Text

Amends 223 & 199 to Licenses DPR-53 & DPR-69,respectively, Incorporating Refs to New C-E TR Describing SG Tube Sleeves, Deleting Refs to Previous C-E TR & Incorporating Sleeve/Tube Insp Scope & Expansion Criterion
	ML20202C356
Person / Time
Site:	Calvert Cliffs
Issue date:	11/18/1997
From:	Bajwa S; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20202C362	List: ML20202C356; ML20202C366;
References
	NUDOCS 9712030317
	Download: ML20202C356 (92)
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t UNITED STATES s

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3066Ho01 l

BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 I CALVERT CLIFFS NUCLEAR PQgER PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

223 License No. DPR-53 1.

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

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated November 30, 1995, as supplemented March 15, 1996, March 6, 1997, and June 27, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) that the activities authorized by this amentent 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. DPR-53 is hereby amended to read as follows:

9712030317 9711tg yDR ADOCK 05000317 PDR

. 2.

Technical Specifications i

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.223, 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 data of its issuance and shall be implemented within 30 days.

j FOR THE NUCLEAR REGULATORY COMISSION S. Singh Bajwa, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 18, 1997

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- UNITED STATES s

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON. D.C. 20006 4 001

%...../

BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

199 License No. DPR-69 1.

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

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated November 30, 1995, as supplemented March 15, 1996, March 6, 1997, and June 27, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; 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. DPR-69 is hereby amended to read as follows:

e

- i 2.

Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.199, are hereby incorporated in the license..The licensee shall o the Technical Specifications. perate the facility in accordance with 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 j'

S. Singh Bajwa, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Oate of Issuance:

November 18, 1997 e

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e ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 223 ' FACIL ITY OPERATING LICENSE NO. DPR-53 AMENDMENT NO. 199 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 Revise Appendix A as follows:

Remove Paaes Insert Paaes 3/4 4-10 3/4 4-10 3/4 4-12 through 3/4 4-43 3/4 4-12 through 3/4 4-45 B 3/4 4-4 through B 3/4 4-12 B 3/4 4-4 through B 3/4 4-12

Pages that did not change, but are overleaf.

e 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION i

3.4.5 Each steam generator shall be OPERA 8LE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERA 8LE status prior to increasing T,,, above 200*F.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERA 8LE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERA 8LE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Tables 4.4-2 and 4.4-3.

The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection.

The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

Where experience in similar plants with similar water chemistry a.

indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

b.

The first-inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1.

All nonplugged tubes that previously had detectable wall penetrations (> 20%), and CALVERT CLIFFS - UNIT 1 3/4 4-10 Amendment No. 223

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) outage 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, an engineering assessment shall be perfonned 'efore operation beyond 24 months a

and shall provide assurance that all tubes will retain adequate

! structural margins againtt 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 AVT conditions, not including the 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 Tables 4.4-2 and 4.4-3 at 40-nonth intervals fall l

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 suosequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a marimum of once per 30 or 40 months, as applicable.

Additional, unscheduled inservice inspections shall be perfonned c.

on each steam generator in accordance with the first sample inspection specified in Tables 4.4-2 and 4.4-3 during the l

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 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 4-12 Amendment No. 223

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REf)UIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

a.. As used in this Specification:

1.

Tubino or Tube means that portion of the tube or sleeve which fonns the primary system to secondary system pressure boundary.

2.

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.

3.

Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

4.

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

5.

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

6.

Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube containing a defect is defective. Any tube which does not pennit the passage of the eddy-current inspection probe shall be deemed a defective tube.

7.

Pluccino or Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging, or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection.

The plugging or repair limit imperfection depths are specified in percentage of nominal wall thickness as follows.

a, ori gi nal tube wa11................................ 40%

b, Westinghouse laser welded sleeve wall............. 40%

c.

ABB-Combustion Engineering Leak Tight Sleeve wa11. 284 8.

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 feedw6ter line break as specified in 4.4.5.3.c, above.

9.

Tube InsDection 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.

CALVERT CLIFFS - UNIT 1 3/4 4-13 Amendrtent No. 223

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_3 3/4.4 REACTOR C00LMT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

10. Tube Repair refers to a process that reestablishes tube serviceability. Acceptable tube repairs will be performed by L

the following processes:

l a) Westinghouse Laser Welded Sleeving as described in the proprietary Westinghouse Reports WCAP-13698, Revision 2

" Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators, Generic Sleeving Report," April 1995; and WCAP-14469, " Specific Application of Laser Welded

. Sleeving for the Calvert Cliffs Power Plant Steam Generators," November 1995, b) ABB-Combustion Engineering Leak Tight Sleeving as t

described in the proprietary ABB-Combustion Engineering -

Report CEN-630-P, Revision 01, " Repair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight Sleeves," August 1996.

A post-weld heat treatment during installation will be perfomed.

Tube repair includes the removal of plugs that were previously installed as a corrective or preventive measure.

A tube inspection per Specification 4.4.5.4.a.9 is required prior to returning previously piug-ed tubes to service.

b.

The steam generator shall be detemint:w OPERABLE after completing.

the corresponding actions (plug or repair all tubes exceeding the pluggin cracks)g or repair limit and all tubes containing through-wall required by Tables 4.4-2 and 4.4-3.

l-4.4.5.5 Reports a '.

Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to.the Com.ission within 15 days pursuant to i

10 CFR 50.4.

b. - The complete results of the steam generator tube inservice inspection during the report period shall be submitted to the Commission prior to March 1 of each year pursuant to 10 CFR 50.4.

This report shall include:

1.

Number and extent of tubes inspected.

2.

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

l 3.

Identification of tubes plugged or repaired.

l.

l l.

CALVERT CLIFFS - UNIT.1 3/4 4-14 Amendment No.-223 J

o i

.7

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Results of steam generator tube inspections which fall into c.

Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to detennine cause of the tube degradation and corrective measures taken to prevent recurrence and shall be submitted within the next 30 days put-suant to 10 CFR 50.4.

CALVERT CLIFFS - UNIT 1 3/4 4-15 Amendment No. 223 1

+

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g TABLE 4.4-1 lg i

r-h N:.NIPENI NINIBER OF SIDWI GENERATORS TO BE

=

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INSPECTED SWRING INSERVICE INSPECTION Preservice Inspection No Yes a

g No. of Stems Generators per Unit Two Thrce Four Two Three Four I

I U

First Inservice Inspection All One Two Two l

Second & Subsequent Inservice Inspections One One One On*3 y

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I TABLE NOTATION:

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I The inservice inspection may be limited to one steam generator on a rotatin 3 lt % of the tubes (where N is the number of steam generators in the plant)g schedule encompassing I

if the results of the i

j first or previous inspections indicate that all steam generators are perfoming in a like menner.

i y

Note that under some circumstances, the operating conditions in one or more steam generators may be i

i 5

found to be more severe than those in other steam generators. Under such circumstances the sample i

sequence shall be modified to inspect the most severe conditions.

g2 The other steaci generator not inspected during the first inservice inspectien shall be inspected.

f g

The third and subsequent inspections should follow the instructions described in I above.

I o

3 Each of the other two steam generators not inspected during the first inservice inspections shall be mO inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.

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n TABLE 4.4-2

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'w2 g

sum as=mune tuer aspecneN

-4 151 5-- d INy cctIUM oup L.

d IW W LcIIUN JINI L.

d Inwttligst sur r-Sample 51ze Nesult Action Required NesuIL Act1on Itegu1 red Nesutt Action Negueree E

M m etntmum or 5 I :;; per c-1 None N/A IE/A N/A N/A h

Q SG.

E-z Plug or resair derecttwe C-1 none R/A N/A l

tubes and inspect C-Z ring or repatr C-1 None additional 25 tubes in defective tubes and h

Q this SG.

linsper* additional c-z ring cr repair p

45 t e s in this defective tubes E

SG.

c-3 Perrom action for 4

C-3 result of

!~

first sample i

C-3 Te, orm act1on ror N

i C-3 result of first N/A N/A E

i sample c-3 Inspect ait tuees in m:1 other

]

this SG, plug or repair SGs are Nvme N/A N/A g

i.

defective tubes and C-1 i

inspect 25 tubes in each j

dW W.

e.*

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N 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal some 56s r, som act1on ror N/A N/A L

m notification to INIC with C-2 but no C-2 ret..t of I

written followup additlocal second sauqile pursuant to 10 CFR 50.4. SG are C-3 mesit1onai Inspect aii tubes SG is C-3 in each SG ard plug or repair defective II/A N/A tubes. 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months R

il verbal notification to INIC with written k"

fo11omme persmant g

to 10 (.rR 50.4.

=

t

, S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators n

P inspected during an inspection M

t' I

e m.,

s

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g TABLE 4.4-3 R.

r-g STEAM GENERATOR REPAIRED TUBE INSPECTION

~*

f2 IST SAMFLE IN5PECIl0M rnu 5 AMPLE IR5PLCIIUM h

i M

Sample Size Result Action Required Result Action Required O

e'n A minimum of 20% of C-1 None N/A N/A E

repaired tubes"H" k

c C-2 Plug defective repaired tubes C-1 None 5

and inspect 100% of the C-2 Plug detective

[

repaired tubes in this SG.

repaired tubes i

C-3 Perform action for C-3 result of first sample O

E C-3 Inspect all repaired tubes Other 56 is C-1 None in this SG, plug defective Other SG is C-2 Perform action for C-2 w

tubes and inspect 20% of the result of first sample i

repaired tubes in the Other SG.

Other 5G is C-3 Inspect alI repaired tubes in each SG and 24-hour verbal notification to plug defective tubes.

NRC with written follow-up, 24-hour verbal pursuant to 10 CFR 50.4.

netification to NRC l

with written follow-up, pursuant to 10 CFR E0.4.

k

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Each repair method is considered a separate population for determination of scope expansion.

E "1

R The inspection of repaired tubes may be performed on tubes from either SG based on outage plans.

a

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTORCOOLANTSYSTEMLEAKMf, Leakaoe Detection Systems LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a.

A Containment Atmosphere Particulate Radioactivity Monitoring

System, b.

The Containment Sump Level Alarm System, and A Containment Atmosphere Gaseous Radioactivity Monitoring System.

c.

APPLICABILITY: MODES 1, 2, 3 and 4 Ell @

a.

With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment etmosphere are obtained and analyzed at least once per 24 hourt w? n either the required Gaseous or Particulate Radioactivity Mordtoring 7ystem is inoperable; otherwise be in at least NOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD $HUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With only one of the above required Leakage Detectica Systems 0PERABLE, operation may continue for up to 7 days provided that:

1.

Grab samples of the containment atmosphere are obtained and analyzed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and 2.

The Reactor Coolant System water inventory balance of Surveillance Requirement 4.4.6.2.c is perfonned at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise be in at least NOT STANDBY within thc next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

o CALVERT CLIFFS - UNIT 1 3/4 4-19 Amendment No. 223 l

3/4.4 REACTOR _ COOLANT SYSTEM

$URVEILLANCE REQUIREMENTS 4.4.6.I The Leakage Detection Systems shall be demonstrated 0PERABLE by:

a.

