Text

Amends 216 & 193 to Licenses DPR-53 & DPR-69,respectively, Revising TS Section 6.0 Administrative Controls to Be Consistent W/Sts & Allow STA to Perform Dual Roles
	ML20117H621
Person / Time
Site:	Calvert Cliffs
Issue date:	08/26/1996
From:	Jeffrey Mitchell; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20117H625	List: ML20117H621; ML20117H632;
References
	NUDOCS 9609090341
	Download: ML20117H621 (97)
	v • d • e

,,um p

1 UNITED STATES g

g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 0001 o%...../

BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO.]

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 216 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 March 15, 1995, as supplemented on June 29, 1995, May 1, 1996 and May 15, 1996 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 (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 4

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:

i 4

9609090341 960826 PDR ADOCK 05000317 P

PDR

, 2.

Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.216, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COPtilSSION han Y" hb Jocelyn A. Mitchell, Acting 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.:

August 26, 1996

p errug UNITED STATES g

4 g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001 o

%4.....,&

j BALTIMORE GAS AND ELECTRIC COMPANY

)

DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 193 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 March 15, 1995, as supplemented on June 29, l

1995, May 1, 1996 and May 15, 1996, 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 t

Comission; C.

There is reasonable assurance (i) that the activities authorized i

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:

l 2.

Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendrent No.193, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY C0ttilSSION ha4{

Jocelyn A. Mitchell, Acting Director Project Directorate I-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation 1

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 26, 1996

1

1>

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 216 FACILITY OPERATING LICENSE NO. DPR-53 DOCKET NO. 50-317 Revise Appendix A as follows:

Remove Paaes Insert Paaes XVI through XVIII XVI 1-3 1-3 2-1 through 2-6 2-1 through 2-7 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 3-35 3/4 3-35 3/4 3-37 3/4 3-37 3/4 4-14 3/4 4-14 5/4 4-16 3/4 4-16 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 3/4 4-34 3/4 4-34 3/4 4-39 3/4 4-33 3/4 5-3 3/4 5-3 3/4 5-7 3/4 5-7 3/4 6-2 3/4 6-2 3/4 6-9 through 3/4 6-9 through 3/4 6-10 3/4 6-10 3/4 6-24 3/4 6-24 3/4 7-31 3/4 7-31 3/4 7-33 3/4 7-33 3/4 7-37 3/4 7-37 3/4 7-40 3/4 7-40 3/4 7-42 3/4 7-42 3/4 7-45 3/4 7-45 l

3/4 7-47 3/4 7-47 6-1 thru 6-32 6-1 thru 6-14 l

1

'4l TABLE OF CONTENTS ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY 5-1 6.2 ORGANIZATION 6.2.1 Onsite & Offsite Organizations 6-1 6.2.2 Unit Staff 6-1 6.3 FACILITY STAFF QUALIFICATIONS............

6-3 6.4 PROCEDURES 6-3 6.5 PROGRAMS AND MANUALS 6.5.1 Offsite Dose Calculation Manual (ODCM) 6-4 9

6.5.2 Post-Accident Sampling 6-5 6.5.3 Primary Coolant Sources Outside Containment....'.

6-5 6.5.4 Technical Specification Bases Control Program....

6-5 6.5.5 Radioactive Effluent Controls Program........

6-6 6.5.6 Containment Leakage Rate Testing Program 6-7 6.6 REPORTING REQUIREMENTS 6.6.1 Occupational Radiation Exposure Report 6-8 6.6.2 Annual Radiological Environmental Operating Report 6-8 6.6.3 Radioactive Effluent Release Report.........

6-9 6.6.4 Monthly Operating Report 6-9 6.6.5 Core Operating Limits Report (COLR).........

6-10 6.6.6 Pressurizer PORY and Safety Valve Report 6-14 CALVERT CLIFFS - UNIT 1 XVI Amendment No. 216

i 1.0 DEFINITIONS CONTROLLED LEAKAGE 1.9 CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals.

CORE ALTERATION 1.10 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.11 The CORE OPERATING LIMITS REPORT is the unit specific document that provides cycle specific parameter limits for the current reload cycle.

These cycle specific parameter limits shall be detennined for each reioad cycle in accordance with Specification 6.6.5.

Plant operation within these l

limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 1.12 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.13

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

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.14 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

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

CALVERT CLIFFS - UNIT 1 1-3 Amendment No. 216

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE i

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1-1.

4 i

APPLICABILITY: MODES 1 and 2.

ACTION:

a.

Whenever the point defined by the combination of the highut i

operating loop cold leg temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in H0T STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

l b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.

c.

The Vice President-Nuclear Energy and the offsite review function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

A Safety Limit Violation Report shall be prepared and submitted to the Commission, the offsite review function and the Vice President - Nuclear Energy within 14 days of the violation.

j REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

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

ACTION:

MODES 1 and 2 a.

Whenever the Reactor Coolant System pressure has exceeded l

2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.

c.

The Vice President - Nuclear Energy and the offsite review function sna11 be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

A Safety Limit Violation Report shall be prepared and submitted to the Comission, the offsite review function and the Vice President - Nuclear Energy within 14 days of the violation.

CALVERT CLIFFS - UNIT 1 2-1 Amendment No. 216

1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded l

a.

2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

b.

The NRC Operations Center shall be notified by telephone as soon '

as possible and in all cases within one hour.

c.

The Vice President - Nuclear Energy and the offsite review function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I d.

A Safety Limit Violation Report shall be prepared and submitted to the Commission, the offsite review function and the Vice President - Nuclear Energy within 14 days of the violation.

4 l

4 i

i i

i t

J l

9 1

1 CALVERT CLIFFS - UNIT 1 2-2 Amendment No. 216

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 3

i i

p

$- 3

- lE

- 5 l

S/#[

- n

~l 8

~ U A g:

ort 00 1,

- S,g h

O Sl

~ E m

is 8

h* 5 l2 llg

- E si 0!

ll l! lglI lsliilIll ll 8l ll!ihlil 4

4 i

4 4

omuvamuommmmumyn FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT CALVERT CLIFFS - UNIT 1 2-3 Amendment No. 216 l

1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS 4

REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

H 4

i i

a i

i CALVERT CLIFFS - UNIT 1 2-4 Amendment No. 216 l

= _ -

i F

g TABLE 2.2-1 p

G g

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LINITS

~

W b

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES m

M i3 1.

Manual Reactor Trip Not Applicable Not Applicable g

2.

Power Level - High

< 10% above THERMAL POWER, with a < 10% above THERMAL POWER, and i5 E

minimum setpoint of 30% of RATED a minimum setpoint of 30% of d

G THERMAL POWER, and a maximum of RATED THERMAL POWER and a 5 107.0% of RATED THERMAL POWER.

maximum of 5 107.0% of RATED E

THERMAL POWER.

g 295}ofdesignreactorcoolant 2 95) of design reactor coolant 3.

Reactor Coolant Flow - Low N flow flow g

h 4.

Pressurizer Pressure - High 5 2400 psia 5 2400 psia 5.

Containment Pressure - High 5 4 psig 5 4 psig A

7 4

6.

SteagGeneratorPressure-2 685 psia 2 685 psia Low g

7.

Steam Generator Water Level -

2 10 inches below top of feed 2,10 inches below top of feed l

Low ring ring

,m E

E a

8-an

.E m

TABLE 2.2-1 (Continued) f*

n?

h REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS W

b FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES m

4 m3 8.

Axial flux offset (3)

Trip setpoint adjusted to not Trip setpoint adjusted to not r-exceed the limits provided in exceed the limits provided in E

i c

the COLR the COLR Q

9.

Thermal Margin / Low Pressure (1)

~

a.

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to be f

Operating exceed the limits provided in not less than the larger of g

the COLR (1) 1875 psia, or (2) the y

limits provided in the COLR g

b.

SteamGeneratorPrgsure 5 135 psid 5 135 psid

[

Difference - High 4

ro m

in

10. Loss of Load NA NA Q

11. Rate of Change of Power - High (*)

5 2.6 decades per minute 5 2.6 decades per minute

%I E

N4=

Fa o.

a

.E o,

l

1' 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TABLE 2.2-1 (Continued)

TABLE NOTATION See Specification 3.2.5, "DNB Parameters," for the design reactor cociant flow.

(I)

Trip may be bypassed below 10-'% OF RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is t 10 -'% of RATED

)

THERMAL POWER.

(2)

Trip may be. manually bypassed below 785 psia; bypass shall be automatically removed at or above 785 psia.

(3)

Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall bc automatically removed when THERMAL POWER is 115% of RATED THERMAL POWER.

l (4)

Trip may be bypassed below 10-'% and above 12% of RATED THERMAL POWER.

i l

a 1

CALVERT CLIFFS - UNIT 1 2-7 Amendment No. 216 l

18 l 3/4.3 INSTRUMENTATION TABLE 3.3-6 (Continued)

TABLE NOTATION Alarm setpoint to be specified in a controlled document (e.g., setpoint control manual).

ACTION STATEMENTS ACTION 14 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 16 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

l ACTION 30 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s),within72 hours,and:

1) either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or 2) prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within 30 days following the l

event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

4 CALVERT CLIFFS - UNIT 1 3/4 3-25 Amendment No. 216

i 3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Meteoroloaical Instrumentation LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a.

With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 within the next l

10 days outlining the cause of the malfunction and the plans for restoring,the channel (s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

~

4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL 1

CHECK and CHANNEL CALIBRATION operations at the frequencies shown in i

Table 4.3-5.

CALVERT CLIFFS - UNIT 1 3/4 3-27 Amendment No. 216

1 3/4.3 INSTRUMENTATION TABLE 3.3-10 (Continued)

ACTION STATEMENTS ACTION 31 -

With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or be in NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 32 -

With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERABLE requirement in Table 3.3-10, operation may proceed provided the inoperable channel is restored to 0PERABLE status at the next outage of sufficient duration.

ACTION 33 -

With the number of OPERABLE post-accident monitoring channels two less than required by Table 3.3-10, either restore one inoperable channel to OPERA 8LE status within 30 days or be in HOT SHUTD0WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 34 -

With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERA 8LE requirement in Table 3.3-10, either restore the system to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within 30 days l

following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERA 8LE status.

ACTION 35 -

With the number of OPERABLE channels two less than required by Table 3.3-10, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring for core and Reactor Coolant System voiding; 2.

Prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 within 30 days following the I

event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and 3.

Restore the system to OPERABLE status at the next scheduled refueling.

CALVERT CLIFFS - UNIT 1 3/4 3-35 Amendment No. 216

i 3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Fire Detection Instrumentation LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the ' fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE.

ACTION: With one or more of the fire detection instrument (s) shown in Table 3.3-11 inoperable:

~

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the a.

zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect the containment at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5; or unless the instrument (s) is located in fire detection zones equipped with automatic wet pipe sprinkler systems alanned and supervised to the Control Room, then within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, inspect the zone (s) with inoperable instruments and.

verify that the Automatic Sprinkler System, including the water flow alarm and supervisory system, is OPERABLE by CNANNEL FUNCTIONAL TEST.

b.

Restore the inoperable instrument (s) to 0PERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within the next 30 days outlining the l

action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status, c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.7.1 At least once per 6 months, at least 25% of the above required

)

fire detection instruments which are accessible during plant operation j

shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST.

