ML20205F174

Amend 94 to License DPR-28,changing Tech Specs to Permit Extended Reactor Operation W/One Recirculation Loop Out of Svc

ML20205F174
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	08/08/1986
From:	Muller D Office of Nuclear Reactor Regulation
To:	Vermont Yankee
Shared Package
ML20205F168	List: ML20205F161 ML20205F174 ML20205F195
References
DPR-28-A-094 NUDOCS 8608190109
Download: ML20205F174 (46)

Text

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h UNITED STATES

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g NUCLEAR REGULATORY COMMISSION j

WASHINGTON, D. C. 20555

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VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. DPR-28 1.

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

A.

The application for amendment by Vermont Yankee Nuclear Power 5

Corporation (the licensee) dated March 12, 1986, as supplemented t

March 27 and May 9, 1986, complies with the standards and g

requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. OPR-28 is hereby amended to read as follows:

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. (2) Technical Specifications The Technical Specifications, contained in Appendix A, as revised through Amendment No. 94

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its 1ssuance.

OR THE NUCLEAR REGULAT Y COMMISSION

(

/

.Af4 Daniel R. Muller, Project Director BWR Project Directorate #2 Division of BWR Licensing

Attachment:

Changes to the Technical 6

Specifications 14 Date of Issuance: August 8, 1986

O ~

Q' ATTACHMENT TO LICENSE AMENDMENT NO. 94 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Revisa the Appendix A Technical Specifications as indicated below. The revised areas are indicated by margional lines.

Pages Deleted Pages Inserted 5

5 5-a 5-a 5-b 5-b

%c 6

6 6-a 6-a 8

8 9

9 14-a 14-a 14-b 14-b 14-c 14-c 19 19 21 21 47 47 48 48 64-a 64-a 65 65 110 110 110-a 110-a i

110-b 110-b 110-c 110-c 110-d 110-d 110-e 110-e 110-f 110-f 110-9 110-h 110-1 110-j 111-c.

124 124 124-a 124-b 125 125 180-a 180-a 180-c 180-c 180-d 180-d 180-h 180-h 180-n 180-n 4

180-n1 180-n1 180-n2 180-n2 180-n3 180-n3 180-n5 180-n5 180-01 180-01

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VYWPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FURL CLADDING INTEGRITY Applicability:

Applicability:

Applies to the interrelated variable associated with fuel thermal behavior.

Applies to trip setting of the instruments and devices which are provided to prevent the nuclear system safety limits from being exceeded.

4 Objective:

Objective:

To establish limits below which the integrity of the fuel cladding is preserved.

To define the level of the process variable at i

which automatic protective action is initiated.

Specification:

A.

Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow >10% of Rated)

A.

Trip Settings

)

ha the reactor pressure is >800 psia and The limiting safety system trip settings core flow is >10% of rated, the existence of shall be as specified below:

a Minimum Critical Power Ratio (MCPR) less than 1.07 (1.08 for single loop operation) 1.

Neutron Flux Trip Settings i

shall constitute violation of the fuel i

cladding integrity safety limit.

a.

APRM Flux Scram Trip Setting (Run i

Mode) h n the mode switch is in the RUN position, the APRM flux scram trip setting shall be as shown on Figure 2.1.1 and shall be:

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j Amendment No. sa, se, 94 5

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t-VYNPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING S

0 1 66(W4 W) + 54%

here:

j S=

setting in percent of rated thermal power (1593 MWt)

W=

percent rated two loop l

drive flow where 100%

rated drive flow is that flow equivalent to 48 x 106 lbs/hr core flow i

AW=

difference between two loop and single loop drive flow at the same core flow. This l

difference must be

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accounted for during single loop operation.

AW = 0 for two loop operation.

In the event of operation with the ratio of MFLPD to FRP greater than 1.0, the APRM gain shall be increased by the ratio: MFLPD l

FRP 1

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Amendment No. 84. 98394 5-a

'O#y VYWPS 1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING,

where:

MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 W/ft l

for 8 x 8 fuel.

FRP = fraction of rated power (1593 lett).

)

In the event of operation with the ratio of NFLPD to FRP equal to or j

less than 1.0, the APEN gain shall be equal to or greater than 1.0.

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For no combination of loop l

recirculation flow rate and core theriaal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated theriaal power.

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Amendment No. sa, is, se,94 5-b i

'94y yyyps 1.1 SAFETY LIMIT 2.1 LIMITIM SAFETY SYSTEM SETTIK l

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Flux Scram Trio Settina (Refuel or Startup and Hot Standby Mode)

When the reactor mode switch is in the REFUEL or STARTUP position, average power range monitor (APRN) scram shall be set down to less than or equal to 15% of rated neutron flux (excePt as allowed by l

Note 12 of Table 3.1.1)'.

The INN flux scram setting shall be set at less than or equal to 120/125 of full scale.

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1.1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING t

3.

Core Theriaal Power Limit (Reactor pressure B.

APRM Rod Block Trip Setting g 800 psia or Core Floor i 10% of Rated) r 1.

The APRM rod block trip setting shall be When the reactor pressure' is 1800 psia or as shown in Figure 2.1.1 and shall be:

core floor 110% of rated, the core thermal power shall not exceed 25% of rated thersel SRB 10.66(W-AW) + 42%

)

power.

where:

C.