Containment Atmosphere Gaseous and Particulate Monitoring Systems-perfomance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b.

Containment Sump Level Alam System-perfomance of C:'ANNEL CALIBRATION at least once per REFUELING INTERVAL.

f CALVERT CLIFFS - UNIT 1 3/4 4-20 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Reactor Coolant Sysf em Leakace LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage thall be limited to:

a.

No PRESSURE B0UNDARY 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 gener? tor, and d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least NOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRES $URE B0UNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least H0T STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

a.

Either:

1.

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

With the gaseous and particulate monitors inoperable, co.iducting the containment atmosphere grab sam)1e analysis in accordance with the ACTION requirements of Tec1nical Specification 3.4.6.1.

CALVERT CLIFFS - UND 1 3/4 4-21 Amendment No. 223 l

3/4.4 REACTH COOLANT SYSTDI

$URVEILLANCE REQUIREMENTS (Continued) b.

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

OPERABLE, c.

Detennining Reactor Coolant System leakage at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when required by ACTIM 3.4.6.1.b. except when 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

CALYLR1 CLIFFS - UNIT 1 3/4 4-22 Amendment No. 223

'l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY 1

LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1.

APPLICABILITY: At all tires.

ACTION:

MODES 1, 2, 3 and 4 a.

With any one or more chemistry parameter in excess of its Steedy State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With any one or more chenistry parameter in excess of its Transient Limit, be in at least NOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer 4

pressure to 5 500 psia, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to M0DE 4.

SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.

4

<CALVERT. CLIFFS - UNIT 1 3/4 4-23 Amendment No. 223 l

3/4.4 REACTOR C00UUfT SYSTEN JA)LE3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN

5 0.10 ppm 5 1.00 ppm CHLORIDE 5 0.15 ppm 5 1.50 ppm FLUORIDE

$ 0.15 ppm 5 1.50 ppm i

Limit not applicable with T.,,5 250*F.

CALVERT CLIFFS - UNIT 1 3/4'4-24 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM TABLE 4.4-3 REACTOR COOLANT SYSTEM E EMISTRY LIMITS SURVEILLANCE REQUIREMENTS Z888!fd[R ANALYSIS FREQUENCY DISSO.VED OXYGEN

At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CHL0f!DE At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months FLUOMDE At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Not required with T.,, s 250'F.

CALVERT CLIFFS - UNIT 1 3/4 4-25 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a.

5 1.0 pC1/ gram DOSE EQUIVALENT I-131, and b.

5 100/E pC1/ gram.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2 and 3*:

a.

With the specific activity of the )rimary coolant > 1.0 pC1/ gram DOSE EQUIVALENT I-131 but within t1e allowable limit (below and to the left of the line) shown on Figure 3.4.8-1, operation may continue for up to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of specification 3.0.4 are not applicable.

b.

With the specific activity of the primary coolant > 1.0 pCi/ gram D0SE EQUIVALENT I-131 for more than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months during one continuous time intervr! or exceeding the limit line shown on Figure 3.4.8-1, be it, et least NOT STANDBY with T.,, < 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With the specific activity of the primary coolant

> 100/E pCi/ gram, be in at least HOT STANDBY with T,,, < 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4 and 5:

d.

With the specific activity of the primary coolant > 1.0 pC1/ gram DOSE EQUIVALENT Z-131 or > 100/E pCi/ gram, perfonn the sampling and analysis requirements of item 4 a) of Table 4.4-4 until tile specific activity of the primary coolant it, restored to within its limits.

Wi th T,, >_ 500*F.

CALVERT CLIFFS - UNIT 1 3/4 4-26 Amendment No. 223 l

_y s.

.r

.y-..

o 3/4.4 REACTOR.C00UFT._ SYSTEM SURVEILUDICE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be detemined to be within the limits by perfonnance of the sampling and analysis program of Table 4.4-4.

CALVERT CLIFFS - UNIT 1 3/4 4-27 Amendment No. 223 l

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a g

TABLE 4.4-4 R

r-N a

5 PRIMARY COOLANT SPECIFIC ACTIVITY SMPLE AND ANALYSIS PROGRAM P

C h

TYPE OF MEASUREMENT SAMPLE AND MODES IN W ICN SAMPLE AND h

AND ANALYSIS ANALYSIS FREQUENCY ANALYSIS REQUIRED e

e 1.

Gross Activity Determination At least once per 1, 2, 3, 4 8

7 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

[

2.

Isotopic Analysis for DOSE EQUIVALENT 1 per 14 days 1

[

~

I-131 Concentration g

3.

Radiochemical for E Determination 1 per 6 months

I h

4.

Isotopic Analysis for Iodine Including a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , l', 2', 3', 4'. S' g

1-131, 1-133, and I-135 whenever the DOSE

=

EQUIVALENT I-131 exceeds 1.0 pCi/ gram, y

and b) One sample between 2

1. 2, 3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a TNEltMAL POWER change exceeding 15 percent of the k

RATED TNERMAL POWER within a one hour a

period.

R E

=

P Until the specific activity of the Primary Coolant System is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since' reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

3/4.4 REACTOR COOLANT SYSTEM s

\\

250

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k

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150 t

OPERATION L

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k K

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ACCEPPABLE k

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a 1

0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED TIERMAL POWER FIGURE 3.4.8-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL PONER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY

>1.0pci/ GRAM DCSE EQUIVALENT I-131 CALVERT CLIFFS - UNIT 1 3/4 4-29 Amendment No. 223 l

4 3/4.4 REACTOR COOLANT SYSTE 3/4.4.9 PRESSURE /TEMPERATURELJMITE Reactor Coolant System (INITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4.9-1 and 3.4.9-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of:

Maximum Allowable Heatuo Rate RCS Tempen igtg 30*F in any one hour period 70'F to 164'F 40'F in any one huur period

> 164'F to 256'F 60'F in any one hour period

> 256'F b.

A maximum cooldown of:

Maximum Allowable Cooldown Rate RCS Temperature 100'F in any one hour period

> 270'F 20'F in any one hour period 270'F to 184'F 10'F in any one hour period

< 184'F c.

A maximum temperature change of 5'F in any one hour period, during hydrostatic testing operations above system design pressure.

APPLICABILITY: At all times.

ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfonn an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least H0T STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,,lhe following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .and pressure to less than 2 respectively, within CALVERT CLIFFS - UNIT 1 3/4 4-30 Amendment No. 223 1

0 3/4.4 REACTOR COOLANT SYSTEM

$URVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and prestm'e shall be detennined to be within the limits at least once per 30 minutes during system heatu operations. p, cooldown, and inservice leak and hydrostatic testing 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to detemine changes in material properties, as required by 10 CFR Part 50, Appendix H.

The results of these examinations shall be used to update Figures 3.4.9-1 and 3.4.9-2.

t 4

CALVERT CLIFFS - UNIT 1 3/4 4-31 Amendment No. 223 l

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100 200 300 400 s00 000 pfDICATED REACTOR COOLANT TEMPWIATURE T, *F C

FIGURE 3.4.9-1 CALVERT CLIFFS UNIT 1 NEATUP CURVE FOR FLUENCE s 2.61 x 10" n/ca' REACTOR COOLANT SYSTEM PRES $URE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 1 3/4 4-32 Amendment No. 223 l

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100 200 300 400 500 000 INDICATED REACTOR COOLANT TEMPERATURE T * *I C

FIGURE 3.4.9-2 CALVERT CLIFFS UNIT 1 C00LDOWN CURVE, FOR FLUENCE s 2.61 x 10" n/ca' REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 1 3/4 4-33 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRES $URE/TEMPERATURELIMIlS Pressurizer LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shs11 be limited to:

a.

A maximum heatup of 100*F in any one hour period, b.

A maximum cooldown of 200'F in any one hour period, and c.

A maximum spray water temperature differential of 400'F.

APPLICABILITY: At all times.

SGI]QN: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutest perforin an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 300 psia within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIR NENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

CALVERT CLIFFS - UNIT 1 3/4 4-34 Amendment No. 223 l

3/4.4 REACTOR 000LANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Overpressure Protection Syste_m1 LIMITING CONDITION FOR OPERATION 3.4.9.3 The following overpressure protection requirements shall be met:

a.

One of the following three Overpressure Protection Systems shall be in place:

1.

Two power-operated relief valves (PORVs) wjth a trip setpoint below the curve in Figure 3.4.9-3 with their associated block valves open, or 2.

A single PORY with a trip setpoint below the curve in Figure 3.4.9-3 with its associated block valve open and a Reactor Coolant System vent of 1 1.3 square inches, or 3.

A Reactor Coolant System (RCS) vent t 2.6 square inches.

disabled by either removing (racking out) pumps' shall be b.

Twohighpressuresafetyinjection(HPSI) their motor circuit breakers from the electrical power supply circuit, or by locking shut their discharge valves, c.

The HPSI loop motor operated valves (MOVs)' shall be prevented from automatically aligning HPSI pump flow to the RCS by placing their hand switches in pull-to-override, d.

No more than one OPERABLE high pressure safety injection pump with suction aligned to the Refueling Water Tank may be used to inject flow into the RCS and when used, it must be under manual control and one of the foPowing restrictions shall apply:

1.

The total high pressure safety injection flow shall be limited to 5 210 gpm, or 2.

A Reactor Coolant System vent of 2 2.6 square inches shall

exist, e.

When not in use, the above OPERA 8LE high pressure safety injection pump shall have its handswitch in pull-to-lock.

APPLICABILITY: When the RCS temperature is 5 365'F and the RCS is vented to < 8 square inches.

When on shutdown cooling, the PORY trip setpoint shall be 5 429 psia.

8 EXCEPT when required for testing.

CV-L'RT CLIFFS - UNIT 1 3/4 4-35 Amendment No. 223 l

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i 3/4.4 REACTOR COOLANT SYSTEM LIMITING CMDITIM FOR OPERATIM (Continued)

ACTION:

With one PORV inoperable in MODE 3 with the RCS temperature a.

5 365'F or in MODE 4, either restore the inopera*J1e PORV to

, OPERABLE status within 5 days or depressurire and vent the RCS

'through a t 1.3 square inch vent (s) within the next 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months s:

maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status, b.

With one PORV inoperable in MODES 5 or 6 either restore the inoperable PORV to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or depressurizeandventtheRCSthrougha>1.3squareinchvent(s) within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; and maintain tee RCS in this vented condition until both PORVs have been restored to 0PERABLE status, With both PORVs inoperable, depressurire and vent the RCS through c.

a 12.6 square inch vent (s) within 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months s: maintain the RCS in a vented condition until either one OPERABLE PORY and a vent of 31.3 scuare inches has been established or both PORVs have been restorec ti 0PERABLE status.

d.

In the event either the FORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Comission pursuant to 10 CFR 50.4 wit 11n 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence, With less than two HPSI pumps' disabled, place at least two HPSI e.

pump handswitches in pull-to-lock within fifteen minutes and disable two HPSI pumps within the next four hours.

f.

