ML20002C684

Amend 61 to License DPR-28,incorporating Limited Operating Conditions Associated W/Cycle 8 Operation

ML20002C684
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	12/18/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20002C685	List: ML20002C684 ML20002C688 ML20002C689
References
NUDOCS 8101100717
Download: ML20002C684 (21)

Text

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NUCLEAR REGULATORY COMMISSION 54 j't WASHING TON, D. C. 20555 t%44

\\. Y. * ',f, VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-28 1.

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

A.

The application for amendment by Vermont Yankee Nuclear Power Corporation (the licensee) dated August 19, 1980, as supplemented October 7 and 23, and November 21, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 regulations of the Comission; C.

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.

DPR-28 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 61, are hereby incorporated in the license.

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

810110 DN

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

FOR THE NUCLEAR REGULATORY COMMISSION 1

I i

Thoma

. Ippolito Chief

{

Operating Reactors Branch #2 Division of Licensing l

Attachment:

Changes to the Technical Specifications Date of Issuance: December 18, 1980 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 61 FACILITY OPERATING LICENSE N0. OPR-28 DOCKET NO. 50-271 Revise Appendix A as follows:

Remove the pages listed and replace with revised pages.

i 1

2 2

2a 5a 5a 6

6 14a 14a 14b 14b 18 18 25 25 31 31 70R 70R 71 71 76 76 77 77 135 135 136 136 179 179 180-01 180-01

l TAllLE OF GNTENTS Page No.

CENERAL 1

1.0 DEFINITIONS 5.0 DESIGN FEATURES 188 6.0 ADMINISTRATIVE C0!(IROLS 190 l

l 6.1 ORGANIZATION -

190 6.2 REVIIM AND AUDIT -

194 l

A.

Plant Operation Review Comunittee -

194 196 B.

Nuclear Safety Audit and Review Consnittee -------

6.3 ACTION TO BE TAKEN IN IllE EVENT OF AN ABNGRMAL 199 OCCURRENCE IN PLANT OPERATION -

6.4 ACTION TO BE TAKEN IF A SAFETY' LIMIT IS EXCEEDED 199 200 6.5 FIANT OPERATING PROCEDURES 207 6.6 PLAlff 0FERATING RECORDS ----

208 6.7 PLANT REPORTING REQUIRIMENTS 6.8 FI RE PROTE CT I ON I NS PE CT I ON ----------- ----------- -- - ------- ------- ----

214 6.9 ENV I RONMEN TAL QU AL I F I C AT I ON - -- ----- ---- --- -- -- - ------ -- - -----------

215 SAFETY LIMITS Page No.

LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 5

2.1 16 2.2 1.2 REACTOR COOLANT SYSTEM Amendment No. 61 1

I

VYNPS G.

Instrument Functional Test - An instrument I..

Operating - Operating means that a system or component - ~

functional test means the injection of a in performing ign intended functions in its required simulated nignal into the inntrument primary manner.

sensor, to veri fy the proper instrument channel

• • v response, alarm, and/or initiating action.

M.

Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent H.

Log System Functional Test - A logic system refueling outage, functional test means a test of all relays and contacts of a logic circuit from sensor to N.

Peaking Factor - The ratio of the fuel rod heat flux 4

cctiviated device to insure all canponents are to the heat flux of an average rod in an identical operable per design intent. Where possible, geometry bundle operating at the average core power.

l cetion will go to completion, i.e.,

pumps will be started and valves opened.

O.

Primary Containment Integrity - Primary containment I.

Minimum Critical Power Ratio - The Minimum integrity means that the drywell and pressure Critical Power Ratio is defined as the ratio of supression chamber are intact and all of the following that power in a fuel assembly which is calculated conditions are satisfied:

to cause some point in that assembly to experience boiling transition as calculated by 1.

All manual containment isolation valves on lines cpplication of the CEXL correlation to tne actual connecting to the reactor coolant system or essembly operating power.

containment which are not required to be open 1

(Reference NEDO-10958) during accident conditions are closed.