Power Transient Sgg =

rod block setting in percent of rated thermal To ensure that the safety limit established power (1593 Indt) i i

in Specification 1.1A and 1.1B is not exceeded, each required scram shall be W = percent rated two loop drive l

initiated by its expected scram signal. The flow where 100% rated drive safety limit shall be assumed to be exceeded flow is that flow equivalent l

when scram is accomplished by means other to 48 x 10' lbs/hr core flow than the expected scram signal.

A W = difference between two loop and single loop drive flow at i

the same core flow. This i

difference must be accounted j

for during single loop operation. A W = 0 for two l

loop operation.

In the event of operation with the ratio of NFLPD to FRP greater than 1.0, the 4

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i Amendment No. 84, se. 94 6

' s~ # y yy,ps 1.1 SAFETY LIMIT 2.1 LIMITING s/JETY SYSTEM SETTINC APRN gain shall be increased by the ratio: IsrLPD FRP where:

MFLPD = maxisman fraction of limiting power density where the limiting power density is 13.4 W/f t for 8 x 8 fuel.

fraction of rated power FRP

=

(1593 ledt).

In the event of operation with the ratio of NFLPD to FRP equal to or less than 1.0. the APRM gain shall be equal to or greater than 1.0.

e Amendment No. 64 se. 94 6-a e

1 Flpro 2.b1 APRM FLOW REFT.ROG SCRAM NG APRM R00 OLOQ( St.TTINGS

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ao to ao so too 130 ftCIRCLA.ATION FLOW (X RATED)

For single loop operation, the APRM Scrom and Rod Block settinge are odjusted occording to Tel.hnical Specificatione 2.tA.to and 2.t8.1 8

AMENDMENT No. g. 94 r

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VYWPS

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Bases:

l 1.1 Fuel Claddina Intearity 4

l Refer to Section S.2 of General Electric Company Licensing Topical Report. " United States Supplement. General Electric Standard Application for Reactor Fuel". NEDE-24011-P-A-US (Most Recent Revision).

The McPR fuel cladding integrity safety limit is increased by 0.01 for single loop operation in order to account l

for increased core float measurement and TIF reading uncertainties, as discussed in " Vermont Yankee Nuclear Pouer j

Station Single Loop Operation". NELO-30060. February 1983.

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j VYNPS I

APRM Flux Scram Trip Setting (Run Mode) 9 1

h scram trip setting must be adjusted to ensure that the LHCR transient peak is not increased for any combination of

)

MFLPD and reactor core thermal power. If the scram requires a change due to an abnormal peaking condition, it will be I

accoeplished by increasing the APEN gain by the ratio in Specification 2.1.A.1.a. thus nasuring a reactor scram at lower than design overpower conditions. For single recirculation loop operation, the APRM flux scram trip setting is reduced in accordance with tha analysis presented in NEDO-30060. February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of l

safety is not reduced during single loop operation.

f i

Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity when' j

the transient is initiated from the operating limit MCPR (Specification 3.11C).

Flux Scram Trip Setting (Re' fuel or Startup and Hot Standbv Mode)

For operation in the startup mode while the reactor is at low pressure, the reduced APRM scram setting to 15% of rated Power provides adequate therasi margin between the setpoint and the safety limit. 25% of the rated.

(During an outage I

when it is necessary to check refuel interlocks, the mode switch must be moved to the startup position. Since the APRM reduced scram may be inoperable at that time due to the disconnection of the LPRMs it is required that the IRN scram and the SRM scram in noncoincidence be in effect. This will ensure that adequate thermal margin is maintained j

between the setpoint and the safety limit.) The margin is adequate to, accommodate anticipated maneuvers associated l

with station startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer.

Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat l

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% of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The reduced APRM scram remains active until the mode switch is placed in the RUE position. This switch can occur when reactor pressure is greater than 800 psig.

l N IRN system consists of 6 chambers, 3 in each of the reactor protection system logic channels. N IRN is a 5-decade instrument, which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRN 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 IBM scram trip setting of 120/125 of full scale is active in each range of the l

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Amendment No. 73, EA, 94 14-a L

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VYNPS I

IBM.

For example, if the instrument were on range 1, the scram setting would be a 120/125 of full scale for that range; likewise, if the instrument were on range 5, the scram would be 120/125 of full scale on that I

range. Thus, as the IRN is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an i

IRM scram would result in a reactor shutdown well before any safety limit is exceeded.

In order to ensure that the IRN provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels.

The most severe case involves an initial condition in which the reactor is just suberitical and the IRN system i

is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IBM channel closest to the withdrawn rod is bypassed. The results of this

+

analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.

B.

APRM Rod Block Trip Setting Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM j

system provides a control rod block to prevent rod withdrawal beyond a given point at the fuel cladding j

integrity safety limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The j

flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the i

I flow decreases for the specified trip setting versus flow relationship, therefore the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As.with the APRM scram trip setting, the APRM rod block trip setting maast be adjusted downward if the ratio of NFLPD to FRP exceeds the specified value. If the APRM rod block requires a change due to abnormal peaking conditions, it will be accomplished by increasing the APRM gain by the ratio in Specification 2.1B, thus ensuring a rod block at lower than design overpower conditions. As with the APRM flux scram trip setting, the APRM rod block trip setting is reduced for singig recirculation loop operation in accordance with the analysis presented in NEDO-30060. February, 1983. This adjustment accounts for j

the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

Amendment No. 81, 94 14-b

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o.

l YYWPS i

l C.

Reacter Low Water Level Scram i

l The reactor low water level scram is set at a point which will prevent reactor operation with the steam i

separators uncovered, thus limiting carry-under to the recirculation loops. In addition, the safety lictit is l

based en a water level below the scram point and therefore this setting is provided.

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I Amendment No. 94 1

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y TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT REOUIREMENTS Modes in Which Minimum Number Required Conditions When Functions Must be Operating Instrument Minimum Conditions For i

Operating

. Channels Per Operation Are Not i

Trio Function Trip Settings Refuel (1) Startup (12) Run Trip System (2)

Satisfied (3) l

1. Mode Switch X

X X

1 A

in Shutdown l

2. Manual Scram I

X X

1 A

i

3. IBM Cigh Flux 1120/125 I

I I(11) 2 A

i INOP X

X I(11) 2 A

4. APlet L

1 66 (W-AW)+54%

I 2

A or B High Flux 0

(flow bias)

(4)

)

Cigh Flux 115%

K I

2 A

(reduced)

INOP X

2(5)

A or B Downscale 12/125 I

2 A or B 1

l

5. Cigh Reactor 1 055 psis I

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2 A

1 j

Pressure 1

l

6. High Drywell 1 5 psig I

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2 A

2 Pressure

7. Reactor Low (6) 1127.0 inches x

x x

2 I'

Water Level i

8. Scram Discharge

-<21 gallons X

X X

2 A

(Per volume)

Volume High Level Amendment No. 39, 88, 78, 78, 79, 99, 94 19 l

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I VYMPS i

3 I

i TABLE 3.1.1 NOTES I

1.

When the reactor is suberitical and the reactor water temperature is less than 212 F, only the following trip 0

functions need to be operable:

a) mode switch in shutdown b) manual scram i

c) high flux IRM or high flux SRM in coincidence i

d) scram discharge volume high water level 2.

Whenever an instrument system is found to be inoperable, the instrument system output relay shall be tripped j

immediately. Except for MSIV and Turbine Stop Valve Position, this action shall result in tripping the trip j

system.

l 3.

When the requirements in the column " Minimum Number of Operating Instrument Channels Per Trip System" cannot be f

met for one system, that system shall be tripped. If the requirements cannot be met for both trip systems, the appropriate actions listed below shall be taken:

i a)

Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

b)

Reduce power level to IRM range and place mode switch in the "Startup/ Hot Standby" position within eight hours.

1 c)

Reduce turbine lead and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, d)

Reduce reactor power to less than 30% of rated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

l 1

4.

"W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 I

j Ibs/hr core flow. AW is the difference between the two loop and single loop drive flow at the same core flow.

This difference must be accounted for during single loop operation. A W = 0 for two recirculation loop operation.

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5.

To be considered operable an APRM must have at least 2 LPRM inputs per level and at least a total of 13 LPRM t

inputs, except that channels A, C, D, and F may lose all LPRM inputs from the companion APRM Cabinet plus one additi,nal LPRM input and still be considered operable.

6.

The top of the enriched fuel has been designated as 0 inches and provides common reference level for all vessel water level instrumentation.

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7.

Channel shared by the Reactor Protection and Primary Containment Isolation Systems.

8.

An alarm setting of 1.5 times normal background at rated power shall be established to alert the operator to abnormal radiation levels in primary coolant.

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Amendment No. ss, is, 94 21 I

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yygps TABLE 3.2.5 I

l ccNTaot ROD BLOCK INSTRUMENTATION 1

Mininum manber of l

Operable Instrsument Modes in Which Function

)

Channels per Trip Must be Operable j

System (Note 1)

Trip Function Refuel Startup Run Trio Setting Startup Range Monitor' Upscale (Note 2)

X X

15 x 105 cps (Note 3) 2 2

b.

Detector Not Fully Inserted X

X (Note 1)

Interinediate Range Monitor 1 08/125 Full Scale 2

a.

Upscale X

X 1

2 b.

Downscale (Note 4)

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1 /125 Full Scale 5

1 2

c.

Detector Not Fully Inserted I

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Average Fower Range Monitor l

2 a.

Upscale (Flow Blas)

X 10.66(W-AW)+42% (Note 5) 1 /125 Full Scale 2

2 b.

Downscale I

Rod Block Monitor (Note 6)

(Note 9)

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Upscale (Flow Blas)(Note 7)

I 1 66(W-AW)+N (Note 5) 0

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1 b.

Downscale (Note 7)

I 1 2/125 Full Scale l

j (Note 8) 1 Scram Discharge Volume X

X X

112 Callons (per volume) 1 Trip System Logic I

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i Amendment No. 64, 73, 76, 99, 94 47 1

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YYNPS l

TABLE 3.2.5 NOTES 1.

There shall be two operable or tripped trip systems for each function in the required operating mode. If the minimum number of operable instruments are not available for one of the two trip systems, this condition may exist for up to seven days provided that during the time the operable system is functionally tested immediately and daily thereafter; if the condition lasts longer than seven days, the system shall be tripped. If the minimum number of instrument channels are not available for both trip systems, the systems shall be tripped.

2.

One of these trips may be bypassed. The SRM function may be bypassed in the higher IRM ranges when the IRM upscale rod block is operable.

3.

This function may be bypassed when count rate is 1100 cps or when all IRM range switches are above Position 2.

4.

IRM downscale may be bypassed when it is on its lowest scale.

5.

"W" is percent rated two loop drive flow where 100% rated drive flow is that flow equivalent to 48 x 106 I

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lbs/hr core flow. Refer to L.C.O. 3.ll.C for acceptable values for N.