With one or more HPSI loo) MOVs' not prevented from automatically aligning a HPSI pump to tie RCS, immediately place the MOV handswitch in pull.to-override, or shut and disable the affected MOV or isolate the affected HPSI header flowpeth within four hours, and implement the ACTIM requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as applicable.

g.

With HPSI flow exceeding 210 gpm while suction is aligned to the RWT and an RCS vent of < 2.6 square inches exists, 1.

Imediately take action to reduce flow to less than or equal to 210.'pm.

EXCEPT when required for testing.

CALVERT CLIFFS - UNIT 1 3/4 4-36 Amendment No. 223 l

3/4.4 REACTOR C00LANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) 2.

Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the given RCS temperature on Figure 3.4.9-1 or Figure 3.4.9-2.

3.

If a pressure limit was exceeded, take action in accordsnce with Specification 3.4.9.1.

h.

The provisions of Specification 3.0.4 are not applicable.

$URVEILLANCE REQUIREMENili 4.4.9.3.1 Each PORV shall be demonstrated OPERA 8LE by:

a.

Perfomance of a CHANNEL FUNCTIONAL TEST on the PORY actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERA 8LE and at least once per 31 days thereafter when the PORV is required OPERABLE.

b.

Perfomance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per REFUELING INTERVAL.

c.

Verifying the PORY block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressurc protection.

d.

Testing in accordar.ce with the inservice test requirements pursuant to Specification 4.0.5.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection.

4.4.9.3.3 All high pressure safety injection pumps, except the above OPERABLE pump, shall be demonstrated ino>erable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the motor circuit brea cers have been removed from their electrical power supply circuits or by verifying their discharge valves are locked shut. The automatic opening feature of the high pressure safety injection loop MOVs shall be verified disabled at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The above OPERABLE pump shall be verified to have its handswitch in pull-to-lock at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Except when the vent pathway is locked, sealed, or otherwise secured in the open position, then verify these vent pathways open at least once per 31 days.

CALVERT CLIFFS - UNIT 1 3/4 4-37 Amendment No. 223 l

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1. 2 and 3 Components LIMITING CONDITION FOR OPERATION 3.4.10.1 The structitral integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1. APPLICABILITY: ALL MODES. ACTION: a. With the structural integrity of any ASME Code Class I component (s) not confoming to the above requirements, restore the structural integrity of the affected comp)onent(s) to within its limit or isolate the affected component (s prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations. b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200'F. c. With the structural integrity of any ASME Code Class 3 component (s) not confoming to the above requirements, restore the structural integrity of the affected comp)onent(s) to within its limit or isolate the affected component (s from service. d. The provisiont of Specification 3.0.4 are not applicable. SURVEILLANCE l.QUIREMENTS 4.4.10.1.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated: a. Per the requirements of Specification 4.0.5, and b. Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2. CALVERT CLIFFS - UNIT 1 3/4 4-39 Amendment No. 223 l l

3/4.4 REACTOR C00UWT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C 4.b of Regulatory Guide 1.14. Revision 1 August 1975. 4.4.10.1.2 Auomented Inservice Ensoection Procram for Main Stram and Main Feedwater Pipina - The unencapsu' ated welds greater than 4 incies in nomina' diameter in the main steam and main feedwater piping runs located outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice inspection program using the applicable rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code, Section XI,1983 Edition and Addenda through Summer 1983, for Class 2 components. Each weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined, cumulatively, during each 10 year inspection interval. The welds to be examined during each inspection period shall be selected to provide a representative sample of the conditions of the welds. If these examinations reveal unacceptable structural defects in one or more welds, an additional 1/3 of the welds shall be examined and the inspection schedule for the repaired welds shall revert back as if a new interval had begun. If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall also be inspected. 4 CALVERT CLIFFS - UNIT 1 3/4 4-40 Amendment No. 223 1

3/4.4 REACTOR COOLANT SYSTEN I 3/4.4.11 CORE BARREL MOVEMENT LIMITING CONDITION FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level. APPLICABILITY: MODE 1. ACTION: a. With the APD and/or SA exceeding their applicable Alert Levels. POWER OPERATION may proceed provided the following actions are taken: 1. APD shall be measured and processed at leasc once per 24 hours, 2. SA shall be measured at least once por 24 hours and shall be processed at least once per 7 days, and 3. A Special Report, identifying the cause(s) for exceeding the a)plicable Alert Level, shall be prepared and submitted to tie Commission pursuant to 10 CFR 50.4 within 30 days of detection. b. With the APD and/or SA exceeding their applicable Action Levels, measure and process APD and SA data within 24 hours to determine if the core barrel motion is exceeding its limits. With the core barrel motion exceeding its limits, reduce the core barrel motion to within its Action Levels within the next 24 hours or be in NOT STANDBY within the following 6 hours. c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. CALVERT CLIFFS - UNIT 1 3/4 4-41 Amendment No. 223 l s ,..--,-,r.--....

3/4.4 REACTOR C0A. ANT SYSTEM $URVEILLANCE REQUIREMENTS 4.4.11 Routine Monitorina Core barrel movement shall be detennined to be less than the APD and SA Alert Levels by using the excore neutron detectors to measure APD and SA at the following frequencies:

a.. APD data shall be measured and processed at least once per 17 days, b.

SA data shall be measured and processed at least once per 31 days. CALVERT CLIFFS - UNIT 1 3/4 4-42 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.12 LETDOWN LINE EXCESS FLOW LIMITING CONDITION FOR OPERATION 3.4.12 The bypass valve for the excess flow check valve in the letdown line shall be closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the above bypass valve open, restore the valve to its closed position within 4 hours or be in at least NOT STANDBY within the next 6 hours and in COLP SHitTDOWN withiri the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.12 The bypass valve for the excess flow check valve in the letdown line shall be determined cic :ed within 4 hours prior to entering M0DE 4 from MODE 5. CALVERT CLIFFS - UNIT 1 3/4 4-43 Amendment No. 223 1

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.13 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.13 One Reactor Coolant System vent path consisting of two solenoid valves in series shall be OPERA 8LE and closed at each of the following locations: a. Reactor vessel head b. Pressurizer vapor space APPLICABILITY: MODES 1 and 2 ACTION: s a. With the reactor vessel head vent path inoperable, maintain the inoperable vent path closed with power removed from the actuator of tlie solenoid valves in the inoperable vent path, and: 1. If the pressurizer vapor space vent path is also inoperable, restore both inoperable vent paths to 0PERABLE status within 72 hours or be in at least HOT STANDBY within 6 hours, or 2. If the pressurizer vapor space vent path is 0PERA8LE, restore the inoperable reactor vessel head vent path to CPERABLE status within 30 days or be in at least HDT STANDBY within 6 hours. b. With only the pressurizer vapor space vent path inoperable, maintain the inoperable vent path closed with )ower removed from the valve actuator of the solenoid valves in tie inoperable vent path, and: 1. Verify at least one PORV and its associated flow path is OPERABLE within 72 hours and restore the inoperable pressurizer vapor space vent path to OPERABLE status prior to entering MODE 2 following the next HOT SHUTDOWN of sufficient duration, or 2. Restore the inoperable pressurizer vapor space vent path to 0?ERABLE status within 30 days, or be in at least HOT STANDBY within 6 hours. c. The provisions of Specification 3.0.4 are not applicable. CALVERT CLIFFS - UNIT 1 3/4 4-44 Amendment No. 223 l a .u

3/4.4 REACTOR C0OLANT SYSTEM i SURVEILLANCE REQUIREMENTS 4.4.13.1 Each Reactor Coolant System vent path shall be demonstrated OPERABLE by testing each valve in the vent path per Specification 4.0.5. 4.4.13.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per REFUELING INTERVAL by: a. Verifying all manual isolation valves in each vent path are locked in the open position. b. Verifying flow thrnugh the Reactor Coolant System vent paths with the vent valves open. 4 CALVERT CLIFFS - UNIT 1 3/4 4-45 Amendment No. 223 1

3/4.4 REACTOR COOLANT SYSTEM BASES adequate 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 end of the fuel cycle or 30 months, whichever comes first. 3. An update of the assessment model, as ap)ropriat 'n comparison of the predicted results of t1e ster + tc. ube integrity assessment w'.th actual inspection resu-frca pt ious inspections. The plant is expected to be operated in a manner such that se s..dary coolant will be maintained within those chemistry limits fo% - sult in negligible corrosion of the steam generator tubes. If the secor uary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant o~eration would be limited by the limitation of steam generator tube leakage between the Primary Coolant System and the Secondary Coolant System (primary-to-secondary lerkage = 1 gallon per minute, total). Cracks having a primarP/ to-secondary leakage less than this limit during operation will have at.dequate margin of safety to withstand the loads imposed during nonnal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of I gallon per minute can readily be detected by radiation monitors of steam generator blowduwn. Leakage in excess of this limi'. will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired. Defective tubes may be repaired by a Westinghouse Laser Welded Sleeve or an ABt3-Combustion Engineering Leak Tight Sleeve. l The technical bau s for Westinghouse Laser Welded Sleeve are described in the proprietary kstinghouse Reports WCAP-13698, Revision 2. " Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators, Generic Sleeving Report," April 1995; and WCAP-14469, " Specific Application of Laser Welded Sleeving for the Calvert Cliffs Power Plant Steam Generators," November 1995. The technical bases for the Combustion Engineering Leak Tight Sleeve are described in the proprietary ABB-Combustion Engineering Report CEN-630-P, Revision 01, " Repair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight Sleeves," August 1996. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it ( will be found during scheduled inservice steam generator tube examinations. Plugging or repair will be required for all tubes with imperfections at or l exceeding the plugging or repair limit of 40% of the original tube nominal wall thickness. If a tube contat:is a Westinghouse Laser Welded Sleeve with l CALVERT CLIFFS - UNIT 1 B 3/4 4-4 Amendment No. 223 l

l 3/4.4 REACTOR COOLANT SYSTEM RASES l imp'trfection exceeding 40% of nominal wall thickness or an ABS-Combustion Engineering Leak Tight Sleeve exceeding 28%, it must be plugged. =The basis for' the sleeve plugging limit is based on Regulatory Guide 1.121 analyses, and is described in the Westinghouse and AB8-Combustion Engineering sleeving technical reports mentioned above.- (Note: The sleeve plugging limit also includes'20% combined allowance for eddy current uncertainty and additional degradation growth.) 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. Repaired tubes are also included in the inservice tube inspection program. 3 Whenever the results of any steam gecrator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Comission prior to the resumption of plant operation. Such cases will be considered by the Comission 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 SYSTEM LEAKAGE 3/4.4.6.1 Leakaae 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. 3/4.4.6.2 Reactor Coolant System Leakaae 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. i The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakege 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 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 with the reconuendations of Generic Letter 91-04.

