Text

Amend to License DPR-3,incorporating Changes to Radiation Monitoring Instrumentation,Main Coolant Sys,Eccs,Containment Sys,Plant Sys & Refueling Operations
	ML19281B034
Person / Time
Site:	Yankee Rowe
Issue date:	04/03/1979
From:	Ziemann D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19281B032	List: ML19281B031; ML19281B034; ML19281B057;
References
	NUDOCS 7904200170
	Download: ML19281B034 (37)
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.

h. [.M""%g)t UNITED STATES NUCLEAR REGULATORY COMMISSION

[ ', i e '

2; WASHINGTON, D. C. 20555

\\ ; '...s'!

YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO. 50-29 YANKEE NUCLEAR POWER STATION (YANKEE-ROWE)

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 58 License No. DPR-3 1.

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

A.

The application for amendment by Yankee Atomic Electric Company (the licensee) dated November 24,1978 (Proposed Chang'e No.139, Supplement No. 3), 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 (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's fegulations; D.

The issuance of this amendment will not be inimical to the comnon 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.

. 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-3 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 58, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Paragraph (3), Fire Protection, Facility Operating License No.

DPR-3 is hereby renumbered to be Paragraph (4).

4.

This license amendment is effective as of the date of its' issuance.

FOR THE NUCLEAR REGULATORY COMMISSION j ' f'O

Ajbvws A Dennis L. Ziemann,vChief Operating Reactors Branch #2 Division of Operating Reactors

Attachment:

Changes to.the Technical Specifications Date of Issuance: April 3,1979 b

ATTACHMENT TO LICENSE AMENDMENT NO. 58 FACILITY OPERATING LICENSE NO. DPR-3 DOCKET NO. 50-29 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages include the captioned amendment number and contain vertical lines indicating the areas of change.

Overleaf pages are included for document completeness.

Remove Insert 3/43-17--3/43-18 3/43-17--3/43-18 3/43-19--3/43-20 3/43-19'-3/43-20 3/4 3-21 -- 3/4 3-22 3/43-21--3/43-22 3/4 4-3 -- 3/4 4-4 3/4 4-3 -- 3/4 4-4 3/4 4-9 -- 3/4 4-10 3/4 4-9 -- 3/4 4-10 3/44-11--3/44-12 3/44-11--3/44-12 3/4 5-7 -- 3/4 5-8 3/4 5-7 -- 3/4 5-8 3/4 6-7 -- 3/4 6-8 3/4 6-7 -- 3/4 6-8 3/46-13--3/46-14 3/46-13--3/46-14 3/4 7-1

-- 3/4 7-2 3/4 7-1

-- 3/4 7-2 3/4 7-3 -- 3/4 7-4 3/4 7-3 -- 3/4 7-4 3/4 7-21 -- 3/4 7-22 3/47-21--3/47-22 3/4 7-23 -- 3/4 7-24 3/4 7-23 -- 3/4 7-24 3/4 9-7 -- 3/4 9-8 3/4 9-7 -- 3/4 9-8 83/4 0-1

-- B3/4 0-2 B3/4 0-1

-- B3/4 0-2 B3/4 7-1

-- B3/4 7-2 B3/4 7-1

-- 83/4 7-2 6-5 6-6 6-5 6-6

INSTRUMENTATION 3/4.3.3 MONITORIfGj 'IRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-4 shall be OPERABLE with their alarm setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-4.

ACTION:

a.

With a radiation monitoring channel alann setpoint exceeding the Alann Setpoint shown in Table 3.3-4, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable, b.

With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-4.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the perfonnance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL MODES and at the frequencies shown in Table 4.3-3.

YANKEE-R0WE 3/4 3-17

TABLE 3.3-4 RADIATION N0filTORING IflSTRUMENTATI0fl

$i

~

I.

Mif11 MUM CilANNELS APPLICABLE ALARM MEASUREMEllT RAflGE ACTION Ill'illttint ill, OPERABLE MODES' SETP0lflT 1.

AlH'A f10fil10RS a.

Spent Fuel Pit Area 1 5 mr/hr or 2 x 0.5 - 50 mr 11 - l

1) i~uel Manipulator 1

Gamma Guard-background, which-ever is greater b.

Containment

1) l~uel Manipulator 1

110 mr/hr or 2 x 0 - 100 mr 12 l Ganina Guard background, whica-ever is greater 2.

l'ROCI:SS f10llITORS a.

Containment 6

9 1)

Mai.i Coolant System 1 1,2, 3, & 4 ftA 10 - 10 cpm 13 Leakage Air Par-ticulate Monitor ij

I b.

Radioactive Gaseous Waste Monitor 7

7,

/g

1) Primary Vent Stack Monitor a) Particulate 0

Monitor 1

At all times

$900 cpm greater 10 - 10 14 u.

than background e

TABLE 3.3-4 (Continued)

.c

[dj, MINIMUM y

CHANNELS APPLICABLE ALARM MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION 6

b) Iodine Monitor 1

At all times

< 700 cpm greater 10 - 10 cpm 14 l

c) Noble Gas 6

Monitors 1

At all times

< 3500 cpm greater 10 - 10 cp, j4 l

than background c.

Radioactive Liquid Monitors II) 4

1) Steam Generator I

1, 2, 3 & 4

< 80 cps or 2 x 10 - 10 cpm, or 15 6

Blowdown Monitor Background, which 10 cpm 6

ever is greater t 9 3.

ACCIDENT-EMERGENCY MONITORS a.

liigh Level Radiation Monitor 1

At all times

< 5 R/hr 0.01 - 1000 R/hr 15 l

5 a

W (D

E

=

33

TABLE 3.3-4 (Continued)

TABLE NOTATION When handling irradiated fuel, control rods, or sources.