4 J.

Mode - The reactor mode is that which is 2.

At least one door in each airlock is closed and l

established by the mode-selector-switch.

sealed.

K.

Operable - A system, subsystem, train, component 3.

All automatic containment isolation valves are or device shall be OPERABLE or have OPERABil.ITY operable or deactivated in the isolated position.

when it is capable of performing its specified function (s).

Implicit in this definition shall 4.

All blind flanges and manways are closed.

be the assumption that all necessary attendant ins t' umentation, controls, normal and emergency P.

Protective Instrumentation Definitions electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are 1.

Instrument Channel - An instrument channel means required for the system, subsystem, train, an arrangement of a sensor and auxiliary component or device to perform its function (s) equipment required to generate an'd'[Tansmit to a cre also capable cf performing their related t rip system a single trip signal related to the support function (s).

plant parameter monitored by that instrument channel.

Anendment flo. 61 2

VYNPS I

I 2.

Trip System - A trip system means an arrangement,

j of instrument channel trip signals and auxiliary l

equipment required to initiate action to accomplish a protective trip sucction. A trip system may require one or more instrument channel trip signals t

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. -34w Amendment No. 61 2a

C 1-4 VYNPS W

Ut 1.1 SAFlirY 1.1 HIT 2.1 1.lHITING SAFETY SYSTEM SETTING Qb In the event of operation with the ratio of HFLPD to FitP greater than 1.0, the APitH gain shall be increased hy the ratio: HFl.PD FitP l

where: 11FLPD = maximum fract. ion of limiting power density where the limiting power density is 18.5 KW/ft for 7 x 7 fuel and 13.84 KW/ft. for 8 x 8 fuel.

FitP

= fraction of rated power (1593 Hwt)

In the event, of operation with the ratio of HELPD to FilP equal to or less than 1.0, the APRH gain shall be equal to or greater than 1.0.

For no combination of loop recirculation flow rate and core thermal power shall the APRH flux scram trip setting be allowed to exceed 120% of rated thermal power.

t b.

Flux Scram To S Setting (Heruel or Startup and llot.

Standby Mode)

When t.he reactor mode switch is in the REFUEL or STAllTUP position, average power range monitor ( APRH) scram shall be set down to less than or equal to 15%

of rated neutron flux. The IHH flux scram sett.ing shall be set at less than or equal to 120/125 of full scale.

B.

Core Thermal Power Limit (Reactor Pressure B.

APitM Hod Block Trip Setting 800 psia or Core Flow 10% of Hated)

When the reactor pressure is 800 psia or The APHH rod block trip setting shall be as core flow 10% of rated, the core thermal shown in Figure 2.1.1 and shall be:

i power shall not, exceed 25% of rated thermal SRD = 0.66W + 182%

power.

5-a Amendment No. 61 w.

VYNPS 1.1 SAFETY 1.lHIT 2.1 1.lHITING SAFETY SYSTEM SE'ITING C.

Power Transient where:

To ensure that the Safety Limit SHD = Hod block tting in percent of established in Specification 1.1 A and rated ther. 1 power 1593 HWL and 1.1B is not exceeded, each required scram shall be initiated by W = percent rated drive flow where 100%

its expected scram signal. The Safety rated drive flow is that flow I.imit shall be assumed t.o be exceeded equivalent, to 48 y 106 lbs/hr core when scrum is accomplished by a means

flow, other than the expected scram sigiuil.

In the event of operation with t.he ratio of HFLPD to FRP l

great.cr than 1.0, the APRH gain shall be increased by the ra tio: HFl.PD FHP where: HFLPD = maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7 x 7 fuel and 13.4 Kw/ft. for 8 e

x 8 fuel.