A W is the difference between the two loop and single loop drive flow at the same core flow. This difference must be accounted for during single loop

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operation. A W = 0 for two recirculation loop operation.

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I 6.

The minimum number of operable instrument channels may be reduced by one for maintenance and/or testing for I

periods not in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day period.

1 7.

The trip may be bypassed when the reactor power is <30% of rated. An RBM channel will be considered inoperable if there are less than half the total number of normal inputs from any LPRM level.

1 j

8.

With the number of operable channels less than required by the minimum operable channels per trip function j

j requirement, place the inoperable channel in the tripped condition within one hour.

l 9.

With one RBM channel inoperable:

Verify that the reactor is not operating on a limiting control rod pattern, and s.

l b.

Restore the inoperable RBM channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next houc[- "r.

l Amendment No. SA, 73, 78, 983 94

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VYNPS 3.2 (Continued)

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease below the fuel cladding integrity safety limit. The trip logic for this function is 1 out of n; e.g., any trip on one of the six APRMs. six IRMs or four SRMs will result in a rod block. The minimum instrument channel requirerants for che IBM may be reduced by one for a short period of time to allow for j

maintenance, testing or calibration. The RBM is an operational guide and aid only and is not needed for rod withdrawal.

?

For single recirculation Icop operation, the RBM trip setting is reduced in accordance with the analysis presented in NEDO-30060 February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced'during j

single loop operation.

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4 Amendment No. 25, 94 64-a 1

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• Ds 48 y

VYNPS 3.2 (Continued)

The APRM rod block trip is flow referenced and prevents a significant reduction in MCPR especially during operation at reduced flow. The APRM provides gross core protection; i.e.,

limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the fuel cladding integrity safety limit. For single recirculation loop operation, the APRM rod block trip setting is reduced in accordance with the analysis presented in NEDO-30060, February 1983. This adjustment accounts for the difference between the single loop and two loop drive flow at the same core flow, and ensures that the margin of safety is not reduced during single loop operation.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst case accident results in rod block action before MCPR approaches the fuel cladding integrity safety limit.

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus control rod motion is prevented.

To prevent excessive clad temperatures for the small pipe break, the HPCI or Automatic Depressurization System must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration and also minimizes the risk of inadvertent operation; i.e.,

only one instrument channel out of service.

e e.

Amendment No. 83, 94 65

' ts 48 y

/

VYNPS

~

3.6 LIMITING CONDITION FOR OPERATION ^

4.6 SURVEILLANCE REQUIREMENT 3.

The baseline data required to evaluate the conditions in Specifications 4.6.F.1 and S

4.6.F.2 shall be acquired each operating i

cycle.

G.

Single Loop Operation C.

Single Loop Operation 1.

The reactor may be started and operated or 1.

With one recirculation pump not.in operation, t

operation may continue with a single core flow between 34% and 45% of rated, and recirculation loop provided that:

core thermal power greater than the limit specified in Figure 3.6.4 (Region 2),

a.

The^ designated adjustments for APRM flux establish baseline APRM and LPRM(1) neutron i

scram and rod block trip settings flux noise levels prior to entering this (Specifications 2.1.A.1.a and 2.1.B.1, region, provided that baseline values have i

Table 3.1.1 and Table 3.2.5), rod block not been established since the last core monitor trip setting (Table 3.2.5), MCPR refueling. Baseline values shall be fuel cladding integrity safety limit and established with one recirculation pump not MCPR operating limits (Specifications in operation and core thermal power less than 1.1.A and 3.11.C), and MAPLHCR limits or equal to the limit specified in Figure (Specification 3.11.A) are initiated 3.6.4.

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either these adjustments must be completed or the reactor brought to Hot Shutdown.

b.

With one recirculation pump not in operation, core thermal power greater than the limit specified in Figure i

3.6.4, and core flow between 34% and 45%

of rated (Region 2 of Figure 3.6.4):

I (1)

Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be monitored.

Amendment No. 29, 94 110 I J

'Rs d' y l

l VYMPS 1

3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT i

I (1) If baseline APRM and LPRMII) l neutron flux noise levels have been established since the last core refueling, initiate action within 15 minutes such that the APRM and LPRM(1) neutron flux noise levels i

are determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and:

I (i)

If the APRM and LPRM(1) l neutron flux noise levels are less than or equal to 3 times i

their established baseline levels, continue to determine the noise levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and within 30 minutes after the completion of a core thermal power increase greater than 5% of rated core thermal power, or l

1 (ii) If the APRM and/or LPRM(1) neutron flux noise levels are greater than 3 times their established baseline levels.

initiate action within 15

' minutes such that the noise levels are restored to within the required limits within 2 I

hours by increasing core flow and/or by initiating an e

orderly reduction of core thermal power by inserting control rods.

I (1)

Detector Levels A and C of one LPRM string per core octanthpidt detector Levels A and C of one LPRM string in l

the center of the core shall be monitored.

l 110-a Amendment No. 94

i

,, 4.,

i j

I i

VYNPS 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT (2) If baseline APRM and LPRM(1) neutron flux noise levels have not i

been established since the last core refueling, initiate action l

within 15 minutes of entering this region (Region 2 of Figure 3.6.4) such that operation is outside this region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> c.

With one recirculation pump not in operation, core thermal power greater than the Itmit specified in Figure 3.6.4, and core flow less than j

34% of rated (Region 1 of Figure 3.6.4),

initiat'e action within 15 minutes of

+

entering this region such that operation is outside this region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

d.

The idle loop is isolated by l

alectrically disarming the breaker to the recirculation pump motor generator set drive motor prior to startup or, if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and until such time as the inactive recirculation loop is to be returned to service.

e.

The recirculation system controls will be placed in the manual flow control mode.

(1)

Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be monitored.

4 l

Amendment No. 94 110-b

' F~ 4* y VYNPS 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT H.

Recirculation System H.

Recirculation System 1.

With two recirculation pumps in operation, 1.

With two recirculation pumps in operation, -

with total core flow less than 45% of rated, total core flow less than 45% of rated, and and core thermal power greater than the limit core thermal power greater than the limit specified in Figure 3.6.4 (Regions 1 and 2):

specified in Figure 3.6.4 (Regions 1 and 2),

establish baseline APRM and LPRM(1) neutron a.

If baseline APRM and LPRM(1) neutron flux noise levels prior to entering these flux noise levels have been estabilched regions, provided that baseline values have since the last core refueling, initiate not been established since the last core action within 15 minutes such that the refueling. Baseline values shall be i

APRM and LPRM(1) neutron flux noise established with core thermal power less than levels are determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or equal to the limit specified in Figure and:

3.6.4

?

l (1) If the APRM and LPRM(1) neutron

{

flux noise levels are less than or i

equal to 3 times their established baseline levels, continue to determine the noise levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and within 30 minutes after the completion of a core thermal power increase greater than 5% of rated core thermal power, or (1)

Detector Levels A and C of one LPRM string per core octant plus detector Levels A and C of one LPRM string in the center of the core shall be monitored.

Amendment No. 29, dde 94 110-c

' Rs 48 y VYNPS 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT (2) If the APRM and/or LPRM(1)

I neutron flux noise levels are greater than 3 times their established baseline levels, initiate action within 15 minutes l

such that the noise levels are restored to within the required l

limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow and/or by initiating an orderly reduction of core thermal power by inserting control rods.

I b.

If baseline APRM and LPRM(1) neutron flux noise levels have not been established since the last core i

refueling, initiate action within 15 minutes of entering these regions (Regions 1 and 2 of Figure 3.6.4) such that operation is outside these regions within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

2.

Operation with one recirculation loop is permitted according to Specification 3.6.G.I.

i 3.

With no reactor coolant system recirculation loops in operation, immediately initiate an orderly reduction in core thermal power to less than or equal to the limit specified in Figure 3.6.4 (Region 3), and initiate measures such that the unit is in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l (1)

Detector Levels A and C of one LPRM string per core octan) pl,us, detector Levels A and C cf' one LPRM string in

)

the center of the core shall be monitored.

Amendment No. 94 110-d

'94 m.

• Rs af t.

VTWp3 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEIt. LANCE REQUIRWENT I.

. Shock _Suppressors (Snubbers 1_

I.

Rhock Suppressgrs (Enubbers)_

1.

Rxcept as noted in 3.6.I.2 and 3.6.I.3 below, 1.

Each sautbec shall be demonstrated operable J

all required safety-related snubbers shall be by perfotiaance of the following inspection operkble avhenever its :upported system is program.

Tequired to be operable.

s.

Visual _ Inspections 2.

With one or more required snubbers inoperab'le, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, replac. or visual inspections shall be performed in restore the snubbeic to operable etstus and accordance with the following dehedule:

perfor1m an eng!neering,.wNunfire pcc Specification 4.6.I.1b end c;, on the No. Inoperable supported component. In all cases, the Snubbers per Next Required reqcited snubbers shm11 he made operchie or Iggp g }on Parloo' Inspection Intervals replaced price to reactor startup.

0 15 nonths 125%

3.

If the requirements of 3.4.I.1 and 3.6.1.2 1

12 senths 125%

cannot be met, the suppceted system shell be 2

6 months 125%

declared inoperebie and the appropriete 3, 4 124 days 125%

action-statement for that system shall be 5,6,7 62 days 125%

followed.

8 or more 31 days 125%

The snubbers may be categorized into two groups: the accessible and those inaccessible during reactor operation.

Each group may be inspected independently in accordance with the above schedule. The inspection interval shall not be lengthened more than one step at a time. Inaccessible snubbers are required to be inspected only if the period of time in Which they'become accessible is greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

110-e Amendment No. 2A, 89, 94

' es 48 y

g 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT b.

Visual Inspection Acceptance criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired operability, and (2) that the snubber installation exhibits no visual indications of detachment from foundations or supporting structures.

Snubbers which appear inoperable as a result of visual inspections may be determined operable for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be i

generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined operable per Specification 4.6.I.c, as applicable. When the fluid port of a

(

hydraulic anubber is found to be l

uncovered, the snubber shall be l

determined inoperable unless it can be determined operable via functional testing far the purpose of establishing the next visual inspection interval.

The functional test, in this case, shall be started with the piston in the as-found condition, extending the piston rod in the tension mode direction.

Amendment No. 2A, 39, GA, se, 94 110-f

' Bs 8 y

s 4

yyypg 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREF.ENT

~

c.

Functional Tests At least once per 18 months during shutdown, a representative sample of 10%

of the snubbers in use in the plant shall be functionally tested either in place or in a bench test. For each j

snubber that does not meet the functional test acceptance criteria of j

Specification 4.6.I.1.d. an additional 10% of the snubbers shall be

]

functionally tested until no more a

failures are found or until all snubbers have been functionally tested.

Snubbers of a rated capacity greater than the capability of the testing machine shall be functionally tested as follows:

(1) the lock up and bleed velocity of the snubber valve shall be verified by testing it on a cylinder that is within the capability of the testing machine, (2) the free stroke of the cylinder shall be checked, and (3) the pressure retaining capability of the cylinder shall be checked.

l 1

e' l

1 Amendment No. 89. 94 110-g l

l l

' ts y#y VYWpS e

i 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT l

l Snubbers identified as especially i

difficult to remove or in high radiation areas shall also be included in the representative sample.

l In addition to the regular sample, j

snubbers which failed the previous l

functional test shall be retested during the next test period unless the root cause for the problem has been determined and corrective actions implemented. If a space snubber has i

been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in i

another position) and the spare snubber shall be retested during the next test period. Failure of these snubbers shall not entail functional testing of additional snubbers.

I If any snubber selected for functional j

testing either fails to lock up or fails j

to move, i.e.,

frozen in place, the cause will be evaluated and if caused by j

manrfacturer or design deficiency, all l

I s'

Amendment No. 89,94 110-h e.

i

4,

• B, 4* y VYNPS 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT 1.

Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

2.

Snubber bleed, or release rate, where required, is within the specified range in compression or l

tension. For snubbers specifically 1

required to not displace under continuous load, the ability of the i

i snubber to withstand load without displacement shall be verified.

i J.

Thermal Hydraulic Stability J.

Thermal Hydraulic Stability 1.

When the reactor mode switch i: in RUN, the reactor shall not intentionally be operated in a natural circulation mode, except as permitted by Specification 3.6.H.3, nor shall an idle recirculation pump be started with i

the reactor in a natural circulation mode.

i e

w.

Amendment No. 2A, 3s, SA, so, se, 94 110-j

' n, f,

1.

~

Figure 3.6.4 1

THERMAL POWER AND CORE ROW UMITS FOR APRM/LPRM MONITORING 70 g____

REG 10N 1 REGION 2

~.

~

~'

__1 APRM/LPRM APRM/LPRM

~::

'~

1:::'

monitoring

J:

monitoring

-_ 4___

60 - ~~:::

required

~~~-~

~

required j

~::::

for dual for dual i

d Ioop operation.

~~:::

and

-~

~~

i 2:::

---~~

~~

single loop n

ps Single loop o

i 50 -

operation not operation.

n

-_l'::-

permitted.

~~:::

::::..._.-::1.

~~~~'

