ML20155D226

Proposed Tech Specs Re Reactor Trip Sys Interlocks & Limiting Safety Sys Settings

ML20155D226
Person / Time
Site:	North Anna
Issue date:	09/30/1988
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20155D219	List: ML20155D215 ML20155D226
References
NUDOCS 8810110136
Download: ML20155D226 (61)

Text

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ATTACliMENT 1 PROPOSED TECilNICAL SPECIFICATION CliANGES (UNIT 1) 1

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6 h310110136880930 DR ADOCK 05000338 FDC

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TABIE 3.3-1 (Continued)

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RFACIDR TRIP SYSTFJi RTITRIOCZS D-AIIOWABIE g

DESIGIATIOl CXXIDITIOT SFHTO DTP VAIUES FUNCTION

-0 P-6 1 of 2 Intemediate range 1 x 10-

<3 x 10 Allows manual block of above setpoint (increasing som:e range reactor trir power level)

-11

-11 2 of 2 Intemediate rarge 5 x 10

>3 x 10 Defeats the block of source below setpoint (decreasing range reactor trip power level)

P-10 2 of 4 Power range above set-10%

<11%

Allows manual block of power point (increasing power level) range (low setpoint) and inter-U mediate range reactor trips and intermediate range rod stop.

Y Blocks source range reactor trip.

3 of 4 Power range below set-8%

>7%

Defeats the block of power point (decreasing power level) range (low setpoint) and inter-mediate range reactor trips ard intaMiate range rod stop.

Input to P-7.

P-7 2 of 4 Power range above set-10%

<11%

Allows reactor trip when any of point the following occur in more than or one loop: low flow, reactor 1 of 2 Ibrbine Inpulse Pressure equiv-

<11%

ocolant ptmp breaker open, chamber pressure above alent to 10%

undervoltage (RCP busses) or setpoint rated turbine underfrequency (RCP busses).

power Also allows reactor trip on:

pressurizar low pressure or pressurizer high level.

(Power level increasing)

TABIE 3.3-1 (Cbntintxd) 2o W$

REACIOR 'IRIP SYSITN DTTERIDCKS l

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Auggq3rz DESIGMTIOT CDNDITIOi SETFODir VAIUES FUNCTIGI ez

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P-7 3 of 4 Fower range below 8%

>7%

Prevents reactor trip when any of the following occur:

(Cont'd) setpoint low flow, reactor coolant ard punp breakers open, under-2 of 2 'Ibrbine Irpulse 8%

>7%

voltage (RCP bne), under-dLL h pressure below frequency (RCP b x w ),

setpoint pressurizer low pressure or (Power level decreasirg) pressurizer high level.

l P-8 2 of 4 Fower range above 30%

<31%

Allows reactor trip when any of v

3 setpoint the following occur: low flow in a single loop, a single reactor w

(Power level increasing) coolant punp breaker open, or a turbine trip.

3 of 4 Power range below 28%

>27%

Prevents reactor trip 1 hen any of setpoint the following occur: low flow in a single loop, a single reactor (Power level decreasing) coolant puup breaker open, or a 4

turbine trip.

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,o LIMITING SAFETY SYSTEM SETTINGS BASES of the Reactor Protection System.

This trip is redundant to the Steam Generator Water Level Low-Low trip.

The Steam /Feedwater Flow Mismatch portion of this trig is activated when the steam flow exceeds the feedwater flow by a 1.616 x 10 lbs/ hour of full steam flow at RATED THERMAL POWER.

The Steam Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltaae and Underfreauency - Reactor Coolant Pumo Busses The reactor trip due to the Undervoltage and Underfrequency on the Reactor Coolant Pump busses provide reactor core protection against DN8 as a result of loss of voltage or underfrequency to more than one reactor coolant pump.

The specifiec set points assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 0.5 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.1 seconds.

The undervoltage and underfrequency trips are automatically blocked when reactor power is below the P-7 setpoint.

Turbine Trio A Turbine Trip causes a direct reactor trip when operating above P-8.

l Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident analyses for operation of these trips.

Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

NORTH ANNA - UNIT 1 B27 l

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4 ATTACllMENT 2 PROPOSED TECHNICAL SPECIFICATION CllANGES (UNIT 2)

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TABIE 3.3-1 (Contiruxd)

N REACIOR 'IRIP SYSTFM INTERIOCFS s

OESIGIATIOT (DNDITIOi SEITOIffP VAIDES FUNCPION cz

-10

-10 N

P-6 1 of 2 Internxliate rarge 1 x 10

<3 x 10 Allows nanual block of above setpoint (increasing source range reactor trip power level)

-11 2 of 2 Intermediate range 5 x 10

>3 x 10~

Defeats the block of source below setpoint (decreasing range reactor trip l

power level)

P-10 2 of 4 Power range above set-10%

<11%

Allows manual block of power point (increasing power level) range (low setpoint) and inter-l mediate range reactor trips and intar=arliate range rod stop.

i Blocks source range reactor l

t trip.

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3 of 4 Power range belo's set-8%

>7%

Defeats the block of power l

point (<tv7 easing power level) range (low setpoint) and inter-mediate range reactor trips and inter =arliate range rod stop.

Input to P-7.

P-7 2 of 4 Iber ranrve above set-10%

<11%

Allows reactor trip when any of point the following occur in more than or one loop: low flow, reactor 1 of 2 'Ibrbine Impulse Pressure equiv-

<11%

coolant pump breaker open, cha:ier pressure above alent to 10%

undervoltage (PCP bw) or setpoint rated turbine underfrequerry A busses).

power Also allows M r trip on:

pressurizer low pressure or pressurizer high level.