PHP

= fraction of rated power (1593 Hwt)

In the event. of operation with the ratio of HFLPD t.o FRP equal to or less than 1.0, t.he APHH gain shall be equal to or greater than 1.0 Amendment No. 61 6

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VYNPS APitM Flux Scrum Trip Set ting (Hun Mode)

The scram t. rip set.t.ing must. he adjust.ed to ensure that, the LHGH transient peak is amt increased for any combination of HiLPD arul reactor core t.hermal power. If the scram requires a change due to an abnormal peaking l

cond it ion, it, will be accomplished by increasing the APitH gain by the rat.io in Specificat.lon 2.1. A.I.a. thus assuring a reactor scram at. lower t,han design overpower condit. ions.

Analyses of the limiting transient.s show that no scram adjustment. is required to assure fuel cladding integrity when the transient. is initiated from the operat.ing limit. MCPH (Specificat.lon 3.11c).

Flux Scram Trip Set,t.ing (Heruel or Startup & Hot. Standby Hode)

For operat. ion in t.he st.artup mode while t.be reactor is at. low pressure, the reduced APitM scram setting to 15 percent. of rat.ed power provides adequate thermal margin between the setpoint, and the safet.y limit, 25 percent of the rated. The margin is adequate to accomodate anticipated maneuvers associated with station startup. Effec ts of increasing pressure at, zero or low void content, are minor, cold uater from sources available during atactup is not, much colder than that alroody in the system, temperature coefficient.s are small, and control rod pai.t. erns are const. rained to be uniform by operating procedures backed up by the rod worth minimzer. Wort.h of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of react.ivity input, uniform control rod withdrawa) is the most, probable cause of significant, power rise. Because the flux distribut.lon associated with uniform rod withdrawals does not. involve high local peaks, and because several rods amist, he moved to change power by a significant, percentage of rated power, t.he rate of power rise is very slow. Generally, the heat, flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and t.he APHH system would be more than adequate to assusu a scram before the power could exceed t.ho safety limit. The reduced APHH scram remains act.ive unt il t.he mode switch is placed in the HUN posit. ion. This switch can occur when reactor pressure is greater than 850 psig.

The IHH system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IHH is a 5-decade instrument which covers the range of power level bet. ween t. hat, covered by the SHH and the APHH. The 5 decades are covered by the IHH by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IHH scram trip sett.ing of 120/125 of full scale is active in each range of the IRH.

For example, if the inst.rument, were on range 1, t.he scram sett.ing would be a 120/125 of full scale To that, range; lik ewise, if the instrument, were on range 5, the scram would be 120/125 of full scale on that range. Thus, as t.he IHH is ranged up to accomodate the increase in power level, the scram trip sett.ing is also ranged up. The most significant sources of react.ivit y change during the power increase are due to control rod wi t hdrawa l. For insequence control rod withdrawal, the rate of change of power is slow enough due t.o t.he physical limitat. ion of withdrawing control rods, that, heat flux is in equilibrium with t.he neut.ron flux arut an IHH scram would result. In a react.or shutdown well before any Safety Limit. is exceeded.

181-a Amendment No. 61

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VYNPS 8.1 StlHVEll.i.ANCE HEQilIHEHENTS h1 1.!HITlHG CONDITIONS FOH OPEHATION 4

l.I ItEACTOH PitOTECTION SYSTEH 3.1 ItFACTOH PI(O'lECTIOli SY3fEH i

Applicability:

Applicability:

Applies to the operbility of plant. Instrumen-Applies to the surveillance of the plant. Instrumen-t.ation arul control systems required for reactor tion azul cont.rol systems required for react.or sa rct y,

sa fe t.y.

Ol>J ec t.t ve:

Objective:

To specify the limit.s imposed on plant, operat. ion To speelry the t.ype and frequency of surveillance by those instrument, and cont.rol systems required to be applied to 1. hose instrument and control for reactor sa fet.y.

syst enu required for reactor safet y.

Sp ee l fl ea t.i on:

3pectrication:

A.

Plant, operat. ion at, any power level shall bu A.