(1)

Per steam generator in a non-isolated loop.

ACTION STATEMENTS ACTION 11 -

With the number of channels OPERA 3LE less than required l by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.

ACTION 12 -

With the number of channels OPERABLE less than required l by the Minimum Cha.nnels OPERABLE requirement, suspend all operations involving CORE ALTERATIONS.

ACTION 13 -

With the number of channels OPERABLE less than required l by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.5.1.

ACTION 14 -

With the number of channels OPERAELE less than required by the Minimum Channels OPERABLE requirement, suspend all planned releases and releases from the evaporator to the atmosphere through the primary vent stack.

ACTION 15 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, provide an OPERABLE temporary cor.tinuous monitor within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

iAN'.EE-ROWE 3/4 3-20 A enirert >;o. 58

TABLE 4.3-3

[:

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS n

NI CHANNEL MODES IN WHICN E2 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE h{

INSTRU[IJ11[

CHECK CALIBRATION TEST REQUIRED

]

1.

AREA MONITORS a.

Spent Fuel Pit Area (1)

Fuel Manipulator Gamma Guard.

S R

M b.

Containment

}{

(1) Fuel Manipulator Gamma Guard S

R M

2.

PROCESS MONITORS a.

Containment Main Coolant System Leakage Air Particulate Monitor S

R M

1, 2, 3, 4

'Ie b.

Radioactive Gaseous T

Waste Monitors F

N (1) Primary Vent Stack r*

Monitors S

R M

At all times E

c.

Radioactive Liquid Monitors ES (1) Steam Generator Blow-Down Monitor S

R M

1, 2, 3, 4 9

I

TABLE 4.3-3 (Continued)

?:

l,;

RADIATION MONITORING IflSTRUMEtiTATI0fl SURVEILL ANCE REQUIREMENT _S_

l'l CilAtlNEL MODES IN WillCH

u9 CHANNEL CilANNEL FUNCTI0flAL SURVEILLANCE INSTRUMEllT CilECK CALIBRATION TEST REQUIRED 3.

ACCIDEN[-EMERGEllCY M0JIITORS a.

liigh Level Radiation Monitor S

R M

At all times kWhen handling irradiated fuel, control rods or sources.

M.

s t,>

I3 y

'S.

{:

g b

o

MAIN COOLANT SYSTEM 3/4.4.5 MAIN COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4,5.1

...e following Pain Coolant System leakage detection systems shall be OPERABLE:

At least one containment atmosphere particulate radioactivity l

a.

monitoring system, b.

The containment drain tank level monitor'79 system. '

The inccre detection system thimble leak alarn system.

c.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the above required radioactivity monitoring leakage detection a.

system inoperable, operation may continue for up to 7 days provided:

1.

Main Coolant System water inventory balance is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

The other above required leakage detection systems are OPERABLE, and 3.

Appropriate grab samples are obtained and analy:ed at least once per hour:

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 the containment drain tank level monitoring system inoperable, restore the inoperable system to OPERABLE status witnin 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 .

With th'e incore detection system thimble leak a'. arm. system inoper-c.

able, restore the leak alarm system to OPEPJ3LE sta:as within 7 days cr cicse all thimble isolation valves; rest:re ?.e "eak alaen syster to OPERABLE status within 31 days or be in a-leas: F3T STAND 5Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD S~-L D:s within the f ollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

IAuKE ~E0

3/

-3

ge: 3.: ne. 58 h

MAIN COOLANT SYSTEM ISOLATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.3 A main coolant loop shall remain isolated until:

a.

The temperature of the cold leg of the isolated loop is within 30'F of the highest cold leg temperature of the operating loops.

b.

The boron concentration of the isolated loop is not less than the main coolant system boron concentration, and c.

The reactor is subcritical by at least 1 perce ak/k APPLICABILITY: All MODES.

ACTION:

With the requirements of the above specification not satisfied,' suspend startup of the isolated loop.

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The isolated loop cold leg temperature shall be determined to te within 30*F of the highest cold leg temperature of the operating l

loops within 30 minutes prior to opening the cold leg stop valve. '

4.4.1.3.2 The isolated loop boren concentration shall be determined to be not less-than the Main Coolant System boron concentration within a hours prior to opening the cold leg stop ealve.

6.,.1.2.3 The reactor shall be determined to be suberitical by at

[. east 1.:ercent _k/k within 30 minutes prior to opening the cold leg r.:p valve.

i,

..KE E :4'..'E 3/2 --

Aren:rer. '.. 13, 58

MAIN COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.5.1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere particulate monitoring system-perfonnance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment drain tank level monitoring system-performance of CHANNEL FUNCTIONAL TEST at least once per 31 days and CHANNEL CALIBRATION at least once per 18 months.

c.

Incore detection system thimble leak alarm system-performance of CHANNEL FUNCTIONAL TEST at least once per 31 days.

YANKEE-ROWE 3/4 4-9

MAIN CDOLANT SYSTEM ODERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Main Coolant System leakage shall be limited to:

a.

No PRESSURE' BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through all steam generators not isolated from the Main Coolant System, d.

4 GPM IDENTIFIED LEAKAGE from the Main Coolant System, and e.

A maximum of two leaking incore detection system thimbles which are valved off and not plugged, when in MODE 1.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a.

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

With any Main Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 .

With > 6 gpm IDENTIFIED LEAKAGE from the Main Coolant System, c.

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

l I: i

.':EE-PC'. E 3/2 -10 Amendrent No. 58 h

MAIN COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued) d.