(Fower level increasing)

_.________m

__._____m

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5 TABIE 3.3-1 (Continued)

U REACIOR 'IRIP SYSTH4 IIIFERIDCKS D

E Allm^aIz DESIGRTICN CONDITIO!

SEITOIfTP VAIIIES FUNCTION e

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P-7 3 of 4 Ibwer range below 8%

>7%

Prevents reactor trip when any of the followirg occur:

(Cbnt'd) setpoint low flow, reactor coolant 7.nd ptmp breakers open, under-2 of 2 'Ibrbine Ig nise 8%

>7%

voltage (HCP bum), under-dunber pressure below fre@ency (RCP bne),

I setpoint pressurizer low presmre or (Power level decreasing) pressurizer high level.

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2 of 4 Ibwer range above 30%

<31%

Allows reactcr trip when any of the setpoint followirq occur: low flow in a e

y single loop, a single reactor e

coolant pump breaker open, or a turbine trip.

(Power Icvel increasing) 3 of 4 Power rarge below 28%

>27%

Prevents reactor trip when any of setpoint the following occar: low flow in a single Icop, a single reactor (Power level decreasing) coolant ptmp breaker open, or a turbine trip.

LIMITING SAFETY SYSTEM SETTINGS BASES Vndervoltaae and Underfreauency - Reactor Coolant Pumo Busses The reactor trip due to the Undervoltage and Underfrequency on the Reactor Coolant Pump busses provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump.

The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For undervoltage, the delay is set so that the time required for a signal to reach the reactor '. rip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 0.5 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.1 seconds.

The undervoltage and underfrequency trips are automatically blocked when reactor power is below the P-7 setpoint.

Turbine Trio A Turbine Trip causes a direct reactor trip when operating above P 8.

I Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident anelyses for operation of these trips.

Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

Safety In.iection Inout from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic uannels will initiate a reactor trip upon any signal which initiates a safety injection.

This trip is provided to protect the core in the event of a LOCA.

The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.

Reactor Coolant Pumo Breaker Position Trio The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of any one pump breaker above P 8 or the opening of two or more pump breakers below P 8.

These trips are blocked below P-7.

The open/close position trips assure a reactor trip signal is generated before the low flow trip set point NORTH ANNA. UNIT 2 B 2-7

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ATTACHMENT 3 SAFETY EVALUATION IN SUPPORT OF DELETION OF REACTOR TRIP ON TURBINE TRIP BELOW 30% OF RATED i

THERH4L POWER NORTH ANNA UNITS 1 AND 2 1

1 Table of Contents List of Tab 1es.............................................

....... 2 ListofFigures....................................................3

1. 0 OISCUSSION OF PROPOSED CHANGES................................. 4 2.0 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION.............. 8 3.0 ANALYSIS OF EFFECTS AND CONSEQUENCES..........................

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4. 0 EVA LUAT ION O F RESU LT S......................................... 19 S. 0 C O N C LU S I O N S................................................... 4 9 6. 0 R E F E R E N C E S.................................................... 51 LIST OF TABLES TABLE TITLE PAGE 3.1.1 Initial Conditions and Key Safety Parameters.....................

16 4.1.1 Time Sequence of Events for Turbine Trip with Pressurizer Control (Minimum Feedback)...............................................

20 4.1.2 Time Sequence of Events for Turbine Trip without Pressurizer Control (Minimum Feedback)...............................................

20 4.2.1 Time Sequence of Events for Turbine Trip with Pressurizer Control (Minimum Feedbac k, Los s of Flow occurs).......................... 36 4.3.1 Time Sequence of Events for Turbine Trip with Pressurizer and Main Steam Bypass Control................

....................... 45 l

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L LIST OF FIGURES Figure Title Page CASE A BOC Minimum Feedback, Pressurizer Control, 40% RTP 4.1.1 Nuclear Power, Fraction of Nominal.......................

21 4.1.2 Pressurizer Pressure (psia)..............................

22 4.1.3 Pressurizer Water Volume, ( f t3).......................... 23 4.1.4 Core Flow, Fract' on of Nomi na1........................... 24 4.1.5 Core Avg. Temperature,

('F)..............................

25 4.1.6 Inlet Temperature,

('F)..................................

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4.1.7 DNBR versus T1me.........................................

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i CASE B BOC Minimum Feedback, No Pressurizer Control, 40% RTP t

4.1.8 Nuclear Power, Fraction of Nominal.......................

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4.1.9 Pre s suri zer P re s sure ( ps t a).............................. 29 i

4.1.10 Pressuri zer Water Volume, ( f t3).......................... 30 4.1.11 Core Flow, Fraction o f Nomi nal........................... 31 4.1.12 Core Avg. Temperature,

('F)..............................

32 4.1.13 Inlet Temperature,

('F)..................................

33 4.1.14 DNBR versus Time.......................................... 34 i

CASE E BOC Minimum Feedback, Pressurizer Control, 40% RTP 4.2.15 Nuclear Power, Fraction of Nominal.......................

37

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4.2.16 Pre s suri ze r Pre s sure (p s t a).............................. 38 4.2.17 Pre s suri zer Wate r Volume, ( f t3).......................... 39 4.2.18 Core Flow, Fracti on of Nomi na 1........................... 40 4.2.19 Co r e Avg. Temp e ra tu re, ( ' F).............................. 41 4.2.20 Inlet Temperature,

('F).................................. 42

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4.2.21 DN B R v e r s u s T i m e......................................... 4 3 t

CASE I BOC Minimum Feedback, No Prassurizer Control 30% RTP, PORV t

4.2.22 Nuclear Power, Fraction of Nominal.......................

46 4.2.23 Pressuri zer Pre ssure (psia).............................. 47 i

4.2.24 Co re Avg. Tempe ra ture, ( ' F).............................. 48 I

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1,0 DISCUSSION OF PROPOSED CHANGES At present, for all power levels above 10% (the P-7 permissive setpoint) of Rated Thermal Power (RTP), the North Anna nuclear reactors are tripped directly on turbine trip from a signal derived from the turbine autostop oil pressure or turbine stop valve position.