Instrument.ation system:i shall be funct.lonally

, permit.ted in accordance wit.h Table 3.1.1.

tested and calibrated as indicat.ed in Tables The system response t.ime from the opening of 84.1.1 and 18'.1.2, respec tively.

the sensor cont.act, up to and including the opening of the scram solenoid relay shall not exceed S0 millisecosuls.

11.

During operat.lon wit.h the ratio of HFl.PD to 11.

Once a day during reactor power operation the FRP great.cr t.han 1.0 either:

maximum fraction of limit.ing power density and fract. ion of rated power shall be determined a.

The APitH System gains shall be adjusted aiul the APIIH system gains shall be adjust.ed by by the ratios given in Technical the ratios given in Technical Specificat.lons Specificat. ions 2.1. A.1 and 2.1.11 or 2.1.A.I.a and 2.1.11.

h.

The power distribution shall he change I to reduce the rat.lo of HFl.PD to FHP.

18 Anendmenl.No.61

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VYNPS TABLE 11.1.2 SCRAM INSTRUMENT CALIBRATION HINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNEL 3 Instrument Channel Group (l)

Calibration Standard (#3)

Hinimun Frequency (2)

High Flux APRH Output Signal B

lleat Balance Once Every 7 Days Output Signal (Reduced)

B 13 eat Balance Once Every 7 Days Flow Bias B

Standard Pressure and Voltage Source Refueling Outage LPRM B(5)

Using TIP System Every 1000 equiv full pwr hr High Reactor Pressure B

Standard Pressure Source Once/ Operating Cycle l

Turbine Control Valve Fast Closure A

Standard Pressure Source Every 3 months 4

High Drywell Pressure A

Standard Pressure Source Every 3 months High Water Level in Scram Discharge Volume A

Water Level Herueling Outage Low Reactor Water Level B

Standard Pressure Source Once/ Operating Cycle Turbine Stop Valve Closure A

(6)

Refueling Outage liigh Hain Steamline Radiation B

Appropriate Radiation Source (3)

Herueling Outage First Stage Turbine Pressure Permissive A

Pressure Source Every 6 months and efter refueling Riine Steam 11ne Isolation Valve Closure A

(6)

Refueling Outage Amendment No. %' 61 Page 25

VYNPS iLine:6:

34. 1 HEACT0lt PI(UTECTION SYSTEM A.

Ttie scram sensor channels list.ed in Tables 4.1.1 and 4.1.2 are divided into three groups: A, B a nd C.

Sensors that make up Group A t.re of the on-off type and will be tested and calibrated at the indicated intervals. Init.lally the tests are more frequent than Yankee experience indicat.es necessary. However, by t.est.ing more frequent. ly, t.he confidence level with this instrumentat. ion will increase and t.esting will provide data to just.ify extending the test. Intervals as experience is accrued.

Group 11 devices ut.111ze an analog sensor followed by an amplifier and bi-stable trip circuit. This t.ype of equipment incorporat.cs control room mount ed indicat. ors and annunciator alarms. A failure in the sensor or amplifier may be detected by an alarm or by an operat.or who observes that, one indicat.or does not track the others in almilar channels. The bi-st.ahic trip circuit. failures are detected by the periodic t.est.ing.

Group C devices are act.1ve only during a given portion nf the operating cycle. For example, the IHH is active during st. art.-up and inaet.ive during full-ppower operation. Test.ing of these instrument.s is only meaningful wit.hin a reasonable period prior to t. heir use.

11.

The rat.lo of MFl.PD to FHP shall be checked once per day to determine if the APHM gains require adjustment..

l llecause few cout.rol rod movements or power changes occur, checking these parameters daily is adequat.e.

t e

Amendment No. 61 31

i VYNPS l

3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS 2.

The control rod drive housing support system 2.

