Text

Amend 66 to License DPR-65,changing Tech Specs to Incorporate Certain Lessons Learned Category a Requirements Re TMI Accident
	ML20126H568
Person / Time
Site:	Millstone
Issue date:	04/07/1981
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20126H571	List: ML20126H568; ML20126H575; ML20126H580;
References
	NUDOCS 8104160060
	Download: ML20126H568 (29)
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UNITED STATES

(_/

,-fj NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

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NORTHEAST NUCLEAR ENERGY COMPANY THE C0!PIECTICUT LIC! T AND POUER COMPANY THE HARTFORD ELECTRIC LICHT COMPANY THE WESTERN t1ASSACHUSETTS ELECTRIC COMPANY DOCKET No. 50-336 MILL' STONE $!UCLEAR POWER STATION, UNIT NO. 2 Af1ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-65 1.

The Muclear Regulatory Commission (the Comrission) has found that:

A.

The application for amendment by Northeast Nuclear Energy Company (the licensee) dated December 27, 1976, 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; 3.

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

. There is reasonable assurance (i) th t the activities authorized by this anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Conmission's regulations; J.

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 amen nent is in accordance stith 10 CFR Part 51 of the Connission's reculations and all applicable requirements nave been satisfied.

8104160 OCO

b 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

!!o. OPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 66, are t,areby 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 TiiE NUCL;AR REGULATORY COMMISSION 5, L.

~~

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

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 7, 1981

ATTACHMENT TO LICENSE AMENDMENT NO. 66 FACILITY OPERATING LICEllSE NO. DPR-65 DOCKET fl0. 50-33,6 Re: lace the following pages of the Appendix A Technical Specifications with the e losed pages. The revised pages are identified by Amendment number and contain vertical lines indicated the area of change. The corresponding overleaf pages a-e also provided to maintain document completeness.

Pages IV X

XVII 3/4 3-46 3/4 3-47 3/4 3-48 3/4 3-49 3/4 4-2 3/4 4 3 3/44-4 83/43-4 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2a 6-1 64 6-24 6-25 m

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY....................

...................... 3/4-0-1 3/4.1 REACTIVITY CONTROL SYSTEF5 3/4.1.1 B0 RATION CONTR0L.....................................

3/4 1-1 Shutdown Margin - T

> 200 F.......................

3/4 1-1 avg Shutdown Margin - T

< 200 F.......................

3/4 1-3 avg -

Boron Dilution......................................

3/4 1-4

~

Mode rator Temperhture Coef 'icir. - (MTC ).............

3/4 1-5 Min imum Temperature fo r Criticality.................

3/4 1-7 3/4.1.2 BORATION SYSTEMS.....................................

3/4 1-8 Fl ow P a t h s - S h u t d ow n................................

3/4 1-8 Fl ow Pa ths - Ope ra ti n g...............................

3/4 1-10 C ha rgi n g Pump - S hu tdown............................. 3/4 1-12 Charging Pumps - 0perating...........................

3/4 1-13 Bo ri c Ac i d Puup s - S hu tdown..........................

3/4 1-14 t

Bori c Acid Pumps - 0perating.........................

3/4 1-15 Ccrated Water Sources - Shutdown.....................

3/4 1-16 Borated Water Sources - Operating...................

3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...........................

3/4 1-20 Full Length CEA Group Position......................

3/4 1-20 Pos i ti on I ndi cator C hannel s..........................

3/4 1-24 3/4 1-26 CEA Drop Time..

Shutdown CEA Insertion Limit.........................

3/4 1-27 Regulating CEA Insertion Limits......................

3/4 1-28 i i

.;...._a......

B

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

INDEX l _IMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS f 'ECTION PAGE-

,3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE....................................... 3/4 2-1 i

T 3/".2.2 TOTAL PLANAR RADIAL PEANING FACTOR - F 3/4 2-6 xy.

T 3/4.2.3 TOTAL INTEGRATED RADI AL PEAKING FACTOR - F 3/4 2-9 r

I

' 3/4.2.4 AZIMUTHAL POWER TILT..................................

3/4 2-10 3/4. 2. 5' FUEL RESI DENC E TIME..................................

3/4 2-12 3/?.2.6 DN B MA RG I N............................................

3/4 2-13 3/4.3 INSTRUMENTATION 3/ 4. 3.1 -

REACTOR PROTECTI VE INSTRUMENTATION....................

3/4 3-1 13/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM I

INSTRUMENTATION.....................................