Detectors selected for testing shall be selected on a rotating basis such CALVERT CLIFFS - UNIT 1 3/4 3-37 Amendment No. 216

1 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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

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 tie Calvert Cliffs Power Plant Steam Generators," November 1995.

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 plugged tubes to service.

b.

The steam generator shall be determined 0PERARLE 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 Table 4.4-2.

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 Commission within 15 days pursuant to 10 CFR 50.4.

l 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 to CFR 50.4.

This report shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thick' ness penetration for each indication of an imperfection.

3.

Identification of tubes plugged or repaired.

c.

Results of steam generator tube inspections which fall into Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 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 detemine cause of the tube degradation and corrective measures taken to prevent recurrence and shall be submitted within the next 30 days pursuant to 10 CFR 50.4.

l CALVERT CLIFFS - UNIT 1 3/4 4-14 Amendment No. 216

i g

TABLE 4.4-2 w

r-a M

S*.EAM GENERATOR TUSE INSPECTION N

ISI SAMPLE INSPLCIIGi1 ZMD SAMPLE IM5PLCIION 3RD SAMPLE IN5PLCIION zu O

Sample Size l Result i Action Required Result Action Required Result Action Required S

Q A minimum of 5 lubes per C-1 Mone M/A N/A N/A N/A Q

]

SG.

C-Z Plug or repair defective C-1 None N/A N/A tubes and inspect C-2 Plug or repair C-1 None n

additional 2S tubes in defective tubes and o

this SG.

inspect additional C-2 Plug or repair o

{

45 tubes in this defective tubes SG.

C-3 Pertom action for g

C-3 result of first sample g

C-3 Perfom action for 4

C-3 result of first N/A N/A g

sample C-3 Inspect all tubes in All other this SG plug or repair SGs are None N/A N/A defective tubes and C-1 inspect 25 tubes in each other SG.

5 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal some SGS Perfonn action for N/A N/A notification to NRC with C-2 but no C-2 result of written followup additional second sample pursuant to 10 CFR 50.4. SG are C-3 l

Additional Inspect all tubes SG is C-3 in each SG and plug or repair defective N/A N/A tubes. 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months y

verbal notification a

to NRC with written E

followup pursuant M

to 10 CFR 50.4.

l o

[ S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection o

m m

E 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 pCi/ gram DOSE EQUIVALENT I-131, and b.

5 100/E pCi/ gram.

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

ACTION:

MODES 1, 2 and 3*:

With the specific activity of the primary coolant > 1.0 pC1/ gram a.

DOSE EQUIVALENT I-131 but within the 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 DOSE 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 interval or exceeding the limit line shown on Figure 3.4.8-1, be in at least H0T 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 STAND 8Y 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 C1/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ 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.

l Wi th T.,, >._ 500 F.

1 CALVERT CLIFFS - UNIT 1 3/4 4-24 Amendment No. 216

l 3/4.4 REACTOR C0OLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be detennined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

CALVERT CLIFFS - UNIT 1 3/4 4-25 Amendment No. 216 l

O i

3/4.4 REACTOR C0OLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

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

y'

< 365 F or in MODE 4, either restore the inoperable PORV to OPERABLE status within 5 days or depressurize and vent the RCS through a 21.3 square inch vent (s) within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; 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 depressurizeandventtheRCSthrougha11.3squareinchvent(s) within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; and maintain the RCS in this vented condition until both PORVs have been restored to OPERABLE status, With both PORVs inoperable, depressurize and vent the RCS through c.

a 2 2.6 square inch vent (s) within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; maintain the RCS in a vented condition until either one OPERABLE PORV and a vent of 21.3 square inches has been established or both PORVs have been restored to OPERABLE status.

d.

In the event either the PORVs 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 l

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 recurrence.

e.

With less than two HPSI pumps' disabled, place at least two HPSI 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 loop MOVs' not prevented from automatically aligning a HPSI pump to the 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 1 3/4 4-34 Amendment No. 216

i i

3/4.4 REACTOR C0OLANT SYSTEM 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.

l 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 least once per 24 hcurs, 2.

SA shall be measured at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and shall be processed at least once per 7 days, and 3.

A Special Report, identifying the cause(s) for exceeding the applicable Alert Level, shall be prepared and submitted to the Commission pursuant to 10 CFR 50.4 within 30 days of l

detection.

b.

With the APD and/or SA exceeding their applicable Action Levels, measure and process APD and SA data within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

CALVERT CLIFFS - UNIT 1 3/4 4-39 Amendment No. 216

3/4.5 EMERGENCY CORE C0OLING SYSTEMS (ECCS) 3/4.5.2 ECCS SUBSYSTEMS - MODES 1. 2 AND 3 (2 1750 PSIA)

LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERA 8LE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety injection pump, b.

One OPERABLE low-pressure safety injection pump, and c.

An OPERA 8LE flow path capable of taking nuction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a 1

Recirculation Actuation Signal.

APPLICABILITY: MODES 1, 2 and 3*.

ACTION:

a.

With one ECCS subsystem inoperable, restore the inoperable subsystem to 0PERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to 10 CFR 50.4 within l

90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

With pressurizer pressure > 1750 psia.

CALVERT CLIFFS - UNIT 1 3/4 5-3 Amendment No. 216

3/4.5 EMERGENCY C0itE COOLING SYSTEMS (ECCS) 3/4.5.3 ECCS SUBSYSTEMS - MODES 3 (< 1750 PSIA) AND 4 LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

One' OPERABLE high-pressure safety injection pump, and a.

b.

An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.

APPLICABILITY: MODES 3* and 4.

ACTION:

a.

With no ECCS subsystem 0PERABLE, restore at least one ECCS subsystem to 0PERABLE status within I hour or be in COLD SNUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Consnission pursuant to 10 CFR 50.4 within l

90 days describing the circumstances of the actuati6n and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

Between 385'F and 365'F, a transition region exists where the OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and restored to automatic status on a heatup. At 365 F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically.

At 365 F and less, HPSI pump use will be conducted in accordance with Technical Specification 3.4.9.3.

With pressurizer pressure < 1750 psia.

CALVERT CLIFFS - UNIT 1 3/4 5-7 Amendment No. 216

i1

?

3/4.6 CONTAllWENT SYSTEMS l

3/4.6.1 PRIMARY CONTAINMENT Containment Leakaae 4

LIMITING CONDITION FOR OPERATION j

3.6.1.2 Containment leakage rates shail be limited to:

A maximum allowab'le containmear leakage rate, L, as specified in a.

Specification 6.5.6,"ContainmtntLeakageRateIestingProgram."

l b.

A combined leakage rate of 5 0,50 L 173,000 SCCM), for all penetrations and valves subjet.t to ty(pe B and C tests when i

i pressurized to P.

i l

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: With either (a) the measured overall integrated containment leakage rate exceeding the acceptance criteria specified in the Containment Leakage Rate Testing Program, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests j

exceeding 0.50 L., restore the overall integrated containment leakage rate to within the acceptance criteria specified in the Containment Leakage Rate j

Testing Program, and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than or equal to 0.50 L, prior i

to increasing the Reactor Coolant System temperature above 200'F.

2 SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the j'

following test schedule and shall be determined in coiiformance with the criteria, methods and provisions specified in 10 CFR Part 50, Appendix J:

a.

Perform required visual examinations and Type A testing in i

accordance with the Containment Leakage Rate Testing Program.

4 i

I i

CALVERT CLIFFS - UNIT 1 3/4 6-2 Amendment No. 216

3/4.6 CONTAllMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) adjacent tendons is found unacceptable, it shall be considered as evidence of possible abnomal degradation of the containment structure.

In addition, more than one unacceptable tendon out of those selected for surveillance (from all three tendon groups) shall be considered as evidence of possible abnormal degradation.

of the containment structure.

If the normalized lift-off force of any single tendon lies below the lower bound individual, the occurrence should be considered as evidence of possible abnormal degradation of the containment structure.

In addition, determining that the avera off forces for each sample population (ge of the normalized lift-hoop, vertical, dome)is equal to or greater than the required average prestress level; h536 kips for hoop tendons, 622 kips for vertical tendons, and 555 kips for dome tendons (reference Figures 3.6.1-1, 3.6.1-2, and3.6.1-3).

If the average is below the required average prestress force, it shall be considered as evidence of possible abnormal degradation of the containment structure.

b.

Removing one wire from each of a dome, vertical and hoop tendon checked for lift-off force, and detennining over the entire length of the wire:

1.

The extent of corrosion, cracks, or other damage. The presence of abnormal corrosion, cracks, or other damage shall be considered evidence of possible abnonnal degradation of the containment structure.

2.

A minimum tensile strength value of 240 Ksi (guaranteed ultimate strength of the tendon material) for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire samples to meet the minimum tensile strength test is evidence of possible abnomal degradation of the containment structure.

c.

Perform a chemical analysis to detect changes in the chemical properties of the sheath filler grease. Any unusual changes in physical appearance or chemical properties that could adversely affect the ability of the filler grease to adhere to the tendon wires or otherwise inhibit corrosion shall be reported to the Comission pursuant to 10 CFR 50.4.within the next 30 days.

l CALVERT CLIFFS - UNIT 1 3/4 6-9 Amendment No. 216 L

3/4.6 CONTAIlWENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i 4.6.1.6.2 End Anchoraaes and Adjacent Concrete Surfaces. The structural 1

integrity of the end anchorages and adjacent concrete surfaces shall be demonstrated by detemining through inspection of a representative sample of tendons (reference Specification 4.6.1.6.1) that no apparent changes have occurred in the visual, appearance of the end anchorages or their adjacent concrete exterior surfaces. Also, inspections of the pre-selected concrete crack patterns adjacent to end anchorages shall be perfomed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment is at its maximum test pressure.

4.6.1.6.3 Containment Surfaces. The exposed accessible interior and exterior surfaces of the containment, including the liner plate shall be visually inspected in accordance with the Containment Leakage Rate Testing Program (reference Specification 4.6.1.2).

4.6.1.6.4 Reports. Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Commission pursuant to 10 CFR 50.4 within the next 30 days.

This l

report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective actions taken.

l l

l CALVERT CLIFFS - UNIT 1 3/4 6-10 Amendment No. 216

~

l 3/4.6 CONTAINMENT SYSTEMS 3/4.6.5

_ COMBUSTIBLE GAS CONTROL Hydrooen Analyzers LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent containment hydrogen analyzers shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTION:

With one hydrogen analyzer inoperable, restore the inoperable a.

analyzer to OPERABLE status within 30 days or:

1.

Verify containment atmosphere grab sampling capability and prepare and submit a special report to the Consnission pursuant to 10 CFR 50.4 within the following 30 days, l

outli.ning the ACTION taken, the cause for the inoperability, and the plans and schedule for restoring the system to l

OPERABLE status, or 2.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With both hydrogen analyzers inoperable, restore at least one inoperable analyzer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

Specification 3.0.4 is not applicable to this requirement.

SURVEILLANCE REQUIREMENTS 4.6.5.1.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least I

bi-weekly on a STAGGERED TEST BASIS by drawing a sample from the Waste Gas System through the hydrogen analyzer.

4.6.5.1.2 Each hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases in accordance with manufacturers' recommendations.

l l

l I

CALVERT CLIFFS - UNIT 1 3/4 6-24 Amendment No. 216 1

J 3/4.7 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Sources in use - At least once per six months for all sealed a.

sources containing radioactive material:

1.

With a half-life greater than 30 days (excluding Hydrogen 3),and 2.

In any form other than gas.

b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

Startup sources and fission detectors - Each sealed startup c.

source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and

{

following repair or maintenance to the source or detector.

4.7.9.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis pursuant to 10 CFR 50.4 if sealed source or l

fission detector leakage tests reveal the presence of > 0.005 microcuries 1

of removable contamination.

j CALVERT CLIFFS - UNIT 1 3/4 7-31 Amendment No. 216

i 3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Sorav and/or Sprinkler Systems i

LIMITING CONDITION FOR OPERATION 3.7.11.2 The spray and/or sprinkler systems shown in Table 3.7-5 shall be OPERABLE:

APPLICABILITY: Whenever equipment in the spray / sprinkler protected areas is required to be OPERABLE.

ACTION:

a.

With one or more of the required spray and/or sprinkler systems inoperable, within one hour establis5 a continuous fire watch with backup fire suppression equipment for those areas in which redundant safe shutdown systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 l

within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path, not locked, sealed or otherwise secured in position, is in its correct position.

b.

At least once per 12 months by cycling each valve in the flow path through at least one complete cycle of full travel.

i c.

At least once per 18 months 1.

By performing a system functional test which includes simulated automatic actuation' of the system, and verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test signal.

CALVERT CLIFFS - UNIT 1 3/4 7-37 Amendment No. 216

l 3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Halon Systems LIMITING CONDITION FOR OPERATION 3.7.11.3 The following Halon Systems shall be OPERABLE with the storage tanks having at least 95% of full charge weight (or level) and 90% of full charge pressure.

a.

Cable spreading room total flood system, and associated vertical cable chase IC, Unit 1.

b.

4160 volt switchgear room 27' & 45' elevation Unit 1.

APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE.

ACTION:

a.

With botn the primary and backup Halon Systems protecting the areas inoperable, within one hour establish an hourly fire watch with backup fire suppression equipment for those areas protected by the inoperable Halon System.

Restore the system to 0PERABLE status within 14 days or prepare and submit a Special Report to.

I the Commission pursuant to 10 CFR 50.4 within the next 30 days l

outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not i

applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.3 Each of the above required Halon Systems shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path is in its correct position.

b.

At least once per 6 months by verifying Halon storage tank weight (level) and pressure.

CALVERT CLIFFS - UNIT 1 3/4 7-40 Amendment No. 216

il 3/4.7 PLANT SYSTEMS l

i 3/4.7.11 FIRE SUPPRESSION SYSTEMS Fire Hose Stations LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.

ACTION:

a.

With one or more of the fire hose ste.tions shown in Table 3.7-6 inoperable, route an additional equivalent capacity fire hose to 1

the unprotected area (s) from an OPERABLE hose station within I hour.

Restore the fire hose station (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Consnission pursuant to 10 CFR 50.4 within the next 30 days I

outliningtheactiontaken,thecauseoftheinoperability(andthe pla OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-6 shall be i

demonstrated OPERABLE:

a.

At least once per 31 days by visual inspection of the station to assure all required equipment is at the station. Hose stations located in the containment shall be visually inspected on each scheduled reactor shutdown, but not more frequently than every 31 days.

b.

At least once per 18 months for hose stations located outside the containment and once per REFUELING INTERVAL for hose stations inside the containment by:

1.

Removing the hose for inspection and re-racking, and 2.

Replacement of all degraded gaskets in couplings.

CALVERT CLIFFS - UNIT 1 3/4 7-42 Amendment No. 216

4 3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Yard Fire Hydrants and Hydrant Hose Houses 4

LIMITING CONDITION FOR OPERATION 3.7.11.5 The following yard fire hydrants and associated hydrant hose houses shall be OPERABLE:

6 yard hydrant and associated hydrant hose house, which provides a.

primary protection for Unit 2 RWT blockhouse.

b.

7 yard ' hydrant and associated hydrant hose house, which provides primary protection for Unit 1 RWT blockhouse.

APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.

ACTION:

a.

With one or more of the yard fire hydrants or associated hydrant hose houses inoperable, within I hour have sufficient additional lengths of 2-1/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression. Restore the hydrant or hose house to OPERABLE status within 14' days or prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 within the next 30 days outlining the action taken, I

the cause of the inoperability, and the plans and schedule for restoring the hydrant or hose house to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i

(

CALVERT CLIFFS - UNIT 1 3/4 7-45 Amendment No. 216 t

l i

3/4.7 PLANT SYSTEMS 3/4.7.12 PENETRATION FIRE BARRIERS LIMITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (i.e., cable penetration barriers, firedoors and fire dampers), in fire zone boundaries, protecting safe shutdown areas shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

With one or more of the above required fire barrier penetrations a.

inoperable within one hour either establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the inoperable fire barrier and establish an hourly fire watch patrol; or verify the operability of Automatic Sprinkler Systems (including the water flow alarm and supervisory system) on both sides of the inoperable fire barrier. Restore the inoperable fire barrier penetration (s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 within the next 30 days outlining the action taken, I

the cause of the inoperable penetration and plans and schedule for restoring the fire barrier penetration (s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be OPERABLE:

a.

At least once per 18 months by a visual inspection.

b.

Prior to returning a fire barrier penetration to functional status following repairs or maintenance by performance of a visual inspection of the affected fire barrier penetration (s).

CALVERT CLIFFS - UNIT 1 3/4 7-47 Amendment No. 216

4 3/4.4 REACTOR C0OLANT SYSTEM BASES penetrated 20% of the original tube wall thickness. Re also included in the inservice tube inspection program. paired tubes are 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 l

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'.

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 recommendations of Generic Letter 91-04.

PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

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

CALVERT CLIFFS - UNIT 1 B 3/4 4-5 Amendment No. 216 i)

l 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall facility 1

operation and shall delegate in writing the succession to this responsibility during'his absence.

6.2 ORGANIZATION 6.2.1 Onsite & Offsite Organizations Onsite and offsite organizations shall be established'for unit operation and corporate management, res)ectively. The onsite and offsite organizations shall include tie positions for activities affecting the safety of the nuclear power. plant.

a.

lines of authority, responsibility and connunication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions.

These relationships shall be documented and updated, as appropriate, in the fonn of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant-specific titles of personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Final Safety Analysis Report (UFSAR).

b.

The plant manager shall be responsible for overall-unit safe l

operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c.

The Vice President - Nuclear Energy shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 Unit Staff The unit staff organization shall include the following:

a.

A total of at least three non-licensed operators shall be assigned to the Units 1 and 2 shift crews.

CALVERT CLIFFS - UNIT 1 6-1 Amendment No. 216

,1 6.0 ADMINISTRATIVE CONTROLS b.

At least one licensed Operator shall be in the Control Room when fuel is in the reactor.

At least two licensed Operators shall be present in the Control c.

Room during reactor STARTUP, scheduled reactor shutdown, and during recovery from reactor trips.

d.

An individual qual'1fied in radiation protection procedures shall be on site when fuel is in the reactor.

e.

A site Fire Brigade of at least five members shall be maintained l

onsite at all times. The Fire Brigade shall not include the minimum shift crew necessary for safe shutdown of both units (four m~ embers) or any personnel re functions during a fire emergency. quired for other essential Fire Brigade training shall meet the requirements of NFPA 27, 1975 edition.

f.

The operations manager shall hold or have held a senior reactor operator license at Calvert Cliffs. The General Su)ervisor, Shift Supervisor and Control Room Supervisor shall told a senior reactor operator license. The Control Room Operator shall hold a reactor operator license.

g.

One Shift Technical Advisor (STA) shall be assigned to the shift crew when either unit is in MODE 1, 2, 3 or 4, and shall be filled as follows:

1.

By the Shift Supervisor or an on-shift Senior Operator License (SOL) holder, provided the individual meets the Commission Policy Statement on Engineering Expertise on Shift; or 2.

By an individual meeting the minimum STA education and training requirement of Specification 6.3.1; or 3.

By an SOL holder previously approved by the NRC as an exception to the minimum STA education requirements of Specification 6.3.1, provided the following conditions are met:

(a) With both units in M0DE 1, 2, 3 or 4, the STA shall be an SOL holder in addition to the two SOL holders required; (b) With one unit in MODE 1, 2, 3 or 4 and the other unit in M0DE 5 or 6, the STA shall be an SOL holder other than the Shift Supervisor; and (c) With one unit in MODE 1, 2, 3 or 4 and the other unit defueled, the STA shall be an SOL holder in addition to the one SOL holder required.

CALVERT CLIFFS - UNIT 1 6-2 Amendment No. 216

6.0 ADMINISTRATIVE CONTROLS h.

Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(1) and Specifications 6.2.2.a and 6.2.2.g for a period of time not to exceed two hours in order to accommodate unexpected absence of on duty shift crew members provided insnediate action is taken to restore the shift crew composition to within the minimum requirements.

i.

Those licensed operators counted toward minimum shift crew composition required by 10 CFR 50.54(m)(2)(1) shall be licensed on both units.

6.3 FACILITY STAFF OVALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Safety Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift i

Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 PROCEDURES l

6.4.1 Written procedures shall be established, implemented and maintained I

covering the activities referenced below:

The applicable procedures reconrnended in Appendix A of Regulatory a.

G'uide 1.33, Revision 2, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c.

Quality assurance for effluent and environmental monitoring; d.

Fire Protection Program implementation; e.

All programs specified in Specification 6.5; and f.

The amount of overtime worked by plant staff members performing l

safety-related functions must be limited in accordance with the NRC Policy Statement on Working Hours (Generic Letter 82-12).

I CALVERT CLIFFS - UNIT 1 6-3 Amendment No. 216

6.0 ADMINISTRATIVE CONTROL 5 6.5 PROGRAMS AND MANUALS The following programs.shall be established, implemented and maintained:

6.5.1 0FFSITE D0SE CALCULATION MANUAL (00CM) 4 a.

The 00CM shall contain the methodology and parameters used in the l

calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program; and i

b.

The 00CM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the infonnation that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports, required by Specifications 6.6.2 and 6.6.3, respectively.

c.

Licensee initiated changes to the 00CM:

)

1.

Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

(a) Sufficient information to support the change (s) together with the ap(propriate analyses or evaluations justifying thechanges);

(b)_A determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and 10 CFR Part 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; 2.

Shall become effective after review and acceptance by the j

onsite review function and the approval of the plant manager; and

\\

3.

Shall be submitted to the NRC in the form of a complete, j

legible copy of the entire ODCM as part of or concurrent with.

the Radioactive Effluent Release Report for the period of the' report in which any change in the 00CM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

1 CALVERT CLIFFS - UNIT 1 6-4 Amendment No. 216

l

!I l

6.0 ADMINISTRATIVE CONTROLS 6.5.2 Post-Accident Samplina l

The licensee shall establish, implement and maintain a program

which will l

ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include -

the following:

i a.

Training of personnel, b.

Procedures for sampling and analysis, and Provisions for maintenance of sampling and analysis equipment.

c.

6.5.3 Primary Coolant Sources Outside Containment The licensee shall implement a program

to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

Provisions establishing preventive maintenance and periodic a.

visual inspection requirements, and i

b.

Leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.5.4 Technical Specification Bases Control Proaram This program provides a means for processing changes to the Technical Specification Bases.

a.

Changes to the Bases of the Technical Specifications shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1.

A change in the Technical Specifications incorporated in the license; or 2.