The control rod drive housing support system i

shall be in place when the reactor coolant system shall be inspected after reassembly and the is pressurized above atmospheric pressure with results of the inspection recorded.

fuel in the reactor vessel unless all operable control rods are fully inserted.

3.

While the reactor is be'ow 20% power, the Rod 3.

Prior to control rod withdrawal for startup the Worth Minimizer (RWM) shall be operating while Rod Worth Minimizer (RWM) shall be verified as moving controls rods except that:

operable by performing the following:

(a)

If af ter withdrawal of at least twelve (a) The Reactor Engineer shall verify that the j

control rods during a startup, the RWM control rod withdrawal sequence for the Rod fails, the startup may continue provided a Worth Minimizer computer is correct.

second licensed operator verifies that the operator at the reactor console is following (b) The Rod Worth Minimizer diagnostic test the control rod program; or shall be performed.

(b)

If all rods, except those that cannot be (c) Out-of-sequence control rods in each moved with control rod drive pressure, are distinct RWM group shall be selected and the fully inserted, no more than two rods may be annunciatior of the selection errors moved.

verified.

(d) An out of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4.

Control rod patterns and the sequence of 4.

The control rod pattern and sequence of withdrawal or insertion shall be established such withdrawal or insertion shall be verified to that the rod drop accident limit of 280 cal /g is comply with Speci fication 3.3.B.4.

not exceeded.

. y..

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Amenduent No. 61 7gg

VYNPS

~w 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEII. LANCE REQUIREMENTS 5.

Control rods shall not be withdrawn for startup 5.

Prior to control rod withdrawal for startup or or refueling unless at least two source range during refueling, verification shall be made that channels have an observed count rate greater than at least two source range channels have an or equal to three counts per second.

observed count rate of at least three counts per second.

6.

During operation with limiting control rod 6.

When a limiting control rod pattern exists, an patterns either:

i ns t rument functional test of the RBH shall be performed prior to withdrawal of the designated (a) Both RBM channels shall be operable; or rod (s) and daily thereafter.

(b) Control rod withdrawal shall be blocked; or (c) The operating power level shall be limited so that the MCPR will remain above the fuel cladding integrity safety limit assuming a single error that results in complete withdrawal of any single operable control rod.

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Amendment No. 61 71

VYNps 5.

The Source Range Monitor (SRM) systesa has no scram functions.

It does provide the operator with a visual

~

indication of neutron level. The consequences of reactivity accidents are a function of the initial neutron flux. The requirement of at least three count s per second assures that any t ransient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from coLJ., conditions. One operable SRH channel is adequate to monitor the approach to criticality therefore, two operable SRH's are specified for added conservatism.

6.

The Rod Block Monitor (RBH) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power densit y during high power level operation. During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rni could result in one or more fuel rods with MCpR less than the fuel cladding integrity safety limit.

During use of such patterns, it is judged that testing of the RBH system prior to withdrawal of such rods will provide added assurance that improper withdrawal does not occur.

It is the responsbility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods.

. ym Amendment flo. 61 77

VYNPS 3.3 (cont'd)

B.

Control Rods 1.

Control rod dropout accidents as discussed in the FSAR can lead to significant core damage.

f coupling integrity is maintained, the possiblity of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

2.

The control rod housing support restricts the outward novement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system.

The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4.

This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

3.

In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWH operable would entail unnecessary risk, continuing to withdraw rods if the RWH fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod.

Withdrawal of rods for testing is permitted with the RWH inoperable, if the reactor is subcritical and all other rods are fully inserted. Above 20% power, the RWH is not needed since even with a single error an operator cannot withdraw a rod with suf ficient worth, which if dropped, would result in anything but minor consequences.

l 4.

Refer to section 5.5.1 of NEDE 240llP-A, latest revision, " Control Rod Drop Accident Evaluation".

. y..