3/ 4 3-10 3/4. 3. 3 MONITORING INSTRUMENTATION...........................

3/4 3-26 l

Radiation Monitoring

...........'....:...................-3/4 3-26 Incore Detectors.....................................

3/4 3-30 Seismic Instrumentation..............................

3/4 3-32 i

Meteorological Instrumentation

..'m..................... 3/4 3-36 Remote Shutdown Instrumentation.......................

3/4 3-39 Chl ori ne Detection Systems.............................

3/4 3-42 Fire Detection Instrumentation

.........................'3/4 3'43 a

l Accident Monitoring...................................

3/4 3-46 l

i i,

1/4.4 REALTOR COOLANT SYSTEM

I/4.4.1 R E ACT O R COOL ANT LOSP S..................................

3 / 4 4-1 3/4.4.2 SAFETY VALVES.......................................... 3/4 4 l v4.4.3 R E t I E r v A t y t S....... -................................

3 / 4 4 1 I

9 "I.' STONE - UNIT 2 IV Amend.T.ent No. 25, 2E,66 h

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i INDEX

-E'SES IECTICN PAGE

< 2'40 APPLICABILITY..........................................

B 3/4 0-1 2'4.1 REACTIVITY' CONTROL SYSTEMS 2'4.1.1 B0 RATION' CONTROL.....................................

B 3/4 1-1 2 4.1.2 BORATION SYSTEMS.....................................

B 3/4 1-2 2 4.1.3 MOVABLE CONTROL ASSEMBLIES...........................

B 3/4 1-3.

~ '4. 2 POWER DISTRIBUTION LIMITS

'4.2.1 LINEAR HEAT RATE.....................................

B 3/4 2-1 T

~ 4.2.2 TOTAL PLANAR RADIAL PEAKING FACTOR - F B 3/4-2-1 Xy.............

T

~ 4.2.3 TOTAL INTLGRATED RADIAL PEAKING FACTOR - F........... B 3/4 2-1 r

2 4.2.4 AZIMUTHAL POWER

1.T..................'...............

B 3/4 2-1

' 4.2.5 FUEL RESIDENCE TIME..................................

B 3/4 2-2

4.2.6 DNB MARGIN....................

... q.................

B 3/4 2-2

~L.3

'iSTRUMENTATION

4.3.1 PROTECTIVE INSTRUNENTATION...........................

B 3/4 3-1 i

L...!

E'!GILEERED SAFETY FEATURE INSTRUMEliTATION............

B 3/4 3-1

0"ITO.] flG - I!'STF.T.ENTATION..

B 3/4 3-2

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INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 R EACTOR COOLANT LOOPS................................... B 3/4 4-1 3/4.4.2 SAFETY VALVES........................................... B 3/4 4-1 l

3/4.4.3 RELIEF VALVES........................................... B 3/4 4-2 l

3/4.4.4 PRESSURIZEP............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS....................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.................,...... B 3/4 4-3 3/4.4.7 CHEMISTRY............................................... B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY....................................... B'3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................. B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY.................................... B 3/4 4-11 3/4.4.11 CORE BARREL MOVEMENT.................................... B 3/4 4-l'2 3/4.5 EMERGENCY C0_RE #^" MING SYSTEMS (ECCS) 3/4.5.1 S AF ET Y I hJ ECTI ON TAN KS.................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBS YSTEMS.............................. B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK (RWST)..................... B 3/4 5-2 MILLSTONE - UNIT 2 X

Amendment No. 66 l

INDEX ADl11NISTRATIVE CONTROLS SECTION

,P..A..G_E_

6.9 REPORTING REQUIREMENTS 6.9.1 ROUTIllE REPORTS AND REPORTABLE OCCURRENCES..............

6-17 6.9.2 SPECIAL REPORTS......................................... 6-21 6.10 RECORD RETENTION.............................................

6-22 6.11 RADIATION PROTECTION PROGRAM......

6-23 6.12 HIGH RADIATION AREA..........................................

6-23 6.13 ENVIRONMENTAL QUALIFICATION 6-24 6.14 SYSTEMS INTEGRITY............................................

6-24 6.15 IODI NE MONI TORING.........

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o FIRE DETECTION INSTRUMENTS z

Heat Smoke

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=.

i Minimum Minimum Total No.

Channels Total No.

Channels N

Instrument location (Zone) of Channels Operable of Channels Operable 5.

Battery Rooms West Battery Room (14'6") (39) 1 1

2 1

East Battery Romv. (14'6") (39) t'.