With more than two leaking incore detection system thimbles which are valved off but not plugged when in MODE 1, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.5.2 Main Coolant System leakages shall be demonstrated to be within each of the above limits by; a.

Monitoring the containment atmosphere particulate radioactiv-ity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when a Main Coolant System water inventory has not been performed within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then monitor the containment atmosphere particulate radioactivity monitor at least once per hour.

b.

Monitoring the containment drain tank monitoring system at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

Perfomance of a Main Coolant System water inventory balance at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during steady state operation, and d.

Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , YA'iKEE-RCWE 3/4 4-11 Amendment No. 38, 58

MAIN COOLANT SYSTEM CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.6 The Main Coolant System chemistry shall be maintained within the limits specified in Table 3.4-1.

APPLICABILITY: At all time 5.

ACTION:

MODES 1, 2, 3 and 4 a.

With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the Parameter to within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT 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 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

b.

With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

At all other times With the concent:-tion of either chloride or fluoride in the Main Coolant System in excess of its Steady State Limit for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in excess of its Transient Limit, reduce the pressurizer pressure to < 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Main Coolant System; determine that the Main Coolant System renains acceptable for continued operations prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.6 The Main Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-1.

YANKEE-RCWE 3/4 4-12

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 9.

Verifying that the charging header flow metering in-l strument is OPERABLE by observing charging flow rate at least once par 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establishing containment integrity, and 2.

Of the areas affected within containment at the comple-tion of each containment entry when containment integrity is established, d.

At least once per 18 months by visual inspection of the con-tainment 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, e.

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

1.

Cycling each power operated (excluding automatic) valve in the flow path through at least one complete cycle of full travel.

2.

Verifying that valve CS-M3V-532 actuates to its correct position on a safety injection signal.

3.

Verifying that each of the following pumps start auto-matically upon receipt of a safety injection signal:

(a) High pressure sifety injection (HPSI) pump (b) Low pressure safety injection (LPSI) pump YANKEE-ROWE 3/4 5-7 Amendment No. 52

EMERGENCY CORE C00LI"G SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 Verifying that two low pressure safety injection pumps develop a combined flow 2180 gpm.

Test every LPSI pump at least once per 36 months.

5.

Verifying that each charging pump stops automatically upon receipt of a safety injection signal.

6.

Verifying that the charging header flow metering instrument is OPERABLE by perfoming a CHANNEL CALIBRATION.

7.

Verifying that each valve listed in Specification 4.5.2.b.3 l

is in its normally open posit ~.on.

8.

Verifying the proper positici;ing of the HPSI throttle valves SI-V-671, 672, 673, and 674 by performing an inspection to insure that:

a)

Each valve locking device is in place and securely Welded to the valve handle and to the valve yoke.

b)

The scribe mark on each valve body aligns with the scribe mark on the valve yoke.

9.

Verifying the proper positioning of hot leg injection throttle valve SI-V-645 at least once per 36 months by flow testing.

f.

At least every 36 months, and/or any time ei+' + test under 4.5.e.8 is failed, by developing a backpressure of 875 psig.

in the high pressure safety injection header with two HPSI pumps operating as follows:

1.

Pressure to the suction of the HPSI pumps to be 170_+

10 ps.i.

2.

LPSI flow is isolated.

3.

Injection flow is to one loop witn the other loops isolated by closing the approoriate inj'e:: ion gate valves CS-40V-535, CS-MOV-537, CS-MOV-53E, and CS-MOV 539.

l

-ne #10., :: :ne inje:: ion loans snail n:t ce less tnan i

23C g:r.

I I

E.

The coeve te:: shall be receate :o in:1uce the operation i.

al' @2: outes.

l'

. c '. EE '.:..'E

E ken: e : N:. /2,JE,52, 58

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Main Coolant System temperature above 200 F.

l SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the vessel and verifying no apparent changes in appearance of the surfaces or other abnormal degr'adation.

An initial report of any abnormal degradation of the containment vessel detected during the above required inspections shall be made within 10 days after detection and the detailed report shall be submitted to the Commission pursuant.to Specification 6.9.1 within 90 days after completion of the surveillance requirements of this specification.

O

AL '.EE-R;.: E 3/4 6-7 Amendment No. 58

CONTAINMENT SYSTEMS CONTINUOUS LEAK MONITORING SYSTEM LIMITING CON 91 TION FOR OPERATION 3.6.1.7 The continuous leak monitoring system shall be OPERABLE within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following establishment of CONTAINMENT INTEGRITY with:

a.

Containment internal pressure 1 0.75 psi;,

b.

At least eight containment temperature detectors, c.

At least one containment pressure detector, d.

Two relative humidity detectors.

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

ACTION:

With the continuous leak monitoring systein inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after closing any containment air lock door, whichever is sooner, or be in at least HOT STANDBY within tne next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN with Main Coolant pressure < 300 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.7 The continuous leak monitoring system shall be demonstrated OPERABLE by:

a.

Verifying containment internal pressure to be 1 0.75 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Calibrating the temperature detectors, the pressure detector (s) and the relative humidity detectors at least once per 18 months.

Baln coolant pressure 1 300 psig.

YANKEE-ROWE 3/4 6-8 Amendment No. 49

TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES 5

M TESTABLE DURING IE VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME (Yes or No)

(Seconds)

E B.

CllECK VALVES (Continued)

SW-V-820*

Serv ce Water to Containment i

Cooler #1 NA NA SW-V-821*

Service Water to Containment Cooler #2 NA NA SW-V-822*

Service Water to Containment Cooler #3 NA NA SW-V-823*

Service Water to Containment Cooler #4 NA NA llc-V-1199*

Steam Supply to Containment lleaters NA NA 1

C.