Historically, a number of reactor trips have been caused by turbine trips at low power, i.e.

below the designed 50% load rejection capability. A direct reactor / turbine trip at low power is unnecessary, has inherent costs due to increased down time, and unduly stresses plant systems. Thus, Virginia Electric and Power Company (Virginia power) is proposing a change which would allow for a block of the direct reactor trip on turbine trip below 30% of rated thermal power.

Currently, Permissive P-8 is used to enable and block reactor trip protection i

for low RCS flow conditions (Iow loop flow or reactor coolant pump breaker open) below 30% power level.

The proposed modification would rewire the Solid State Protection System so that Permissive P-8 is also used to block the reactor trip on turbine trip instead of permissive P-7.

A Virginia Power review of historical trip data shows that the most commonly occuring reactor trip on turbine trip events are well below 30% of rated thermal power.

Thus it was concluded that the use of the existing P-8 bistable to block the direct reactor trip on turbine trip will be an effective means of eliminating unneeded low power reactor trips. Direct reactor trip on turbine trip will l

still be available above 30% power. The plant's designed load rejection

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capability is 50% of full load.

Virginia Power is proposing the following Technical Specification Changes which support blocking of reactor trip on turbine trip below 30% of rated j

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thermal power. A safety evaluation of these changes is presented in subsequent sections.

Each change is discussed separately in the following paragraphs. The Specifications that follow and the changes discussed are identical for Units 1 and 2.

Discussion of Technica1' Specification Changes 3/4.3.1 Table 3.3-1 REACTOR TRIP SYSTEM INTERLOCKS Permissive P-7 is currently interlocked to block the direct reactor trip on turbine trip below 10% of RTP, Therefore, the proposed Technical i

Specification changes delete reactor trip on turbine trip from the list of 7

trips associated with Permissive P-7.

In addition, the function of P-7 has been reworded slightly for clarity.

Permissive P-8, at present, blocks the reactor trip on low flow or reactor coolant pump breaker open in a single loop below 30% of RTP.

The proposed Technical Specification change allows a block of the reactor trip on turbine f

trip below 30% of RTP, Thus, the function column of Table 3.3.1 for Permissive P-8 is changed to read as follows-0 Allows reactor trip when any of the following occurs: low flow In a single loop, a single reactor coolant pump breaker open, or i

a turbine trip (Power levels > 30% RTP increasing) i 1

Prevents reactor trip when any of the following occurs: low flow j

in a single loop, a single reactor coolant pump breaker open, or l

l a turbine trip (Power levels < 28% RTP decreasing)

Bases page B 2-7 is changed to reflect the above change in reactor trip J

system interlock P-8.

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Methodology The following three items have been considered from a safety analysis f

standpoint and are addressed in this evaluation:

a.

It has been demonstrated that the loss of external load accident initiated from 30% power would be acceptable and meet ANS Condition i

II Criteria. Thus, this analysis was performed to demonstrate the adequacy of the primary and secondary side pressure relieving l

devices and to show that the minimum DNBR is above the limit value.

b.

An analysis has been performed to show that the occurence of a loss of flow event during a loss of load, which could result from a i

i failure of a fast bus transfer to offsite power after a 30 second turbine generator motoring delay or similar events, would be a

acceptable.

Loss of load could lead to elevated inlet temperature at the time of the loss of flow; the impact on DNBR has therefore been analyzed.

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c.

An analysis has been performed to demonstrate on a best estimate basis that a turbine trip without reactor trip at reduced power will not challenge the pressurizer Power Operated Relief Valves (PORV).

Virginia Power's response to NUREG-0737 Post-TMI requirements, submitted in Reference 1, committed to a program of reducing the probability of a small break LOCA due to a stuck open PORV such that i

i it is not a significant contributor to the probability of a small break LOCA due to all causes. The results of this program were j

documented by the Westinghouse Owners Group in Reference 2. Based 1

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on the results of our analysis, the stuck open PORV will remain an insignificant contributor to the small break LOCA frequency.

Section 2 discusses the ideratification of causes and accident description.

Section 3 presents the analysis assumptions and transient response of these accidents. Section 4 discusses the results of the analyses.

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t 2.0 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION

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Major load loss on the pl int can result from loss of external electrical load or from a turbine trip.

For either case, offsite power is available for the i

l continued operation of plant components such as the reactor coolant pumps.

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j For a loss of external electrical load without subsequent turbine trip, no t

direct reactor trip signal would be generated.

The plant is designed to accept a 50% step loss of load without actuating a reactor trip.

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automatic steam bypass system (stea5 dump valves) with 40% steam dump capacity to the condenser is able to accommodate this load rejection by

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1 reducing the transient imposed upon the reactor coolant system. The reactor i

I power is reduced to the new equilibrium power level at a rate consistent with the capability of the rod control system. The pressurizer relief valves may be actuated, but the, pressurizer safety valves and the steam generator safety valves do not lift for the 50% load rejection with steaie, dump.

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Should the steam dump valves fail to open, or should their capacity be b

exceeded following a large loss of load, the steam generator safety valves may lif t and the reactor may be tripped by the high pressurizer pressure 1

j signal, the high pressurizer water level signal, or the overtemperature delta T signal.