A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

It is acceptable if the licensee maintains details of the program in plant operation manuals (e.g., chemistry procedures, training instructions,maintenanceprocedures,ERPIPs),

i CALVERT CLIFFS - UNIT 1 6-5 Amendment No. 216

~

i i

6.0 ADMINISTRATIVE CONTROLS c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes to the Technical Specifications incorporated in i

the license or proposed changes to the UFSAR or Bases that involve an unreviewed safety question shall be reviewed and approved by the NRC prior to implementation. Changes to the i

Bases implemented' without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

6.5.5 Radioactive Effluent Controls Procram This program conforms to 10 CFR 50.36a for the control of radioactive a

effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall 4

include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a.

Limitations on the functional capability of radioactive liquid i

and gaseous monitoring instrumentation, including surveillance tests and setpoint determination, in accordance with the methodology in the 0DCM; b.

Limitations on the concentrations of radioactive material j

released in liquid effluents to unrestricted areas, conforming to i

10 CFR Part 20, Appendix B. Table II, Column 2;

~

c.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents, in accordance with 10 CFR 20.1302, and with the methodology and parameters in the ODCM; d.

Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR Part 50, Appendix I; e.

Detennination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year, in accordance with the methodology and parameters in the ODCM, at least every 31 days; f.

Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR Part 50, Appendix I; 4

4 2

g.

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary i

conforming to the dose associated with 10 CFR Part 20, l

Appendix B, Table II, Column 1; I

CALVERT CLIFFS - UNIT 1 6-6 Amendment No. 216

6.0 ADMINISTRATIVE CONTROLS h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR Part 50, Appendix I; 1.

Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR Part 50, Appendix I; and i

j.

Limitations on the annual dose or dose commitment to any member i

of the public due to releases to radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR Part 190.

6.5.6 Containment Leakaae Rate Testina Program l

A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163,

" Performance-Based Containment Leak-Test Program," dated September 1995, as modified by approved exceptions.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, P., is 49.4 psig.

The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, L,, shall be 0.20 percent of containment air weight per day at P,.

Containment leakage rate acceptance criterion is 51.0 L,.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria is 5 0.75 L, for Type A tests.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

J The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

CALVERT CLIFFS - UNIT 1 6-7 Amendment No. 216

6.0 ADMINISTRATIVE CONTROLS 6.6 REPORTING RE0VIREMENTS The following reports shall be submitted in accordance with 10 CFR 50.4.

6.6.1 Occupational Radiation Exposure Report

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures > 100 mrem /yr 1

and their associated man rem exposure according to work and job functions 1

(e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignment to various duty functions may be estimates based on pocket dosimeter, electronic personal dosimeter or thermoluminescent dosimeter. Small exposures totalling < 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. The report shall be submitted prior to March 31 of each year.

6.6.2 Annual Radioloaical Environmental Operatina Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year, j

The report shall include summaries, interpretations, and analyses of trends i

of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the ODCM, and in 10 CFR Part 50, Appendix I, Sections IV.B.2, IV.B.3 and IV.C.

The report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the fonnat of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

A single submittal may be made for Calvert Cliffs. The submittal should combine those sections that are common to both units.

Occupational dose from the Independent Spent Fuel Storage Installation will be reported separately.

CALVERT CLIFFS - UNIT 1 6-8 Amendment No. 216

i 6.0 ADMINISTRATIVE CONTROLS 6.6.3 Radioactive Effluent Release Report

l l

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a (i.e., time between submittal of the reports must be no longer than 12 months).

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units.

The material provided shall be consistent with the objectives outlined in the 00CM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

Licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous and solid) shall be reported to the Commissicn in the Radioactive Effluent Release Report for the period in which the modification to the waste system is completed. The discussion of each change shall contain:

a.

A description of the equipment, components and processes involved; and b.

Documentation of the fact that the change including the safety analysis was reviewed and found acceptable by the onsite review function.

The report shall also include changes to the ODCM, in accordance with Specification 6.5.1.c.

6.6.4 Mon' thly Operatino Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis, no later than the 15th of each month following the calendar month covered by the report.

i A single submittal may be made for Calvert Cliffs, since the Radwaste Systems are common to both units.

CALVERT CLIFFS - UNIT 1 6-9 Amendment No. 216

l l

6.0 ADMINISTRATIVE CONTROLS 6.6.5 Core ODeratino Limits Report (COLR) l a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

2.2.1 3.1.1.1 3.1.1.2 3.1.1.4 3.1.3.1 3.1.3.6 3.2.1 3.2.2.1 3.2.3 l

3.2.5 i

3.9.1 b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC; specifically, those described in the following documents:

i (1)

CENPD-199-P, Latest Approved Revision, "C-E Setpoint Methodology: C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems,"

January 1986 (2) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 1: C-E Calculated Local Power Density and Thermal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," December 1979 (3) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 2: Combination of System Parameter Uncertainties in Themal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980 (4) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2," March 1980 (5) CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2,"

December 1981 (6)

Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr.

BG&E),datedJune 24, 1982,' Unit 1 Cycle 6 License Approval Amendment No. 71 to DPR-53 and SER)

CALVERT CLIFFS - UNIT 1 6-10 Amendment No. 216

6.0 ADMINISTRATIVE CONTROLS (7) CEN-348(B)-P, " Extended Statistical Combination of Uncertainties," January.1987 (8)

Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated October 21, 1987, Docket Nos. 50-317 and 50-318, " Safety Evaluation of Topical Report CEN-348(B)-P, Extended Sta,tistical Combination of Uncertainties" (9) CENPD-161-P-A,'" TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986 (10) CENPD-162-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 1. Uniform Axial Power Distribution" (11) CENPD-207-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-Unifonn Axial Power Distribution" (12) CENPD-206-P-A, Latest Approved Revision, " TORC Code, Verification and Simplified Modeling Methods" (13) CENPD-225-P-A, Latest Approved Revision, " Fuel and Poison Rod Bowing" (14) CENPD-266-P-A, Latest Approved Revision, "The ROCS and DIT.

Computer Code for Nuclear Design" (15) CENPD-275-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Gadolinia - Urania Burnable Absorbers" (16) CENPD-382-P-A, Latest Approved Revision -"C-E Methodology for Core Designs Containing Erbium Burnable Absorbers" (17) CENPD-139-P-A, Latest Approved Revision, "C-E Fuel Evaluation Model Topical Report" (18) CEN-161-(B)-P-A, Latest Approved Revision, " Improvements to Fuel Evaluation Model"'

(19) CEN-161-(B)-P, Supplement 1-P, " Improvements to Fuel Evaluation Model," April 1989 (20) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated February 4, 1987, Docket Nos. 50-317 and 50-318, " Safety Evaluation of Topical Report CEN-161-(B)-P, Supplement 1-P, Improvements to Fuel Evaluation Model" (21) CEN-372-P-A, Latest Approved Revision, " Fuel Rod Maximum Allowable Gas Pressure" CALVERT CLIFFS - UNIT 1 6-11 Amendment No. 216 l

l i

6.0 ADMINISTRATIVE CONTROLS (22) Letter from Mr. A. E. Scherer (CE) to Mr. J. R. Miller (NRC), dated December 15, 1981, LD-81-095, Enclosure 1-P, "C-E ECCS. Evaluation Model Flow Blockage Analysis" (23) CENPD-132, Supplement 3-P-A, Latest Approved Revision,

" Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS" (24) CENPD-133, Supplement 5, "CEFLASH-4A, a FORTRAN 77 Digital 1

l Computer Program for Reactor Blowdown Analysis," June 1985-(25) CENPD-134, Supplement 2 "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985 l

(26) Letter from Mr. D. M. Crutchfield (NRC) to Mr. A. E. Scherer (CE),datedJuly 31, 1986, " Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports" (27) CENPD-135, Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 (28) Letter from Mr. R. L. Baer (NRC) to Mr. A. E. Scherer (CE),

dated September 6,1978, " Evaluation of Topical Report CENPD-135, Supplement 5" (29) CENPD-137, Supplement 1-P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977 (30) CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977 (31) Letter from Mr. K. Kniel (NRC) to Mr. A. E. Scherer (CE),

dated September 27,'1977, " Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P" (32) CENPD-138, Supplement 2-P, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 (33) Letter from Mr. C. Aniel (NRC) to Mr. A. E. Scherer, dated April 10,1978, " Evaluation of Topical Report CENPD-138, Supplement 2-P" (34) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. J. R. Miller (NRC) dated February 22,1985, "Calvert Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, Amendment to Operating License DPR-53, Eighth Cycle License Application" CALVERT CLIFFS - UNIT 1 6-12 Amendment No. 216 l

I 4

I l

6.0 ADMINISTRATIVE CONTROLS I

(35) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr.

(BG&E), dated May 20, 1985, " Safety Evaluation Report Approving Unit 1 Cycle 8 License Application" (36) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. R. A. Clark (NRC), dated September 22, 1980, " Amendment to Operating License No. 50-317, Fifth Cycle License Application" (37) Letter from Mr. R. A. Clark (NRC) to Mr. A. E. Lundvall..Jr.

(BG&E), dated December 12, 1980, " Safety Evaluation Report Approving Unit 1, Cycle 5 License Application" (38) Letter from Mr. J. A. Tiernan (BG&E) to Mr. A. C. Thadani (NRC), dated October 1,1986, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2, Docket Nos. 50-317 & 50-318, Request for Amendment" (39) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated July 7, 1987, Docket Nos. 50-317 and 50-318, Approval of Amendments 127 (Unit 1) and 109 (Unit 2)

(40) CENPD-188-A, Latest Approved Revision, "HERMITE: A Multi-Dimensional Space-Time Kinetics Code for PWR Transients" 1

(41) The Full Core Power Distribution Monitoring System referenced in Specifications 3.1.3.1, 3.2.2.1, 3.2.3, and the BASES is described in the following documents:

(a) CENPD-153-P, Latest Approved Revision, " Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System" (b) CEN-199(B)-P, "BASSS, Use of the Incore Detector System to Monitor the DNB-LCO on Calvert Cliffs Unit 1 and Unit 2," November 1979 (c)

Letter from Mr. G. C. Creel (BG&E) to NRC Document Control Desk, dated February 7,1989, "Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket 50-318, Request for Amendment, Unit 2 Ninth Cycle License Application" (d)

Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. G. C. Creel (BG&E), dated January 10, 1990, " Safety Evaluation Report Approving Unit 2 Cycle 9 License Application" (42)

Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton (BGE), dated May 11, 1995, " Approval to Use Convolution Technique in Main Steam Line Break Analysis -

Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 (TAC Nos. M90897 and M90898)

CALVERT CLIFFS - UNIT 1 6-13 Amendment No. 216 l

6.0 ADMINISTRATIVE CONTROLS The core operating (e.mits shall be determined such that all li c.

a)plicable limits g., fuel thermal mechanical limits, core tiermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.6.6 Pressurizer PORV and Safety Valve Report A report shall be submitted prior to March 1 of each year documenting all failures and challenges to the pressurizer PORVs or safety valves.

