ML20086A255

Proposed Tech Specs Supporting Proposed Amend Re BWR Thermal Hydraulic Stability

ML20086A255
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	06/22/1995
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20086A252	List: ML20086A249 ML20086A255
References
NUDOCS 9507030127
Download: ML20086A255 (15)

Text

C:

. i I

VYNPS <

BASES:

2.1 FUEL CLADDING INTEGRITY A. Trip Settinos The bases for individual trip settings are discussed in the following

' - paragraphs.

1. Neutron Flux Trip Settings

a. APRM Flux Scram Trip Settino (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1593 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses are performed to demonstrate that the APRM flux scram over the range of settings from a maximum of 120% to the minimum flow biased setpoint of 54% provide protection from the fuel safety limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the AFRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.

APRM Flux Scram Trip Settino (Run Mode)

The scram trip setting must be adjusted to ensure that the LHGR 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 accomplished by increasing the APRM gain by the ratio in Specification 2.1.A.l.a, thus assuring a reactor scram at lower than design overpower conditions. For single i recirculation loop operation, the APRM flux scram trip J 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 l safety is not reduced during single loop operation. '

Analyses of the limiting transients show that no scram adjustment is required to assure fuel cladding integrity j when the transient is initiated from the operating limit i MCPR defined in the Core Operating Limits Report. l Amendment No. M . & M , M , M . M , 14, 14 ;

9507030127 950622 l PDR ADOCK 05000271  :

P PDR

Oi ,, .

-> .- 1

• VYNPS _,

'i 3.6 . LIMITING CONDITIONS.FOR 4.6' SURVEILLANCE REQUIREMENTS ['

OPERATION

3. 'The indicated core flow 3. 'The surveillance-is the sum of the flow _ requirements of 4.6.F.1 indication from each of .and 4.6.F.2 do not apply

the twenty jet pumps. to the idle loop and

'. If flow indication associated jet pumps failure occurs.for two when in. single loop- ';

or more jet pumps, operation.

'in1 mediate corrective ,

action shall be taken. 4. -The baseline data-If flow indication for required to evaluate the all but one jet pump ccnditions in cannot be obtained Specifications 4.6.F.1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> an and 4.6.F.2 shall be '

orderly shutdown shall. acquired each operating be initiated and the ' cycle. Baseline data reactor shall be in a for evaluating 4.6.F.2-cold shutdown' condition :while in single loop within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. operation shall-be -

updated as soon as-G. Sinole Looo Operation practical after entering.

single loop operation.

1. The reactor may be started and operated or operation may continue with a single recirculation loop -e provided that:

a. The designated i adjustments for APRM flux scram and rod block trip settings (Specifi- '

cations 2.1.A.l.a-and 2.1.B.1, Table 3.1.1 and ,

Table 3.2.5), rod block monitor trip I

setting (Table 3.2.5), MCPR fuel cladding j integrity safety I limit (Specifi- ]

cation 1.1.A), and 1 MCPR operating I limits and MAPLHGR limits, provided in the Core Operating Limits Report, are initiated 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.

(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. 44, 94, ++4, +++, 122-

]

)

r l

~*

i- i i

VYNPS~

J 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION

b. The requirements for avoiding ,

potentially unstable thermal t hydraulic ,

conditions defined in Technical Specification 3.6.J are met.

c. The idle loop is isolated by electrically 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.

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

l 1

-i t

l l

4 l

Amendment No. 94, 123 l

l

r 3 f ..

4 WNPS t

3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMDITS OPERATION r

h I

This page has been deleted.

I I

Amendment No. 94, 124

l. .

4 VYNPS-3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMD4TS OPERATION

! i ,

This page has been deleted.

1 l

l 1

Amendment No. 94, 125

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMDITS OPERATION H. Recirculation System

1. Operation with one -

i recirculation loop is l

permitted according to '

l Specification 3.6.G.I. '

2. With no Reactor Coolant System recirculation loops in operation, initiate measures such that the unit is in hot k=

shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. _

=

Amendment No. M, 92, 94, - 126

f

... j VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS 1 OPERATION 1

]

l l-This page has been deleted.

l 1

I Amendment No. 94, - 127

'4 VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION i

l I. Shock Suppressors (Snubbers) I. Shock Suppressors (Snubbers)

1. Except as noted in 1. Each snubber shall be l *. 3.6.I.i and 3.6.I.3 demonstrated operable by

! below, all required performance of the safety-related snubbers following inspection i shall be operable program. ,

I whenever its supported  ;

system is required to be a. Visual Inspections operable.