The steam generator shell side pressure and reactor coolant

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i temperatures will increase rapidly. The pressurizer safety valves and steam generator safety valves are, however, sized to protect the reactor coolant j

system and steam generator against overpressure for all load losses without

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assuming the operation of t.he steam dump valves and steam generator PORVs, i

pressurizer spray, pressurizer power operated relief valves, automatic rod i

cluster control assembly control, or direct reactor trip resulting from J

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l turbine trip.

This capability is demonstrated by the analysis presented in Section 15.2.7, Chapter 15 of the UFSAR.

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The steam generatoe safety valve capacity is sized to remove the steam flow r

j at' the engineered safeguards design rating (2910 MWt) from the steam l

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generator without exceeding 110% of the steam system design pressure.

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pressurizer safety valve capacity is based on a complete loss of heat sink i

i with the plant initially operating at the maximum calculated turbine load along with operation of the steam generator safety valves.

The pressurizer safety valves are then able to maintain the reactor coolant system pressure l

within 110% of the reactor coolant system design pressure without taking j

credit for direct (i.e. on turbine trip) or immediate reactor trip action, l

Consequently, this incident is not sensitive to initial pressurizer level, therefore the programmed level versus power is assumed.

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For a turbine trip, the reactor would be tripped directly (unless below approximately 10% power for the current design) from a signal derived from the turbine autostop oil pressure or turbine stop valves. The auto steam dump 4

f valves would accommodate the excess steam generation. Raactor coolant I

i temperatures and pressure do rot significantly increase if the steam dump 1

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system and pressurizer pressure control system are functioning properly. If the turbine condenser was not available, the excess steam generation would j

be dumped to atmosphere through the steam generator PORVs and safety valves.

1 Additionally, main feedwater flow would be lost if the turbine condenser was j

not available. For this situation, feedwater would be maintained by the t

j auxiliary feedwater system.

Turbine trip initiation signals include:

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1. Generator Trip

2. Low Condenser Vacuum

3. Loss of Lubricating 011 l

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4. Turbine Thrust Bearing Failure j

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5. Turbine Overspeed I

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6. Manual Trip

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7. Reactor Trip i

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Normal power for the reactor coolant pumps is supplied through station service busses from station service transformers connected to the generator t

j 22KV bus. When a generator trip occurs on Unit 2, these busses are 4

automatically transferred to the reserve station service transformers supplied from offsite power, and the pumps continue to supply coolant flow 4

to the core. When a generator trip occurs on Unit 1, the generator breaker i

a opens and these busses remain powered from the station service transformers which receive offsite power. Should the generator breaker on Unit 1 fail or power be lost, these busses will automatically transfer to the reserve 1

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station service transformers. The case of a direct generator trip on turbine trip is bounded by the current analysis of a loss of offsite power to the i

station auxiliaries presented in Section 15.2.9 of the UFSAR.

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Following any turbine trip where there are no faults which require immediate tripping of the generator from the network, the generator remains connected s

j to the network for approximately 30 seconds. The reactor coolant pumps remain l

connected to this generator, thus ensuring flow for 30 seconds before any l

i transfer is made. Should a reactor trip not occur in the first 30 seconds

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fo1*owing the turbine trip (the period during which the generator is 1

i motored), a complete loss of forced reactor coolant flow should be assumed t

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l to occur due to a postulated failure in the fast bus transfer to offsite l

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j power for Unit 2.

The immediate effect of loss of coolant flow is a rapid

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increase in the reactor *oolant temperature superimposed on the already increased coolant temperature resulting from the turbine trip. This increase l

could result in DNB with subsequent fuel damage if the reactor were not tripped within a short time period following the loss of flow.

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i The following signals provide the necessary protection during a complete loss of flow accident.

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1. Reactor coolant pump power undervoltage and underfrequoney.

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2. Reactor coolant pump breakers open.

3. Low reactor coolant loop flow.

I For the analysis of this event, the two important signals are reactor coolant pump undervoltage and low reactor coolant loop flow.

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The reactor trip on reactor coolant pump undervoltage it provided to protect I

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against conditions which can cause a loss of voltage to all reactor coolant l

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pumps i.e.

station blackout, This function is blocked below approximately

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10 percent power '(Permissive P-7).

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The reactor trip on low reactor coolant loop flow is provided to protect 1

against loss of flow conditions which affects one or more reactor coolant i

loop.

Between approximately 10% power (Permissive P-7) and 30%

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(Permissue P-8), low flow in any two loops will actuate a reactor trip.

j Above 30% power (Permissive P-8), low flow in any single loop will actuate r reactor trip, 11

0 3.0 ANAL.YSIS OF EFFECTS AND CONSEQUENCES l

3.1 THE LOSS OF LOAD /TUR8INE TRIP EVENT 4

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This event is described in Section 15.2.7, Chapter 15 of the North Anna UFSAR. This transient was analyzed at 100% of rated thermal power, where it is limiting. The existing analysis shows acceptable results for the complete I

load rejection from 100% power without taking credit for the direct reactor

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t trip on turbine trip, and will bound the results at 30% power. However, an explicit analysis at 30% power was performed to support this Technical Specification change.

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In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from 40% of full power (30% of rated thermal power plus 10%

l instrumentation uncertainties) without a direct reactor trip on turbine i

trip, primarily to show the adequacy of the pressure-relieving devices and also to demonstrate core protection margins. The assumptions delay reactor j

i trip until conditions in the RCS result in a trip due to a signal other than l

turbine trip.

Thus, the analysis assumes a worst transient.