^ n. =3/4.4: REACTOR CMLANT SYSTEM 4 - BASES PRES $URE BOURARY LEAKAGE of any magnitude 1s unacceptable since it may be indicative of an impending gross failure of the pressure boundary. c Therefore, the presence of any PRESSURE BONDARY LEAKAGE requires the unit s i to be promptly placed in COLD $NUTDOW. 3/4.4.7 CHEMISTR1 i The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant 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 associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be cont' m A with contaminant concentration levels in excess of the Steady State E ts, up to the Transient Limits, for the specified limited time interveis Mthout having a significant effect on the structural integrity of the Reetor Coolant System. The time interval pemitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actiers 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. '3/4.4.8 SPECIFIC ACTIVITY i The limitations on the specific activity of the primary coolant ensure that-: the resulting 2 hour doses at the SITE 80HDARY 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 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 thei NRC of typical site locations. These values are conservative in that ~ specific site parameters of the Calvert Cliffs site, such as SITE B0HRARY' 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 sacific activity limits of this site. This reevaluation may result in hig1er limits. The ACTION statement pemitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram i DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4.8-1, accommodates possible iodine spiking phenomenon which may t occur following r.hanges in-THERMAL POWER. Operation with specific activity, levels exceeding 1.0 pCi/ gram D0SE EQUIVALENT I-131 but within the limits k 1 CALVERT CLIFFS - UNIT 1 B 3/4 4-6 Amendment No. 223 tl. 4 m .-'w. _...E-,_---,_-- ,-,,,_.,.,y-__,.,.m ..r,_,-+4_ ,e_ ,,,_,.,,,_,,_.,.._ i,,, ,._-r+, ,-,,7,,,,

3/4.4 REACTOR COOLANT SYSTEM BASES 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 Figure 3.4.8-1 increase the 2 hour 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 relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Infonnation obtained on iodine saiking will be used to assess the parameters associated with spiking plenomena. A reduction in frequency of isotopic analyses following power changes may be pemissible if justified by the data obtained. 3/4.4.9 PRESSU_RE/ 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 nontal 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 raNs of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are cor,sistent 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 reactor vessel against fracture is duced by combinative themal and pressure stresses. As the vessel is subjected to increasing fluence, the toughness of the limiting material continues to decline, and ever more restrictive Pressure / Temperature limits must be observed. The current limits, Figures 3.4.9-1 and 3.4.9-2, are for a p'eak neutron fluence to the inner surface of the reactor vessel of < 2.61x10 N/cm' (E > 1 MeV). This fluence corresponds to the Pressurized Thermal Shocx Screening Criteriu defined in 10 CFR 50.61 for weld 2-203 A, B, C. The reactor vessel materials have been tested to determine their initial RT ; 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. The actual shift in RTm of the vessel material will be established 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 1 B 3/4 4-7 Amendment No. 223 l

3/4.4 REACTOR COOLANT SYSTEM aASES The shift in the material fracture toughness, as represented by RTa, is calculated using Regulatory Guide 1.99, Revision 2. For a fluence of 2.61x10"N/cm, the adjusted reference temperature (ART) value at the r 1/4 T position is 241.4'F. At the 3/4 T position the ART value is 181.0'F. These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code, Section III, 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 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thermal rate. Then 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. The composite pressure-temperature limit for the cooldown transient is developed similarly. The Appendix G limits in Figures 3.4.9-1 and 3.4.9-2 assume the following number of RCPs are runnir.g: HEATUP Indicatf_d RCS Temperabtra Maximum Number of RCPs Operatino 70*F to 330*F. 2 > 330'F 4 C00LDOWN Indicated RCS Temperature Maximum Number of RCPs Doeratino > 350*F 4 350*F to 150*F 2 < 150'F 0 Both 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 detemined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for a fluence of < 2.61x10"N/cm on Figures 3.4.9-1 and 3.4.9-2. 2 Similarly,10 CFR Part 50 specifies that core critical limits be 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. CALVERT CLIFFS - UNIT 1 B 3/4 4-8 Amendment No. 223 i

I 3/4.'4 REACTOR C0OLANT SYSTEM . nasts - The Lowest Service Temperature is the minimum allowable temperature at p(625 psia).ressures above 20% of the pre-operational system hydrostatic test pressure This temperature is defined as equal to the most limiting RT. for the balance of the Reactor Coolant System components plus 100'F, per Article NB 2332 of Section III of the ASME Boiler and Pressure Vessel Code. The horizontal line between the' minimum boltup temperature and the Lowest Service Tem)erature is defined by the ASME Boiler and Pressure Vessel Code as 20% of tie pre-operational hydrostatic test pressure. The change in the line at 150*F on Figure 3.4.9-2 is due to a cessation of RCP flow induced pressure deviation, since no RCPs are pemitted to operate during a cooldown below 150*F. The minimum boltup temperature is the minimum allowable temperature at 1 pressures below 20% of the pre-operational system hydrostatic test pressure. The minimum is defined as the initial RT for the material of the higher stressed region of the reactor vessel pl,us any effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was detemined'using the certified material test reports and Branch Technical Position MTEB 5-2. The maximum initial RT associated with the stressed region of the closure head flange is -10*F. However, in order to comply with the 10 CFR 50, Appendix G limits, the minimum allowable reactor vessel temperature with the reactor head attached is 70*F. Hence, the minimum boltup temperature used in Figures and 3.4.9-1 and 3.4.9-2. The Low-Temperature Overpressure Protection (LTOP) System' consists of administrative controls coupled with low-pressure setpoint PORVs. The administrative controls provide the first line of defense against overpressurization events; the PORVs provide a backup to the administrative controls. The following section discusses the bases for the PORV setpoint and administrative controls. Low-Temperature Overpressure Protection uses a variable PORY setpoint to take advantage of the increased Appendix G limits at higher RCS temperatures. Reactor Coolant System temperature is measured at the cold leg RTDs. This provides an accurate temperature indication during forced - circulation, and is also adequate for natural circulation. However, the TtbYflowstream.RTDs are not accurate when on shutdown cooling because they are not in-For this reason, the lowest PORY setpoint is maintained whenever on shutdown cooling. This setpoint, which is independent of RCS temperature, is manually set when shutdown cooling is initiated and maintained until forced circulation is established after the RCPs are started. The PORY setpoint is chosen to protect the most limiting of the heatup or cooldown Appendix G limits. Figure 3.4.9-3 shows the maximum PCRV opening pressure. This includes corrections for static and dynamic head, and /

w I 3/4.4 REACTOR COOLANT SYSTEM BASES -pressure overshoot to account for PORY response time and the maximum ptessurization rate. The actual PORV set >oint is controlled by procedure and accounts for device uncertainty, cali > ration uncertainty and loop drift. The design basis events in the low temperature region are: An RCP start with hot steam generators; and. An inadvertent HPSI actuation with concurrent charging. These transients are most severe when the RCS is initially water solid. Any measures which will prevent or mitigate the design basis events are sufficient for any.less severe incidents. Therefore, this section will discuss the results of the RCP start and mass addition transient analyses. Also discussed is the effectiveness of a pressurizer steam bubble and a single PORY relative to mitigating the design basis events. The RCP start transient is a severe LTOP challenge that can quickly exceed the Appendix G limits for a water solid RCS. Therefore, during water solid operations all four RCPs are tagged out of service and their motor circuit breakers are disabled. However, the transient is adequately mitigated by restricting three parameters:

1) the initial water volume in the pressurizer to 170 inches (indicated), thereby providing a volume for the 1

primary coolant ~ to expand into; 2) the indicated secondary water

3) perature for each steam generator to 30 F above the RCS temperature; and tem the initial pressure of the pressurizer to 300 psia.

With these restrictions in place, the transient is adequately controlled without the assistance of the PORVs. Failure to maintain one of the initial conditions could cause the PORVs to open following an RCP start. The mass addition transient from HPSI or multiple ciiarging pumps is a severe LTOP challenge for a water solid system due to PORY response time. To preclude this event frcm happening while water solid, all HPSI pumps and two charging pumps are tagged out-of-service during water solid operations. Analyses were perfomed for a HPSI mass addition transient with concurrent charging and the expansion of the RCS water volume following loss of decay heat removal, assuming one PORV available (due to single-failure criteria). This mass addition, detemined at the point when the RCS reached water solid conditions, must be less than the capability of a single PORV to limit the LTOP event. Sufficient oversressure protection results when the - equilibrium pressure does not exceed tie limiting Appendix G curve pressure..Because the equilibrium pressure exceeds the minimum Appendix C limit for-full HPSI flow, HPSI flow is throttled to no more than 210 gpm . indicated when the HPSI pump is used for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS-expansion following loss of decay heat removal. The HPSI flow is injected through oniy one HPSI loop MOV to limit instrumentation uncertainty. No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition. CALVERT CLIFFS - UNIT 1 B 3/4 4-10 Amendment No. 223 l

c 3/4.4 REACTOR COOLANT SYSTEM RASES Three 100% capacity HPSI pumps are installti at Calvert Cliffs. Procedures will require that two of the three HPSI puy be disabled (breakers racked out) at RCS temperatures less than or equal to 365*F and that the remaining HPSI pump handswitch be placed in pull-to-lock. Additionally, the HPSI pump normally in pull-to-lock shall be throttled to less than or equal to 210 gpm when used to add mass to the RCS. Exceptions are provided for ECCS testing and for response to LOCAs. To provide single failure protection against a HPSI pump mass addition transient when in MPT enable, the HPSI loop MOV handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actions, described in the ACTION statement, are to disable the e.'fected MOV (by racking out its motor circuit breaker or equivalent), or to isolate the affected HPSI header. Examples of HPSI header isolation actions include; (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge valves; and (3) disabling all thre2 HPSI pumps. RCS temperature, as used in the applicability statement, is determined as follows: (1) with the RCPs running, the RCS cold leg temperature is the I appropriate indication, (2) with the Shutdown Cooling System in operation, the shutdown cooling temperature indication is appropriate (3) if neither the RCPs or shutdown cooling is in operation, the core exit thennocouples are the appropriate indicators of RCS temperature. The allowed out-of-service times for degraded low temperature overpressure protection system in M00ES 5 and 6 are based on the guidance provided in Generic Letter 90-06 and the time required to conduct a controlled, deliberate cooldown, and to depressurize and vent the RCS under the ACTION statement entry conditions. 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for the ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code. 3/4.4.11 CORE BARREL MOVEMENT This specification is provided to ensure early detection of excessive core barrel movement if it should occur. Core barrel movement will be detected by using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Spectral Analysis (SA). Baseline core barrel movement Alert Levels and Action Levels will be confinned during each reactor startup test program following a core reload. CALVERT CLIFFS - UNIT 1 B 3/4 4-11 Amendment No. 223 il

4; . -d -3/4.4 REACT M CMLANT SYSTEM -BASES Data from these detectors is to be reduced in two foms. Root mean square (RMS) values are computed from the APD of the signal amplitude..These RMS magnitudes include variations due both to various neutronic effects and internals motion. Consequently, these signals alone can only provide a gross measure of core barrel motion. A more accurate assessment of core barrel motion is obtained from the Auto and Cross Power Spectral Densities (PSD, XPSD), phase (6) and coherence (C0H) of these signals. These data result from the SA of the excore detector signals. A modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of the affected parts during the plant shutdown at the end of any fuel cycle, y L 3/4.4.12 LETDOWN LINE EXCESS FLOW This specification is provided to ensure that the bypass valve for the excess flow check valve in the letdown line will be maintained closed during plant operation. This bypass valve is required to be closed to ensure that the effects of a pipe rupture downstream of this valve will not exceed the accident analysis assumptions. 3/4.4.11' REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the Primary System that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer vapor space ensures the capability exists to perfom this function. The-valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation'while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolatior..of the vent path. The function, caphilities, and testing requirements of the Reactor Coolant System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirrents," November 1980. m. CALVERT CLIFFS - UNIT-1 83/44-12 Amendment No. 223 l-

3/4.4 ELACTOR C00lMT SYSTEM-3/4,4.5 STEAM GENERATORS LIMITING S0NDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. A[f!ON: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERA 8LE status prior to increasing T.,, above 200*F. SURVEILLANCE ret)UI6.EMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam ge 4erator shall be determined OPERABLE during shutdown by selecting and inspecting.at least the minimum number of steam generators specified in Table 4.4-1, 4.4.5.2 Steam Generator Tube Sampic Selection and Inspection - The steam generator tube minimum samp>0 size, inspection result classification, and the corresponding action required shall be as specified in Tables 4.4-2 and 4.4-3. The inservice inspection of steam generator tubes shall be perforned at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these-inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes insp ;cted shall be fran these critical areas, b. The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include: 1. All nonplugged tubes that previously had detectable wall penetrations (>20%), and CALVERT CLIFFS - UNIT 2 3/4 4-10 Amendment No. 199

= 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCEREQUIREMENTS(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 perfomed before operation beyond 24 months and shall provide assurance that all tubes will retain adequate structural margins against burst throughout nonnal 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 t1e 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 Tables 4.4-2 and 4.4-3 at 40-month intervals fall l 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 Tables 4.4-2 and 4.4-3 during the i shutdown subsequent to any of the following conditions: 1. Primary-to-secondary tube leaks (not including leaks originatir.g 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 atply for extending the frequency for perfonning inservice inspections as specified in Specifications 4.4.5.3.a and b. CALVERT CLIFFS - UNIT 2 3/4 4-12 Amendment No. 199

c 3/4.4 REACT 0A_C00LML1YSTM $URVEILLANCE REQUIREMENTS (Continued)- 4.4.5.4 Acceptance-Criteria a. As used in this Specification: 1. Tubina or Tube means that portion of the tube or sleeve which forms the primary system to secondary system pressure boundary. 2-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. 3. Deoradation means a service-induced cracking, wastage, wsar or general corrosion occurring on either inside or outside of a tube. 4. Deoraded Tube neans c tube contsining imperfections t 20% of the nor.inal wall thickness caused by degradation. 5. % Deoradation means thr. percentage of the tube wall thickness affected or removed by degradation. 6. Defec" means an imperfection of such severity that it exceeds the p' ugging or repair 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. 7. Plucaina or Repair Limit means the imperfection depth at. or beyond which the tube shall be removed from service by plugging, or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection. The plugging or repair limit imperfection depths-are specified in percentage of nominal wall thickness as follows: a. ori gi nal tube wa11................................ 40% b. Westinghouse l aser wel ded sleeve wa11............. 40% c. ABB-Combustion Engineering Leak Tight-Sleeve wall. 28% -l 8. 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, above. 9. Tube Lnspection means an inspection of the steam generator-tube " rom the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. o l CALVERT CLIFFS - UNIT 2 3/4 4-13 Amendment No. 199 l iL --

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

10. Tube Repair refers to a arocess that reestablishes tube serviceability. Accepta)1e tube repairs will be performed by the following processes:

l a) Westinghouse Laser Welded Sleeving as described in the proprietary Westinghouse Reports WCAP-13698, Revision 2, " Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators, Generic Sleeving Report," April 1995; and WCAP-14469, "S)ecific Application of Laser Welded Sleeving for t1e Calvert Cliffs Power Plant Steam Generators," November 1995, b) ABB-Combustion Engineering Leak Tight Sleeving as described in the proprietary ABB-Combustion Engineering Report CEN-630-P, Revision 01, " Repair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight Sleeves," August 1996. A post-weld heat treatment during installation will be performed. Tube repair includes the removal of plugs that were previously installed as a corrective or preventive measure. A tube inspection per Specification 4.4.5.4.a 9 is required prior to returning previcusly plugged tubes to service, b. The steam generator shall be determined OPERA 8LE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit and all tubes containing through-wall cracks) required by Tables 4.4-2 and 4.4-3. l 4.4.5.5 Reports a. Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Comission within 15 days pursuant to 10 CFR 50.4. b. The complete results of the steam generator tube inservice inspection during the report period shall be submitted to the Comission prior to March 1 of each year pursuant to 10 CFR 50.4. This 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. Identiff bation of tubes plugged or repaired. CALVERT CLIFFS - UNri 2 3/4 4-14 Amendment No. 199

3/4.4 j[ ACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c. Results of steam generator tube inspe.ctions which fall into Category C-3 require verbal notification of the NRC Regional Ad.ninistrator by telephone within 24 hours )rior to resumption of plant operation. The written followup of t11s report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective reasures taken to prevent recurrence and shall be subraitted within the next 30 days pursuant to 10 CFR 50.4. i CALVERT CLIFFS - UNIT 2 3/4 4-15 Amendment No. 199 l'

4 i. g TABLE 4.4-1 R '. l t;; g MINIMlpt NU15ER OF STEAft GENERAT0lb TO BE INSPECTED DURIIIG IIISERVICE Ill5PECTI0ll p h a M, Q o. j

^

Preservice Inspection No Yes g No. of Steam Generators per Unit Two Three Four Two Three Four I-First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections One Og.1 One* One !Q I l 3 g E R u o". TABLE 110TATIO11: l 1 The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicete that all steam generators are perfoming in g a like manner. Note that under some circumstances, the operating conditions in one or more steam a generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions. [ ?+ 2 The other steam generator not inspected during the first inservice inspectian shall be inspected. p The third and. subsequent inspections should follow the instructions described in 1 above. 3 Each of the other two steam generators not inspected during the first inservice inspections shall E be inspected during the second and third inspections. The fourth and subsequent inspections shall follow th instructions described in 1 above. i

g TABLE 4.4-2

Q r-STEAM GENERATOR TUBE IllSPECTION m

5 ui wa Imrtuaos oss smurtE anrti uus Jus L_ _ a zertuauss ur h Sasple Size Result Action Required Result Action Recuired Result Action Required S M A mininum of 5 Tr. des per C-1 Rone N/A N/A N/A N/A Q Q SG. E-Z Flug or repair defective C-1 None N/A N/A tubes and inspect C-Z Plug or repair C-1 None n additional 25 tubes in defective tubes and g g-this SG. Inspect additional Plug or repatr { 45 tubes in this C-2 defective tubes ss. Perfom action w N C-3 for C-3 result of first sangle C-3 Perfom action for M C-3 result of first N/A N/A Q san!ple C-J Inspect all tubes in All other 55s this SG, plug or repair are C-1 None N/A N/A R defective tubes and inspect 25 tubes in each other SG. U 24 hour verbal Sc~e 55s C-Z Perfom action for notification to NRC with but no C-2 result of N/A N/A written followup additional second sangle pursuant to 10 CFR 50.4. SG are C-3 Additional Inspect all tubes SG is C-3 in each SG and pl'ag or repair defective N/A N/A tubes. 24-hour g verbal notification to NRC with written aa followup pursuant ,1 to 10 CFR 50.4. A S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators 2 .o inspected during an inspection M

h g TA8LE 4.4-3 y r-j STEAM f4hTRATOR REPAIRED TUSE INSPECTION I p IST SAMPLE INSFt.CTION ZMD 5 AMPLE IM5PtLII0g a g M Sample Size Result Action Required Result Action Required l Q 3 A minimum of 20% of C-1 Mone N/A N/A se repaired tubes"M" i n c C-2 Plug defective repaired tubes C-1 None E 5 and inspect 100% of the C-2 Plug defective h [ repaired tubes in this SG. repaired tubes C-3 Perform act. ion for C-3 result of first sample $9 C-3 Inspect all repaired tubes Other 55 is C-1 Mone in this SG, plug defective Other SG is C-2 Perfor1n action for C-2 tubes and inspect 20% of the result of first sample w2 repaired tubes in the a other SG. Other SG is C-3 Inspect all repaired .L tubes in each SG and 24-hour verbal notification to plug defective tubes. NRC with written follow-up, 24-hour verbal pursuant to 10 CFR 50.4. notification to NRC with written follow-up, pursuant to ) 10 CFR 50.4. k "I Each repair method is considered a separate population for determination of scope expansion. m The inspection of repaired tubes may be performed on tubes from either SG based on outage plans. 5 .if

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakaoe Detection Systems LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE: a. A Containment Atmosphere Particulate Radioactivity Monitoring

System, b.

The Containment Sump Level Alam System, and c. A Containment Atmosphere Gaseous Radioactivity Monitoring System. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: a. With only two of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours when either the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise be in at least NOT STAN08Y within the next 6 hours and in COLS SHUTDOWN within the following 30 hours. b. With only one of the above required Leakage Detection Systems OPERABLE, operation may continue for up to 7 days provided that: 1. Grab samples of the containment atmosphere are obtained and analyzed at least once per 12 hours, and 2. The Reactor Coolant System water inventory balance of Surveillance Requirement 4.4.6.2.c is performed at least once per 24 hours. Otherwise be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. CALVERT CLIFFS - UNIT 2 3/4 4-19 Amendment No. 199 l

3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.1 - The Leakage Detection Systems shall be demonstrated OPERABLE by: a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-perfomance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b. "ontainment Sump Level Alarm System-perfomance of CHANNEL CALIBRATION at least once per REFUELING INTERVAL. J CALVERT CLIFFS - UNIT 2 3/4 4-20 Amendment No. 199 l

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Reactor Coolant System Leakaae LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: a. No PRESSURE BOUNDARY LEAKAGE, b. 1 GPM UNIDENTIFIED LEAKAGE, c. 1 GPM total primary-to-secondary leakage through all steam generators and 100 gallons-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 STAND 8Y within 6 hours and in COLD SHUTDOWN within the following 30 hours, 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTD0WN within the following 30 hours. 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 hours, 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 4-21 Amendment No. 199 l

1 3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b. Monitoring the containment sump discharge frequency at least once per 1? hours, when the Containment Sump Level Alann System is

0PERABLE, c.