Visual inspections I

2. With one or more shall be performed required snubbers in accordance with  :

inoperable, within the following 72 hourr, replace or schedule:

restore the snubber to operable status and perform an engineering No.

evaluation per Inoperable Specification 4.6.I.lb Snubbers Per Next Required' and c, on the supported Inspection Inspection component. In all Period Intervals cases, the required snubbers shall be made 0 18 months 125%

operable or replaced 1 12 months 25%

l prior to reactor 2 6 months 125%

startup. 3, 4 124 days 125%

5, 6, 7 62 days 125%

3. If the requirements of 8 or more 31 days 125%

3.6.I.1 and 3.6.I.2 cannot be met, the The snubbers may be support ed system shall categorized into be declared inoperable two groups: the and the appropriate accessible and l action statement for those inaccessible l that system shall be during reactor followed. operation. Each group may be l inspected independently in accordance with the above schedule.

l Amendment No. -G4, 69, 94, 128

, 11

, Hs i 4

c. .!

i

. VYNPS

'. l t

3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION  !

t ~ ' J. Thermal Hydraulic Stability J. Thermal Hydraulic Stability i

1. When the reactor' node  ;

h .,

switch is in RUN - l

' i

a. Under normal (

operating i

,- conditions the  !

reactor shall not I intentionally be i operated within.the  !

l' power flow exclusion region I defined in Core .

Operating Limits  ;

Report (COLR).

I

b. If the reactor has  :,

entered the power  !

flow exclusion  ;

region (COLR), the I operator shall immediately insert

) -control rods and/or-increase 1 recirculation flow  !

to establish ,

operation outside  !

of the region, j i

-f P

'I t

i

'I I

-l

.i f

f t

i I

l l

I j

i

)..

- Amendment No. 39, 64, 94, 94, 134 l

l i

, - - - - , --v - , ,-, ,,-,- , , .-, -l

, , \

I i

WNPS i 1

l l

\

j t

This page has been deleted.

t 1

t-.

t k

t

?

t i

i J

t i

i i

l t

i e

v i

Amendment No. 94, 138 l

l I

. VYNPS BASES: 3.6 and 4.6 (Cont'd)

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 summation 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 (except in the caso of single loop operation when reverse flow Thus, the indicated l is subtracted from the total jet pump flow). q 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.

c. Sincle Loop operation Continuous operation with one recirculation loop was. justified in

" Vermont Yankee Nuclear Power Station Single Loop Operation",

NEDO-30060, February 1983, with the adjustments specified in Technical Specification 3.6.G.l.a.

l 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.

Under single loop operation, the flow control is placed in the manual mode to avoid control oscillations which may occur in the recirculation flow control system under these conditions.

H. Fecirculation System Twelve hours is a reasonable period of time to reach hot shutdown conditions. Operation of the reactor may not occur without forced recirculation flow.

Amendment No. 44, 92, 94, i41, 144

l

s. . .- :

. VYNPS j

- BASES: 3.6 and 4.6 (Cont'd)-

.I Shock Suppressors (Snubbers)

I. ,

All snubbers are required operable to ensure that the structural integrity of the Reactor Coolant System and all other safety-related ,

systems is maintained during and following a seismic or other event L

4 . initiating dynamic loads.

-The visual inspection frequency is based upon maintaining a constant

. level of.anubber 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 determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen 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 snubbers 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 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 snubber 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 mode of failure 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 testing of additional units.

Amendment No. G4, 49, 69, 94, 145.

- VYNPS BASES: 3.6 and 4.6 (Cont'd)

J. Thermal Hydraulic Stability The reactor design criteria is such that therwal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of an instability, a power / flow exclusion region to be avoided during normal operation is calculated using the approved methodology as stated in Specification 6.7.A.4. Since the exclusion region may change each fuel cycle, the limits are contained in the Core Operating Limits Report. Specific directions are provided to avoid operation in this region and to immediately exit upon an entry.

Entries into the exclusion region are not part.of normal operation.

An entry may occur as a result of an abnornal event, such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though each operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow-biased scram function is designed to suppress global oscillations, the most likely mode of oscillation, prior to exceeding the fuel safety limit. While global oscillations are the most likely mode, protection from out-of-phase oscillations are provided through avoidance of the exclusion region and administrative controls on reactor conditions which are primary factors affecting reactor stability.

Amendment No. - 145a

e i

. VYNPS The dope assignment to various duty functions may be estimates based on pocket dosimeter, TLD or film badge measurement. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

3. Monthly Statistical Report Routine reports of operating statistics and shutdown '

experience shall be submitted on a monthly basis to the Office of Management Information and Program Control, U.S.

