.

CALVERT CLIFFS - UNIT 2 3/4 4-18 Amendment No. 150-

. - - - - - ~. -

3/4.4 REACTOR C00lANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) h.

Monitoring the containment sump discharge frequency at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, when the Containment Sump Level Alarm System is

OPERABLE, c.

Detennining the Reactor Coolant System water leakage at leaet l

once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when required by ACTION 3.4.6.1.b. except when j

operating in the shutdown cooling mode, and d.

Monitoring the reactor vessel head closure seal Leakage Detection-System at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

l l

.CALVERT. CLIFFS - UNIT 2-3/4 4-19 Amendment No. 150

t 3/4.4 M ACTOR COOLANT SYSTEM BASES a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizir;g the nature and cause of any tube degradation so that corrective measures can be taken.

An engineering assessment of steam generator tube integrity will confirm that no undue risk is associated with plant operation beyond 24 months of the previous steam generator tube inspection. To provide this confimation, the assessment would demonstrate that all tubes will retain adequate structural margins against burst during all nomal operating, transient, and accident conditions until the end of the fuel cycle.

This evaluation would include the following elements:

1.

An assessment of the flaws found during the previous inspections.

2.

An assessment of the structural margins relative to the criteria of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," that can be expected before the end of the fuel cycle or 30 months, whichever comes first.

3.

An update of the assessment model, as appropriate, based on comparison of the predicted results of the steam generator tube integrity as:essment with actual inspecticn results from previous inspections.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Primary Coolant System and the Secondar Coolant System (prirnary-to-secondary leakage = 1 gallon )er minute, total)y Cracks having i primary-to-secondary leakage less than t11s limit during operation will have an adequate inarhtf safety to withstand the loads imposed during normal operation an6 W c 'stulated accidents. Operating plants have demonstrated that primary-tb-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator CALVERT CLIFFS - UNIT 2 8 5/4 4-3 Amendment No. 150 l

3/4.4 _ REACTOR C0OLANT SYSTEM BASES blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes wib be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develo) in service, it will be found during scheduled inservice steam generator tuae examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whencver the results of any ster.m generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specifications 6.9.2 prior to the resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEH LEAKAGE 3/4.4.6.1 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 recommendations of Regulatory Guide 1.45, ' Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973.

3/4.4.6.2 Reactor Coolant System 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.

The total steam generator tube leakage limit of 1 GPM for all steam generators 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 11; eak.

The 1 GPM limit is consistent with the assumptions used in tnt analysis of these accidents.

The 100 gallon-per-day leakage limit per steam generator ensures that steam generator tube integrity is maintained in accordance with the recommendations of Generic Letter 91-04.

CALVERT CLIFFS - UNIT 2 B 3/4 4-4 Amendment No. 150

3/4.4 REACTOR C0OLANT SYSTEN BASES PRES $URE B0UNDARY 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 Cooltnt System is minimized and reduce 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 associatedeffects 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 Stcte Limits, up to the Transient Limits, for the specified limited time intwvals 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.

z 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the SITE B0UNDARY will not exceed an ap)ropriately small fraction of Part 100 limits following a steam gevrator tu)e rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Calvert Cliffs site, such as SITE B0UNDARY location and meteorological conditions, were not considered in this evaluation.

The NRC is finalizing site specific criteria which will be 1

used as the basis for the reeve,uation of the s)ecific activity limits of this site. This reevaluation may result in hig1er limits.

The ACTION statement permitting POWER OPERATION to continue-for limited tine periods with the pririary r,oolant's specific activity > 1.0 pCi/ gram DOSE EQUIVAF.ENT I-131, but within the allowable limit shown on Figure 3.4.3-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pC1/ gram D0SE EQUIVALENT I-131 but within the limits shown on Figure 3.4.8-1 must be restricted to no more than 10 percent of CALVERT CLIFFS - UNIT 2 B 3/4 4-5 Amendment No. M9,150

3/4.4 REACTOR COOLANT SYSTEM sAS45 the unit's year, operating time since the activity levels allowed by Figure 3.4.8-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the SITE B0UNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T.,, to < 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam reitef valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to t3ke corrective action.