In addition l

I no credit is taken for condenser steam dumps or steam generator PORVs. Main l

j feedwater is terminated at the time of turbine trip, with no credit taken i

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i for auxiliary feedwater to mitigate the consequences of the transient, i

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This accident is analyzed by using the computer program RETRAN (Reference I

3).

The program simulates the neutron kinetics, reactor coolant system,

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l pressurizer, pressurizer relief and safety valves, pressurizer spray, steam i

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generator, and steam generator safety valves. The program computes pertinent 0

plant variables, including temperatures, pressures, and power level.

The l

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. _ = _ -.

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i COBRA (Reference 4) code is then used to calculate the minimum DNBR during l

the transient based upon the heat flux, flow, temperature, and pressure from l

RETRAN.

The WRB-1 CHF correlation is used (Reference 5).

l Assumptions are:

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1.

Initial Operating Conditions - The initial reactor power is assumed at j

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40 percent of nominal full power.

The initial reactor coolant system i

j pressure and temperatures are assumed at their nominal value consistent j

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with the steady-state 40% power level. Allowances for calibration and i

instrument errors are incorporated into the ONBR limit value as described j

in Reference 6.

These assumptions result in the maximum power difference i

for the load loss, and the minimum margin to core protection limits at l

the initiation of the accident.

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Moderator and Doppler Coefficients of Reactivity - The total loss of load l

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is analyzed for both beginning-of-life and end-of-life conditions.

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j positive moderator temperature coefficient at beginning of life and a f

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large (absolute value) negative value at end of life are used.

Least j

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negative and most negative Doppler power coefficients are used for i

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j minimum and maximum feedback, respectively.

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3.

Reactor Control - From the standpoint of the maximum pressures attained, 1

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it is conservative to assume that the reactor is in manual control.

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Steam Release - No credit is taken for the operation of the condenser 1

t steam dump system or steam generator power-operated relief valves.

The

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steam generator pressure rises to the safety valve setpoint, where steam

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release through safety valves limits secondary steam pressure at the 13

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setpoint value (the analysis assumes the valves open at the highest safety valve lift point of 1150 psia).

5.

Pressurizer Spray and Power-Operated Relief Valves - Two cases for both the beginning and end of life are analyzed:

a.

Full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant

pressure, b.

No credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure.

6.

Feedwater Flow - Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow following the reactor trip since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed to occur.

The auxiliary feedwater flow would remove core decay heat following plant stabilization.

Reactor trip is actuated by the first reactor protection system trip setpoint reached, with no credit taken for the direct reactor trip on the turbine trip.

In summary, four cases were analyzed; Case A BOC Minimum Feedback with pressurizer control Case B BOC Minimum Feedback without pressurizer control Case C EOC Maximum Feedback with pressurizer control 14

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Case D EOC Maximum Feedback without pressurizer control 3.2 THE LOSS OF LOAD /TUR8INE TRIP WITH LOSS OF FLOW EVENT l

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This is not an original design basis accident. This scenario is analyzed in i

detail and documented in this evaluation.

1 An attempted fast bus transfer is assumed 30 seconds following the loss of

)

j steam load. The transfer to an external power source is assumed to fail which a

results in a complete loss of flow transient initiated from the loss of load i

condition, i

The assumptions used in this analysis are the same as given in the previous

{

section.

l l

l In summary, four cases were analyzed; i

Case E BOC Minimum Feeaback with pressurizer control r

t Case F BOC Minimum Feedback without pressurizer control r

Case G EOC Maximum Feedback with pressurizer control

(

l Case H EOC Maximum Feedback without pressurizer control I

t The initial conditions and key safety parameters used in the analysis are given in Table 3.1.1.

l l

15

TABLE 3.1.1 INITIAL CONDITIONS AND KEY SAFETY PARAMETERS AND ASSUMPTIONS Core Power 40% of Rated Thermal Power 1157.2 Mwt Thermal Design Flow (GPM) 289200 Reactor Coolant Temperature Vessel Outlet, 'F 577 Vessel Inlet,

'F 549.6 Steam Generator Steam Temperature, 'F 537 Pressure, PSIA 924 Moderator Temperature Coef ficients

+6 pcm/F (BOC)

-59 pcm/F (EOC)

Doppler Temperature Coefficients

-1.4 pcm/F (BOC) l

-2.9 pcm/F (E0C)

Doppler Power Coefficients Least Negative (Hot Zero Power to Hot Full Power) -10.2 to -6.7 pcm/% of power Most Negative (Hot Zero Power to Hot Full Power) -20 to -14 pcm/% of power Beta-eff

.0075(BOC)

.0043(E0C)

Normalized Trip Reactivity 4% dk/k l

Generator Motoring Time 30.0 seconds i

Undervoltage Trip Delay Time 1.2 seconds 1

1 l

16

~

3.3 PRESSURIZER PORV RESPONSE EVALUATION I

J L

To demonstrate that the pressurizer PORVs are not challenged during the loss of load transient without reactor trip, more of a best estimate analysis was

[

pe rf ormed.

b i

NUREG-0737 required that the frequency of a small break LOCA caused by a i

t stuck-open PORV be reduced and that it be demonstrated not to be a

[

[

significant contributor to the probability of a small break LOCA. The loss

)

of load and loss of flow both have the potential of causing the PORV to open j

(Reference 2). Thus, an analysis was also performed to demonstrate that the

]

PORVs are not normally challenged during this event at reduced power.

The following case was analyzed; 3

j CASE I BOC Minimum Feedback with pressurizer control and credit i

taken for steam dump.

l i

i i

]

A most positive moderator temperature coefficient was used in the transient and no credit was taken for the heat transfer to the reac6or coolant system i

I metal.

Main feedwater was terminated at the time of turbine trip, with r.a credit taken for auxiliary feedwater to mitigate the consequences of the 5

a transient. Credit was taken for pressurizer spray, the condenser steam dumps d

'l and the steam generator PORVs.

i 1

In this analysis, loss of flow is not assumed. The reactor trips eventually i

on low steam generator mass. The transient is more severe than the case with 17 1

subsequent loss of flow with respect to the PORV challenge in that the latter case would result in an early reactor trip.

These assumptions make this a conservative analysis with respect to maximizing the pressure increase.

Figure 4.3.2 shows that the PORVs are not challenged using these limiting parameters.

'l 1

)

s I

f I

l 18

4.0 EVALUATION OF RESULTS 4.1 TURBINE TRIP / LOSS OF LOAD The transient responses for a total loss of load from 40% power level were analyzed in detail for four cases: two cases for the beginning of core life and two cases for the end of core life. Two cases of these analyses (minimum feedback with and without pressurizer control) were found to be limiting with respect to the other two. Thus the results of these analyses are shown in Figures 4.1.1 through Figures 4.1.14 Figures 4.1.1 thru 4.1.6 show the transient response for the turbine trip /

loss of load transient with minimum feedback. Full credit was taken for the pressurizer spray and pressurizer power-operated relief valves.

No credit is taken for the condenser steam dump and steam generator PORVs. The minimum DNBR remains well above the design limit value as shown in Figure 4.1.7 and is bounded by the loss of load analysis delineated in UFSAR Section 15.2.7.

Figures 4.1.8 thru 4.1.13 show the transient response due to the turbine trip / loss of load transient with minimum feedback. All the assumptions are the same as in the previous case except credit is not taken for the pressurizer spray and pressurizer power operated relief valves. The minimum DNBR is always above the design limit value as shown in Figure 4.1.14 and is bounded by the loss of load analysis delineated in UFSAR Section 15.2.7.

The peak pressure remains well within the design limit.

The time secuences of events for these two cases are shown in Tables 4.1.1 and 4.1.2.

19

~__

i i

i i

Table 4.1.1 Time Sequence of Events for Turbine Trip with Pressurizer Control l

l i

EVENT TIME (s) j l

Minimum Feedback (BOC)

{

Turbine Trip 0.0 i

Peak pressurizer pressure occurs 14.5

[

Initiation of steam release j

from steam generator i

safets; valves 21.1 i

)

8.ow Steam Generator mass trip l

l (simulates low-low level) 106.3 l

i j

Rod begins to fall 108.3

[

4 Mintnum DNBR occurs 108.5 i

i i

l i

i 4

j Table 4.1.2 Time Sequence of Events for a Turbine Trip w/o Pressurizer i

Control i

1 i

EVENT TIME (s) i Minimum Feedback (BOC) i j

Turbine Trip 0.0 Hi pressurizer pressure trip j

setpoint reached 12.3

]

i Rod begins to fall 14.3 l

Minimum DNBR occurs 16.0 1

Peak pressurizer pressure occurs 16.0

)

{

Initiation of steam release j

from steam generator j

safety valves 21.0 4

k l

i e

20

FIGURE 4.1.1 NA TURBINE TRIP / LOSS OF IDAD NUCLEAR POTER. FRect10N OF NOMDEL MDitMW FREDSACE PREMURIIIR Colfm0L 1.O l

N l

Uc l

kO.5 2 A

R l

p O

W E

R O.6 l

e F

R A

a T

l e

O.4 0N l

k N

l oO.2 M

IN l

A L

l 1

0.0 0

20 40 SO 80 100 120 TIME IN SECON05 womewww reseeaca 21

FHMt3 4.1J NA TURBINE TRIP / LOSS OF LOAD PRESUEZIR P6 PSE nENilMdFeenaarr F M CONTROL 2600 -;

2500 -

A RE 2400 -

SS l

2,300 -

s R

A 2200-RE i

E

^

g 2100-c A

2000-S I

4 1900 -

1

\\

isOO O

20 40 80 50 i00 i20 TIME IN SECONOS i

J ris.....

1 22 l

9 l

i l

FIGURE 4.1.3 NA TURBINE TRIP / LOSS OF IDAD PRRESUREM WATER TOLUME. FT3 MDRMUM FEpSeCE PN CONT 90L l,

4 1400 l,

1300 -

i l

1 1200-i

)

A 4

RZ 1100 -

r

.i R

W A

1000 -

r h

900 V

l O

4 L.

I 800 t

F T

3 700 j

i SOO j

SCO O

2b Ab Sb Sb iNO 120 i

TIME IN SECONOS I

l

....w. r e s e n.

I 23 l

e a

FMlURE 4.1.4 NA TURBINE TRIP / LOSS Of LDAD COM FIM. FRACTION OF Wellm&L I

IGNIMUM FtrDS&CE, PMMURIER CON 1ROL 1.2 C

1.O J

O R

E bO.5 W

l F

R A

gOS l

O N

l k O.4 N

l C

M l

t N

A O.2 L

l l

l O.O O

2h 4b ed ad ido i20 TIME IN SECONOS wi=iwww rassenes 24

i I

1 d

r m its 4.t.s NA TURBINE TRIP / LOSS OF LOAD COREATG TEMPERATU M,F MDitMUM FREDSACE. PRENURIEER ColfTROL I

650 ---

640 -

i t

S30 1

C N 820 E

a AV SiO G

.l T

)

E SOO M

A t

E I

R 590 i

A s

t 4

y 550 r

l F 570 i

k l

SSO 1

550

)

O 20 40 SO 50 100 120 i

TIME IN SECONDS 25

o i

'l 6

FNIURS 4.1.4

[

f NA TURBINE TRIP / LOSS OF LOAD INI2T TERIPM&TVRR.F MINIRMd FREDGE% FM C0ft190L I

i 600 590 550 i

1 r

i i

N 570 4

l 550 T

i E

N 550 S

A

)

T 540 N

E 530 F*

i 520 510 a

I J

500 O

2b Ab ed

'g [ '

kdo ido I

TIME IN SECONOS I

l 1

I wi=iwww ressemos 26 I

i

L

.,e h

4 9

(

4 i

i i

i i

i i

i l

?