Nuclear. Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, to arrive no later than the fifteenth of each month following the calendar month covered by the report. These reports shall include a narrative summary of operating experience during the report period which describes the operation of the facility.

4. Core Operatino Limits Report The core operating limits shall be established and documented in the Core Operating Limits Report (COLR) before each reload cycle or any remaining part of a reload cycle for the following: (a) The Average Planar Linear Heat Generation Rates (APLHGR) for Specifications 3.11.A and 3.6.G.la, (b) The Kr core flow adjustment factor for Specification 3.11.C., (c) The Minimum Critical Power Ratio (MCPR) for Specifications 3.11.C and 3.6.G.la, (d) The Linear Heat Generation Rates (LHGR) for Specifications 2.1.A.la, 2.1 B.1, and 3.11.B, and (e) The Power / Flow Exclusion Region for Specifications 3.6.J.la and 3.6.J.1b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

Report, E. E. Pilat, " Methods for the Analysis of Boiling Water Reactors Lattice Physics," YAEC-1232, December 1980 (Approved by NRC SER, dated September 15, 1982).

Report, D. M. VerPlanck, " Methods for the Analysis of Boiling Water Reactors Steady State Core Physics,"

YAEC-1238, March 1981 (Approved by NRC, SER, dated September 15, 1982).

Report, J. M. Holzer, " Methods for the Analysis of Boiling Water Reactors Transient Core Physics," YAEC-1239P, August 1981 (Approved by NRC SER, dated September 15, 1982).

t Report, S. P. Schultz and K. E. St. John, " Methods for the Analysis of Guide Fuel Rod Steady-State Thermal Effects (FROSSTEY): Code /Model Description Manual," YAEC-1249P, April 1981 (Approved by NRC SER, dated September 27, 1985).

I I

Amendment No. 49, 44, 64, 95, 414, 267

C -

3 -- g

's i . 5 s .[ .

VYNPS.

sh Letter from L'.A. Tremblay,'Jr. (VYNPC) to USNRC,

• Supplemental.Information to VYNPC April 19, 1990 Response i Regarding FROSSTEY-2 Fuel Performance Code,",EVY 90-054,

-j dated May 10, 1990 (Approved by NRC SER, dated September 24, ,

1992).  !

. ' Letter from L.-A. Tremblay,' Jr. (VYNPC) to USNRC, " Responses t 'to Request for Additional Information on FROSSTEY-2 Fuel f- Performance Code," BVY 91-024, dated March 6, 1991 (Approved h by NRC SER, dated September 24, 1992).

N Letter from L. A. Tremblay, Jr. (VYNPC) to USNRC, "LOCA . l Related Responses to Open Issues en FROSSTEY-2 Fuel Performance Code,* BVY 92-39, dated March 27, 1992 (Approved

.by NRC SER, dated September 24, 1992).

Letter from L. A.'Tremblay, Jr. (VYNRC) to USNRC, 'FROSSTEY-- .

2 Fuel-Performance Code - Vermont Yankee Response to Remaining Concerns," BVY 92-54, dated May 15, 1992 (Approved by NRC SER, dated September 24, 1992).

Report, ' Loss-of-Coolant-Accident Analysis for Vermont Yankee Nuclear Power Station," NEDO-21697, August 1977, as amended (Approved by NRC SER, dated November 30, 1977).

Report, " General Electric Standard Application for Reactor Fuel (GESTARII)," NEDE-24011-P-A, GE Company Proprietary (the latest NRC-approved version will be listed in the COLR). i Report, General Electric Nuclear Energy,-"BWR Owner's Group Long-Term Solutions Licensing Methodology," NEDO-31960, June 1991 (Approved by NRC SER, dated July 12, 1993).

Report, General Electric Nuclear Energy, "BWR Owner's Group '

Long-Term Solutions Licensing Methodology," NEDO-31960, i

' Supplement 1, March 1992 (Approved by'NRC SER,' dated July 12, 1993).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic' limits, ECCS limits,. nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are' met'. The COLR, including-any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

B. Reportable Occurrences This section' deleted.

C. Unique Reporting Requirements

1. Annual Radioactive Effluent Release Report

a. Within 90 days after January 1 of each year,'a report shall be submitted covering the radioactive content of, effluents released to unrestricted areas during the previous calendar year of operation.

P I

l

< Amendment No. M, M, %, 4%, -1M, 4M, H4, 270-t .

. . . .. , .. _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _