Text

Amends 75 & 56 to Licenses DPR-53 & DPR-69,respectively, Correcting Typographical Errors Re Inservice Insp & Valve Testing & Allowing Start Up of Reactors W/Inoperable But Isolated Power Operated Relief Valves
	ML20071N291
Person / Time
Site:	Calvert Cliffs
Issue date:	09/20/1982
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20071N297	List: ML20071N291; ML20071N300; ML20071N303;
References
	NUDOCS 8210070117
	Download: ML20071N291 (34)
	v • d • e

h UNITED STATES **

E%-

'i NUCLEAR REGULATORY COMMISSION

{.;1

sf c ]n,E WASHINGTON, D. C. 20555 BALTIM0RE GAS AND ELECTRIC COMPANY

~

DOCKET N0. 50-317

~

CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE t

Amendment No. 75 License No. DPR-53 1.

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

A.

The application for amendment by Baltimore Gas & Electric Company (the licensee) dated August 2,1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

33 3.,

OM 5IEAL ff'M 0$L Gertified By _

(

8210070117 820920 PDR ADOCK 05000317 P

PDR

. 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:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No'. 75, 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.

FOR THE NUCLEAR REGULATORY COMMISSION-

/

ober

. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 20, 1982 e

e e

i

.m

4 s

ATTACHMENT TO LICENSE AMENDMENT N0. 75 FACILITY OPERATING LICENSE h0. DPR-53

!l DOCKET N0. 50-317

~..~

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by

. Amendment number and contain vertical lines indicating the area of change.

i The corresponding overleaf pages are also provided to maintain document

i. completeness.

i Pages s-3/4 4-4 3/4 4-28 3/4 5-Sa 3/4 6-2 3/4 6-3 3/4 6-4 3/4 6-18 6-13 e

l 1

I N

...----..-w

o.

s REACTOR COOLANT SYSTEM SAFETY VALVES LIMITING CONDITION FOR OPERATION

.1 The following pressurizer code safety valves'shall be'6PERABLE:

a.s.

Valve Lift Settings ( 1%1 RC-200 2500 psia RC-201 2565 psia APPLICA3ILITY:

MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ..

3.4.2.2 At least one of the above pressurizer code safety valves shall be OPERABLE:*

APPLICABILITY:

MODES 4 and 5.

ACTION:

l With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity. changes and place an OPERABLE shutdown cooling loop into operation.

SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

I

Both valves may be removed in MODE 5 provided at least one valve is replaced by a spool piece which allows the pressurizer to relieve directly to the quench tank.

CALVERT CLIFFS - UNIT 1 3/4 4-3 Amendment No.34, 53 I

~

REACTOR COOLANT SYSTEM L

RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3 Two power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With one or more PORV(s) inaperable, within.1' hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the block.

valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With one or more block valve (s) closed and power removed from the block valve (s) to satisfy a. or b. above, the provisions of Specifica-tion 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.3.1 Each' PORY shall be demonstrated OPERABLE:

a.

At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, in accordance with Table 4.3-1. Item 4.

b.

At least once per 18 months by performance of a CHANNEL CALIBRATION.

4.4.3.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel, i

This demonstration is not required if a PORV block valve is closed and power.

removed to meet Specification 3.4.3 a. or b.

CALVERT CLIFFS - UNI'T 1 3/4 4-4 Amendment No. 53,75 A

A -

a

.-..__,-,---,we,.

,,,,,,m,w,_

~

__..y,.

,.-... -.,,..,n,e_,,n m,,-.

- - - 9

,-w, y

y--9 p

ye-,

REACTOR C00LAN1 SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 AND 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 com-ponents shall be maintained in accordance with Specification 4.4.10.1.

APPLICABILITY: ALL MODES.

ACTION:

a.

With the structural integrity of any ASME Code Class 1 com-ponent(s) not conforming to the above requirements, restore tha structural integrity of the affected component (s) to witMn its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50"F above the minimum temperature required by NDT considera-tions.

b.

With the structural integrity of any ASME Code Class 2 com-ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.

c.

With the structural integrity of any ASME Code: Class 3 com-l ponent(s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l

l 4.4.10.1.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated:

l a.