6.

Electrical Penetration Rooms T

3 2

East (14'6") (20) 2 1

West (14'6") (17) 7.

Diesel Generators t

?

1 1

Diesel 1221 (30)-

1 1

F Diesel 1321 (32) ru 8.

Main Exhaust Equipnent Room

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1 o"

Room (El 38'6") (5)

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JUSTRUMENTATION ACCIDENT MONITORING LIf11 TING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3.11 shall be OPERABLE.

i APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

Actions per Table 3.3-11.

b.

The provisions of Specification 3.0.4 are not applicable.

I

'5URVEILLANCE REQUIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

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I MILLST]1E - UNIT 2 3/4 3-46 Amendment No. 66

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C TABLE 3.3-11 ACCIDENT !10 NIT 0_ RING INSTRU?1EflTATION t11NI:40t4 TOTAL NO.

CHAN!!ELS E

INSTRU:1EilT OF CHANNELS OPERABLE ACTION 1

Z ro 1.

Pressurizer Water Level 2

1 1

2.

Auxiliary Feeduater Flow Rate 1/S. G.

1/S. G.

l 3.

RCS Subcooling flargin Monitor 1

1 2

4.

PORY Position Indicator l

Acoustic Flow Monitor 1/ valve 1/ valve 3

i 1

5.

PORV Block Valve Position i

Indicator 1/ valve 1/ valve 3

u 6.

Safety Valve Position Indicator i

Acoustic Flow Manitor 1/ valve 1/ valve 3

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E 1

i a

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i TABLE 3.3-11 (Continued)

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE. channels less than required by Table 3.3-11, either restore the inoperable channel (s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 - With the subcooling margin monitor INOPERABLE, determir,a the subcooling margin once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 - With any individual valve position indicator inoperable, ebtain. quench tank temperature, level and pressure infor-mation, and monitor discharge pipe temperature once per shift to determine valve position.

'<ILLSTONE - UNIT 2 3/4 3-48 Arendment No. 65

E5 TABLE 4.3-7 l-El ACCIDENT f10NITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION E

Z 1.

Pressurizer Water Level M

R ro 2.

Auxiliary Feedwater Flow Rate il R

3.

Reactor Coolant System Subcooling Margin f1onitor R

4.

PORV Position Indicator M

R 5.

PORV Block Valve Position Indicator M

R 4{

6.

Safety Valve Position Indicator M

R Y

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'4.4 REACTOR COOLANT SYSTEM

IA; TOR COOLANT LOOPS L:P.:T:N3 CONDITION FOR OPERATION 2.4.1 Tour reactor coolant pumps shall be in operation.

PLICABILITY

As noted below, but excludit.g MODE 6,*

l A:T:0:

V:' DES 1 and 2:

' 'th less than four reactor coolant pumps in operation, be in H0T S A',D5Y within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

V"; DES 3, 4** and 5**:

C:eration may proceed provided at least one reactor coolant loop is in

eration with an associated reactor coolant pump or shutdown cooling

m

.= The provisions of Specifications 3.0.3 and 3.0.4 are not 5: p '. i ca bl e.

S.T/E:LLANCE REQUIREMENTS

.4.1 The Flow Dependent Sel:.ctor Switch stiall be determined to be in

e 4 pump position within 15 minutes prior t'o making the reactor critical 5-d at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

-Sea Special Test Exception 3.10.4.

-A reactor coolant pump shall not be started with one or more of the R:3 ccic leg temperatures < 275 F unless 1) the pressurizer water v:lume is less than 600 cuEic feet or 2) the secondary water

emperature of each steam generator is less than 43 F (31 F when reasured by a surface contact instrument) above the coolant t+m;trature in the reactar vessel-

1 reactor coolant pumps and shutdown cooling pumps may be

=

ae-ene gized for up to I hour, provided no operations are permitted

-ich :ould cause dilution of the reactor coolant system boron c:qcentration.

.. 3 7,E - UhiT 2 3/4 4-1 Amendment No, 50

REACTOR COOLANT SY5 TEM SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 PSIA 1 %.

APPLICABILITY: MODES 4 and 5.

ACTION:

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

3.4.2.2 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIA 1 %.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the

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inoperable valve to OPERABLE status within 15 minutes or be ir HOT. SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2 Each pressurizer code safety valve shall be demonstrated OPERABLE with a lift setting of 2500 PSIA 1%, in accordance with Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code, dated July 1,1974.