Manual Valves NA h

SC-MOV-551+553*

Shutdown Cooling - In No NA SC-MOV-552+554*

Shutdown Cooling - Out No NA Cil-MOV-522*

MC Feed to Loop Fill Header NA NA CS-V-601 Shield Tank Cavity Fill NA NA CA-V-746*

Containment Air Charge NA NA IIV-V-5*

Containment 112 Vent System NA NA llV-V-6*

Containment il2 Vent System NA NA CA-V-688 Containment H2 Vent System Air Supply NA NA CS-MOV-500 Fuel Chute Lock Valve No NA

Not subject to Type C tests

TABLE 3.6-1 (Continpr '

CONTAINMENT ISOLATION VALVES y

TESTABLE DURING VALVE NUMBER FUNCTION PLANT OPERATION ISOLATION TIME x

(Yes or No)

Seconds y

C.

Manual Valves (Cont'd) m CS-CV-215 Fuel Chute Equalizing NA NA CS-CV-216 Fuel Chute Dewatering NA NA Pump Discharge VD-V-752*

Neutron Shield Tank-Outer Test NA NA VD-V-754*

Neutron Shield Tank-Inner Test NA NA BF-V-4-1 Air Purge Inlet NA NA BF-V-4-2 Air Purge Outlet NA NA HC-V-602 Air Purge Bypass HA NA m

SI-MOV-516 ECCS Recirculation No NA y

SI-MOV-517 ECCS Recirculation No NA BF-CV-1000*

3G#1 Feedwater Regulator No 30 BF-CV-1100*

SG#2 Feedwater Regulator No 30 BF-CV-1200*

SG#3 Feedwater Regulator No-30 BF-CV-1300*

SG#4 Feedwater Regulator No 30 2i g

PR-V-610 Main Coolant Heise Pressure Gauge NA NA l n.

2 PU-V-543 Purification System Containment

?+

Sump Suction HA NA z

PU-V-544 Purification System Containment P

Sump Suction NA NA VD-V-1093 SG#1 Emergency Feed (SI)

No NA g

VD-V-1094 SG#2 Emergency Feed (SI)

No NA VD-V-1095 SG#3 Emergency Feed (SI)

No NA VD-V-1096 SG#4 Emergency Feed (SI)

No NA g

Not subject to Type C tests

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator of an unisolated main coolant loop shall be OPERABLE.

APPLICABILITY: MODES 1, and 3.

ACTION:

a.

With 3 ar 4 main coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable:

1.

Operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either:

a)

The inoperable valve (s) is restored to JPERABLE status, or b)

Three Power Range Neutron Flux channels are OPERABLE **

with:

1)

The Power Range coincidence selector switch in the single position, 2)

The trip setpoints reduced per:

(a) Table 3.7-1 for 4 loop operation, or (b) Table 3.7-2 for 3 loop operation.

3)

One Intermediate Power Range Neutron Flux channel in the tripped condition.

2.

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.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition, and Addenda through Summer, 1975.

Operation in the 3-loop mode is not permitted until appropriate LOCA analyses for this mode have been approved by the NRC.

- 0ne Power Range Neutron Flux channel may be made inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance per Specification 4.3.1.1.

YANKEE-ROWE 3/4 7-1 Amendment No. 58

TABLE 3.7-1

[;l MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGil SETPOINT l

'. 3 WITil INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION f81 Itax imum flumber of Inoperable Safety Maximum Allowable fleutron Fiux liigh Setpoint valve _s on. Any_ Opera ting Steam Genera tor (Percent of RATED TilERMAL POWE_R).

1 27 tl r.

~

o

'r e 1

a

J R

f a

e

~^

c

~, ~ '

's-s, g,,-

s-6:'.

s 1-:

.------.a

.~ -.

s TABLE 3.7-2 r

MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGil SETPOINT l

i.,

WITil INOPERABLE STEAM LINE SAFETY VALVES 90 RING 3 LOOP OPERATION

.n n

sei J;

Maximum thunber of Inoperable Safety Maximum Allowable Neutron Flux liigh Setpoint Valyes on Any_0pe_ rating Steam Generator *

(Percent af RATED THERMAL POWER) s I

20 t

Y 4 s

r, m,

(.D t u i

.i.

a

ht' Teast Tw~o~ safety valves shall be OPERABLE on the non-operating steam generator.

r 3

[*

q.

'I (J

=

,.e 4

F 4

v y-m 4

q.

.,..,.-.~---7,sp.y

.._e w

h

<v m

y

1 I

3 ',.

3 b

. v..

r

,~

e x-

+

e

TABLE 4.7-1 2

,y STEN 1 LINE SAFETY VALVES PER LOOP VALVE Ntif1BER LIFT SETTING (I 3%)

ORIFICE SIZE

a.

SV-409 E, F, G or ti 935 psig K

b.

SV-409 A, B, C or D 985 psig K2 c.

SV-409 I, J, K or L 1035 psig Q

R.

z l'

t.

K = 1.838 square inches K = 2.545 square inches 2

Q = 11.05 square inches

PLANT SYSTEMS 3/4.7.7 WASTE EFFLUENTS RADI0 ACTIVE SOLID WASTE LIMITING CONDITION FOR OPERATION 3.7.7.1 Radioactive solid waste shall not be disposed of at the site.

l APPLICABILITY: At all times ACTION:

With any radioactive solid waste disposed of at the site, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.6 within 90 days describing the circumstances of the disposal and outlining plans for removal of the waste. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.7.1 Not applicable.

l YANKEE-ROWE 3/4 7-21 Amendment No. 49

WMamPWl _

1

,_L

--f T

PLA': SYSTEMS

_R_;DI0 ACTIVE LICUID WASTE LIMITIr;G CONDITION FOR OPERATION

~

ct 3.7.7.2 Radioactive liquid waste shall be discha 'ged only when t*e activity of the waste, together with the activity being released from steam generator bicwdown, is less than the maximum permissible concen-tration established in 10 CFR Part 20.