Per the requirements of Specification 4.0.5, and l

l b.

Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2.

.[

CALVERT CLIFFS - UNIT 1 3/4 4-27 l

b.

n REACTOR COOLANT SYSTEM i

SURVEILLANCE' REQUIREMENTS (Continued)=

In addition to the requirements of Specification 4.0.5, each Reactor Ceolant Pump flywheel shall be inspected per the recommendations. of_

Regalatory Position C.4.b of. Regulatory Guide 1.14 Revision 1, August -

1975.

f 4.4.10.1.2 Augmented :nservice Inspection Program'for Main Steam and i4ain Feedwater Piping - The unencapsulated welds greater than-4 inches in nominal diameter in the main steam and main feedwater piping runs located outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice in-spection program using the applicable rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code,Section XI,1974 Edition and Addenda through Summer 1975, for Class l

[

2 components.

a.

System integrity and baseline data shall be established by performing a 100% volumetric examination of each weld prior to exceeding 18 months of operation.

b.

Each weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined, cumulatively, during each 10 year in-spection interval. *The welds to be examined'during.each inspection period shall be. selected to provide a repre -

sentative sample of the conditions of the welds.

If-these examinations reveal unacceptable structural defects in one or more welds, an additional 1/3 of.the welds shall be examined and the inspection schedule for the repaired welds shall revert back to the first 10 year inspection program.

If additional unacceptable defects'are detected in the second sampling, the remainder of the welds shall also'be inspected.

i l

l I

l f

l cal. VERT CLIFFS - UNIT 1 3/4 4-28 Amendment No. 75 i

l-

i 4 -

. EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued).

4 e.

At least once per 18 months by:

~~

~

1.

Verifying ' automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300.

psia.

i 2.

A visual inspection.of the containment sump and verifying

i that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

3.

Verifying that a minimum total of 100 cubic feet of l

solid granular trisodium phosphate dodecahydrate (TSP) i.

is contained within the TSP storage baskets, t

4.

Verifying that when a representative sample of 4.0 + 0.1 grams of TSP from a TSP storage. bas.ket is subme,rged-yithout 0

agitation'.lin 3.5'* 071' liters'of'777 10 F borated. water from the RWT, the pH of the mixed solution is. raised to 4

[

> 6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i t.

At least once per 18 months, during shutdown, by:

2 1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation test signal.

j..

2.

Verifying that each of the following pumps start auto-matically upon receipt of a Safety Injection Actuation Test Signal:

a.

High-Pressure Safety Injection pump.

b.

Low-Pressure Safety Injection pump.

g.

By verifying the correct position of each electrical position i

stop for the following Emergency Core Cooling System throttle valves:

1.

During each performance of valve cycling required by Specification 4.0.5 by observation of valve position on the control boards.

l CALVERT CLIFFS - UNIT 1 3/4 5-5 AmendmentNo.#,48

..-,-.-.,..a,

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

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE. REQUIREMENTS (Continued) 2.-

- Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of maintenance ~c7i the valve or its operator by measurement of stem travel when the ECCS subsystems are required to be OPERABLE.

HPSI SYSTEM Valve Number Valve Number MOV-616 MOV-617 MOV-626 MOV-627 MOV-636 MOV-637-MOV-646 MOV-647 l

l h.

By perfonning a flow balance test during shutdown following i

completion of HPSI system modifications that alter system i

flow characteristics and verifying the following flow rates:

t HPSI System Single Pump 170 + 5 gpm to each injection leg.

1 CALVERT CLIFFS - UNIT 1 3/4 5-5a Amendment No.M. 75 D

-.p--..

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAlhMEt!T CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ..

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that:

1.

All penetrati6ns* not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated e.utomatic valves secured in

-their positions, except'as provided in Table 3.6-1 of Specification 3.6.4.1, and 2.

All equipment hatches are closed and sealed.

b.

By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

Except valves, blind flanges, and deactivated automatic valves which are located inside the c'ontainment and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTCOWN except that such verification need not be performed more i

I often than once per 92 days.

b a

CALVERT CLIFFS - UNIT 1 3/4 6-1

-e--

CONTAINMENT SYSTEMS CONTAINM'ENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 ' Containment leakage rates shall be limited to:'

~

a '.