4 MILLSTONE - UNIT 2 3/44-2 Amendment No, 66

1.E I CT00 COOLANT SYSTEM

E_IEF VALVES

_:"ITING CONDITION FOR OPERATION

...2 Two power operated relief valves (PORVs) and their associated block ta".vts shall be OPERABLE.

J;L:CABILITY: MODES 1, 2 and 3.

IC I";N :

a.

With one or more PORV(s) inoperable, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 wit.iin 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 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either restore the block valve (s) to OPERABLE status or close the block valve (s) and ', emove 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 folicwing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

_.f:VE I'_'_ANCE REQUIREMENTS

.2.'

Each P3RV shall be demonstrated OPERABLE:

Once per 31 days by performance of a. CHANNEL FUNCTIONAL a.

TEST, exclud; g valve operation,tod s'

5.

Once per 18 cenths by performance of a CHANNEL CALIBRATION.

..:.2 Each c1:cr. valve shall be demonstrated OPERABLE once per 92 days by

. : a r ? ' c. ; the /Elve through one complete cycle of full travel.

._:-~"I - UN:T 2 3/4 "-3 Amend ~ent No. 45 6

n

I iREACTOR COOLANT SYSTEli PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble and with et-least 130 kw of pressurizer heater capacity capable of being supplied by emergency power.

The pressurizer level shall be within 5% of its programmed value.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

A.

Ilith the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters e'*her restore the inoperable emergency power supply within 72 h:

's or '? in at least HDT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT 5HUTD0' within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B.

Ilith the pressurizer otherwise inoperable, be in at least HOT STANDBY with the rear. tor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREliENTS 4.4.4 The pressurizer water level shall be determined to be within 5% of its prograrred value at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

!!ILLSTONE - UNIT 2 3/4 4-4 Amendment No. 66

5 F.Ji'ENTATION i;5 ES

2.3.3.2 liCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum

oralecent of equipment ensures that the measurements obtained from use

f this system accurately represent the spatial neutron flux distribution

f -he reactor core.

\\

I/4.3.3.3 SEISMIC INSTRUMENTATION 7te OPERABILITY of the seismic instrumentation ensures that suffi-

ient capability is available to promptly determine the magnitude of a j.

e'sric event and evaluate the response of those features important to

a#etj.

This capability is required to permit comparison of the measured esocqse to that used in the design basis for the facility.

.2.2.4.

METEOROLOSICAL INSTRUMENTATION i

Tre OPERABILITY of the meteorological instrumentation ensures that

i

1. ier.: neteorological data is available for estimating potential acia: ion doses to the public as a result of routine or accidental

'a' es te of radioactise materials to the athosphere. This capability is s:Ji-ei to evaluate the need for initiating protective measures to

::e:t tne health and safety of the public. This instrumentation is

3rsistent with the reconmiendations of Regulatory Guide 1.23 "0nsite i

e.e: c '. cgical Programs. "

.2.2.5 REMOTE SHUTDOWN INSTRUMENTATION

~re OPERABILITY of the remote shutdown instrumentation ensures' that

'

1. 'en capability is available to permit shutdown and maintenance of 5-1 ?7#: of the facilit" from locations outside of the control room.

ascoility is require _ in the event control room habitability is

. -- is ::msistent witn General Design Criteria 19 of 10 CFR 50.

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1i INSTRUMENTATION BASES 3/4.3.3.6 CHLORINE DETECTION SYSTEMS The operability of the chlorine detection systems ensures that an accidental chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel.

Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically isolate the control room and initiate its operation in the recirculation mode of operation to provide the required protection.

The chlorine detection systems required by this specification are consistent with the recommendations of Regulatory Guide 1.95, ~" Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release",

5 February 1975.

3/4.3.3.7 FIRE DETECTION INSTRUMENTATIL.

OPERABILITY of the fire. detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation.is inoperable, the establishment of frequent fire. patrols in the affected areas is required to provide detection capability until the inoperable instrumenta-tion is restored to OPERABILITY.

3/4.3.3.8 ACCIDENT MONITORP:G INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an 2:cident.

This capability is consistent with the recontendations of NUREG-05 '., "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".

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MILLSTONE - UNIT 2 8 3/4 3-4 Amendment No. 3E, /9, 66

?

3/1.4 REACTOR COOLANT SYSTEM 4

3!SES 3/ 4. 4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above i.30 dut ing all normal operations and anticipated transients.

STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power

_e/el-High, Reactor Coolant Flow-Low, and Thermal Margin / Low Pressure rips have been reduced to their specified values.