CALVERT CLIFFS - UNIT 1 6-14 Amendment No. 216

i ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 193 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NO. 50-318 Revise Appendix A as follows:

Remove Paaes Insert Paaes XVI through XVIII XVI a

1-3 1-3 2-1 through 2-6 2-1 through 2-7 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 3-35 3/4 3-35 3/4 3-37 3/4 3-37 3/4 4-14 3/4 4-14 3/4 4-16 3/4 4-16 3/4 4-24 through 3/4 4-24 through 3/4 4-25 3/4 4-25 3/4 4-34 3/4 4-34 3/4 4-39 3/4 4-39 4

3/4 5-3 3/4 5-3 3/4 5-7 3/4 5-7 3/4 6-2 3/4 6-2 3/4 6-9 3/4 6-9 3/4 7-31 3/4 7-31 3

3/4 7-33 3/4 7-33 3/4 7-37 3/4 7-37 3/4 7-40 3/4 7-40 4

3/4 7-42 3/4 7-42 3/4 7-45 3/4 7-45 3/4 7-47 3/4 7-47 B 3/4 4-5 B 3/4 4-5 6-1 through 6-32 6-1 through 6-14 l

)

6

TABLE OF CONTENTS ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6.2.1 Onsite & Offsite Organizations 6-1 6.2.2 Unit Staff 6-1 6.3 FACILITY STAFF QUALIFICATIONS...........

6-3 6.4 PROCEDURES 6-3 6.5 PROGRAMS AND MANUALS 6.5.1 Offsite Dose Calculation Manual (0DCM) 6-4 6.5.2 Post-Accident Sampling 6-5 6.5.3 Primary Coolant Sources Outside Containment....

6-5 6.5.4 Technical Specification Bases Control Program...

6-5 6.5.5 Radioactive Effluent Controls Program.......

6-6 6.5.6 Containment Leakage Rate Testing Program 6-7 6.6 REPORTING REQUIREMENTS 6.6.1 Occupational Radiation Exposure Report 6-8 6.6.2 Annual Radiological Environmental Operating Report 6-8 6.6.3 Radioactive Effluent Release Report........

6-9 6.6.4 Monthly Operating Report 6-9 6.6.5 Core Operating Limits Report (COLR)........

6-10 6.6.6 Pressurizer PORV and Safety Valve Report 6-14 CALVERT CLIFFS - UNIT 2 XVI Amendment No. 193

i l

l 1.0 DEFINITIONS i

CONTROLLED LEAKAGE 1.9 CONTROLLED LEAKAGE shall be the water flow from the reactor coolant pump seals.

CORE ALTERATION l

l 1.10 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of C0RE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

l CORE OPERATING LIMITS REPORT 1.11 The CORE OPERATING LIMITS REPORT is the unit specific document that provides cycle specific parameter limits for the current reload cycle.

These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.6.5.

Plant operation within these l

limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 l

1.12 D0SE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance l

Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.13 E shall be the average (weighted in proportion to the conentration of each radionuclide in the reactor coolant at the time of samp(Ng) of the sum of the average beta and gama energies per disintegration in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

1 ENGINEERED SAFETY FEATURE RESPONSE TIME 1.14 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.).

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

CALVERT CLIFFS - UNIT 2 1-3 Amendment No. 193

r 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest l

a.

operating loop cold leg temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STAND 8Y within I hour.

b.

The NRC Operations Center shall be notified by telephone as soon as 4

possible and in all cases within one hour.

c.

The Vice President-Nuclear Energy and the offsite review function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

A Safety Limit Violation Report shall be prepared and submitted to the Commission, the offsite review function and the Vice President -

Nuclear Energy within 14 days of the violation.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

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

ACTION:

MODES 1 and 2 a.

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, l

be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour, c.

The Vice President-Nuclear Energy and the offsite review function shall be r.otified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

A Safety Limit Violation Report shall be prepared and submitted to the Commission, the offsite review function and the Vice President -

Nuclear Energy within 14 days of the violation.

CALVERT CLIFFS - UNIT 2 2-1 Amendment No. 193 JL

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS MODES 3, 4 and 5 a.

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, l

reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.

j c.

The Vice President-Nuclear Energy and the offsite review function shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

J d.

A Safety Limit Violation Report shall be prepared and submitted to the Commission, the offsite review function and the Vice President -

Nuclear Energy within 14 days of the violation.

i f

3 i

j i

i 4

CALVERT CLIFFS - UNIT 2 2-2 Amendment No. 193

".=

mMm4 EEM >2O rm3Hd2* M>Qy y4E wm$gD 0

2 i

f A 1

E IBArN i

eO f

6 I

L sT cA cR R

AE E

NP W

iuO i 4 O L PLA M

E R

IN E

i 9

B 1

2 I

A L H 3

TP D

E E

3 C

I

\\

C A

A O R i

$i i

L F O

S N

S O

E T

I I

i M

MI t A C E

E O A C

L H

R N

D A D

F F

N -

AR N U EIO S WO A Q T

E mE DR S B M NV IOL 6

E 1

i S

lO P

R.

O1 LDA I

0 P

l A

A R A G 2

NA V R

H N NB L

OE SIA OUF I

N T KIE HICE OY i

z-CL NN L ADR A A F

AIE U

R A t

0 D N IE S 4

S 8

A O T M N XP S t

A G

EYS O

^

LN NUI ALEF PB Y I

T RAIN OFR i

N*

RE t

CO OUA O DS O O 8D R(fA I

ED C SU F A R R A ENI DRUE T

N i

0 6

IS CNO 2

P R TIC I

I MRU LDC0 AAO OP IUL AOA EI B 1

FO LFF VRFU RHS 0

a 0

0 0

0 g

8 0

4 2

0 8

5 5

6 5

5 4

m *-

g3 gg ymca" N,a -

-e mm$4o" nomm 4%

F g g E m>M Q.m

r OG!-a1 n[m] i g 4N m8" g" $.

w

{

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

i ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with i

the Trip Setpoint value.

i 1

l l

CALVEPT CLIFFS - UNIT 2 2-4 Amendment No. 193 l

l

g TABLE 2.2-1 m

Gg REACTOR PKJTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS W

b FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 54 m3 1.

Manual Reactor Trip Not Applicable Not Applicable g

2.

Power Level - High

< 10% above THERMAL POWER, with a < 10% above THERMAL POWER, and i5 E

minimum setpoint of 30% of RATED a minimum setpoint of 30% of d

Z THERMAL POWER, and a maximum of RATED THERMAL POWER and a 5 107.0% of RATED THERMAL POWER.

maximum of 5 107.0% of RATED E

m THERMAL POWER.

g

>95)ofdesignreactorcoolant E95)ofdesignreactorcoolant i!

3.

Reactor Coolant Flow - Low N y

Tlow flow g

4.

Pressurizer Pressure - High 5 2400 psia 5 2400 psia 5.

Containment Pressure - High 5 4 psig 5 4 psig A

m 6.

SteagGeneratorPressure-E 685 psia 1 685 psia Low g

7.

Steam Generator Water Level -

t 10 inches below top of feed 2,10 inches below top of feed y

Low ring ring

,m E

E iir a

an w

Q TABLE 2.2-1 (Continued)

,m

<g REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E

~

b FUNCTIONAL UNIT TRIP SETiOINI ALLOWABLE VALUES m

M m3 8.

Axial flux offset N Trip setpoint adjusted to not Trip setpoint adjusted to not r-exceed the limits provided in exceed the limits provided in E

i the COLR the COLR Q

ez m

N Q

9.

Thermal Margin / Low Pressure g

a.

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to be f

Operating exceed the limits provided in not less than the larger of, g

the COLR (1) 1875 psia, or (2) the y

limits provided in the COLR g

Steam Generator Pr 5 135 psid 5 135 psid

[

b.

Difference - High gsure 4

m m

a

10. Loss of Load NA NA 4

N

11. Rate of Change of Power - High 5 2.6 decades per minute 5 2.6 decades per minute E

N4 EU E

3 a,

rt

.E w

.u

-m.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TABLE 2.2-1 (Continued)

TABLE NOTATION See Specification 3.2.5, "DNB Parameters," for the design reactor coolant flow.

(1)

Trip may be bypassed below 10-'% OF RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is t 10-*% of RATED THERMAL POWER.

(2)

Trip may be manually bypassed below 785 psia; bypass shall be automatically removed at or above 785 psia.

(3)

Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 215% of RATED THERMAL POWER.

(4)

Trip may be bypassed below 10-'% and above 12% of RATED THERMAL POWER.

CALVERT CLIFFS - UNIT 2 2-7 Amendment No. 193 l

~

3/4.3 INSTRUMENTATION l

TABLE 3.3-6 (Continued) t TABLE NOTATION j

Alarm setpoint to be specified in a controlled document (e.g., setpoint control manual).

ACTION STATEMENTS ACTION 14 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 16 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.

l ACTION 30 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable channel (s) to 0PERABLE status within 7 days of the event, or 2) prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within 30 days following the l

event, outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

l l

CALVERT CLIFFS - UNIT 2 3/4 3-25 Amendment No. 193

1 3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Meteorolooical Instrumentation LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a.

With one or more required meteorological monitoring channels 1

inoperable for more than 7 days, prepare and submit a Special

)

Report to the Commission pursuant to 10 CFR 50.4 within the next l

1 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5.

i 1

CALVERT CLIFFS - UNIT 2 3/4 3-27 Amendment No. 193 j

i 3/4.3 INSTRUMENTATION TABLE 3.3-10 (Continued)

ACTION STATEMENTS ACTION 31 -

With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 32 -

With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERABLE requirement in Table 3.3-10, operation may proceed provided the inoperable channel is restored to OPERABLE status at the next outage of sufficient duration.

ACTION 33 -

With the number of OPERABLE post-accident monitoring channels two less than required by Table 3.3-10, either restore one inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 34 -

With the number of OPERABLE post-accident monitoring channels one less than the Minimum Channels OPERABLE requirement in Table 3.3-10, either restore the system to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within 30 days l

following the event, cutlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to 0PERABLE status.

ACTION 35 -

With the number of OPERABLE channels two less than required by Table 3.3-10, either restore the inoperable channel (s) to 0PERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring for core and Reactor Coolant System voiding; 2.

Prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within 30 days following the l

event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring i

the system to OPERABLE status; and 3.

Restore the system to OPERABLE status at the next scheduled refueling.

CALVERT CLIFFS - UNIT 2 3/4 3-35 Amendment No. 193

,j

3/4.3 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Fire Detection Instrumentation LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE.

ACTION: With one or more of the fire detection instrument (s) shown in Table 3.3-11 inoperable:

Within I hour establish a fire watch patrol to inspect the a.

zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect the containment at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5; or unless the instrument (s) is located in fire detection zones equipped with automatic wet pipe sprinkler systems alarmed and supervised to the Control Room, then within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, inspect the zone (s) with inoperable instruments and verify that the Automatic Sprinkler System, including the water flow alarm and supervisory system, is OPERABLE by CNANNEL FUNCTIONAL TEST.

b.

Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within the next 30 days outlining the l

action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status, The provisions of Specifications 3.0.3 and 3.0.4 are not c.

applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.7.1 At least once per 6 months, at least 25% of the above required fire detection instruments which are accessible during plant operation i

shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST Detectors selected for testing shall be selected on a rotating basis such I

I CALVERT CLIFFS - UNIT 2 3/4 3-37 Amendment No. 193

_ -- -=_- -.

i 3/4.4 REACTOR C0OLANT SYSTEM 4

l j

SURVEILLANCE REQUIREMENTS (Continued)

10. Tube Reoair refers to a )rocess that reestablishes tube i..

serviceability. Accepta)1e tube repairs will be perfonned by the following process:

4 i

a) Westinghouse Laser Welded Sleeving as described in the proprietary Westinghouse Reports WCAP-13698, Revision 2, 1

" Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam i

i Generators, Generic Sleeving Report " April 1995; and j

WCAP-14469, " Specific Application of Laser Welded Sleeving for the Calvert Cliffs Power Plant Steam Generators," November 1995.