An overall integrated leakage rate of:

I 1 L,pe(346,000 SCCM), 0.20 percent by weight of the containment 1.

air r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,, ~50 psig, or 2.

<Lt (61,600 SCCM), 0.058 percent by weight of the containment l

air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of P. 25 psig.

t b.

A combined leakage rate of 5 0.60 L (207,600 SCCM), for all penetra-l tions and valves subject to Type D $nd C te:;ts when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L SCCM) or 0.75 L (46,200 SCCM), as applicable, l

or (b) with the Mea (259,500sured combined leakage fate for all penetrations and va subject to Types B and C tests exceeding 0.60 L, restore the leakage rate (s).

to within the limit (s) prior to increasing the keactor 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 specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI-N45.4 -

1972:-

a.

Three Type A tests (Overall Integrated Containment Leakage Rate) shall.

be conducted at 40 + 10 month intervals during shutdown at either P a (50 psig) or at P 25 psig) during each 10-year service period.

t The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.

CALVERT CLIFFS - UNIT 1 3/4 6-2 Amendment No. 75

CONTAINMENT SYSTEMS SURVEILLANCE' REQUIREMENTS (Continued) b.

If any periodic Type.A test fails to meet either.75 L (259,500SCCM) t (46,200 SCCM), the test schedule for subsequlnt Type A tests or.75 L shall be reviewed and approved by the Commission.

If.two consecutive (46,200 SCCM), a Type A test shall be her(formed at least every Type A tests fail to meet = either.75 L 259,500 SCCM)or~.75L tl8

- months until two consecutive Type A tests meet either.75 L (259,500 SCCM) or.75 Lt (46,200 SCCM) at which time the above test $chedule may be resumed

- c.

The accuracy of each Type A test shall be verified by a supplemental test which:

1.

Confirms the accuracy of the Type A. test by verifying that the difference between supplemental and Type A test data is within 0.25 L, (86,500 SCCM) or 0.25 Lt (15,400 SCCM).

l 2.

Has a duration sufficient to establish accurately the change in leakage between the Type A test and the supplemental test.

3.

Requires the quantity of gas 1.njected into the containment or bled from the containment during the supplemental test to' be equivalent to at least 25 percent of the total measured leakage rate at P, (50 psig) or Pt (25 psig).

d.

Type B and C tests shall be conducted with gas at P (50 psig) at intervals no greater than 24 months except for test $ involving aip' locks.

e.

Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f.

All test leakage rates shall be calculated using observed data' converted to absolute values.

Error analyses shall be performed to select a balanced integrated leakage measurement system.

4 4

CALVERT CLIFFS - UNIT 1 3/4 6-3 Amendment No. 75 i

a o

--v,--nn-,

-,_,n.,

,n...,

.._,,__,,n,

-s,_. -

-.n.--,

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

s.

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION

~

=-

3.6.1.3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of < 0.05 L, (17,300 SCCM) at P,,

l 50 psig.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With an air lock inoperable, except as a result of an inoperable door gasket, restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With an air lock inoperable due to an inoperable door gasket:

1.

Maintain the remaining door of the affected air lock closed and sealed, and 2.

Restore the air lock to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS i

4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.* After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verifying that the seal leakage is < 0.0002 L (69.2SCCM)asdetermined by precision flow measurement when the volume between the door seals is pressurized to a constant pressure of 15 psig, 4

Exemption to Appendix "J" of 10 CFR 50.

CALVERT CLIFFS - UNIT 1 3/4 6-4 Amendment No.75 6

,.,-,-..---.n,,

,,s,

o.

CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4.1 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the isolation valve (s) specified in Table 3.6-1 inoperable, either:

a.

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or c.

Isolate the affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at

(

least one closed manual valve or blind flange; or

~

d.

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

\\

l l

SURVEILLANCE REQUIREMENTS l

4.6.4.1.1 The isolation valves specified in Table 3.6-1 shall be demon-

.strated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time.