Infomation obtained on iodine s)iking will be used to assess the parameters associated with spiking pienomena.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 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 operation. The various categories of load cycles used for design purposes are provided in Section 4.1.1 of the UFSAR.

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.

Operation within the appropriate heatup and cooldown curves assures the integrity of the reactrr vessel against fracture induced by combinative thermal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and even more restrictive Pressure / Temperature limits must be observed. The current limits, Figures 3.4.9-1 and 3.4.9-2, are for up to and including 12 Effective Full Power Years (EFPY) of operation.

The reactor vessel materials have been tested to detemine their initial RTm; the results of these tests are shown in Section 4.1.5 of the UFSAR.

Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RT, material will be establishedi. The actual shift in RTm of the vessel periodically during operation by removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4-13 and are approved by the NRC prior to implementation in compliance with the requirements of 10 CFR Part 50, Appendix H.

CALVERT CLIFFS - UNIT 2 B 3/4 4-6 Amendment No. 149', 150

=

3/4.4 Ma.CTOR COOLANT SYSTEM 1

BASCS The t.hift in the material fracture toughness, as represented by RT., is calculated u:;ing Regulatory Guide 1.99, Revision 2.

For 12 EF0Y, at the 1/4 T position, the adjusted reference temperature.(ART) value is 171'F.

At the 3/

T position the ART value is 125'F. These values are used with orocedur. developed in the ASME Boiler and Pressi:re Vessel Code, JuW.II, Appendix G to calculate heatup and cooldown limits in

t.. e ece with the requirerents of 10 CFR Part 50,. Ap)endix G.

To awlop composite pressure-temperature limits for tie heatup transient, the isothermal,1/4 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thennal rate.

Then the most restrictive pressure-W ar":re limits are combined over the complete temperature' interval ig a in a composite limit curve for the reactor vessel br itline for r

.nt.

.ll

. omposite pressure-tempeuture limit for the cooldown event, J

w nh 9:1 pressure-temperature liinit must be calculated. The isott _ ' aressure-temperature limit'is then compared to the press e -

tem r ~

a limit associated with a cooling rate and the more restrk.t;ve al,

. ;ressure-temperature limit is chosen resulting in a composite limit 1 for the reactor vessel beltline.

Both 10 CFh Part 50, Appendix G and ASME, Code Appendix G require the deulopment of pressure-temperature limits which are applicable to uservice hydrostatic tests. The minimum temperature _ for the inservice hydrostatic test pressure can be determined by entering f.he curve at the test pressure (1.1 times normal operating pressure) and locating the i

corresponding temperature. This curve is shown for 12 EFPY on l

Figures 3.4.9-1 and 3.4.9-2.

Similarly,10 CFR Part 50 specifies that c, ore critical limits be established bhsed o 9at t al considerations. This limit is shown on the heatup curve Five

.6 0 -1.

Note that tlis limit does not_ consider the core reach.:vity safety aralyses that actual.y control the temperature at which tu core can be brought critical.

The Lowest Service temperature is the minimum-allowable temperature at p(625 psia).ressures abave 20% of the pre-operational. system hydrostatic test a ihis temperature is defined as equal to t.he most~ limiting RTm i

for the balance of the Reactor Coolant System components plus 100'F per Article NB 230 of Se: tion III' of tne ASME. Boiler and Pressure Vessel Code.

The horizontal line between the minimum boltup temperature and the Lowest 4

Service Temperature is defined by the ASME Boiler and Pressure Vessel Code as 20* cf the pre-operatonal hydrestatic test pressure.. The change ir the i

line at 150 F on the cooldown curve is due to a cassation of RCP flow induced pressure deviation. since no RCPs are pennitted to operate during !.

cooldown below 150 F.

i 4

CALVERT CLIFFS - UNIT 2 B 3/4 4-7 Amendment No. Y49,- 150

=

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