Reducing these trip se points ensures that the DNBR will be maintained above 1.30 during

-hree pump operation and that during two pump operation the core void "raction will be limited to ensure parallel channel flow stability within

he core and thereby prevent premature DNB.

A single reactor coolant loop with its steam generator filled above

he low level trip setpoint provides sufficient heat removal capability

'or core cooling while in MODES E and 3; however, single failure consider-l stions require plant cooldown if component repairs and/or corrective

.' ac-ions cannot be made within the allowable out-of-service time.

TherestrictionsonstartingaReactogCoolantPumpduringMODES4 and 5 with one or more RCS cold legs < 275 F are provided to prevent RCS

ressure transients, caused by energy additions from the secondary system,

-hich ceuld exceed the limits of Appendix G to 10 CFR Part 50.

The RCS

..i'1 be protected against overpressure transients and will not exceed the

'i.-its of Appendix G by either (1) restricting the water volume in the

ressurizer and thereby providing a volume for the primary cooltat to Ex;ar.d into or (2) by restricting starting of the RCPs to when the g

e:gedary water temperature of each steam generator is less than 43 F

'3" F when measured by a surface contar.t ins,trument) above the coolant

e perature in the reactor vessel.

l 3/2.L.2 SAFETY VALVES i

The pressurizer code safety valves operate to prevent the RCS from

ef n; pressurized above its Safety Limit of 2750 psia.

Each safety valve

's ds3igned to relieve 296,000 lbs per hour of saturated steam at the valve Ie: point.

The relief capacity of a single safety valve is adequate to e ieve any overpressure condition which could occur during shutdown.

In

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-5s e /ent that no safety valves are OPERABLE, an operating shutdown

]: ling loop, connected to the RCS, provides overpressure relief capability in: will prevent RCS overpressurization.

__I ;NE - UNIT 2 B 3/4 4-1 Amendment No. E0, 66

REACTOR COOLANT SYSTEM BASES During operation, all pressurizer coder saiety valves must be OPERABLE to prevent the RCS from being pressurized above; its safety limit of 2750 psia.

The combined relief capacity of these valves-is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

I 3/4.4.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emer-gency power source to ensure the ability to seal this possible RCS leakage path.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer with the level as programmed ensures l

i that the RCS is not a hydraulically solid system and is capable of accommo-dating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief.

The power operated relief valves function to relieve RCS l

pressure during all design transients.

Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer-Pressure-High signal, minimizes the undesirable opening of the spring loaded pressurizer code safety valves.

The requirement that 130 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of off-site power condition to maintain natural circulation at H0T STANDBY.

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

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

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is MILLSTONE - UNIT 2 B 3/4 4-2 Amendment No. 22, 37, 52, 66 1

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evidence of mechanical damage or progressive degradation due to design,.

manufacturing errors, or inservice conditions that lead to corrosion.

In-service inspection of. steam generator tubing also provides a means of

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characterizing the nature and cause of any. tube degradation.so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary

.ccolant will be maintained within those chemistry limits found to result in.

negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

l The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system l

and the secondary coolant system (primary-to-secondary leakage = 0.5 GPM, per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an. adequate margin of safety to withstand the i

loads imposed during normal operation and by postulated accidents..0perating plants have demonstrated that primary-to-secondary leakage of 0.5 gallon per minute can readily be detected by radiation monitors of steam generator blow-down.

Leakage in excess of this limit will require plant shutdown and an 1

unscheduled inspection, during which the leaking tubes will be' located and pl ugged.

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

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the

- E plugging limit of 40% of the tube nominal wall thickness.

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

ja Whenever the results of any steam generator tubing inservice inspection 1

fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may reruit in a requirement 1or analysis, laboratory examina-tions, tests, ado 1tioral eddy-current inspection, and revision of the Technical Specifications, if necessary.

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MILLST0h'E - UNIT 2 B 3/4 4-2a Amend ent "o. 22, 7J f 2, 66

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5.3 ADMINISTRATIVE CONTROLS 5.1 RESP 0NSIBILITY

, 5.1.1 The Station Superintendent shall be responsible for overall operation

, of the Millstone Station Site while the Unit Superintendent shall be respon-st ole for operation of the unit. The Station Superintendent and Unit Super-irtendent shall each delegate in writing the succession to these responsi-

, oilities during their absence.

5.2 ORGANIZATION f3FFSITE I 5.2.1 The offsite organization for facility management and technical support

. stall be as shown on Figure 6.2-1.