2 APPLICABILITY: At all times ACTION:

7 With discharge of radioactive liquid waste in excess o' the limits, immediately suspend the discharge.

The provisions of J.pecifica-ion 3.0.4 are not applicable.

2 S;

SURVEILLANCE REQUIREMENTS

(

4.7.7.2.1 Radioactive liquid waste shall be determined to be within the

}

above limits by radioactivity analysis prior to discharge.

A 4.7.7.2.2 Steam generator blowdown radioactivity shall be analy::ed at I[

1 east every 7 days whenever steam generator blowdcwn is ir. progres:.

l N

lll

'y*

lYP':E E 0.C.lE 3/2 7-22 A ert sr.: ';;. 58 l

'i 4K 1--

Y

PLANT SYSTEMS RADI0 ACTIVE GASE0US WASTE LIMITING CONDITION FOR OPERATION 3./.7.3 Concentration of radioactive gaseous wastes c:ischarged, as determined at the point of discharge from the pritaary vent stack and averaged over a period not exceeding one year, shall not exceed 1000 times the limits specified in 10 CFR Part 20, App;ndix B Table II, except that, for isotopes of Iodines and particulates with half-lives

> 8 days, the MPC values of Table II shall be reduced by a factor of 1/700.

APPLICABILITY: At all times ACTION:

With discharge of radioactive gaseous waste in excess of limits, im-mediately suspend the discharge.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.7.3 Radioactive gaseous waste discharge concentrations shall be determined to be within the limits, by continuously monitoring the primary vent stack effluent and analyzing for tritium, particulates, iodines, and tission, and activation gases at least once per 31 days.

YANKEE-ROWE 3/4 7-23

FLAllT SYSTEMS 3/4.7.8 ENVIROMMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.7.8 The environmental monitoring program shall be performed in accordance with Table 4.7-4.

APPLICABILITY: At all times ACTION:

With the sampling and analysis program speciffsd in Table 4.7-4 not j

satisfied, a special report shall be prepared and submitted to the Commission pursuant to Specification 6.9.6 within 90 days describing the circumstances of the violation and outlining plans to prevent re-occurrence of the violation. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.8 The er/vironmental monitoring samples shall be collected and analyzed in accordance with the requirements of Table 4.7-4.

i A.';KEE :C'..'E 3/4 7-24 Amendment No. /;,58 l

REFLELIMG GPEP.*TIONS SF:E'_D TAMi: C A'.' ITY MAM: ULATOR CRANE OPERASILITY LIMITING CONDITION FOR OPERATION 3.9.5 Control rods and fuel assemblies shall be har: led one-by-cne witn an OPERAELE snield tank cavity maniculator crane a.c universal handling tooi with:

a.

A minimum capacity of 900 pounds, and b.

An overload cut off limit 1 4800 pounds.

l APPLICABILITY: During movement of control rods er fusi assemblies within the reactor pr essure vessel.

ACTION:

With the requirements for crane and handling tool OPERABILITY not satisfied, suspend use of the incperable manipulator crane or handlir.g tool from operations involving the movement of cont'rol rods and fuel assemblies within the reactor pressure vessel. The provisiens of Sps:ification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.6 The manipulator crane and handling tool used for movement of

control rods or fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of at least 900 pouncs, demon-strating an automatic load cut off when the crane load exceeds 4800 pounds, and verifying proper opertation of the handling t:ol.

l-e e

4

REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL PIT LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 900 pounJs shall be prohibited from travel over the spent fuel pit except for the:

a.

Spent fuel pit building roof hatches, b.

Spent fuel inspection stend, c.

Fuel handling equipment d.

Spent fuel racks.

e.

Temporary gate, and f.

Shielding panels.

APPLICABILITY: With fuel assemblies in the cpent fuel pit.

~

ACTION:

With the requirements of the above specification not satisfied, place the crane load in a safe condition.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.7 Loads in excess of 900 pounds shall be prevented from traveling over the spent fuel pit by administrative control except that the:,

Spent fuel pit building roof hatches, the spent fuel inspection a.

stand, the fuel handling equipment, the spent fuel racks, the temporary gate support brackets, and the shielded work platform may travel over the spent fuel pit in accordance with approved written procedures.

b.

Spent fuel storage racks, the spent fuel inspection stand, the temporary ga'te support brackets, and the shielded work platform shall be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control, and c.

Fuel handling equipment when moved for maintenance shall be pre-vented from traveling over fuel assemblies in the spent fuel pit by administrative control.

P.EE-ROWE 3/4 9-8 Amendement No. 29, D,57

3/2.0 APPLICASILITY 5ASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operati:n and Surveil-lance Requirements within Section 3/4.

3.0.1 This specification defines the applicability of each speci-fication in terms of defined OPERATIONAL MODES or other s:ecified ccnditions and is provided to delineate specifically when eacn specific-ation is applicable.

3.0.2 This spec'ification defines those conditions recessary to ccnstitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3.9.3 This specification delineates. the ACTION to be taken for circur stances not directly provided for in the ACTION state' ents and whose occurrence would violate the intent of the specification.

For example, Specification 3.1.2.6 calls for two charging p;m:s t: be OPEPABLE and provides explicit ACTION requirements when only cne charging pump is OPERABLE. Under the terms of Specification 3.0.3, if no charging pumps are OPERABLE, the facility is required to be in at l

least HOT STANDBY within i hour and in COLD SHUTCOWN within the follow-ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or com:enents OPER-AELE and (b) all other par: meters as specified in the Limiting Conditicns for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTICN statements.

The intent of this provision is to insure that facility :peration is n;t initiated with either required equipment :r syste s in;;erable :r rer specified limits being exceeded.

t Exce::icas t: :b's Orovisicn have been :-: 'ded #:

a 'i-ite:

[~.:eE:fs:e-i'itati:ns'.pnstar:;pwi:nin::eaiee:.:e;.y;1:

i

a-ec ::an sa f e::.

inese e>ceotiens are s a e: "n :he.....

I s a:erents f tre a:; ::riate specifications.

I l!

\\

Y

.i t

};

y

- 'IE Ol..E E 3/2' Amendment No. 58

APPLICABILITY BASES 4.0.1 This specification provides that surveillance activities ne;essary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Conditions for Operation are applicable.

Provisions for additional surveillance activities to be performed without regard to the a9plicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements.

4.0.2 The provisions of this specification provide allowable toler-ances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations.

The tolerance values, taken either individually or consecutively over 3 test intervals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not signifi-cantly degrcded beyond that obtained from the nominal specified interval.

4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements.

4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been perfomed within the specified time interval prior to entry into an OPERATIONAL MODE or other applicable condition.

The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.

Under the tems of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.

YANKEE-ROWE B 3/4 0-2

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line ccde safety valves ensurds that the secondary system pressure will be limited to within its design pressure of 1035 psig during the.most severe anticipated system opera-tional transient. The maximum relitving capacity is associated with a turbine trip from 100% RATED THERMAu POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section VIII of the ASME Boiler and Pressure Code, 1966 Edition. The total religving capacity for all valves on all of the steam lines is 3.1 x 10 1bs/hr which is 129 percent 6

of the total secondary steam flow of 2.4 x 10 lbs/hr at 100% RATED TH:RMAL POWER. A minimum of 2 OPERABLE safety valves per OPERABLE steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Tabics 3.7-1 and 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requircments on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron l Flux channels. The reactor trip setpoint reductions are derived on the following bases:

For 4 loop operation 3p, (X) - (Y)(V) x (108)

X For 3 loop operation SP = ( ) -

II x (81) x

'here:

t i

SP = ieduced reactor trip set:oint in cercent cf RATEJ THERMAL POL'ER l

i V = maximum number of in' operable safety valves per stea-generater lg YA' ;EE :!..E 3 3/4 7-1 Amendren ';: 58

4 PLANT SYSTEMS BASES U

maximum number of inoperable safety valves per operating

=

steam generator (108)

Power Range and InterTnediate Power Range Neutron Flux-

=

High Trip Setpoint for a loop operation (81)

Maximum percent of RATED THERMAL POWER permissible

=

for 3 loop operation.

X Total relieving capacity of all safety valves per steam

=

generator in lbs/ hour Y

Maximum relieving capacity of any one safety valve in

=

lbs/ hour 3/4.7.1.2 EMERGENCY BOILER FEEDWATER SYSTEM The OPERABILITY of the emergency boiler feedwater system ensures that the Main Coolant System can be cooled down to less than 330*F from normal operating conuitions in the event of a total less of off-site power.

The steam driven emergency boiler feedwater pump is capable of delivering a total feedwater flow of 80 gpm at a pressure of 950 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Main Coolant System temperature to less than 330 F when the Shutdown Cooling System may be placed into operation.

The monthly testing interval of the steam generator emergency boiler feed pump verifies ii.3 operability by recirculating water to the demineralized water tank.' Proper functioning of the steam turbine and the emergency boiler feed pump will be made by direct visual observa-tion.

3/4.7.1.3 PRIMARY AND DEMINERALIZED WATER STORAGE TANK The OPERABILITY of the primary and demineralized water storage tanks with the minimum combined water volume ensures that sufficient water is available to maintain the Main Coolant System at HOT STANDBY in excess of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with steam discharge to the atmosphere concurrent with total loss of off-site power.

YANKEE-ROWE B 3/4 7-2

TA,BLE 6.2-1 Ta.;1r$MjHIFTCo.EWCOMPOSITION=

LICENSE APPLICABLE l'3 DES CATEGORY 1, 2, 3 & 4 5&6 SOL 1

1*

~

OL 2

1 Non-Licensed 2**

1

Does not include the licensed Sen:or Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising Core ALTERATIONS after the initial fuel loading.

- 0ne additional non-licensed operator is required for MODE 2 except when restarting within four hours of a shutdown for which the cause has been clearly established.

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 in order to acccamadate unexpected absence on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

. ;EE-RC'.:E 5-5 A e-d cr.t ';. ;g, 58

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff listed below shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.

a.

Plant Superintendent b.

Assistant Plant Superintendnet c.

Chemistry and Health Physics Supervisor d.

Operations Supervisor e.

Reactor Supervison f.

Maintenance Supervisor g.

Instru=ent and Controls S.ipervisors h.

Shift Supervisors 1.

Plant He'alth Physicist 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the require =ents and rece=mendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of. a me=ber of the plant staff appointed to perform the duties of Fire Protection Coordinator and shall meet or exceed the requirements cf Section 27 of the NFPA Code-1975 except for fire brigade training sessions which shall be held at least quarterly.

