ML20205S312

Proposed Tech Specs Revised Bases Pages 90,227,164 & 221a, Accounting for Change in Reload Analysis,Reflecting Change in Condensation Stability Design Criteria & Accounting for More Conservative Calculation

ML20205S312
Person / Time
Site:	Vermont Yankee
Issue date:	04/15/1999
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20205S307	List: ML20205S303 ML20205S312
References
NUDOCS 9904260177
Download: ML20205S312 (10)

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April 15,1999 BVY 99-55 Revision of Technical Specification Bases Pages Marked-up Basea Pages l

9904260177 990415 i PDR ADOCK 05000271 h P PDR n.

r VTNPS BASES: 3.3 & 4.3 (Cont'd)

2. The control rod housing support restricts the outward movement 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 RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM 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 RWM inoperable, if the reactor is suberitical and all other rods are fully inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything but minor consequences, l 4. h

/e n [ Pe r[ [An_a s Reppfrt

5. The Source Range Monitor (SRM) system 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 I the initial neutron flux. The requirement of at least three counts J per second assures that any transient, should it occur, begins at or above the initial value of 10~8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During i' reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod 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 i testing of the RBM system prior to withdrawal of such rods will provide added assurance that improper withdrawal does not occur.

It is the responsibility 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.

< v v U c N'* * ""

Refer % Ac. Geserst Eteew' %Ad ^M Fun ( GE m R. It), " 14ED E. - 2Aott - P - A , (h Ab* IMC- *M*s -

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h Amendment No. e4, 49, 64, 44, 444; EVy91-56 90

4 VYNPS BASES: 3.7 (Cont'd)

In conjunction with the Mark I Containment Long-Term Program, a plant unique analysis was performed (see Vermont Yankee letter, dated April 27, 1984, transmitting Teledyne Engineering Services Company Reports, TR-5319-1, Revision 2, dated November 30, 1983 and TR-5319-2, Revision 0) which demonstrated that all stresses in the suppression chamber structure, including shell, external supports, vent system, internal structures, and attached piping meet the structural acceptance criteria of NUREG-0661. The maintenance of a drywell-suppression chamber differential pressure of,1.7 psid and a suppression chamber water level corresponding to a downcomer ,

submergence range of 4.29 to 4.54 ft. will assure the integrity of l the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.

Using a 50'F rise (Section 5.2.4 FSAR) in the suppression chamber water temperature and a minimum water volume of 68,000 ft', the 170'F temperature which is used for complete condensation would be approached only if the suppression pool temperature is 120*F prior to the DBA-LOCA. Maintaining a pool temperature of 100'F will assure that the 170*F limit is not approached.

Expe mental da a indic e that pxy ssive at condensi be oided if he peak temperaturf of the s pression ol is loadscan]

ma ntained b ow 160* during a period relief va e operati w th sonic nditio at the scharge e it. Speci cations ha e een place on the nvelope reactor perating c nditions so that he react can be depressur zed in a imely mann r to avoid he regime of potentia ly high uppression chamber loadings.

[In dditjhn to/the l$thits 4n temp 6rature 4f the /suonr/ssion /hmmWr3 O ol wat'erJgieratino procedures define the action to be taQ)1n the event a relief valve inadvertently opens or sticks opend This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers,  ;

(3) initiate reactor shutdown, and (4) if other relief valves are 1 used to depressurize the reactor, their discharge,shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

Double isolation valves are provided on lines which penetrate the l primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of the isolation valves are discussed in Section 5.2 of the FSAR.

Manual primary containment isolation valves that are required to be closed by the definition of Primary Containment Integrity _ may be opened intermittently under administrative controls. These controls consist of stationing a dedicated operator, with whom Control Room communication is immediately available, in the immediste vicinity of the valve controls. In this way, the penetration'can be rapidly isolated when a need for primary containment isolation is indicated.