~

Tube repair includes the removal of plugs that were l

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 plugged tubes to service.

j b.

The steam generator shall be determined OPERABLE after completing i

the corresponding actions (plug or repair all tubes exceeding the l

plugging or repair limit and all tubes containing through-wall 1

^

cracks) required by Table 4.4-2.

i 4.4.5.5 ReDorts j

.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.

l b.

The complete results of the steam generator tube inservice i

inspection during the re) ort period shall be submitted to the i

Comission prior to Marc 11 of each year pursuant to 10 CFR 50.4.

This report shall include:

j 1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each j

indication of an imperfection.

j.

3.

Identification of tubes plugged or repaired.

i c.

Results of steam generator tube inspections which fall into Category C-3 require verbal notification of the NRC Regional Administrator by telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to resumption of 3

j plant operation. The written followup of this report shall provide a description of investigations conducted to determine i

cause of the tube degradation and corrective measures taken to i

prevent recurrence and shall be submitted within the next 30 days j

pursuant to 10 CFR 50.4.

l i

i CALVERT CLIFFS - UNIT 2 3/4 4-14 Amendment No. 193 1

g TABLE 4.4-2 w

r-

+

STEAM GENERATOR TURE INSPECTION m

d 15T 5AlWLE IN5PLEIION Me 5ARFLE ImrtETIUN JHU i___i.E In rtLIIgu up h

Sample Size Result Action Required Result Action Required Result Action Required S

y A minimum of 5 Tubes per C-1 Mone N/A N/A N/A-N/A Q

y SG.

C-2 Plug 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

C this SG.

inspect additional Plug or repair 5

45 tubes in this C-2 defective tu'ws

[

H SG, Perfom action y

N C-3 for C-3 result of first saimle g

C-3 Perfor1n action for

-4 C-3 result of first N/A N/A E

sample C-3 Inspect all tubes in All other SGs this SG, plug or repair are C-1 None N/A N/A defective tubes and inspect 25 tubes in each other SG.

m 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal Some SGS C-2 Perfot1R action for notification to NRC with but no C-2 result of N/A N/A written followup additional second sample pursuant to 10 CFR 50.4.

SG are C-3 l

Additional Inspect all tubes SG is C-3 in each SG and plug or repair defective N/A N/A tubes. 24-hour

{

verbal notification to NRC with written 3

ct followup pursuant 5

to 10 CFR 50.4.

l A

S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators 2

P inspected during an inspection G

w

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 DOSE EQUIVALENT I-131, and b.

5 100/E pCi/ 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 1

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 1

DOSE 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 interval or exceeding the limit line shown on Figure 3.4.8-1, 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 .

c.

With the specific activity of the primary coolant > 100/E Ci/ gram, be in at least H0T 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 I-131 or > 100/E pCi/ 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.

With T.,, > 500 F.

CALVERT CLIFFS - UNIT 2 3/4 4-24 Amendment No. 193

3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

l i

i i

l CALVERT CLIFFS - UNIT 2 3/4 4-25 Amendment No. 193 l

3/4.4 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

l a.

With one PORY inoperable in MODE 3 with RCS temperature < 301 F or in M0DE 4, either restore the inoperable PORV to OPERABLE status within 5 days or depressurize and vent the RCS through a 1 1.3 square inch vent (s) within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; 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 depressurize and vent the RCS through a t 1.3 square inch vent (s) within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; and maintain the RCS in this vented condition until both PORVs have been restored to 0PERABLE status.

c.

With both PORVs inoperable, depressurize and vent the RCS through a 2 2.6 square inch vent (s) within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; maintain the RCS in a vented condition until either one OPERABLE P0RV and a vent of 1 1.3 square inches has been established or both PORVs have been restored 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 l

within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on j

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 loop MOVs' not prevented from automatically aligning a HPSI pump to the RCS, immediately 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 i

to 210 gpm.

Except when required for testing.

CALVERT CLIFFS - UNIT 2 3/4 4-34 Amendment No. 193

4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.11 CORE BARREL MOVEMENT l

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 least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.

SA shall be measured at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and shall be processed at least once per 7 days, and 3.

A Special Report, identifying the cause(s) for exceeding the applimble Alert Level, shall be prepared and submitted to the Coanission pursuant to 10 CFR 50.4 within 30 days of l

detection.

b.

With the APD and/or SA exceeding their applicable Action Levels, measure and process APD and SA data within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H0T STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

4 CALVERT CLIFFS - UNIT 2 3/4 4-39 Amendment No. 193

i 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.2 ECCS SUBSYSTEMS - MODES 1, 2 AND 3 (2 1750 PSIA)

LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high-pressure safety injection pump, b.

One OPERA 8LE low-pressure safety injection pump, and c.

An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a.

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to 10 CFR 50.4 within l

90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

i With pressurizer pressure > 1750 psia.

CALVERT CLIFFS - UNIT 2 3/4 5-3 Amendment No. 193

3/4.5 EMERGENCY C0RE COOLING SYSTEMS (ECCS) 3/4.5.3 ECCS SUBSYSTEMS - MODES 3 (< 1750 PSIA) AND 4 LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

One' OPERABLE HPSI pump, and a.

i b.

An OPERABLE flow path capable of taking suction from the i

refueling water tank on a Safety Injection Actuation Signal and cutomatically transferring suction to the containment sump on a Recirculation Actuation Signal.

APPLICABILITY: MODES 3* and 4.

ACTION:

a.

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN l

within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects water into the RCS, a Special Report shall be prepared and submitted to the Commission pursuant to 10 CFR 50.4 within 90 days describing the l

circumstances of the actuation and the total accumulated actuation cycles to date.

l SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of Specification 4.5.2.

Between 325'F and 301 F, a transition region exists where the OPERABLE HPSI pump will be placed in p'ull-to-lock on a cooldown and restored to automatic status on a heatup. At 301*F and less, the required OPERABLE HPSI pump shall be in pull-to-lock and will not start automatically. At 301 F and less, HPSI pump use will be conducted in accordance with Technical Specification 3.4.9.3.

With pressurizer pressure < 1750 psia.

CALVERT CLIFFS - UNIT 2 3/4 5-7 Amendment No. 193 L

l l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Leakaae LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

l a.

A maximum allowable containment leakage rate, L,, as specified in Specification 6.5.6, " Containment Leakage Rate Testing Program."

l b.

A combined leakage rate of 5 0.50 L, (173,000 SCCM), for all penetrations and valves subject to Type B and C tests when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: With either (a) the measured overall integrated containment leakage rate exceeding the acceptance criteria specified in the Containment Leakage Rate Testing Program, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.50 L., restore the overall integrated containment leakage rate to within the acceptance criteria specified in the Containment Leakage Rate Testing Program, and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than or equal to 0.50 L, prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria, methods and provisions specified in 10 CFR Part 50, Appendix J:

a.

Perform required visual examinations and Type A testing in accordance with the Containment Leakage Rate Testing Program.

CALVERT CLIFFS - UNIT 2 3/4 6-2 Amendment No. 193

3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.3 Containment Surfaces. The exposed accessible interior and exterior surfaces of the containment, including the liner plate shall be visually inspected in accordance with the Containment Leakage Rate Testing Program (reference Specification 4.6.1.2).

4.6.1.6.4 Reports. Any abnormal degradation of the containment structure detected during the above required tests and inspections shall be reported to the Commission pursuant to 10 CFR 50.4 within the next 30 days. This l

report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective actions taken.

CALVERT CLIFFS - UNIT 2 3/4 6-9 Amendment No. 193

l 3/4.7 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a.

Sources in use - At least once per six months for all sealed sources containing radioactive material:

1.

With a half-life greater than 30 days (excluding Hydrogen 3),and 2.

In any form other than gas, b.

Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months.

Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c.

Startuo sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected _ to core flux or installed in the core and following repair or maintenance to the source or detector.

4.7.9.1.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis pursuant to 10 CFR 50.4 if sealed source or l

fission detector leakage tests reveal the presence of > 0.005 microcuries of removable contamination.

CALVERT CLIFFS - UNIT 2 3/4 7-31 Amendment No. 193

3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System LIMITING CONDITION FOR OPERATION 3.7.11.1 The Fire Suppression Water System shall be OPERABLE with:

a.

Two high pressure pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header, b.

Two water supplies, each with a minimum contained volume of 300,000 gallons, and c.

An OPERABLE flow path capable of taking suction from the Pretreated Water Storage Tanks Numbers 11 and 12 and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser required to be OPERABLE per Specifications 3.7.11.2, 3.7.11.4, and 3.7.11.5.

APPLICABILITY: At all times.

ACTION:

a.

With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or prepare and submit a Special Report to the Conmission pursuant to 10 CFR 50.4 within the next 30 days outlining the plans and l

procedures to be used to provide for the loss of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b.

With the Fire Suppression Water System otherwise inoperable:

1.

Establish a backup Fire Suppression Water System within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.

Submit a Special Report in accordance with 10 CFR 50.4:

l a) By telephone witM n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , CALVERT CLIFFS - UNIT 2 3/4 7-33 Amendment No. 193

3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Sorav and/or Sprinkler Systems LIMITING CONDITION FOR OPERATION 3.7.11.2 The spray and/or sprinkler systems shown in Table 3.7-5 shall be OPERABLE:

APPLICABILITY: Whenever equipment in the spray / sprinkler protected areas is required to be OPERABLE.

ACTION:

a.

With one or more of the required spray and/or sprinkler systems inoperable, within one hour establish a. continuous fire watch with backup fire suppression equipment for those areas in which redundant safe shutdown systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

Restore the system to 0PERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to 10 CFR 50.4 - l within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, i

power-operated or automatic) in the flow path, not locked, sealed 1

or otherwise secured in position, is in its correct position.

b.

At least once per 12 months by cycling each valve in the flow path through at least one complete cycle of full travel.

c.

At least once per 18 months:

1.

By performing a system functional test which includes

]

simulated automatic actuation of the~ system, and verifying j

that the automatic valves in the flow path actuate to their correct positions on a simulated test signal.

CALVERT CLIFFS - UNIT 2 3/4 7-37 Amendment No. 193

3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Halon Systems LIMITING CONDITION FOR OPERATION 3.7.11.3 The following Halon Systems shall be OPERABLE with the storage tanks having at least 95% of full charge weight (or level) and 90% of full charge pressure.

a.

Cable spreading room total flood system, and associated vertical cable chase IC, Unit 2.

b.

4160 volt switchgear room 27' & 45' elevation Unit 2.

APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE.

ACTION:

a.

With both the primary and backup Halon Systems protecting the areas inoperable, within one hour establish an hourly fire watch with backup fire suppression equipment for those areas protected by the inoperable Halon System.

Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to 10 CFR 50.4 within the next 30 days I

outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE

status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.3 Each of the above required Halon Systems shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated or automatic) in the flow path is in its correct position.

b.

At least once per 6 months by verifying Halon storage tank weight (level) and pressure.

CALVERT CLIFFS - UNIT 2 3/4 7-40 Amendment No. 193

3/4.7 PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS Fire Hose Stations LIMITING CONDITION FOR OPERATION j

3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.

ACTION:

a.