Detemining the Reactor Coolant System water leakage at least once per 72 hours during steady state operation and at least once per 24 hours when required by ACTION 3.4.6.1.b, except when operating in the shutdown cooling mode, and d. Monitoring the reactor vessel head closure seal Leakage Detection System at least once per 24 hours. CALVERT CLIFFS - UNIT 2 3/4 4-22 Amendment No. 199 l

l 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1. APPLICABILITY: At all times. ACTION: MODES 1, 2, 3 and 4: a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Stecdy State Limit within 24 hours or be in at least NOT STAND 8Y-within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least NOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES 5 and 6: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to 5,500 psia, if applicable, and perfom an engineering evaluation to detemine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; detemine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to M0DE 4. SURVEILULNCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be detemined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. CALVERT CLIFFS - UNIT 2 3/4 4-23 Amendment No. 199 l

l 3/4.4 REACTOR C0OLANT SYSTEM TABLE 3._43 REACTOR C0OLANT SYSTEM , CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETEd LIMIT LIMIT OISSOLVED OXYGEN

5 0.10 ppm 5 1.00 ppm CHLORIDE 5 0.15 ppm 5 1.50 ppm FLUURIDE 5 0.15 ppm 5 1.50 ppm t

't Limit not applicable with T,,5 250*F. CALVEPJ: C!.IFFS - UNIT 2 3/4 4-24 Amendment No. 199 l

3/4.4 REACTOR C00LANT SYSTEM TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLMCE REQ l[IREMENTS PARAMETER ANALYSIS FREQUENCY DISSOLVED OXYGEN

At least once per 72 hours CHLORIDE At least once per 72 hours FLUORIDE At least once per 72 hours Not required with I.,, < 250*F.

CALVERT CLIFFS - UNIT 2 3/4 4-25 Amendment No. 199 l

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to: a. 5 1.0 Ci/ gram D0SE EQUIVALENT I 1L31, and b. 5 100/E pCi/ gram. APPIICABILITY: M00ES 1, 2, 3, 4 and 5. ACTION: MODES 1, 2 and 3*: a. With the specific activity of the primary coolant > 1.0 pCi/ gram D0SE EQUIVALENT I-131 but within t1e allowable limit (below and to the left of the line) shown on Figure 3.4.8-1, operation may continue for up to 100 hours provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable. b. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 100 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least NOT STANDBY with T.,, < 500 F within 6 hours. c. With the specific activity of the primary coolant > 100/l pC1/ gram, be in at least HOT STAND 8Y with T.,, < 500*F within 6 hours. MODES 1, 2, 3, 4 and 5: d. With the specific activity of the primary coolant > 1.0 Ci/ gram DOSE EQUIVALENT I-131 or > 100/E Ci/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. I With T.,, >_ 500*F, CALVERT CLIFFS - UNIT 2 3/4 4-26 Amendment No. 199 l 1 l

-3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be detennined to be within the limits by perfonnance of the sampling and analysis program of Table 4.4-4. I CALVERT CLIFFS - UNIT 2 3/4 4-27 Amendment No. 199 I

g TABLE 4.4-4 R r-N ~ 4 PRIMRY C000Hli SPECIFIC ACTIVITY SAMPLE E AND ANALYSIS PROGRAM P E ~ h TYPE OF MEASUREMENT SAMPLE AND MODES IN WICE SAfrLE Am AND ANALYSIS _ _ ANALYSIS FREQUENCY ANALYSIS RE08 RES e n8 e i 5 1. Gross Activity Detemination At least once per 72 1,2,3.4 E hours 5 ro 1 2. Isotopic Analysis for DOSE 1 per 14 days 1 E EqWIVALENT I-131 Concentration 4 3. Radiochemical for E Detemination 1 per 6 months

1 w

4. Isotopic Analysis for Iodine a) Once per 4 hours, l', 2' 3', 4' 5' 1 Including I-131,1-133, and I-135 whenever the DOSE EQUIVALENT T 131 = h exceeds 1.0 pC1/ gram, and b) One sagle between

1. 2, 3 3

2 and 6 hours following a TREMAL POWER change exceeding 15 percent of the k RATED TIhtRd4L POWER k" within a one hour period. f P Until the specific activity of the Primary Coolant System is restored within its limits. 3 Samole to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elspsed since i reac' tor was last subcritical for 48 hours or longer. I l

3/4.4 REACTOR COOLANT $YSTEM t i 5 1 i 260 k .( T. T L i k k h i UNAOCEPTABIE 's OPERAMON ( 150 'i -1 T L g 't g 300 i k \\ -2

AOCEPTABLE _

L \\,o -- OPERAMON i 50 0 90 30 40 50 60 70 80 90 100 PERCENT OF RATEDTHERMAL POWER FIGURE 3.4.8-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VER$US PERCENT 0F RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY ~l.0pC1/ GRAM DOSE EQUIVALENT I-131 r CALVERT CLIFFS - UNIT 2 3/4 4-29 Amendment No. 199 l s- - = = w ,,,y r. r.. p ,4-u-- g' 4---yg y v

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS RtM.tqt Coolant System LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordence with the limit lines shown on Figures 3.4.9-1 and 3.4.9-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: a. A maximum heatup of: Maximum Allowable Heatun Ratt RCS Temperature 30'F in any one hour period 70'F to 156'F 40'F in any one hour period > 156'F to 246'F 60'F in any one hour period > 246'F b. A maximum cooldown of: MaFimum Allowable C00ldown Rate RCS Temperature 100'F in any one hour period > 200'F 40'F in any one hour period 200*F to 176'F 15'F in any one hour period < 176'F c. A maximum temperature change of 5'F f n any one hour period, during hydrostatic testing operations above system design pressure. APPLICABILITY:.it all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfonn an engineering evaluation to detenaine the effects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; detennine that the Reactor Coolant System remains acceptable for continued operations or be in at least NOT liTAN0tY within the next 6 hours and reduce the RCS T.,lhe following 30 hours.and pressure to less than 200'F respectively, within CALVERT Cl1FFS - UNIT 2 3/4 4-30 Amendment No. 199 l

3/4 4 REACTOR C00tMT. $YSTD( 0 $URVEILLANCE REQUIREMENTS 4 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be detemined to be within the limits at least once per 30 minutes during system heatu operations. p, cooldown, and inservice leak and hydrosti. tic testing 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to detemine changes in material properties. as required by 10 CFR Part 50, Appendix H. The results of these examinations shall be used to update Figures 3.4.9-2 and 3.4.9-2. CALVERT CLIFFS - UNIT 2 3/4 4-31 Amendment No. 199 l

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~ ' ' ' ' ' ' ' ~ ~ ' O 100 200 300 400 500 000 INDICATED REACTOR COOLANT TEMPERATURE, Tc, 'F FIGURE 3.4.9-2 CALVERT CLIFFS UNIT 2 C00LDOWN CURVE, for FLUENCE 5 4.0x10" n/ca' REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS CALVERT CLIFFS - UNIT 2 3/44-33 Amendment No. 199 l

e 3/4.4 REACTOR C00 TANT SYSTEM ^ 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Pressurizer LIMITING CON 0! TIM FOR OPERATIM 3.4.9.2 The pressurizer temperature shall be limited to: a. A maximum heatup of 100'F in any one hour period, b. A maximum cooldown of 200'F in any one hour period, and c. A maximum spray water temperature differential of 400'F. APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perfonn an engineering evaluation to deMmine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; detemine that the pressurizer remains acceptable for continued operation or be in at least NOT STANDAY within the next 6 hours and reduce the pressurizer pressure to less than 300 psia within the following 30 hours. $URVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be detemined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be detemined to be within the limit at least once per 12 hours during auxiliary spray operation. CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.9 ?!tj$50RE/TEMPERATURELIMITS Overpressure Protection Systems LIMITING CONDITI M FOR OPERATION 3.4.9.3 The following overpressure protection requirements shall be mets a. One of the following three overpressure protection systems shall be in place: 1. Two power-operated relief valves (PORVs) with a trip setpoint below the curve in Figure 3.4.9-3* with their associated block valves open, or 2. A single PORV with a trip setpoint below the curve in Figure 3.4.9-3* with its associated block valve open and a Reactor Coolant System vent of t 1.3 square inches, or 3. A Reactor Coolant System (RCS) vent t 2.6 square inches, b. Two high 3ressure safety injection (HPSI) pumps' shall be disabled )y either removing (racking out) their motor circuit breakers from the electrical power supply circuit, or by locking shut their discharge valves. c. The HPSI loop motor operated valves (MOVs)' shall be prevented from automaucally aligning HPS! pump flow to the RCS by placing their handswitches in pull-to-override. d. No more than one OPERABLE high pressure safety injection pump with suction aligned to the Refueling Water Tank may be used to inject flow into the RCS and when u:ed, it must be under manual control and one of the following restrictions shall apply: 1. The total high pressure safety injection flow shall be limited to 5 210 gpm OR 2. A Reactor Coolant System vent of t 2.6 square inches shall exist. e. When not in use, the above OPERABLE HPSI pump shall have its handswitch in pull-to-lock. APPLICABILITY: When tne RCS temperature is 5 301'F and the RCS is vented to < 8 square inches. When on shutdown ooling, the PORY trip setpoint shall be 5 443 psia. Except when required for testing. CALVERT CLIFFS - UNIT 2 3/4 4-35 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTDi LIMITING CONDITION FOR OPERATION (Continued) ACTION: a. With one PORV inoperable in MODE 3 with RCS temperature 5 301'F or in MODE 4, either restore the inoperable PORY to OPERABLE status within 5 days or depressurize and vent the RCS through a > 1.3 square inch vent (s) within the next 48 hourst maintain the ICS in a vented condition until both PORVs have been restored to 0PERABLE status, b. With one PORV inoperable in F1006 5 or 6, either restore the inoperable PORY to OPERABLE status within 24 hours, or depressurize and vent the RCS through a > 1.3 square inch vent (s) within the next 48 hourst and maintain tee RCS in this vented condition until both PORVs have been restored to OPERABLE status. c. With both PORVs inoperable, depressurize and vent the RCS through a 12.6 square inch vent (s) within 48 hourst maintain the RCS in a vented condition until either one OPERABLE PORY and a vent of 11.3 scuare inches has been established or both PORVs have been restorec to OPERABLE status, d. In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Comission pursuant to 10 CFR 50.4 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurt ence. With less than two HPSI pumps' disabled, place at least two HPSI e. pump handswitches in pull-to-lock within fifteen minutes and disable two HPSI pumps within the next four hours. f. With one or more HPSI loo) MOVs' not prevented from automatically aligning a HPSI pump to tie RCS, imediately place the MOV handswitch in pull-to-override, or shut and disable the affected MOV or isolate the affected HPSI header flowpath within four hours, and implement the action requirements of Specifications 3.1.2.1, 3.1.2.3, and 3.5.3, as applicable, g. With HPSI flow exceeding 210 gpm while suction is aligned to the RWT and an RCS vent of < 2.6 square inches exists, 1. Immediately take action to reduce flow to less than or equal to 210 gpm. Except when required for testing. CALVERT CLIFFS - UNIT 2 3/4 4-36 Amendment No. 199 l

3/4.4 REACTOR C0OLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) 2. Verify the excessive flow condition did not raise pressure above the maximum allowable pressure for the given RCS temperature on Figure 3.4.9-1 or Figure 3.4.9-2. 3. If a pressure limit was exceeded, take action in accordance with Specification 3.4.9.1. h. The provisions of Specification 3.0.4 are not applicable. $URVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORY is required OPERABLE. b. Perfonnance of a CHANNEL CALIBRATION on the PORY actuation channel at least once per REFUELING INTERVAL. c. Verifying the PORV block valve is open at least once per 72 hours when the PORV is being used for overpressure protection, d. Testing in accordance with the inservice test requirements pursuant to Specification 4.0.5. 4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours

when the vent (s) is being used for overpressure protection.

4.4.9.3.3 All high pressure safety injection pumps, except the above OPERABL: pump, shall be demonstrated inoserable at least once per 12 hours by verifying that the motor circuit breacers have been removed from their electrical power supply circuits or by verifying their discharge valves are locked shut. The automatic opening feature of the high pressure safety injection loop MOVs shall be verified disabled at least once per 12 hours. The above OPERABLE pump shall be verified to have its handswitch in pull-to-lock at least once per 12 hours. Exce)t when the vent pathway is locked, sealed, or otherwise secured in t1e open position, then verify these vent pathways open at lea;,t once per 31 days. CALVERT CLIFFS. UNIT 2 3/4 4-37 Amendment No. 199 l

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-= => = ,0--- r_ ~ ---t-, 100 200 3'J0 400 SOO 900 ACTUAL REACTOR COOLANT TEMPERATURE 1,,'F FIGURE 3.4.9-3 CALVERT CLIFFS UNIT 2. for FLUENCE 5 4.0x10" n/ca' MAXIMUM PORY c?ENING PRESSURE vs TEMPERATURE CALVERT CLIFFS - UMIT 2 3/4 4-38 Amendment No. 199 l

3/4.4 REACTOR COOLA,fT SYSTDi 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1. 2 and 3 Components i LIMITIM CMOITIM FOR OPERATIM 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1. APPLICABILITY: ALL MODES. ACTION: a. With the structural integrity of any ASME Code Class I component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations. b. ifith tne structural integrity of 'iy ASME Code Class 2 component (s)notconformingtotheaboverequirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F. c. With the structural integrity of any ASME Code Class 3 component (s)notconformingtotheaboverequirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service. d. The provisions of Sgcification 3.0.4 are not applicable. $URVEILUUICE REQUIREMENTS 4.4.10.1.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated: a. Per the requirements of Specification 4.0.5, and b. Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2. CALVERT CLIFFS - UNIT 2 3/4 4-39 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTDi $URVEILLANCE REQUIRDENTS (Continued) In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the reconmendation of Regulatory Ppsition C.4.b of Regulatory Guide 1.14. Revision 1, August 1975 4.4.10.1.2 Auomented Inservice

nspection Prooram for Main Steam and Main Feedwaner Pipina - The unencapsu' ated welds greater than 4 inches in nomina' diameter in the main steam and main feedwater piping runs located outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice inspection program using the applicable rules, acceptance criteria, and repir procedures of the ASME Coller and Pressure Vessel Code Section XI,1983 Edition and Addenda through Sunner 1983, for Class 2 components.

Each weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined, cumulatively, during each 10-year inspection interval. The welds to be examined during each inspection period shall be selected to provide a representative sample of the conditions of the welds. If these examinations reveal unacceptable structural defects in one or more welds, an additional 1/3 of the welds shall be examined and the inspection schedule for the repaired welds shall revert back as if a new interval had begun. If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall also be inspected. t i Reactor coolant pump flywheel inspections for the first inservice inspection interval may be completed during Unit 2 Refueling Outage No. 9 in conjunction with the reactor coolant pump motor overhaul program. CALVERT CLIFFS - UNIT 2 3/4 4-40 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.11 CORE BARREL MOVEMENT LIMITIM CONDITIM FOR OPERATIM 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level. APPLICABILITY: MODE 1. ACTIQM: a. With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION, may proceed provided the following actions are taken: 1. APD shall be measured and processed at least once per 24 hours, 2. SA shall be measured at least once per 24 hours and shall be processed at least once per 7 days, and 3. A Special Report, identifying the cause(s) for exceeding the a)plicable Alert Level, shall be prepared and submitted to t1e Comission pursuant to 10 CFR 50.4 within 30 days of detection. b. With the APD and/or SA exceeding their applicable Action Levels, measure and process APD and SA data within 24 hours to determine if the core barrel motion is exceeding its limits. With tiie core barrel motion exceeding its limits, reduce the core barrel motion to within its Action Levels within the next 24 hours or be in NOT STAN08Y within the following 6 hours, c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. 4 CALVERT CLIFFS - UNIT 2 3/4 4-41 Amendment No. 199 1

i 3/4.4 REACTOR COOLANT $YSTEM SURVEILLANCE RE4i!REMENTS 4.4.11 Routine Vonitorino Core barrel movement shall be detemined to be less than the APD and SA Alert Levels by using the excore neutron detectors to measure APD and SA at the following frequencies: a. APD data shall be measured and processed at least or.ce per 7 days. b. SA data shall be measured and processed at least once per 31 days. 6 1 0 D l \\ 1 CALVERT CLIFFS - UNIT 2 3/4 4-42 Amendment No. 199 l m ,.__...y

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.12 LETDOWN LINE EXCESS FLOW LIMITING CONDITION FOR OPERATIM 3.4.12 The bypass valve for the excess flow check valve in the letdown line shall be closed. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the above bypass valve open, restore the valve to its closed position within 4 hours or be in at least NOT STANDRY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. $URVEILLANCE REQUIREMENTS 4.4.12 The bypass valve for the excess flow check valve in the letdown line shall be detennined closed within 4 hours prior to entering MODE 4 ' from M00E 5. CALVERT CLIFFS - UNIT 2 3/4 4-43 Amendment No. 199 l

3/4.4 REACTOR C0OLANT SYSTEM 3/4.4.13 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.13 One Reactor Coolant System vent path consisting of two solenoid valves in series shall be OPERABLE and closed at each of the following locaticits: a. Reactor vessel head b. Pressurizer vapor space APPLICABILITY: MODES 1 and 2. ACTION: a. With the reactor vessel head vent path inoperable, maintain the inoperable vent path closed with power removed from the actuator of the solenoid valves in the inoperable vent path, and: 1. If the pressurizer vapor space vent path is also inoperable, restore both inoperable vent paths to OPERAkLE status within 72 hours or be in at least NOT STAND 8Y within 6 hours, or 2. If the pressurizer vapor space vent path is OPERABLE, restore the inoperable reactor vessel head vent path to 0PERABLE status within 30 days or be in at least NOT ST/.NDBY within 6 hours. b. With only the pressurizer vapor space vent path inoperable, maintain the inoperable vent path closed with power removed from the valve actuator of the solenoid valves in the inoperable vent path, and: 1. Verify at least one PORV and its associated flow path is OPERABLE within 72 hours and restore the inoperable pressurizer vapor space vent path to OPERABLE status prior to entering N00E 2 following the next NOT SHUTDOWN of sufficient duration, or 2. Restore the inoperable pressurizer vapor spece vent path to CPERABLE status within 30 days, or be in at least NOT STAND 8Y within 6 hours, c. The provisions of Specification 3.0.4 are not applicable. CALVERT CLIFFS - UNIT 2 3/4 4-44 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.13.1 Each Reactor Coolant System vent path shall be demonstrated OPERA 8LE by testing each valve in the vent path per Specification 4.0.5 4.4.13.2 Each Reactor Coolant System vent path shall be demonstrated OPERA 8LE at least once per REFUELING INTERVAL by: a. Verifying all manual isolation valves in each vent path are locked in the open position. b. Verifying flow through the Reactor Coolant System vent paths with the vent valves open. CALVERT CLIFFS - UNIT 2 3/4 4-45 Amendment No. 199 l

3/4.4 REACTOR.. COOLANT SYSTEM BASE 5 adequate 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 end of the fuel cycle or 30 months, whichever comes first. 3. An update of the assessment model, as appropriate, based on i comparison of the predicted results of the steam generator tube i integrity assessment with actual inspection results from previous 4 inspections. The plant is expected to be operated in a manner sur.h that the secondary coolant will be mair,tained 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 (primary-to-secondary leakage = 1 gallon >er minute, total)y Cracks having a primary-to-secondary leakage less than tais limit during operation will have an adequate margin of safety to withstand the loads imposed during nonnal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired. Defective tubes may be repaired by a Westinghouse Laser Welded Sleeve or an ABS-Combustion Engineering Leak Tight Sleeve. The technical bases for Westinghouse Laser Welded Sleeve are described in the proprietary Westinghouse Reports WCAP-13698, Revision 2. " Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring April 1995; and WCAP-144 Type and Westinghouse Preheater Steam Generators, Generic Sleeving Report, " Specific Application of Laser Welded Sleeving for the Calvert Cliffs Power Plant Steam Generators," November 1995. The technical bases for the Combustion Engineering Leak Tight Sleeve are described in the proprietary ABB-Combustion Engineering Report CEN-630-P, Revision 01, " Repair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight Sleeves " August 1996. -j 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 or repair will be required for all tubes with imperfections at or exceeding the plugging or repair limit of 40% of the tube original-nominal wall thickness. If a tube contains a Westinghouse Laser Welded Sleeve with l CALVERT CLIFFS - UNIT 2 B 3/4 4-4 Amendment No. 199 l _... _ _ _ _ _ -. _,... _ _ ~ _ _ _ _ _. _ _ _....... -. - _ _ _ _.,. _,.. _

3/4.4 REACTOR C00UdlT $YSTEM BASES imperfection exceeding 40% nominal wall thickness or an ABS-Combustion. Engineering Leak Tight Sleeve exceeding 28%, it must be plugged. The basis for the sleeve plugging limit is based on Regulatory Guide 1.121 analyses, and is described in the Westinghouse and ABB-Combustion Engineering sleeving technical reports mentioned above. (Note: The sleeve plugging limit also includes 20% combined allowance for eddy current uncertainty and additional degradation growth.) Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect t degradation that has penetrated 20% of the original tube wall thickness. Repaired tubes are also included in the inservice tube inspection program. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Comission prior to the resumption of plant operation. Such cases will be - considered by the Comission 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. S/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakaae Detection Systems The RCS Leakage Detection Systems required by this specification are provided to monitor and detict 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. 3/4.4.6.2 Reactor Coolant System Leakaae 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 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 with the recomendations of Generic Letter 91-04. i CALVERT CLIFFS - UNIT 2-B 3/4 4-5 Amendment No. 199 l

e 3/4.4 REACTOR COOLANT SYSTEM BASES PRES 5URE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD $NUTDOWN. 3/4.4.7 CHEMISTRY The limitations on RCS chemistry ensure that corrosion of the RCS is minimized and reduce the r,tential for RCS leakage or failure due to stress corrosion. Maintaining '.ie chemistry within the Steady State Limits provides adequate cor* on protection to ensure the structural integrity of the RCS over the ' of the plant. The associated effects of exceeding the oxygen, chloride id fluoride limits are time and temperature dependent. Corrosion 'udies show that operation may be continued with contaminant concentrat) .) levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval pemitting 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. 3/4.4.8 JPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the reculting 2 hour doses at the SITE B0UNDARY will not exceed an ap)ropriately small fraction of Part 100 limits following a steam generator 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 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 pemitting POWER OPERATION to continue for limited time periods wi. 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, acconinodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 pCi/ gram DOSE 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 4-6 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTEM BASES the unit's yearly operating time since the activity levels allowed by Figure 3.4.8-1 increase the 2 hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated si/am 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 relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the' primary coolant will be detected in sufficient time to take corrective action. Infonnation obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. 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 RCS 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 STARTUF 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 reactor vessel against fracture induced by combinative themal 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, F,1,gures 3.4.9-1 and 3.4.9-2, are for u) to and including a fluence of 4.0x10 n/cm (E > 1 Mev) at the clad / vessel interface, which corresponds to approximately 30 Effective Full Power Years. The reactor vessel materials have been tested to determine their initial RT.,; 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 material will be establisheI The actual shift in RT., of the vessel cause an increase in the RT 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. The shift in the material fracture toughness, as represented by RT.1, is calculated using Regulatory Guide 1.99, Revision 2. For a fluence of 4.0x10" n/cm', the adjusted reference temperature (ART) value at the 1/4 T position is 177.1*F. At the 3/4 T position the ART value is 146.8'F. CALVERT CLIFFS - UNIT 2 B 3/4 4-7 Amendment No. 199 l