Amendment No. M, 60, 44,1.tr dtd '/1/S&, W DVY 93"bb 164

VYNPS j&ggg,, 3.10 (Cont'd)

In the event that one off-site power source and one emergency diesel generator are unavailable,-adequate power is available to operate both emergency safeguards buses fross the operable off-site power source and to operate 100% of the minimum emergency safeguards loads from the operable diesel generator. In addition, the station blackout alternate ac source of power is capable of supplying power to the bus with the inoperable diesel generator. 'Iherefore, continued operation is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one off-site power source and one emergency diesel generator unavailable.

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 indication 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 l is considered inoperable if 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 77'F.

The Battery Room is ventilated to prevent accumulation of hydrogen gas. With a complete loss of the ventilation system, the M lation of hydrogen would not exceed 4 percent concentration in 5' days. Therefore, on loss of Battery Room ventilation, the use of table 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-day period.

Amendment No. M, M, 4W, d"E 'is-51.) Vf % 55' 221a (Nit 6 Ntt[ [7/h

I -

l '

VYNPS

~

ngg.

3.11 FUEL RODS A. Averace Planar Linear Heat Generation (APLHGR)

Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.

(Note: All exposure increments in this Technical Specification j section are expressed in tenns of megawatt-days per short  ;

con.)

The MAPLHGR reduction factor for single recirculation loop  ;

operation is based on the assumptzon that the coastdown flow from the  !

unbroken recirculation, loop would not be available during a postulated large break in the active recirculation loop.fa dis ss pan,Dp*3034rg, f* er.ypht YphkeeAfuc44ar Pper Atatpon .ydng Lo j (werarlorv.

• Fepruary 1985./ See h ym Luth ReF+ 9e A* '

Linear Heat Generation Rate (LHGR) cic),,_3fut  : c. r e du c+ta n A c++ f -

B.

Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.

C. Minimum Critical Power Ratio (MCPR) j l'

Operatine 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 justified by the analyses, the results of whirh are presented in the current cycle's(Geee- '.9. Report. Refer to the appropriat topical reports listed in Specification 6.7.A.4 for analysis a hods. The increase in MCPR operating limits for i single loo operation accounts for increased core flow l l measuremenu and TIP reading uncertainties.

i

,,.- Su km M R.e\..a L tceasty l Amendment No. u, o, M, H, H, 94. He, H4,450, svY 99-55 227

1 April 15,1999 BVY 99-55 Revision of Technical Speci0 cation Bases Pages Revised Bases Pages l

l i

L

~

VYNPS l

BASES: 3.3 & 4.3 (Cont'd)

~

l 2. The control rod housing support restricts the outward movement of a l control rod to less than 3 inches in the extremely remote event of l 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 I 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 RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence. Continuing the startup ir. creases core powcr, reduces the rod worth and reduces the consequences of i dropping any rod. Withdrawal of rods for testing is permitted ~s- t i the RWM inoperable, if the reactor is subcritical and all other i rods are fully inserted. Above 20% power, the RWM is r.ot needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in anything  ;

but minor consequences. j i

4. Refer to the " General Electric Standard Application for Reactor Fuel (GESTAR II) ," NEDE-24 0ll-P-A, (the latest NRC-approved version will be listed in the COLR).

5. The Source Range Monitor (SRM) system 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 counts per second assures that any transient, should it occur, begins at or above the initial value of 10 4 of rated power used in the analyses of transients from cold conditions. One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single' control rod 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 RBM system prior to withdrawal of such rods will provide a'ded assurance that improper withdrawal does not occur.

It is the responsibility 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, l

Amendment No. G4, 44, 64, 44, 444, BVY 99-55 90

a l

VYNPS BASES: *3.7 (Cont'd)

In conjunction with the Mark I Containment Long-Term Program, a plant unique analysis was performed (see Vermont Yankee letter, dated April 27, 1984, transmitting Teledyne Engineering Services Company Reports, TR-5319-1, Revision 2, dated November 30, 1983 and TR-5319-2, Revision 0) which demonstrated that all stresses in the suppression chamber structure, including shell, enternal supports, vent system, internal structures, and attached piping meet the structural acceptance criteria of NUREG-0661. The maintenance of a drywell-suppression chamber differential pressure of 1.7 psid and a suppression chamber water level corresponding to a downcomer submergence range of 4.29 to 4.54 ft. will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.