CALVERT CLIFFS - UNIT 1 3/4 6-17

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.4.1.2 Each isolatior.. valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD THUTDOWN or REFUELING MODE at least once per 18 months by:

Verifying that un each containment isolation Channel A or bhannel B

'~

a.

test signal, each required isolation valve actuates to its isolation

. position.

b.

Verifying that on each Containment Radiation-High Test Channel A or. Channel B test-signal, both required containment purge valves actuate to their. isolation position.

c.

Verifying that on each Safety Injection Actuation Channel A or Channel B test signal, each required isolation valve actuates to its isolation' position.

4. 6.'4.1. 3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be determined to be within.its limit when tested pursuant to Technical Specification 4.0.5.

1-

.4.6.4.1.4 Containment purge isolation valves shall be demonstrated OPERABLE at least once every 6 months by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Technical Specification 4.6.1.2.d for all other Type B or C penetrations, the combined leakaSe rate is 7 ess than or' equal to 0.60 L (207,600SCCM). The. leakage rate for the

-]

containment purge isolation val,es,shall also b.e compared to-the previously -

v l

measured leakage rate to' detect excessive valve degradation.

CALVERT CLIFFS - UNIT ?

3/4 6-18 Amendment No, gr$. 7F 1

AD_MINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The facility shall be placed in at least HOT STANDBY within one hour.

b.

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

The Manager - Nuclear Power Department and the OSSRC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the POSRC.

This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the OSSRC and the Manager - Nuclear Power Department, within 14 days of the violation.

6.8 PRGCEDURES 6.8.1 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 1.50, Revision 2, February 1978.

l b.

Refueling. operations.

c.

Surveillance and test activities of safety related equipment.

d.

Securit3 Plan implementation.

e.

Emargency Plan implementation, f.

Fire Protection Program implementation.

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes l

thereto, shall be reviewed by the POSRC and approved by the Plant Superin-tendent. prior to implementation and reviewed periodically as set forth in administrative procedures.

CALVERT CLIFFS - UNIT 1 6-13 Amendment No. 25, J3, 75 l

w-2

o ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-vided:

The intent of the originial procedure is ' dot altse[

a.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

The change is documented, reviewed by the POSRC and approved by' c.

the Plant Superintendent within 14 days of implementation.

6.9 REPORTING REOUIREMENTS 1

ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements o# Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

~

STARTUP REPORT, 6.9.1.1 A sumary report of plant startup and power escalation testing-shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thennal, or hydraulic perfor-mance of the plant.

6.9.1.2 The startuo report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions, or characteristics obtained during.the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

l CALVERT CLIFFS - UNIT 1 6-14 AmendmentNo.[f,43

j#"%4 UNITED STATES c'

'n NUCLEAR REGULATORY COMMISSION

... j WASHINGTON, D. C. 20555 BALTIM0RE GAS AND ELECTRIC COMPANY

_ DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE 1

Amendment No. 56 License No. DPR-69 1.

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

'A. The application for amendment by Baltimore Gas & Electric Company (the licensee) dated August 2, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

~

C.

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

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

j of the public; and E.

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

l l

DESIGN D ORIGINAL CortifiedBy[

y NN-bCLA-

/

l l

. 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:

2.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56, 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.

FOR THE NUCLEAR REGULATORY COMMISSION 1

Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 20, 1982 i

i l

n

4 1

ATTACHMENT TO LICENSE AMENDMENT NO. 56 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET N0. 50-318 Replace the follwing pages of the Appendix A Technical Specifications with the enclosed pages as indicated.

The revised pages are identified'by.

Amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Pages 3/4 4-4

-3/4429

,3/4 5-Sa

' 3/4 6-2 3/4 6-3 3/4 6-4 3/4 6-18 6-13 i.

t 4

9 4

-,a

,g,

6 s

REACTOR COOLANT SYSTEM SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 The following pressurizer code safety valves shall be 0PERABLE:

Valve Lift Settings (11%)

RC-200 2500 psia RC-201 2565 psia APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status witain 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.4.2.2 At least one of the abcVe pressurizer code safety valves shall be OPERABLE:*

APPLICABILITY:

MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.

SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

Both valves may be removed in MODE 5 provided at least one valve is replaced by a spool piece which allows the pressurizer to relieve directly to the quench tank.

3/4 4-3 CALVERT CLIFFS - UNIT 2 Amendment No. 75, 36

REACTOR COOLANT SYSTEM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3 Two power operated relief valves (PORVs) and their associafid block I

valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With one or more PORY(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated. block valve (s) and remove power. from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDEY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , c,

With one or more block valve (s) closed and power removed from the block valve (s) to satisfy a. or b. above, the provisions of.Specifica-tion 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.3.1 Each PORY shall be demonstrated OPERABLE:

a.

At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, in accordance with Table 4.3-1, Item 4.

b.

At least once per 18 months by performance of a CHANNEL. CALIBRATION.

4.4'.3.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one completescycle of full travel.

3' '

This denonstration is not required if a PORV block valve is closed and power renoved to meet Specification 3.4.3 a. or b.

CALVERT CLIFFS - UNIT 2

-3/4 4-4 Amendment No. 36, 56' v.v

- =

w w,

vw

-- - - ------w--

-rw---+

-e-w 4-re-e------:-

w4

~

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

H In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of-Regulatory Position C.4.b of Regulatury Guide 1.14, Revision 1, August 1975.

4.4.10.1.2 Augmented Inservice Inspection Program for Main Steam and Main ceedwater Piping - The unencapsulated welds greater than 4 inches in nominal diameter in the main steam and main feedwater piping runs located outside the containment and traversing safety related areas or located in compartments adjoining safety related areas shall be inspected per the following augmented inservice in-spection program using the applicable rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code,Section XI,1974 Edition and Addenda through Summer 1975, for Class l

2 components.

a.

System integrity and baseline data shall be established by performing a 100% volumetric examination of each weld prior to exceeding 18 months of operation.

4 b.

Each weld shall be examined in accordance with the above ASME Code requirements, except that 100% of the welds shall be examined,, cumulatively, during each 10 year in-spection interval. The welds to be examined during each inspection period shall be selected to provide a repre-sentative sample of the conditions of the welds.

If these examinations reveal unacceptable structural defects in one-or more welds, an additional 1/3 of the welds shall be i

examined and the inspection schedule for the repaired welds shall revert back to the first 10 year inspection program.

If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall also be inspected.

i I

b t.

l CALVERT CLIFFS - UNIT 2 3/4 4-29 Amendment No. 56 t

e

~

6 REACTOR COOLANT SYSTEM CORE BARREL MOVEMENT LIMITING CONDITION _ FOR OPERATION 3.4.11 Core barrel movement shall be limited to less than the Amplitude Probability Distribution (APD) and Spectral Analysis (SA) Alert Levels for the applicable THERMAL POWER level.

APPLICABILITY: MODE 1.

ACTION:

a.

With the APD and/or SA exceeding their applicable Alert Levels, POWER OPERATION may proceed provided the following actions are taken:

1.

APD shall be measured and processed at 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 applicable Alert Level, shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within 30 days of detection.

l b.

With the APD and/or SA exceeding their applicable Action Levels, measure ar.d 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 .

The provisions of Spe'cifications 3.0.3 and 3.0.4 are not'

~

c.

applicable.

l CALVERT CLIFFS - UNIT 2 3/4 4-30 Amendment No. 39

d. -

EMERGENCYCORECOOLINGSYSiEMS SURVEILLANCE REQUIREMENTS (Continued) i i

e.

' At least once per 18 months by:

1.

Verifying automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psia.

2.

A visual inspection of.the. containment sump and verifying' that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

1 3.

Verifying that a minimum total of 100 cuoic feet of l

solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4.

Verifying that when a representative sample of 4.0 + 0.1 grams of TSP from a TSP storage basket is submergedT without l

agitation, in 3.510.1 liters'of'77' 10 F borated water from the RWT, the pH of the mixed solution is raised to

> 6 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f.

At least once per 15 months, during shutdown,L by:

~

1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation test signal.

2.

Verifying that each of the following pumps start auto-matically upon receipt of a Safety Injection Actuation Test Signal:

a.

High-Pressure Safety Injection pump.

i b.

Low-Pressure Safety Injection pump.

g.

By verifying the correct position.of. each electrical position stop for the following Emergency Core Cooling System throttle valves:

I 1.

During each performance of valve cycling required by Specification 4.0.5 by observation of valve position on the control boards.

CALVERT CLIFFS ~- UNIT 2 3/4 5-5 Amendment No. 76,31 g

e---,

-e.-.

.-..~,,

..r

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of maintenancW oW~ the valve or its operator by measurement of stem travel when the ECCS subsystems are required to be OPERABLE.

HPSI SYSTEM-Valve Number Valve Number _

MOV-616 MOV-617 i

MOV-626 MOV-627 MOV-636 MOV-637 f

MOV-646 MOV-647

!I h.

By performing a flow balance test during shutdown folle 3'g c

completion of HPSI system modifications that alter system flow characteristics and verifying the following flow rates:

HPSI System Single Pump i

170 + 5 gpm to each injection leg.

CALVERT CLIFFS - UNIT 2 3/4 5-Sa Amendment No. JA, 56 4

e

o-3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

~

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

I Without primary. CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY I

within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following'30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

l a.

At least once per 31 days by verifying that:

1.

All penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, l

blind flanges, or deactivated automatic valves secured in l

their positions, except as provided in Table 3.6-1 of l

Specification 3.6.4.1, and 2.

All equipment hatches are closed and sealed.

b.

By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.

j

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise l

l secured in the closed position.

These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need i

I not be performed more often than once per 92 days.

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

1

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of:

1.

<La (346,000 SCCM), 0.20 percent by weight of the containment I

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

50 psig, or 2.

<Lt (44,600 SCCM), 0.042 percent by weight of the containment I

air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of P. 25 psig.

t b.

A combined leakage rate of 10.60 L (207,600 SCCM) for all penetra-l tions and valves subject to Type B $nd C tests when pressurized to P,.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate e::ceeding 0.75 L SCCM), or 0.75 L (33,400 SCCM), as applicable, l

c,r (b) with the Mea (259,500sured combined leakage rkte for all penetrations and valves subject to Types B and C tests exceeding'0.60 L, Reactor Coolant System

, restore the leakage rate (s) to within the limit (s) prior to increasing the temperature above 200*F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the followino test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 - 1972:

a.

Three Type.A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at

~

either P,The third test of bac(h set shall be conducted during the25 psig)

(50 psig) or at P period.

shutdown for the 10-year plant inservice inspection.

CALVERT CLIFFS - UNIT.2 3/4 6-2 Amendment No. 56 I

I

CONTAlfiMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

If any periodic Type A test fails to meet either.75 L (259,500SCCM) or.75 Lt (33,400 SCCM), the test schedule for subsequ$nt_ Type A tests shall be reviewed and approved by the Commissio~n.

If two consecutive (33,400 SCCH),-a Type A test shall be $er(259,500 formed at leasti eve Type A tests fail to meet either.75 L SCCM), or.75 L months until two consecutive Type A tests meet either.75 L (259,500 SCCM) or.75 Lt (33,400 SCCM) at which time the above test $chedule may be resumed.

l c.

The accuracy of each Type A test shall be verified by a supplemental test which:

1.

Confirms the accuracy of the Type A test by verifying that the I

difference between supplemental and Type A test data is within O.25 L7(86,500 SCCM) or 0.25 Lt (11,100 SCCM).

l 2.

Has a duration sufficient to establish accurately the change in leakage between the Type A test and the supplemental test.

i 3.

Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at P, (50 psig) or Pt (25 psig).

d.

Type B and C tests shall be conducted with gas at P (50 psig) at.

intervals no greater than 24 months except for test $ involving air j

locks.

e.

Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f.

All test leakage rates shall be c.alculated using observed data converted to absolute values.

Error analyses shall be performed to select a balanced integrated leakage measurement system.

l 1

t CALVERT CLIFFS - UNIT 2 3/4 6-3 Amendment No. 56

=_.

c.

CONTAINMENT SYSTE' _

CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of < 0.05 L, (17,300 SCCM), at P,,

l.

50 psig.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a.

With an air lock inoperable, except as a result of an inoperable door gasket, restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following~30 hours.

b.

With an air lock inoperable due to an inoperable door gasket:

1.

Maintain the remaining door of the affected air lock close'd and sealed, and 2.

Restore the air lock to OPERABLE status within 7 days or be.

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

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.* After each opening, except when the airlock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verifying that the seal leakage is < 0.0002 L (69.2 SCCM) as determined l

by precision flow measurement when the volume between the door seals is pressurized to a constant pressure of 15 psig,

Exenption to Appendix "J" of 10 CFR 50.

CALVERT CLIFFS - UNIT 2 3/4 6-4 Amendment No. 56

CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4.1 The containment isolation valves specified-in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.

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

ACTION:

l With one or more of the isolation valve (s) specified in Table 3.6-1 i

inoperable, either-I Restore the inoperable valve (s) to OPERABLE status within 4 a.

hours, or l

b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at f

least one deactivated automatic valve secured in the isolation

+

position, or c.

' Isolate the affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange; or d.

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

,l, i

SURVEILLANCE REQUIREMENTS I

4.6.4.1.1 The isolation valves specified in Table 3.6-1 shall be demon-strated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time.

CALVERT CLIFFS - UNIT 2 3/4 6-17 n

a

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENT 3 (Continued) 4.6.4.1.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated; OPERABLE during the COLD SHUTDOWN or REFUELING MODE at 1.. east once per 18 months by:

Verifying that on each containment isolation Channel A or' Channel B

- a.

test signal, each required isolation valve actuates to its isolation position.

b.

-Verifying that on each Containment Radiation-High Test Channel A or Channel B test signal, both required containment purge valves actuate to their isolation position._

c. ' Verifying that on each Safety Injection Actuation Channel A or Channel B test signal, each re' uired isolation valve actuates to q

its isolation position.

4. 6. 4.1. 3 The isolation time of each power operated or automatic valve of i.

Table 3.6-1 shall be determined to be within its limit when tested pursuant to Technical Specification 4.0.5.

4.6.4.1.4 Containment purge isolation valves shall be demonstrated OPERABLE at-least once every 6 months by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Technical Specification 4.6.1.2.d for all other Type B or C penetrations, the combined leakage rate.

is less than or equal' to 0.60rl, (207,600 SCCM)".' The leakage rate for the.

containment purge. isolation valves shall also be compared to'the previously measured leakage rate to detect excessive valve degradation.

~

t CALVERT CLIFFS.- UNIT 2 3/4 6-18 Amendment No. //7, 56 w

v y,,---

e-

...m

C*

ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-

~

vided:

a.

The intent of the origintal procedure is not alte d.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected, The change is documented, reviewed by the POSRC and aooroved by c.

the Plant Superintendent within 14 days of implementation.

l 6.9 RE%.flNG REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6. 9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unicss otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary repo'rt of plant startup and power escalation testin shall be submitted following (1) receipt of an operating license, (2) g.

i amendment to the license involving a planned increase in power level, i

(3) installation of fuel that has a different design or has been manu-l factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-i mance of the plant.

6.9.1.2_ The s'.artup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with de',ign predictions and speci fications. Any corrective actions that were required to obtain satisfactory operation shall also be describeJ. Any additional specific details required in license conditions based on other commitments shall be included in this report.

AmendmentNo./[26 CALVERT C'LIFFS - UNIT 2 6-14 I

ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The facility shall be placed in at least HOT STANDBY within one hour.

b.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Manager - Nuclear Power Department and the OSSRC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the POSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recu rrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the OSSRC and the Manager - Nuclear Power Department within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

The applicable procedures recommended ~in Appendix "A: of

~

Regulatory Guide 1.33, Revision 2, February 1978.

l b.

Refueling operations.

c.

Surveillance and test act'vities of safety related equipment.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

1:

6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the POSRC and approved by the Plant Superin-tendent prior to implementation and reviewed periodically as set-forth in administrative procedures.

CALVERT CLIFFS - UNIT 2 6-13 Amendment No. 77, 26, 56

.

ML20071N291

Text

Attachment:

Attachment:

Navigation menu

Search