d e

._-y

-y

~

~~~

a 40 -

g REGION 3

..-_y No APRM/LPRM monitoring 6

- ___.____.3f required for dual or single loop d

M

.::::::::1 operation.

t W

Et:

. _.p O

J----

0 30 --

-- ---7

~

__ 3 p

i t..____

20 20 30 40 50 60 70 CORE ROW (% RATED)

Amendment No. 94 111-c

' ts #

t-

~

VYWPS 3.6 & 4.6 (continued)

The following factors form the basis for the surveillance requirements:

A break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.

The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.

The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pusy operability in the event that the jet pumps fail the tests in Specifications 4.6.F.1 and 2.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet pump flow. The indicated total core flow is a susmation of the flow indications for the twenty individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow. Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump. Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure. occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure.

124 Amendment No. 29, 92, 94

' 8% 48 y

VYNpS 4

~

i 3.6 & 4.6 (continued)

G.

Single Loop Operation Continuous operation with one recirculation loop was justified in " Vermont Yankee Nuclear Power Station l.

Single Loop Operation", NEDO-30060, February 1983, with the adjustments specified in Technical i'

Specification 3.6.G.1.a.

1 APRM and/or LpRN oscillations in excess of those specified in Section 3.6.G.1.b could be an indication that a condition of thermal hydraulic /neutronic instability exists and that appropriate remedial action should be taken. By restricting core flow to greater than or equal to 34% of rated, which corresponds to the core flow at tho 80% rod line with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most likely to occur is avoided (Region 1 of Figure 3.6.4).

These specifications are based upon the guidance of GE SIL #380, Revision 1, dated February 10, 1984.

During single loop operation, the idle recirculation loop is isolated by electrically disarming the recirculation pump motor generator set drive motor, until ready to resume two loop operation. This is done to prevent a cold water injection transient caused by an inadvertent pump startup.

i Under single loop operation, the flow control is placed in the manual mode to avoid control osci!1ations I

which may occur in the recirculation flow control system under these conditions.

1 i

4 e

Amendment No. 94 124-a S

N 9

9

' Ds #y VYNPS 3.6 & 4.6 (continued)

H.

Recirculation System The largest recirculation break area assumed in the ECCS evaluation was 4.14 square feet.

APRM and/or LPRM oscillations in excess of those specified in Section 3.6.H.1 could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken.

These specifications are based upon the guidance of CE SIL #380, Revision 1, dated February 10, 1984.

Specification 3.6.H.3 restricts reactor operation under natural circulation conditions in order to avoid potential thermal hydraulic /neutronic instabilities.

t I

I I

l l

l l

\\;

l l

e Amendment No. 94 124-b

i YYNpS

~

3.6.I & 4.6.I SHOCK SUpPRESSORS (SNUBBERS)

I i

All snubbers are required operable to ensure that the structural integrity of the Reactor Coolant System and all other l

octety-related systems is maintained during and following a seismic or other event initiating dynamic loads.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems.

Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to deterinine the next inspection. However, the results of such early inspections 1

l performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen j

the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

[

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other enubbers that may be generically susceptible, and verified by functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are (1) of a specific make or model, (2) of the same design, and (3) similarly located or exposed to the same environmental conditions such as temperature,

{

radiation, and vibration. These characteristics of the snubber installation shall be evaluated to determine if' l

further functional testing of similar snubber installations is warranted.

When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the l

cnubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber : node of i

fcilure has imparted a significant effect or degradation on the supported component or system.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested once each operating cycle. Observed failures of these sample snubbers shall require functional tacting of additional units.

l l

3.6.J THERMAL HYDRAULIC STABILITY I

I Cst allowing startup of an idle recirculation pump from natural circulation conditions prevents the reactivity in:ertion transient that would occur.

Amendment No. 2A, 39, 89, 94 125 l

g, 4

• ps 48 y

VYNPS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l

3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES l

Applicability:

Applicability:

The Limiting Conditions for Operation associated The Surveillance Requirements kpply to the I

with the fuel rods apply to these parameters which parameters which monitor the fuel rod operating monitor the fuel rod operating conditions.

conditions.

j Objective:

Objective:

I The Objective of the Limiting Conditions for The Objective of the Surveillance Requirements is Operation is to assure the performance of the fuel to specify the type and frequency of surveillance rods.

to be applied to the fuel rods.

Specifications:

Specifications:

A.

Averate Planar Linear Heat Generation Rate A.

Averate Planar Linear Heat Generation Rate (APLHGR)

(APLHGR)

During steady state power operation, the The APLHCR for each type of fuel as a APLHGR for each type of fuel as a function of function of average planar exposure shall be average planar exposure shall not exceed the determined daily during reactor operation at limiting values shown in Tables 3.11-1A

>25% rated thermal power.

through G.

For single recirculation loop operation, the limiting values shall be the values from Tables 3.11-1B through E and Table 3.11-1G listed under the heading

" Single Loop Operation." These values are obtained by multiplying the values for two loop operation by 0.83.

If at any time during steady state operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within Amendment No. SA, Se, 94 180-a

'Ds #

T-VYNPS l

I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.

Minimum Critical Power Ratio (MCPR) 1.

During steady-state power operation, the McPR Operating Limit shall be equal or greater than the values shown on Table 3.11-2.

For single recirculation loop operation, the MCPR Limits at rated flow are increased by 0.01.

For core flows other than rated, the Operating MCPR Limit shall be the above value multiplied by Kg where Kg is given by Figure 3.11-2.