With one or more of the fire hose stations shown in Table 3.7-6 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Restore the fire hose station (s) to 0PEPACLE status within 14 days or prepare and submit a Special Repcrt to the Commission pursuant to 10 CFR 50.4 within the next 30 days l

outliningtheactiontaken,thecauseoftheinoperability(and the plans and schedule for restoring the fire hose station s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-6 shall be demonstrated OPERABLE:

a.

At least once per 31 days by visual inspection of the station to assure all required equipment is at the station. Hose stations located in the containment shall be visually inspected on each scheduled reactor shutdown, but not more frequently than every 31 days.

b.

At least once per 18 months for hose stations located outside the containment and once per REFUELING INTERVAL for hose stations inside the containment by:

1.

Removing the hose for inspection and re-racking, and 2.

Replacement of all degraded gaskets in couplings.

CALVERT CLIFFS - UNIT 2 3/4 7-42 Amendment No. 193

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

3/4.7 PLANT SYSTEMS j

3/4.7.11 FIRE SUPPRESSION SYSTEMS Yard Fire Hydrants and Hydrant Hose Houses i

l LIMITING CONDITION FOR OPERATION 3.7.11.5 The following yard fire hydrants and associated hydrant hose houses shall be OPERABLE:

1 l

a.

6 yard hydrant and associated hydrant hose house, which provides primary protection for Unit 2 RWT blockhouse.

7 yard hydrant and associated hydrant hose house, which provides 1

u.

primary protection for Unit 1 RWT blockhouse.

I APPLICABILITY:

Whenever equipment in the areas protected by the yard fire 1

hydrants is required to be OPERABLE.

i l

ACTION:

a.

With one or more of the yard fire hydrants or associated hydrant hose houses inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient additional lengths of 2-1/2 inch diameter hose located in an adjacent i

OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant hose 3

9 house is the primary means of fire suppression. Restore the hydrant or hose house to OPERABLE status within 14 days or i

prepare and submit a Special Report to the Commission pursuant to 10.CFR 50.4 witnin the next 30 days outlining the action taken, l

i the cause of the inoperability, and the plans and schedule for restoring the hydrant or hose house to 0PERABLE status.

i j

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not j

applicable.

I T

i l

i 1;

i CALVERT CLIFFS - UNIT 2 3/4 7-45 Amendment No. 193

3/4.7 PLANT SYSTEMS 3/4.7.12 PENETRATION FIRE BARRIERS LIMITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (i.e., cable penetration barriers, fire doors and fire dampers), in fire zone boundaries, protecting safe shutdown areas shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a.

With one or more of the above required fire barrier penetrations inoperable within one hour either establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the inoperable fire barrier and establish an hourly fire watch patrol; or verify the operability of Automatic Sprinkler Systems (including the water flow alarm and supervisory system) on both 1

sides of the inoperable fire barrier.

Restore the inoperable fire barrier penetration (s) to OPERABLE status within 7 days or i

prepare and submit a Special Report to the Commission pursuant to j

10 CFR 50.4 within the next 30 days outlining the action taken, l

the cause of the inoperable penetration and plans and schedule for restoring the fire barrier penetration (s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be OPERABLE:

a.

At least once per 18 months by a visual inspection.

b.

Prior to returning a fire barrier penetration to functional status following repairs or maintenance by perfonnance of a visual inspection of the affected fire barrier penetration (s).

CALVERT CLIFFS - UNIT 2 3/4 7-47 Amendment No. 193

l 3/4.4 REACTOR COOLANT SYSTEM BASES l

penetrated 20% of the original tube wall thickness. Repaired tubes are also included in the inservice tube inspection program.

1 Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the I

Commission prior to the resumption of plant operation. Such cases will be l

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

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakaae Detection Systems The RCS Leakage Detection Systems required by this specification are i

provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations I.

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 recommendations of Generic Letter 91-04.

PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of.the pressure boundary.

Therefore, the presence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3/4.4.7 CHEMISTRY The limitations on RCS chemistry ensure that corrosion of the RCS is minimized and reduce the potential for RCS leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits CALVERT CLIFFS - UNIT 2 B 3/4 4-5 Amendment No. 193

i 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall facility l

operation and shall delegate in writing the succession to this responsibility during his absence.

j 6.2 ORGANIZATION 6.2.1 Onsite & Offsite Oraanizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility and communication shall be 1

established and defined for the highest management levels through 2

intermediate levels to and including all operating organizatior positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent fonns of documentation. These requirements, including the plant-specific titles of personnel fulfilling the responsibilities of the positions delineated in these Technical i

Specifications, shall be documented in the Updated Final Safety Analysis Report (UFSAR).

b.

The plant manager shall be responsible for overall unit safe l

4 operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c.

The Vice President - Nuclear Energy shall have corporate responsibility for overall plant nuclear safety and shall take 1

any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who train the operating staff and those who carry-out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 Unit Staff The unit staff organization shall include the following:

a.

A total of at least three non-licensed operators shall be assigned to the Units 1 and 2 shift crews.

CALVERT CLIFFS - UNIT 2 6-1 Amendment No. 193 e

6.0 ADMINISTRATIVE CONTROLS b.

At least one licensed Operator shall be in the Control Room when fuel is in the reactor.

4 c.

At least two licensed Operators shall be present in the Control Room during reactor STARTUP, scheduled reactor shutdown, and i

during recovery from reactor trips.

d.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

i e.

A site Fire Brigade of at least five members shall be maintained l

1 onsite at all times. The Fire Brigade shall not include the minimum shift crew necessary for safe shutdown of both units (four members) or any personnel required for other essential i

functions during a fire emergency.

Fire Brigade training shall i

meet the requirements of NFPA 27, 1975 edition.

f.

The operations manager shall hold or have held a senior reactor operator license at Calvert Cliffs. The General Supervisor, Shift Supervisor and Control Room Supervisor shall hold a senior reactor operator license. The Control Room Operator shall hold a reactor operator license.

i l

g.

One Shift Technical Advisor (STA) shall be assigned to the shift crew when either unit is in MODE 1, 2, 3 or 4, and shall be filled as follows:

1.

By the Shift Supervisor or an on-shift Senior Operator License (SOL) holder, provided the individual meets the Comission Policy Statement on Engineering Expertise on Shift; or 2.

By an individual meeting the minimum STA education and training requirement of Specification 6.3.1; or 3.

By an SOL holder previously approved by the NRC as an exception to the minimum STA education requirements of Specification 6.3.1, provided the following conditions are met:

(a) With both units in MODE 1, 2, 3 or 4, the STA shall be an SOL holder in addition to the two SOL holders required; (b) With one unit in MODE 1, 2, 3 or 4 and the other unit in MODE 5 or 6, the STA shall be an SOL holder other than the Shift Supervisor; and (c) With one unit in MODE 1, 2, 3 or 4 and the other unit defueled, the STA shall be an SOL holder in addition to the one SOL holder required.

f CALVERT CLIFFS - UNIT 2 6-2 Amendment No. 193

6.0 ADMINISTRATIVE CONTR0LS h.

Shift crew composition ma of 10 CFR 50.54(m)(2)(1) y be less than the minimum requirements and Specifications 6.2.2.a and 6.2.2.g for a period of time not to exceed two hours in order to

{

accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew 1

composition to within the minimum requirements.

1.

Licensed operators shall be licensed for both units.

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Safety Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 PROCEDURES l

6.4.1 Written procedures shall be established, implemented and maintained l

covering the activities referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c.

Quality assurance for effluent and environmental monitoring; d.

Fire Protection Program implementation; e.

All programs specified in Specification 6.5; and f.

The amount of overtime worked by plant staff members performing i

safety-related functions must be limited in accordance with the NRC Policy Statement on Working Hours (Generic Letter 82-12).

l CALVERT CLIFFS - UNIT 2 6-3 Amendment No. 193

4

)

6.0 ADMINISTRATIVE CONTROLS 6.5 PROGRAMS AND MANUALS The following programs shall be established, implemented and maintained:

6.5.1 0FFSITE DOSE CALCULATION MANUAL (ODCM) a.

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous i

and liquid effluents, in the calculation of gaseous and liquid i

effluent monitoring alann and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program; and b.

The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports, required by Specifications 6.6.2 and 6.6.3, respectively.

c.

Licensee initiated changes to the ODCM:

1.

Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

(a) Sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying thechange(s);

4 (b) A determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, i

40 CFR Part 190, 10 CFR 50.36a, and 10 CFR Part 50, Appendix I, and ~not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; 2.

Shall become effective after review and acceptance by the onsite review function and the approval of the plant manager; and 3.

Shall be submitted to the NRC in the fonn of a complete, legible copy of the entire ODCM as part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

CALVERT CLIFFS - UNIT 2 6-4 Amendment No. 193

6.0 ADMINISTRATIVE CONTROLS 6.5.2 Post-Accident Samplina l

The licensee shall establish, implement and maintain a program

which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall inciude the following:

a.

Training of personnel, b.

Procedures for sampling and analysis, and c.

Provisions for maintenance of sampling and analysis equipment.

6.5.3 Primary Coolant Sources Outside Containment The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

a.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and b.

Leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.5.4 Technical Specification Bases Control Program This program provides a means for processing changes to the Technical Specification Bases.

a.

Changes to the Bases of the Technical Specifications shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

1.

A change in the Technical Specifications incorporated in the license; or 2.

A change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

It is acceptable if the licensee maintains details of the program in plant operation manuals (e.g., chemistry procedures, training instructions, maintenance procedures, ERPIPs).

i CALVERT CLIFFS - UNIT 2 6-5 Amendment No. 193 I

l 6.0 ADMINISTRATIVE CONTROLS c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes to the Technical Specifications incorporated in d

the license or proposed changes to the UFSAR or Bases that involve an unreviewed safety question shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC a) proval shall be provided to the NRC on a frequency consistent witi 10 CFR 50.71(e).

6.5.5 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from j

radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial-actions to be taken whenever the program limits are i

exceeded. The program shall include the following elements:

a.

Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation, including surveillance

{

tests and setpoint determination, in accordance with the methodology in the ODCM; I

b.

Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to j

i 10 CFR Part 20, Appendix B. Table II, Column 2 i

c.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents, in accordance with 10 CFR 20.1302, and with the methodology and parameters in the ODCM; d.

Limitations on the' annual and quarterly doses or dose comitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, 2

confonning to 10 CFR Part 50, Appendix I; e.

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and i

current calendar year, in accordance with the methodology and parameters in the ODCM, at least every 31 days; f.

Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of 4

radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose comitment, confonning to 10 CFR Part 50, Appendix I; g.

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR Part 20 Appendix B, Table II, Column 1; CALVERT CLIFFS - UNIT 2 6-6 Amendment No. 193

~,

6.0 ADMINISTRATIVE CONTROLS h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, confoming to 10 CFR Part 50, 1

Appendix I; i.

Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and.all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the j

4 site boundary, conforming to 10 CFR Part 50, Appendix I; and j.

Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR Part 190.

6.5.6 Containment Leakaae Rate Testino Proaram l

A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50,-Appendix J.

Option B.

This program shall be in accordance with.the guidelines contained in Regulatory Guide 1.163, "Perfomance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, P,, is 49.4 psig. The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, L, shall be 0.20 percent of containment air weight per day at P.

Containment leakage rate acceptance criterion is 51.0 L.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria is 5 0.75 L, for Type A tests.