Using a 50'F rise (Section 5.2.4 FSAR) in the suppression chamber water temperature and a minimum water volume of 68,000 ft', the 170'F temperature which is used for complete condensation would be approached only if the suppression pool temperature is 120'F prior to the DBA-LOCA. Maintaining a pool temperature of 100'F will assure that the 170'F limit is not approached.

In the event a relief valve inadvertently opens or sticks open, operating procedures define the action to be taken. This action would include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

Double isolation valves are provided on lines which penetrate the primary containment and open to the free space of the containment.

Closure of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident. Details of'the isolation valves are discussed in Section 5.2 of the FSAR.

Manual primary containment isolation valves that are required to be closed by the definition of Primary Containment Integrity may be opened intermittently under administrative controls. These controls consist of stationing a dedicated operator, with whom Control Room communication is immediately available, in the immediate vicinity of the valve controls. In this way, the penetration can be rapidly '

isolated when a need for primary containment isolation is indicated.

I Amendment No. 44, 64, 44, Lt r d:d '/1/9 5, 446, BVY 99-55 164

, l I

VYNPS e l BASES: 3.10 (Cont'd) l In the event that one off-site power source and one emergency diesel generator are unavailable, adequate power is available to operate both emergency safeguards buses from the operable off-site power source and .

to operate 100% of the minimum emergency safeguards loads from the j operable diesel generator. In addition, the station blackout I alternate ac source of power is capable of supplying power to the bus with the inoperable diesel generator. Therefore, continued operation is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one off-site power source and one emergency diesel generator unavailable.

]

Either of the two station batteries has enough capacity to energize the vital bukes and supply d-c power to the other energency equipment i 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 I Instrumentation batteries supply power to instruments that provide )

automatic initiation of the ECCS and some reactor pressure and level l indication in the Control Room.

l 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 tima for reasonable repairs. A station battery, ECCS Instrumentation battery, or an Uninterruptible Power System battery is considered inoperable if 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 77'F. 4 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 5 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.

l C. The minimum diesel fuel supply of 25,000 gallons will supply one l diesel generator for a minimum of seven days of operation satisfying '

the load requirements for the operation of the safeguards equipment. l Additional fuel can be obtained and delivered to the site from nearby sources within the seven-day period.

i l

Amendment No. &&, 64, 444, "VY 09-52, BVY 99-55 221a

r-l l

VYNPS l

\;

ll BASES:

-3,11 FUEL RODS l A. Average Planar Linear Heat Generation Rate (APLHGR)

Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analyses methods.

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

l The MAPLHGR reduction factor for single recirculation loop operation is based on the assumption . hat the coastdown flow from the unbroken recirculation loop would no: be available during a postulated large l break in the active recirculation loop. See Core Operating Limits l Report for the cycle-specific reduction factor. j B. Linear Heat Generation Rate (i.HGR)

Refer to the appropriate topicaA reports listed in l Specification 6.7.A.4 for analyses methads. ,

l C. Minimum Critical Power Ratio (MCPR) l Operating Limit MCPR 1 l

1. The MCPR operating limit is a cycle-dependent parameter which can j 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 occurrenct s. The MCPR 1 operating limits are justified by the analyses, the reeults of '

which are presented in the ca. rent cycle's Supplemental n91oad  !

Licensing Report. Refer to the appropriate topical reports listed in Specification 6.7.A.4 for analysis methods. The increase in MCPR operating limits for single loop operation accounts for increased core flow measurement and TIP reading uncertainties.

i Amendment No, M, M, M, M, M, M, MO, M4, MO, BVY 99-55 227