P90 U R E 4.1.7 NA TURBINE TRIP / LOSS OF LOAD Mana==

MIIIIMUM FRESteCE. PIRESURRER C00ffROL 1,

10 1

9 S

1 7

3 1

j M,

e-N y

i 9

I M

l e

t u

5 i

\\

O t

N 4

l

)

3 l

2 l

1 i

O -

l O

20 dh 80 ab tbb 120 TIME IN SECONOS i

,u.....

1 27 1

l FIGURE 4.1 A NA TURBINE TRIP / LOSS OF LOAD NUClJAR POWM, PRMl20N OF MIENAL 1sNtaMd FREREMI. M PM CONTEDL 1.O l

N l

U C

l kO.5 2 A

R l

A C

l O6 l

r R

A lo-l M N

k N

l O O2 M

4N l

O.O O

20 O

ed ab 100 TIME IN SECONOS winewww rarosaan 28

FMPJRR 4.1.9 NA TURBINE TRIP / LOSS OF LOAD MIER PRM. PEla MDmWM FERDBMX. MS PM CollTROL i

2600 2500 p

2400 R

l

' 2300 g 2200 E

S 2100 P 2000 5

i A

1900 1800 O

20 40 SO SO 100 TIME IN SECONOS wiseawww ressenes 29

P l

l t

'1 t

l l

1 1

J l

i j

4 i

f l

i l

FIGUM 4.1.10 L

NA TURBINE TRIP / LOSS OF LOAD l

j PIRESURRER WATES TOLUME, FTt f

i EDmfUM FESDS&CE. NO PM C0trTROL 1

r i

t l

q-1400 I

1300 i

l 1

j i

1200 1

l A

i R

l g 1100

{

[

W j

t A 1000 T

900 J

v J

Q k

G aOO -

l l

r T

f 3

700 -

]

L

]

1 i

eOO -

I I

500 i,

O 2O 40 eO a0 100 j

TIMC iN SECONOS I

i a

I t

L t

4

.I 30 t

i 1

1

l t

i i

I 1

1 I

i

(

i i

\\

\\

i j

l 4

r10U18 4.1.11

)

NA TURBINE TRIP / LOSS OF LOAD come now, ruaction or noussat 4

temit'Is FEE 8 Sam NO PIMBURIBER COIITROL i

l i.2

)

i l

?

l r

l t

g 1. O i f

m l

t E

l r

t60.8 i W

6 i

l r

q A

l 1,

9 i

O j

]

j N

kO.4 i

r 1

N l

[

t o

M l

[

I i

N l

i j

A O.2 L

l i

.l l

l l

r r

l i

4 0.0 l

O 20 40 ed 50 i00 J

TIME IN SECONOS 1

1 winiwww ressense i

I 31 l

i 1

i f

4 j

i i

l FMME 4.1.18 I

NA TURBINE TRIP / LOSS OF LOAD CMB ATS TEMPER &TVM. F MDmtW FRESSAG. MS PM C0ffTIWL I

650

,f 640 f

S30 C

k 620 E

I j

AV 610 4

G l

T l

E SCO u

A 4

)

k seO AT k 550 i

E

]

F 570 i

560 i

550 J

1 O

20 40 SO 80 100 l

TIME IN SECONDS i

j i

1 wi=iwww resseAsa l

I 32 i

t F10 U R E 4.1.13 NA TURBINE TRIP / LOSS OF LOAD i

1 IN12T TEMPWTURE.F MDmWM FERDSEE NO PN CONTML J

S00 i

[

.i 590 4

350 t

)

t l

l N 570 l

4 J

560 T

l E

MA 550 k

'i A

1 T 540 I

%E 530 i

F i

)

520 a

1 510

]

)

500 i

0 20 40 ed 30 100 j

l fIME

!N SECONOS q

l I

wimiwww reseeace i

1 33

0 I

l i

i I

t i

{

nous 4.1.14 NA TURBINE TRIP / LOSS OF IAAD l

unmam awan MgmfUM FREDSMX, MO 6 CONTML iO i

9 i

4 1

5 j

1 7 -

]

M t

6

(

N I

i M

l 1

U 5

1 M

1 o

N 4

k j

3 l

1 i

2 t

1 I

O O

2b 40 eb 50 1OC l

TIME IN SECONOS 34 1

1

4.2 TURBINE TRIP / LOSS OF LOAD WITH LOSS OF FLOW j

The transient responses for a total loss of load with subsequent loss of flow from 40% power level were analyzed in detail:

two cases for the beginning of core life and two cases for the end of core life.

The case with minimum reactivity feedback and with pressurizer control is most limiting with respect to DNBR. The results of that case are presented here.

i i

i Figures 4.2.1 through 4.2.6 show the transient response. For this case, full credit was taken for the pressurizer spray and pressurizer power-operated relief valves.

No credit was taken for the condenser steam dump and steam l

l generator FORVs.

It was' also assumed that a fast bus transfer fails and a loss of flow event is initiated 30 seconds after turbine trip.