6.5 RE'.'IEW AND AUDIT 6.5.1 PLANT OPERATION REVIEW COMMIT'EE FUNCTION 6.5.1.1 The Plant Operation Review Co==ittee (PORC) shall function to advise the Plant Superintendent on all matters related to nuclear safety.

YANKIE-ROWE 6-6 Arendment No. (6, 49

4 UNITED STATES r

j 3"

'4 NUCLEAri HEbJLATORY COMMisslON

l

{ 7.!

j WASHINGTON, D. C. 20555 0 ~

?

\\%

/

- .+

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. OPR-3 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION (YANKEE-ROWE)

DOCKET NO. 50-29 Introduction By application dated November 24,1978 (Proposed Change No.139, Supplement No. 3), Yankee Atomic Electric Company (the licensee) requested an amendment to Facility Operating License No. DPR-3 for the Yankee Nuclear Power Station (Yankee-Rowe). The amendment would allow 16 miscellaneous changes to the Technical Specifications to update its provisions.

Discussion The 16 proposed changes relate to:

a.

Correction of typographical or editorial errors in four specifications; b.

Correction of errors or omissions in seven specifications; c.

Revision of five specifications as a result of changes proposed by the licensee.

We have proposed five additional changes whic. are acceptable to the licensee.

Each of the changes are evaluated below and are vrganized in accordance with the above groupings.

Evaluation a.

The following proposed changes relate only to the correction of typographical or editorial errors and are therefore acceptable.

7 904 200 Sto

l

. 1.

Proposed Change #9, Page 3/4 4-22, Table 4.3-3, Item 3.b Item 3.b was deleted from Table 3.3-4 by Amendment No. 49 and should have been deleted from the associated Table 4.3-3 by the same amendment.

2.

Proposed Change #10, Page 3/4 4-4 The temperature limit of 200F was changed to 300F in Specification 3.4.1.3.a by Amendment No. 49.

The same change should have been made to the associated Surveillance Requirement 4.4.1.3.1.

3.

Proposed Change #14, Page 3/4 7-24, ACTION -- Ta bl e 3. 7-4 should be renumbered Table 4.7-4.

4.

Proposed Change #16, Page B3/4 0-1, Specification 3.0.3 The word " inoperable" should read "0PERABLE".

b.

The following proposed changes correct errors or omissions made in the identified specifications.

1.

Proposed Change #3, Page 3/4 3-18, Table 3.3-4 These Yankee-Rowe technical specifications incorrectly identify the Loop Seal Monitor as a monitor which would detect a radioactive gaseous discharge to the environment.

This error occurred because this possible flow path direct to the atmosphere had previously existed.

Proposed Change No. 127 dated August 13, 1975, proposed a modification to the ventilation system which, among other changes, would route any release from this source through the primary vent stack where it would be monitored by particulate, iodine and noble gas monitors. The change to the ventila-tion system was approved by Amendment No.18, dated November 12, 1975.

The Loop Seal Discharge Monitor then became a local monitor which did not require a technical specification.

These technical specifications require the primary vent stack particulate, iodine and noble gas monitors to be operable.

Celetion of the Loop Seal Monitor from the Technical Specifications is consistent with previously approved changes to the ventilation system and is therefore acceptable.

. 2.

Proposed Change #5, Pages 3/4 3-18 and 3/4 3-19, Table 3.3-4 During the development of these Yankee-Rowe technical specifications, the numbers for the Action Statements were incorrectly assigned not to be sequential after the Action Statement numbers of Tables 3.3-1 and 3.3-2.

Reassignment of the numbers to be sequential is an acceptable correction.

3.

Proposed Change #6, Page 3/4 3-20, Table 3.3-4 Action Statement numbers are to be changed to be consistent with changes made to Table 3.3-4 in b.2, above. This is an acceptable correction.

4.

Proposed Change #7, Page 3/4 3-20, Table 3.3-4 and Page 2/4 3-22, Table 4.3-3 The footnote "** With radioactive effluent in the Waste Gas Surge Drum" should be deleted as it refers to item 2.b.1 of Table 3.3-4 which was deleted in item b.1, above.

Therefore, deletion of the footnote is an acceptable correction.

5.

Proposed Change #8, page 3/4 3-21, Table 4.3-3 The Loop Seal Monitor was deleted from Table 3.3-4 by item b.1, above.

Therefore, its deletion from this associated table is an acceptable correction.

6.

Proposed Change #13, Page 3/4 7-22, Specification 4.7.7.2.2 During development of the Yankee-Rowe Technical Specifications, an error was made in requiring sampling and analysis of a steam generator "every 7 days, whenever any steam generator contains water". A steam generator can only be sampled for activity during blowdown. The Yankee-Rowe steam generators are blown-down continuously during operation in a trickle stream through a blowdown monitor which is required to be operable by the Technical Specifications. To be consistent, change of the blowdown radioactivity sampling and analysis requirement to "every 7 days whenever blowdown is in progress" resolves an apparent inconsistency and is therefore acceptable.

7.

Proposed Change #15, Page 3/4 9-7 Specification 3.9.6.b specified "an overload cutoff limit < 4800 pounds above base load." The base load, 2200 pounds, should have been included in the 4800 pound limit to operate the load cell within its range, 0-5000 pounds, and to avoid exceeding the tool boom design overload limit of 6200 pounds. To include the base load in the 4800 pound limit is more conservative than a 4800 limit above base load, and assures an appropriate safety margin for the tool boom design limit. Therefore, this change is acceptable.

- We conclude that the above group of changes to correct errors or omissions made in the development of the Technical Specifications are administrative in nature, do not decrease the level of safety of the facility, and are therefore acceptable.

c.

The following changes to the Technical Specifications were proposed by the licensee.

1.