3/4.4 REACTOR COOLANT SYSTEM -aAsts These values are used with procedures developed in the ASME Boiler and Pressure Vessel Code Section III, 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,1/2 T heatup, and 3/4 T heatup pressure-temperature limits are compared for a given thennal rate. Then 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. The composite pressure-temperature limit for the cooldown transient is developed similarly. The Appendix G limits in . Figures 3.4.9-1 and 3.4.9-2 assume the following number of RCPs are running: Heatup Indicated RCS Temperature Maximum Number of RCPs Operatina 70'F to 308'F 2 > 308'F 4 Cooldown Indicated RCS Temperature Maximum Number of RCPs Operatina > 350*F 4 350'F to 150'F 2 < 150'F 0 Both 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 detennined by entering the curve at the test pressure (1.1 times nonnal operating pressure) and locating the corresp'onding temperature. This curve is shown for a fluence of 4.0x10 n/cm' on Figures 3.4.9-1 and 3.4.9-2. Similarly,10 CFR Part 50 specifies that core critical limits be 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 e.nalyses that actually control the temperature at which the core can be brought critical. The Lowest Service Temperature is the minimum allowable temperature at ressures above 20% of the pre-operational system hydrostatic test pressure p(625 psia). This temperature is defined as equal to the most limiting RTm for the balance of the RCS components plus 100 F, per Article NB 2332 of Section III of ^.'e ASME Boiler and Pressure Vessel Code. -CALVERT CLIFFS - UNIT 2 8 3/4 4-8 Amendment No. 199. l

3/4 4 KACT0t_C00LANT SYSTEM 0 BASES The horizontal Tine between the minimum boltup temperature and the Lowest Service Tem >erature is defined by the ASME Boiler and Pressure Vessel Code as 20% of tie pre-operational hydrostatic test pressure. The change in the line at 150'F on Figure 3.4.9-2 is due to a cessation of RCP flow induced pressure deviation, since no RCPs are permitted to operate during a cooldown below 150'F. The minimum boltup temperature is the minimum allowable temperature at pressures below 20% of the pre-operational system hydrostatic test pressure. The minimum is de'ined as the initial RT the higher stressed region of the reactor vessel pl for the material of us any effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel Code. The initial reference temperature of the reactor vessel and closure head flanges was detennined using the certified material test reports and Branch Technical Position MTEB 5-2. The maximum initial RT, associated with the stressed region of the closure head flange is 30'F. However, for conservatism, a minimum boltup temperature of 70*F is utilized in the analysis to establish the low temperature PORY lift setpoint. The Low-Temperature Overpressure Protection (LTOP) system consists of administrative controls coupled with low-pressure setpoint PORVs. The administrative controls provide the first line of defense against overpressurization events; the PORVs pravide a backup to the administrative controls. The following section discusses the bases for the PORV setpoint and administrative controls. Low-Temperature Overpressure Protection uses a variable PORV setpoint to take advantage of the increased Appendix G limits at higher RCS temperatures. Reactor Coolant System temperature is measured at the cold leg Resistance Temperature Detectors (RTDs). This provides an accurate temperature indication during forced circulation, and is also adequate for natural circulation. However, the T,,a RTDs are not accurate when on SDC because they are not in the flow stream. For this reason, the lowest PORY setpoint is maintaincd whenever on SDC. This set)oint, which is independent on RCS temperature, is manually set wien SDC is initiated and maintained until forced circulation is established after the RCPs are started. The PORY setpoint is chosen to protect the most limiting of the heatup or cooldown Appendix G limits. Figure 3.4.9-3 shows the maxir.un PORY opening pressure. This includes corrections for static and dynamic head, and pressure overshoot to account for PORV response time and the maximum pressurization rate. The actual PORY setpoint and the Minimum Pressure and Temperature Enable setpoint are controlled by procedure and account for all associated uncertainties. CALVERT CLIFFS - UNIT 2 B 3/4 4-9 Amendment No. 199 l

o 3/4.4 REACTOR C00Vuff SYSTEM-BASES The design basis events.'n the low temperature region are: An RCP start with hot steam generators; and. An inadvertent HPSI actuatior, with concurrent charging. These transients are most severe when the RCS is initially water solid. Any measures which will prevent or mitigate the design basis events are sufficient for any less severe incidents. Therefore, this section will discuss the results of the RCP start and mass addition transient analyses. Also discussed is the effectiveness of a pressurizer steam bubble and a single PORY relative to mitigating the design basis events. The RCP start transient is a severe LTOP challenge that can quickly exceed the Appendix G limits for a water solid RCS. Therefore, during water solid operations all four RCPs are tagged out of service and their motor circuit breakers are disabled. However,)the transient is adequately mitigated by 1 the initial water volume in the restricting three parameters: pressurizer to 170 inches (indiceted), thereby providing a volume for the priinary coolant to expand into; 2) the indicated secondary water temperature for each steam generator to 30'F above the RCS temperature; and

3) the initial pressure of the pressurizer to 320 psia. With these restrictions in place, the transient is adequately controlled without the assistance of the PORVs.

Failure to maintain one of the initial conditions could cause the PORVs to open following an RCP stert. The mass addition transient from HPSI or multiple charging pumps is a severe LTOP challenge for a water solid system due to PORY response time. To preclude this event from happening while water solid, all HPSI pumps and two charging pumps are tagged out-of-service during water solid conditions. Analyses were perfomed for a mass addition transient with concurrent charging and the expansion of the RCS water volume following loss of decay heat removal, assuming one PORY available (due to single-failure criteria). This mass addition, determined at the point when the RCS reached water solid conditions, must be less than the capability of a single PORV-to limit the LTOP event. Sufficient over)ressure protection results when the equilibrium pressure does not exceed tie limiting Appendix G curve pressure. Because the equilibrium pressure exceeds the minimum Appendix G limit for full HPSI flow HPSI flow is throttled to no more than 210 gpm indicated when the HPSI pump is used for mass addition. The HPSI flow limit includes allowances for instrumentation uncertainty, charging pump flow addition and RCS expansion following loss of decay heat removal. The HPSI flow is injected through only one HPSI loop MOV to limit instrumentation uncertainty. No more than one charging pump (44 gpm) is allowed to operate during the HPSI mass addition. Three 100% capacity HPSI pumps are installed at Calvert Cliffs. Procedures will require that two of the three HPSI pumps' be disabled (breakers racked out) at RCS temperatures less than or equal to 301'F and that the remaining HPSI pump handswitch be placed in pull-to-lock. Additionally, the HPSI CALVERT CLIFFS - UNIT 2 B 3/4 4-10 Amendment No. 199 l

l 3/4.4 REACTOR C0OLANT SYSTEM BASES pump nonnally in pull-to-lock shall be throttled to less than or equal to 210 gpm when used to add mass to the RCS. Exceptions are provided for ECCS testing and for response to LOCAs. To provide single failure protection against a HPSI pump mass addition transient, the HPSI loop M0V handswitches must be placed in pull-to-override so the valves do not automatically actuate upon receipt of a SIAS signal. Alternative actions, described in tne ACTION Statement, are to disable the affected MOV (by racking out its motor circuit breaker or equivalent), or to isolate the affected HPSI header. Examples of HPSI header isolation actions include; (1) de-energizing and tagging shut the HPSI header isolation valves; (2) locking shut and tagging all three HPSI pump discharge MOVs; and (3) disabling all three HPSI pumps. P.CS temperature, as used in the applicability statement, is detennined as (1) with the RCPs runni' g, +he RCS cold leg temperature is the follows: n appropriate indication, (2) with the SDC System in operation, the SDC temperature indication is appropriate, (3) if neither the RCPs or SDC is in operation, the core exit thennocouples are the appropriate indicators of RCS temperature. The allowed out-of-service times for degraded low temperature overpressure protection system in MODES 5 and 6 are based on the guidance provided in Generic Letter 90-06 and the time required to conduct a controlled, deliberate cooldown, and to depressurize and vent the RCS under the ACTION statement entry conditions. 3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for the ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code. 3/4.4.11 CORE BARREL MOVEMENT This specification is provided to ensure early detection of excessive co'e barrel movement if it should occur. Core barrel movement will be detected by using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Spectral Analysis MA). Baseline core barrel movement Alert Levels and Action Levels will be confinned during each reactor STARTHP test program following a core reload. Data from these detectors is to be reduced in two fonns. Root mean square (RMS) values are computed from the APD of the signal amplitude. These RMS magnitudes include variations due both to various neutronic effects and internals motion. Consequently, these signals alone can only provide a gross measure of core barrel motion. A more accurate assessment of core barrel motion is contained from the Auto and Cross Power Spectral Densities CALVERT CLIFFS - UNIT 2 B 3/4 4-11 Amendment No. 199 l

o 3/4.4 REACTOR COOLANT sVSTEM sAsEs (PSD,XPSD), phase (4)andcoherence(COH)ofthesesignals. The:e data result from the SA of the excore detector signals. A modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of the affected parts during the plant shutdown at the end of any fuel cycle. 3/4.4.12 LETDOWN LINE EXCESS FLOW This specification is provided to ensure that the bypass valve for the excess flow check valve in the letdown line will be maintained closed during plant operation. This bypass valve 1e oquired to be closed to ensure that the effects of a pipe rupture downstream of this valve will not exceed the accident analysis assumptior.s. 3/4.4.13 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the Primary System that could inhibit natural circulation core cooling. The OPERASILITY of at least one RCS vent path from the reactor vessel head and the pressurizer vapor space ensures the capability exists to perfonn this function. The valve redundancy of the RCS vent paths serves to minimize the probability of inadvertent or irreversib% actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the RCS vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Require:nents," November 1980. CALVERT CLIFFS - UNIT 2 B 3/4 4-12 Amendment No. 199 'l __}}

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