The provisions.of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

CALVERT CLIFFS - UNIT 2 6-7 Amendment No. 193

j

)

6.0 ADMINISTRATIVE CONTROLS 6.6 REPORTING REOUIREMENTS 1

The following reports shall be submitted in accordance with 10 CFR 50.4.

I 6.6.1 Occupational Radiation Exposure Report

l j

A tabulation on an annual basis of the number of station, utility, and 1

other personnel (including contractors) receiving exposures > 100 mrem /yr and their associated man rem exposure according to work-and job functions (e.g... reactor operations and surveillance, inservice inspection, routine j

maintenance, special maintenance [ describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignment to various duty functions may be estimates based on pocket dosimeter, electronic personal dosimeter or f

thermoluminescent dosimeter.

Small exposures totalling < 20% of the individual total dose need not be accounted for. -In the aggregate, at

{

1 east 80% of the total whole body dose received from external sources i

should be assigned to specific major work functions. The report shall be submitted prior to, March 31 of each year.

}

{

6.6.2 Annual-Radioloaical Environmental Operatina Report

?

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted l-prior to May 1 of each year.

1 The report shall include summaries, interpretations, and anaryses of trends of the results of the Radiological Environmental Monitoring Program for the

)

i reporting period. The material provided shall be consistent with the objectives dutlined in the ODCM, and in 10 CFR Part 50, Appendix I, i

Sections IV.B.2, IV.B.3 and IV.C.

The report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken j

during the period pursuant to the locations specified in the table and figures in the ODCM, as well as suninarized and tabulated results of these

[

analyses and measurements in the fonnat of the table in the Radiological j

Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted ncting and explaining the reasons for '

j the missing results. The missing data shall be submitted as soon as l

possible in a supplementary report.

l l

i A single submittal may be made for Calvert Cliffs.

The submittal should combine those sections that are common to both units.

Occupational dose from the Independent Spent Fuel Storage Installation will be reported separately.

CALVERT CLIFFS - UNIT 2 6-8 Amendment No. 193 l

6.0 ADMINISTRATIVE CONTROLS 6.6.3 Radioactive Effluent Release Report

The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a (i.e., time between submittal of the reports must be no longer than 12 months).

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units. The material provided shall be consistent with the objectives outlined in the 00CM and Process Control Program and in confonnance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

Licensee initiated major changes to the Radioactive Waste Svstems (liquid, gaseous and solid) shall be reported to the Comission in t!.e Radioactive Effluent Release Report for the period in which the modification to the waste ' system is completed. The discussion of each change shall contain:

a.

A description of the equipment, components and processes involved; and b.

Documentation of the fact that the change including the safety analysis was reviewed and found acceptable by the onsite review function.

The report shall also include changes to the 00CM, in accordance with Specification 6.5.1.c 6.6.4 Monthly Operatina Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis, no later than the 15th of each month following the calendar month covered by the report.

)

A single submittal may be made for Calvert Cliffs, since the Radwaste Systems are comon to both units.

CALVERT CLIFFS - UNIT 2 6-9 Amendment No. 193

6.0 ADMINISTRATIVE CONTROLS 6.6.5 Core Operatino Limits Report (COLR) l Core operating limits shall be established prior to each reload a.

cycle, or prior to ariy remaining portion of a reload cycle, and s

shall be documented in the COLR for the following:

2.2.1 3.1.1.1 3.1.1.2 3.1.1.4 3.1.3.1 3.1.3.6 3.2.1 3.2.2.1 3.2.3 3.2.5 3.9.1 b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC; specifically, those described in the following documents:

(1) CENPD-199-P, Latest Approved Revision, "C-E Setpoint Methodology: C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems,"

i January 1986 (2) CEN-124(B)-P, " Statistical Combination of Uncertair. ties Methodology Part 1: C-E Calculated Local Power Density and i

Thermal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II," December 1979 (3) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 2: Combination of System Parameter Uncertainties in Thermal Margin Analyses for Calvert Cliffs Units 1 and 2," January 1980 (4) CEN-124(B)-P, " Statistical Combination of Uncertainties Methodology Part 3: C-E Calculated Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Calvert Cliffs Units 1 and 2," March 1980 (5) CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981 (6)

Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr.

(BG&E), dated June 24, 1982, Unit 1 Cycle 6 License Approval (Amendment No. 71 to DPR-53 and SER)

(7) CEN-348(B)-P, " Extended Statistical Combination of Uncertainties," January 1987 CALVERT CLIFFS - UNIT 2 6-10 Amendment No. 193

6.0 ADMINISTRATIVE CONTROLS (8)

Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated October 21, 1987, Docket Nos. 50-317 and 50-318, " Safety Evaluation of Topical Report CEN-348(B)-P, Extended Statistical Combination of Uncertainties" (9)

CENPD-161-P-A, " TORC Code A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986 (10) CENPD-162-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 1. Uniform Axial Power Distribution" (11) CENPD-207-P-A, Latest Approved Revision, " Critical Heat Flux Correlation of C-E Fuel Assemblies with Standard Spacer Grids Part 2, Non-Uniform Axial Power Distribution" (12) CENPD-206-P-A, Latest Approved Revision, " TORC Code.

Verification and Simplified Modeling Methods" (13) CENPD-225-P-A, Latest Approved Revision " Fuel and Poison Rod Bowing" (14) CENPD-266-P-A, Latest Approved Revision, "The ROCS and DIT Computer Code for Nuclear Design" (15) CENPD-275-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Gadolinia - Urania Burnable Absorbers" (16) CENPD-382-P-A, Latest Approved Revision, "C-E Methodology for Core Designs Containing Erbium Burnable Absorbers" (17) CENPD-139-P-A, Latest Approved Revision, "C-E Fuel Evaluation Model Topical Report" (18) CEN-161-(B)-P-A, Latest Approved Revision, " Improvements to Fuel Evaluation Model" (19) CEN-161-(B)-P, Supplement 1-P, " Improvements to Fuel Evaluation Model," April 1989 (20) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated February 4, 1987, Docket Nos. 50-317 and 50-318, " Safety Evaluation of Topical Report CEN-161-(B)-P, Supplement 1-P, Improvements to Fuel Evaluation Model" (21) CEN-372-P-A, Latest Approved Revision, " Fuel Rod Maximum Allowable Gas Pressure" (22) Letter from Mr. A. E. Scherer (CE) to Mr. J. R. Miller (NRC), dated December 15,1981, LD-81-095, Enclosure 1-P, "C-E ECCS Evaluation Model Flow Blockage Analysis" CALVERT CLIFFS - UNIT 2 6-11 Amendment No. 193 l

6.0 ADMINISTRATIVE CONTROLS (23) CENPD-132, Supplement 3-P-A, Latest Approved Revision,

" Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W1 Designed NSSS" (24) CENPD-133, Supplement 5. "CEFLASH-4A, a FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," June 1985 (25) CENPD-134, Supplement 2 "COMPERC-II, a Program for Emergency Refill-Reflood of the Core," June 1985 (26) Letter from Mr. D. M. Crutchfield (NRC) to Mr. A. E. Scherer (CE), dated July 31,1986,. " Safety Evaluation of Combustion Engineering ECCS Large Break Evaluation Model and Acceptance for Referencing of Related Licensing Topical Reports" (27) CENPD-135, Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 (28) Letter from Mr. R. L. Baer (NRC) to Mr. A. E. Scherer (CE),

dated September 6,1978, " Evaluation of Topical Report CENPD-135,. Supplement 5" (29) CENPD-137, Supplement 1-P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," January 1977 (30) CENPD-133, Supplement 3-P, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident," January 1977 (31) Letter from Mr. K. Kniel (NRC) to Mr. A. E. Scherer (CE),

dated September 27, 1977, " Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P" (32) CENPD-138, Supplement 2-P, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," January 1977 (33) Letter from Mr. C. Aniel (NRC) to Mr. A. E. Scherer, dated April 10,1978, " Evaluation of Topical Report CENPD-138, Supplement 2-P" (34) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. J. R. Miller (NRC) dated February 22,1985. "Calvert Cliffs Nuclear Power Plant Unit 1; Docket No. 50-317, Amendment to Operating License DPR-53, Eighth Cycle License Application" (35) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr.

(BG&E), dated May 20, 1985, " Safety Evaluation Report Approving Unit 1 Cycle 8 License Application" CALVERT CLIFFS - UNIT 2 6-12 Amendment No. 193 l

i I

l 6.0 ADMINISTRATIVE CONTROLS l

i (36) Letter from Mr. A. E. Lundvall, Jr. (BG&E) to Mr. R. A. Clark (NRC), dated September 22, 1980, " Amendment to Operating License No. 50-317, Fifth Cycle License i

Application" (37) Letter from Mr. R. A. Clark (NRC) to Mr. A. E. Lundvall, Jr.

1 (BG&E), dated December 12, 1980, " Safety Evaluation Report Approving Unit 1, Cycle 5 License Application" (38) Letter from Mr. J. A. Tiernan (BG&E) to Mr. A. C. Thadani (NRC), dated October 1,1986, "Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2, Docket Nos. 50-317 & 50-318, Request for Amendment" (39) Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. J. A. Tiernan (BG&E), dated July 7, 1987, Docket Nos. 50-317 and 50-318, Approval of Amendments 127 (Unit 1) and 109 (Unit 2)

(40) CENPD-188-A, Latest Approved Revision, "HERMITE: A Multi-l Dimehsional Space-Time Kinetics Code for PWR Transients" (41) The Full Core Power Distribution Monitoring System i

referenced in Specifications 3.1.3.1, 3.2.2.1, 3.2.3, and the BASES is described in the following documents:

(a) CENPD-153-P, Latest Approved Revision, " Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed Incore Detector System" (b) CEN-199(B)-P,"BASSS,UseoftheIncoreDetectorSystem to Monitor the DNB-LCO on Calvert Cliffs Unit I and Unit 2," November 1979 (c)

Letter from Mr. G. C. Creel (BG&E) to NRC Document Control Desk, dated February 7, 1989, "Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket 50-318, Request

)

for Amendment, Unit 2 Ninth Cycle License Application" (d)

Letter from Mr. S. A. McNeil, Jr. (NRC) to Mr. G. C. Creel (BG&E), dated January 10, 1990, " Safety Evaluation Report Approving Unit 2 Cycle 9 License Application" (42)

Letter from Mr. D. G. Mcdonald, Jr. (NRL) to Mr. R. E. Denton (BGE), dated May 11,199f, " Approval to Use Convolution Technique in Main Steam Line Ereak Analysis -

Calvert Cliffs Nuclear Power Plant, Unit Kos. I and 2 (TAC Nos. M90897 and M90898)

CALVERT CLIFFS - UNIT 2 6-13 Amendment No. 193 l

f jl 6.0 ADMINISTRATIVE CONTR0LS l

c.

The core operating (e.mits shall be determined such that all li a)plicable limits g., fuel thermal mechanical limits, core tiennal hydraulic limits, Emergency Core Cooling Systems (ECCS) i limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall j

be provided upon issuance for each reload cycle to the NRC.

i 6.6.6 Pressurizer PORV and Safety Valve Report A report shall be submitted prior to March 1 of each year documenting all failures and challenges to the pressurizer PORVs or safety valves.

l l

i l

l 1

CALVERT CLIFFS - UNIT 2 6-14 Amendment No. 193 l

t

ML20117H621

Text

Attachment:

Attachment:

Navigation menu

Search