Proposed Change #;, Pages 3/4 3-18 and 3/4 3-19 Table Process Monitors 2.a.1, 2.b.l.a (formerly 3.3-4 2.b.2.a), 2.b.l.b (formerly 2.b.2.b), 2.b.l.c (fonnerly 2.b.2.c) and 2.c.1 have been upgraded by installation of new improved instruments.

The licensee proposed a new alarm setpoint for item 2.a.1, Main Coolant System Leakage Air Parti ~' late Monitor.

However, to be consistent with the Standard Tet mical Specifications (STS) for Westinghouse plants, which are applicable to Yankee-Rowe, the setpoint for the Main Coolant System Leakage Air Particulate Monitor need not be specified in the Technical Specifications. This is because the leakage rate correlation versus leak detector setpoint varies with Main Coolant System activity and t,cckground radioactivity. Regulatory Guide 1.45, May 1973, addresses leakage detector setpoint, calibration and response time. The licensee has agreed to changing the alarm setpoint for item 2.a.1 to NA (Not Applicable).

This is consistent with the Westinghouse STS, including the guidance contained in Regulatory Guide 1.45, and is therefore acceptable.

Furthermore, as a result of our telephone discussions, the licensee withdrew the proposed new setpoints for items 2.b.l.a (formerly 2.b.2.a), 2.b.l.b (formerly 2.b.2.b),

2.b.l.c (formerly 2.b.2.c) and 2.c.l.

2.

Proposed Change #2, Page 3/4 3-18, Table 3.3-4, item 2.a.1 The measurement range of the main coolant system }eakage air 10-10gulate monitor has been increased from 10-10 parti cps to cpm as a result of upgrading the instruments.

3.

Proposed Change #4, Page 3/4 3-19 Table 3.3-4, item 2.c The measurement range of the sjeam generator blowdown monigor 4

has been increased from 1 x 10 cps to 10-10 cpm or 10-10 cpm as a result of upgrading the instrument.

. 4.

Proposed Change #11, Page 3/4 4-8 -- Specification 3.4.5.1.a has been changed to read "At least one" because a redundant containment atmosphere particulate radioactivity monitor has been added.

The 5.

Proposed Change #12, Pages 3/4 4-10 and 3/4 4-11 licensee has dpgraded the Main Coolant System Leakage Air Particulate Monitor. The licensee proposed this change based on its proposed setpoint for the air particulate monitor, but Item c.1, above, deleted the proposed setpoint

.for this new instrument from the Technical Specifications.

By deleting the setpoint, Action "d" was made redundant to Action "b".

The Westinghouse plant STS, which are applicable to Yankee-Rowe, require an identical Action "b" but no further actions.

Therefore, deletion of Action "d" from the Yankee-Rowe Technical Specifications is acceptable.

We conclude that the above group of changes do not decrease the level of safety of the facility and are thercfore acceptable.

The following changes were identified by us and have been accepted by the licensee.

1.

Page 3/4 5-8, Surveillance Requirement (SR) 4.5.2.e.7 Amendment No. 52 renumbered SR 4.5.2.b.4 to 4.5.2.b.3 -

SR 4.5.2.e.7 refers to SR 4.5.2.b.4 and should therefore be corrected to refer to SR 4.5.2.b.3.

2.

Page 3/4 6-7, Specification 3.6.1.6, Action -- The U

typographical error "200kF" should be corrected to "200 F".

The Main Coolant Heise 3.

Page 3/4 6-14, Table 3.6-1.

Pressure Gage isolation valve number should be changed from PR-V-623 to PR-V-610 to reflect the correct valve number.

4.

Page 3/4 7-1, Specification 3.7.1.1, Action; Page 3/4 7-2, Table 3.7-1 title; and Page 3/4 7-3, Table 3.7-2 title --

Amendment No. 47 recognized that the Reactor Protective System Intermediate Power Range Neutron Flux channel setpoints cannot be reduced in the same manner as the Power Range Neutron Flux channel setpoints. This knendment deleted the requirement for the Intermediate Power Range channel setpoints to be reduced

in Specifications 3.10.3 and 3.10.4, but required that three Power Range Neutron Flux channels must be operable with reduced setpoints. Specification 3.7.1.1 Actions "a" and "b" also require the Power Range and Intermediate Power Range Neutron Flux channel setpoints to be reduced as indicated in Table 3.7-1 or 3.7-2 when main steam line code safety valves are inoperable. Because the Intermediate Power Range channel setpoints cannot be reduced to the setpoints specified, Actions "a" and "b" of Specification 3.7.1.1 would be changed to require only the three Power Range channels to be operable and to have their setpoints reduced as indicated in Table 3.7-1 or 3.7-2, as applicable.

In addition, one Intermediate Power Range Neutron Flux channel would be required to be in the tripped condition and the Power Range Neutron Flux channels coincidence logic would be required to be in the " Single" position such that tripping of any one of the three Power Range Neutron Flux channels at the reduced setpoint will trip the reactor. These changed requirements would be consistent with the requirements of the Westinghouse plant STS, which are applicable to Yankee-Rowe, and are acceptable. The titles of Tables 3.7-1 and 3.7-2 would be changed to delete reference to the Intermediate Power Range channels. A footnote would be added to the LC0 action statement to prohibit operation with less than 4 reactor coolant loops in service until NRC approval of appropriate LOCA analyses.

5.

Page 6-5, footnote

the misspelling of ALTERATIONS would be corrected.

We conclude that the above group of changes do not reduce the level of safety of the facility and are acceptable.

Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

. Conclusion We have concluded, based on the considerations discussed above, that (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date:

April 3,1979

ML19281B034

Text

Attachment:

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