Anenduent No. 61 76

TABLE 4.7.2.e

_ PRIMARY CONTAINMENT ISOLATION VALVES VALVES SUBJECT TO TYPE C LEAKAGE TESTS Issistion Number of Power Maximum Action on Operated Valven Operating Normal Initiating Group (Note 1)

Valve Identification Inboard Outboard Time (Sec)

Position Signal 1

Main Steam Line Isolation (2-80A, D & 2-86A, D) 4 4

5(note 2)

Open CC 1

Main Steam Line Drain (2-74, 2-77) 1 1

35 Closed SC 1

Recirculation Loop Sample Line (2-39, 2-40) 1 1

5 Closed SC 2

RHR Discharge to Radweste (10-57, 10-66) 2 25 Closed SC 2

Drywell Floor Drain (20-82, 20-83) 2 20 Open CC 2

Drywell Equipment Drain (20-94, 20-95) 2 20 Open CC 3

Drywell Air Purge Inlet (16-19-9) 1 10 Closed SC 3

Drywell Air Purge Inlet (16-19-8) 1 10 Open CC 3

Drywell Purge & Vent outlet (16-19-7A) 3 1

10 Closed SC Drywell Purge & Vent Outlet Bypass (16-19-6A) 3 1

10 Closed SC Drywell & Suppresalon Chamber Main Exhaunt (16-19-7) 1 10 Closed SC 3

Suppression Chamber Purge Supply (16-19-10) 1 10 Closed SC 3

Suppression Chamber Purge & Vent Outlet (16-19-78) 3 1

10 Closed SC Suppression Chamber Purge & Vent Outlet Bypass (16-19-6B) 3 1

10 Open CC Exhaust to Standby Cas Treatment System (16-19-6) 3 1

10 Open CC Containment Purge Supply (16-19-23) 1 10 Open CC 3

Containment Purge Makeup (16-20-20,16-20-22A,16-20-22b) 3 NA Closed SC 5

Reactor Cleanup System (12-15, 12-18) 1 1

25 Open CC 5

Reactor Cleanup System (12-68) 6 HPCI (23-15, 23-16 1

45 Open CC 1

1 55 Open GC 6

RCIC (13-15, 13-16) 1 1

20 Open CC Primary / Secondary Vacuum Relief (16-19-11A,16-19-118) 2 NA Closed SC Primary / Secondary Vacuum Relief (16-19-12A,16-19-12B) 2 NA Closed Process Control Rod Hydraulic Return Check Valve (3-181) 3 Containment Air Sampling (VC 23, VC 26,109-76A&B)

NA Open Process 4

5 Open CC 135.

Amendment flo. pI 61

WMS Table 4.7.2.b PRIMARY COMAINMENT IS0lJLTION VALVES VALVES NOT StilLIECT TO TYPE C I.EAKAGE TESTS Number of Power Maximum Action om Isolation Operated Valves Operating Normal Initiating Group (Note 1)

Valve Identification Inboard outboard Time (sec)

Position Signal 2

RHR Return to Supprension Pool (10-39A,5) 2 70 closed SC 2

RilR Return to Suppression Pool (10-34A,B) 2 120 Cloned SC 2

RIIR Drywell Spray (10-26A, B & 10-31A,5) 4 70 Closed SC 2

RilR Suppression Chamber Spray (10-38A,B) 2 45 Closed SC 3

Containment Air Compressor Suction (72-38A,B) 2 20 Open CC 4

RIIR Shutdown Cooling Supply (10-18,10-17) 1 1

28 Closed SC 4

RilR Reactor IIcad Cooling (10-32, 10-33) 1 1

25 Closed SC Feedwater Check Valves (2-28 A,B) 2 2

NA Open Proc.

Reactor Head CoolinR Check Valve (10-29) 1 NA Closed Proc.

Standby Liquid Control Check Valves (11-16, 11-17) 1 1

NA Closed Proc.

liydrogen Monitoring (109-75 A,1-4; 109-75 B-D,1-2) 10 NA NA fiA I

Sampling Valves - Inlet Ilydrogen Monitoring (VG-24, 25, 33, 34) 4 NA NA NA

• These valves are remote manual sampling valves which do not receive an isolation signal.