+ A:I'_ITY STAFF 5.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

c.

At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips..

d.

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

e.

All CORE ALTERATIONS after the ini,tial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f.

A site Fire Brigade of at least 3 members shall be maintained onsite at all times.

The Fire Brigade shall not include 2 members of the minimum shift crew necessary fe ' safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.

F;:ILITY STAFF QUALIFICATIONS

'.1.~

Each member of the facility staff sttil meet or exceed the minimum

aiifications of ANSI G18.1-1971 for comparable positions, except for (1) the

-til:, Physics Supervisor who shall meet or exceed the qualifications of

e;f story Guide 1.8, Revir' n 1 and l2) the Shift Technical Advisor who shall

.e a Bachelor's Degree or equivalent in a scientific or engineering

.::' aline with specific training in plant design, and response and analysis

- s :ltnt for tecnsients and accidents.

._'T:NE - UNIT 2 6-1 Amendment No. 23, 25, fE, 66

NORTHEAST NUCLEAR ENERGY COMPANY r-N s2 G;

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- Provides Operating anel Engencering Support by Contractual Arrangement n

if3 Figure 6.21 Offsite Organization for Facility Management and Technical Support 5

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Figure 622 Facility Organization - Millstone Nuclear Power Station - Unit 2.

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e TABLE 6.2 MINIMUM SHIFT CREW COMPOSITION APPLICABLE MODES LICENSE CATEGORY 1, 2, 3 & 4 5&6 SQL 1

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1 Shift Technical Advisor 1

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Does not include the licansed Senior Reactor or. Senior Reactor Operator Limited to Fuel Handling individual supervision CORE ALTERATIONS after the initial fuel loading.

a' Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to accommodate injury or sickness occurring to on duty shift crew members.

MILLST0:!E-U"IT 2 6-4 Amendment I!o. 66

o

't ADMINISTRATIVE CONTROLS

+

g.

Records of training and qualification for current. members ~ of the.

plant staff.

h.

Records of inservice inspections performed pursuant to these-Technical Specifications.

i.

Records of quality assurance activities required by the QA Manual.

j. Records of reviews' performed for changes made to procedures or

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equipment or reviews of tests and experiments pursuant.to 10 CFR. Part 50.59.

l i

k.

Records of meetings of the PORC, the NRB, the 'S0RC ano the SNRB.

1.

Records of Environmental Qualification which are covered under-the. provisions of paragraph 6.13.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation. protection shall be prepared consistent t

with.the. requirements of 10 CFR Part 20 and shall be approved, maintained.

and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of. the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a'high r

radiation area and entrance thereto shall be controlled by. requiring issuance of a Radiation Work Permit *. Any individual.or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area, b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset inte.

grated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate lev'el in.the area has been established and personnel have been made knowl -

edgeable of, them.

Health Physics personnel or personnel escorted by Health Physics per-sonnel shall be exainpt from the RWP issuance requirement during the

. performance'of their assigned radiation protection duties, provided they comply with approved radiation protection procedures.for entry

)

into high radiation areas.

l MILLSTONE - UNIT 2 6-23 Order' dated October 24, 1980 1

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ADMINISTRATIVE CONTROLS D

c.

An individual qualified in radiation protection procedures who is equipped with a; radiation dose rate monitoring device.

This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency spacified in the Radiation Work Permit..The surveillance frequency shall i

be established by the Health Physics Supervisor.

6.12.2 The requirements of 6.12.1, above, shall also apply te each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr'.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Health Physics Supervisor.

6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June-30. 1982 all safety-related electrical equipment-I in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors " Guidelines for Evaluating Environmental Qualifications of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979.

Copies of these documents are attached to Order for Modification of License i

DPR-65 dated October 24, 1980.

l 6.13.2 By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the 00R Guidelines or NUREG-0583.

Thereafter, such records should be updated and maintained current as equipment is replaced, further tested,i or otherwise further qualified.

6.14 SYSTEMS INTEGRITY The licensee shal~. 4mplement a program to reduce leakage from systems outside containment that would 6r could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.

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

MILLSTONE - UNIT 2 6-24 DWBf / Midi OfAlMMNM6 Ah6D Amendment No. 66 l

i

ADMINISTRATIVE CONTROLS 6.15 10DINF MONITORING The. licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1.

Training of personnel,

- i 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

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MILLSTONE - UNIT 2 6-25 Amendment No. 66 v--

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