If at any time during steady-state operation it is detennined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the steady-state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor powcr shall be brought to shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

Amendment No. A7, 83, 94 180-c

• Ds #

t-VYNPS Bases:

3.11 Fuel Rods 3.11A Averate Planar Linear Heat Generation Rate (APLHGR)

Refer to Section 5.5.2 of NEDs-24011P, Amendment 3, dated March 1978.

(Note: All exposure increments in this Technical Specification Section are expressed in terms of megawatt-days per short ton).

A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1.

The MAPLHGR reduction factor of 0.83 for single recirculation loop operation is based on the assumption that the coastdown flow from the unbroken recirculation loop would not be available during a postulated large break in the active recirculation loop, as discussed in NEDO-30060, " Vermont Yankee Nuclear Power Station Single Loop Operation", February, 1983.

i l

v.

Amendment No. A7, 78 s 94 180-d l

j8 Ps y

VYNPS Bases:

3.11C Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR 1.

The MCPR Operating Limit is a cycle-dependent parameter which can be determined for a number of different combinations of operating modes, initial conditions, and cycle exposures in order to provide reasonable assurance against exceeding the Fuel Cladding Integrity Safety Limit (FCISL) for potential abnormal occurrences.

The MCPR operating limits are presented in Appendix A of the current cycle's Core Performance Analysis Report. The 0.01 increase in MCPR operating limits for single loop operation accounts for increased core flow measurement and TIP reading uncertainties, as discussed in NEDO-30060,

" Vermont Yankee Nuclear Power Station Single Loop Operation", February, 1983.

e e

e Amendment No. A7, 78, 83

• 94 180-h l

4 L

VYMPS

.a Table 3.11-1B l

MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE l

i Plant: Vermont Yankee Puel Type: 8D219 Average Planar MAPLHGR (kW/ft)

Exposure Two Loop

• Single Loop PCT Oxidation (MWD /t)

Operation Operation b

Fraction 200.0 11.4 9.5 2053.

0.021 e

1,000.0 11.5 9.5 2061.

0.021 5,000.0 11.9 9.9 2117.

0.023

,7 h

10,000.0 12.1 10.0 2164.

0.026 15,000.0 12.3 10.2 2192.

0.029 20,000.0 12.1 10.0 2189.

0.029

's 25,000.0 11.3 9.4 2077.

0.020 30,000.0 10.2 8.5 1933.

0.012 35,000.0 9.6 8.0 1704.

0.004 i

Source: NEDO-21697, Augurt 1977 (revised)

• MAPLHGR for single loop operation is obtained by multiplying MAPLHCR for two loop operation by 0.83.

Amendment No. 3e, AZ, 7e, 94 180-n h 'a,

4

C VYNPS Table 3.11-1C MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE e

Plant: Vermont Yankee Fuel Type: 8D274L Average Planar MAPLHCR (kW/ft)

Exposure Two Loop

• Single Loop PCT Oxidation j

(MWD /t)

Operation Operation

( F)

Fraction 200.0 11.2 9.3 2060.

0.019 1,000.0 11.3 9.4 2064.

0.019 5,000.0 11.9 9.9 2133.

0.024 10,000.0 12.1 10.0 2129.

0.023 15,000.0 12.2 10.1 2159.

0.025 20,000.0 12.1 10.0 2167.

0.026 25,000.0 11.6 9.6 2118.

0.023 30,000.0 10.9 9.0 2028.

0.017 35,C00.0 9.9 8.2 1896.

0.010 40,000.0 9.3 7.7 1812 0.007 Source: NEDO-21697, August 1977 (revised)

• MAPLHGR for single loop operation is obtained by smaltiplying MAPLHGR for two loop operation by 0.83.

Amendment No. 31, 94 180-n1

• $'n,

. D ' f t-a VYWPS Table 3.11-1D MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Vermont Yankee Fuel Type: 8D274H i

Average Planar MAPLHCR (kW/ft)

Exposure Two Loop

• Single Loop PCT Oxidation (MWD /t)

Operation Operation

( F)

Fraction 200.0 11.1 9.2 2052.

0.019 I

1,000.0 11.2 9.3 2050.

0.018 5,000.0 11.8 9.8 2113.

0.022 1

10,000.0 12.1 10.0 2131.

0.023 15,000.0 12.2 10.1 2161.

0.026 20,000.0 12.0 10.0 2164.

0.026 i

25,000.0 11.5 9.5 2112.

0.022 30,000.0 10.9 9.0 2029.

0.017 i

I 35,000.0 10.0 8.3 1900.

0.011 I

i 40,000.0 9.3 7.7 1815.

0.008 Source: EEDO-21697, August 1977 (revised)

• MAPLHCR for single loop operation is obtained by;mulpiplying MAPLHGR for two loop operation by 0.83.

Amendment No. 57, 94 180-n2

• 9#y e

VYNPS Table 3.11-1E MAPLHOR VERSUS AVERACE PLANAR EXPOSURE Plant: Vermont Yankee Fuel Type: 8D274 (High Gd)

Average Planar MAPLHCR (kW/ft)

Exposure Two Loop

• Single Loop PCT Oxidation l

(MWD /t)

Operation Operation

( F)

Fraction 200.0 11.1 9.2 2053.

0.019 1,000.0 11.1 9.2 2044.

0.01G 5,000.0 11.6 9.6 2092.

0.021 10,000.0 12.1 10.0 2141.

0.024 15,000.0 12.2 10.1 2165.

0.026 20,000.0 12.1 10.0 2170.

0.027 25,000.0 11.6 9.6 2119.

0.023 30,000.0 10.6 8.8 1993.

0.015 35,000.0 10.0 8.3 1751.

0.005 40,000.0 9.4 7.8 1671.

0.004 Source: NEDO-21697, August 1977 (revised)

• MAPLHGR for single loop operation is obtained by multiplying MAPLHCR for two loop operation by 0.83.

Amendment No. AZ, 7e, 94 180-n3

e VYNPS i

i a

Table 3.11-1G MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Plant: Vermont Yankee Fuel Type: 8DPS289 & P8DPB289 Avera e Planar MAPLHGR (kW/ft) 1 Exposure Two Loop

• Single Loop PCT 0xidation (MWD /t)

Operation Operation b

Fraction 200.0 11.2 9.3 2126.

0.027 1,000.0 11.2 9.3 2119.

0.026 i

5,000.0 11.8 9.8 2178.

0.030 10,000.0 12.0 10.0 2185.

0.030 15,000.0 12.1 10.0 2200.

0.032 i

i 1

20,000.0 11.8 9.8 2187.

0.031 1

25,000.0 11.3 9.4 2120.

0.025 30,000.0 11.1 9.2 2095.

0.023 35,000.0 10.4 8.6 1862.

0.008 40,000,0 9.8 8.1 1784.

0.006 Source: NEDO-21697, August 1977 (revised)

I

• MAPLHGR for single loop operation is obtained by multiplying MAPLHGR for i

j two loop operation by 0.83.

I Amendment No. 55, 7e, 94 180-n5 l

4

1. 'a,

1

' Ds M t'

Tcbis 3.11-2 MCPR Operating Limits (3)

MCPR Operating Limit for Value of "N" in RBM Average Control Rod Cycle Fuel Type (2)

Equation (1)

Scram Time Exposure Rany 8x8 8x8R P8xER 42%

Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1.29 than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.29 1.29 1.29 3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.29 1.29 1,29 than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C.I.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 41%

Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25 3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C.I.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35

<40%

Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25 than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.26 1.25 1.25 i

3.3 C.1.1 EOC-1 GWD/T to EOC 1.30 1.30 1.30 Equal or better BOC to EOC-2 GWD/T 1.25 1.25 1.25

)

than L.C.O.

EOC-2 GWD/T to EOC-1 GWD/T 1.33 1.31 1.31 3.3 C.I.2 EOC-1 GWD/T to EOC 1.36 1.35 1.35 (1) The Rod Block Monitor (RBM) trip setpoints are determined by the equation shown in Table 3.2.5 of the Technical Specifications.

(2) The current analyses for MCPR Operating Limits do not include 7x7 fuel. On this basis, further evaluation of MCPR Operating Limits is required before 7x7 fuel can be used in reactor power operation.

J (3) MCPR Operating Limits are increased by 0.01 for single loop operation.

e Amendment No. 72, 9e, 94 180-01 e