Only one valve in each line is required to be operable.

Amendment No. JS 61 136

3 10 (cont'd)

In the event that both startup transformers are lost, adequate power is available to operate the emergency safeguards equipment from either of the emergency diesel generators or from either of t.he delayed-access offsite power sources.

Also, in the event that both emergency diesel generators are lost, adequate power is available immediately to operate t.he emergency safeguards equipment from at least. one of the startup transfonners or from either of the delayed-access offsite power sources within six hours. The plant is designed to accept one hundred percent load rejection without adverse effects to the plant or the transmission system. Network stability analysis studies indicate that the loss of the Vermont Yankee unit will not cause instability and consequent tripping of the connecting 345 kv and 115 kv lines. The Vernon feed is an independent source. Thus, the availability of the delayed-access offsite power sources is assured in the event of a turbine trip. There fore, reactor operation is permitted with the startup transformers out of service and with one diesel generator out of service provided the NHC is notified immediately of the event and restoration plans.

Either of the two station batteries has enough capacity to energize the vital buses and supply d-c power to the other emergency equipment for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without being recharged.

In addition, two 24 volt ECCS Instrumentation batteries supply power to instruments that provide automatic initiation of the ECCS and some reactor pressure and level indicat, ion in the control room.

Due to the high reliability of battery systems, one of the two batteries may be out of service for up to three days.

This minimizes the probability of unwarranted shutdown by providing adequate time for reasonable repairs. A station battery, ECCS Instrumentation battery, or an Uninterruptible Power System battery is considered inoperable if more than one cell is out of service. A cell will be considered out of service if its float voltage is below 2.13 volts and the specific gravity is below 1.190 at 770F.

The battery room is ventilated to prevent accumulation of hydrogen gas.

With a complete loss of the ventilation system, the accumulation of hydrogen would not exceed 4 percent concentration in 16 days. Therefore, on loss of battery room ventilation, the use of portable ventilation equipment and daily sampling provide assurance that potentially hazardous quantities of hydrogen gas will not accumulate.

C.

The minimum diesel fuel supply of 25,000 gallons will supply one diesel generator for a minimum of seven days of operation satisfying the load requirements for the operation of the safeguards equipment. Additional fuel can be obtained and delivered to the site from nearby sources within the seven-das period.

4.10 AUXILIAHY ELECTHICAL POWEH SYSTEMS Bases:

A.

The monthly tests of the diesel generators are conducted to check for equipment failures and deterioration. The test of the undervoltage automatic starting circuits will prove that each d.esel will receive a start signal if a loss of voltage should occur on its emergency bus.

The loading of each diesel generator is conducted to demonstrate proper operation at less than the continuous rating and at equalibrium operating conditions. Generator experience at other generator stations indicates that the testing frequency is adequate to assure a high reliability of operation should the system be required.

Amendment No. #

170

VYNPS Table 3.11-2 MCPR OPERATING LIMITS Value of "N" Fuel Type in Exposure Range RBM Equation (1) 8x8 8 x ER P8x8R BOC to EOC-2 CWd/t 42 1.21 1.26 1.27 41 1.21 1.22 1.23 l

40:

1.21 1.21 1.22 39 1.21 1.21 1.21 EOC-2 CWd/ t to EOC-1 GWD/t 42 1.26 1.26 1.28 41 1.26 1.26 1.28

<40 1.26 1.26 1.28 EOC-1 GWd/t to EOC 42%

1.29 1.29 1.31 41 1.29 1.29 1.31

._ 0 1.29 1.29 1.31

<4 (1) The Rod Block Monitor trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.

(2) The current analysis for MCPR Operating Limits do not include 7 x 7 fuel.

On this basis further evaluation of MCPR operating limits is required before 7 x 7 fuel can be used in Reactor Power Operation.

Amendment No. 61 180-01

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