The minimum CN8R remains well above the limit value as shown in Figure a.2.7 and is bounded by the loss of load analysis delineated in UFSAR Section 15.2.7.

j The time sequence of events is shown in Table 4.2.1.

1 1

1 1

35

Table 4.2.1 Time Sequence of Events for Turbine Trip with Pressurizer Control EVENT TIME (s)

Minimum Feedback (BOC)

Turbine Trip 0.0 Peak pressurizer pressure occurs 11.0 Initiation of steam release from steam generator safety valves 21.1 Fast bus transfer failure, flow coastdown begins 30.0 Undervoltage sensor reactor trip 30.025 Rod begins to fall 31.225 Minimum DNBR occurs 33.0 4

36

1 f

l 4

N 1

I il l

1

(

i I

rNIURE 4.8.1 a

i NA TURBINE TRIP / LOSS OF FLOW i

mrtna revan, rnenow or wonoun t

MDGIMI PREDGEE. PIEDBURIIER CofffRDL i

j i.O I

l L

i l

N i

L U

C l

hO.8 2 i

A J.

R l

J pc l

O.S 2

l 1

r t

R I

A g

i O.4 i

e i

o N

l i

r k

4 N

l l

O O.2 M

i a

i N

l 4

A L

l

]

O.O l

O 20 40 Mb SO 100 TIME IN SECONOS i

4 l

i womewww asseeats 1

)

37 1

l l

4

.i a

FIGURE 4.3J l

i NA TURBINE TRIP /IASS OF FLOW P M P M.PEla MDRNUM FEEBS&CE PM C0fff90L i.

i 1

2500 i

f t

2500 i

l A

R E 2400 i

2 W

6 2300 l

I A 2200 R

E S

i g 2100 j

E l

l A 2000 5

)

A I

1900 1

1 i

1800 l

C 20 Ab SO 80 ibO TIME IN SECONOS 1

)

winiwww reseeme=

i i

i 38

9 4

P I

f FIGURE 4.3.3 NA TURBINE TRIP / LOSS OF FLOW PFESSURIZER WATER TOLUMI, Ff3 MDi1WUM FREDSACE, PRESSURIZER CONTROL 1400 1300 1200 -

P RZ 1100 R

W A 1000 e

R 900 VO L

500 P

T3 700 500 4

500 O

20 4b Sb 50 ibo TIME IN SECONOS wiuiwww resosana 39 m

FIGUltE 4.3.4 NA TUP.BINE TRIP / LOSS OF FLOW ColtE FIAW. FitaLTION OF NotsNAL MDilMUM FIEDSACE. FitBESURI2EE CON 11t0L 1.2 C

1.0 O

RE F"

LO O.8 I'

W PR A

gO.S -

I O

N

)

OF O.4 NO M

f N

A O.2 L

n O O 0

20 40 SO SO 100 TIME IN SECONDS i

wi w e ww 4 ressemos 40 t

FIGURE 4.2.5 NA TURBINE TRIP /IDSS OF FLOW CORE AfG TEMPERATURE. F MDf1 MUM FEIDEACE PRENBURIIER ColffROL 650 640 830 C

OR 820 -

E Av S10 G

~

TE 800 M

P ER 590 UR 550 E

F 570 I

560 550 0

20 40 Sb Sb 100 TIME IN SECONOS l

uswswww regeeAos 41

+ _

k 4

O FIGURE 4.3.4 NA TURBINE TRIP / LOSS OF FLOW I m m m r m m E.r MDf1 MUM FREDRACE. PRERBURI1EE CONTROL 800 590 -

l 580 N 570 -

SSO -

E MP 550 -

ERAT 540 URE 530 -

P 520 4

510 4

500 0

20 4O Sb sb 100 TIME IN SECONOS wiwawww reassaou 42 i

I i

i l

FIGURE 4.3.7 NA TURBINE TRIP / LOSS OF FLOW MDmRM DNEA MLGWUM FEIDSACE, PREssURI2ER CONTROL 1O 9 -

l l

8 -

?

M i

S N

IM U

5 -

I M

0N 4 -

c3R 3

2 1

4 0

O 20 40 so 50 10 C l TIME IN SECONOS i

)

wiwiwww reseeaan 43

O 4.3 PRESSURIZER PORV RESPONSE EVALUATION The transient was analyzed to confirm that on a best estimate basis, the pressurizer PORVs are not challenged during this transient.

Figures 4.3.1 through 4.3.3 shows the results of this transient.

The time sequence of events during the transient is given in Table 4.3.1.

l 1

l 44

Table 4.3.1 Time Sequence of Events for a Turbine Trip with Pressurizer Control EVEki TIME (s)

Minimum Feedback (BOL)

Credit is taken for the condenser steam dump and steam generator PORVs Turbine Trip 0.0 Peak pressurizer pressure occurs 2,?. 7 (2331.68 psia)

Lo Steam Generator mass trip 150.4 (simulates low-low level)

Rod begins to fall 152.4 45

e FHiUltB 4.3.1 NA TURBINE TRIP / PORY NUC12AR PCTIII. FIl&CTION OF NolGNAL IGNWUM FREDE&CE. PitBEURIZER CONTIt0L 1.0 NU CkO.S AR P

O W

ER O.6 F

R A

1 C

0.4 O

N o

P NO O.2 M

1NA O.O i

O 40 50 120 ISO 200 TIME IN SECONDS uinewww ressemos e

46

O t

PIG 4DtB 4.3.5 NA TURBINE TRIP /PORV PREB5URr5R PRal!MtE. P51A MDmmi FEEDE&CE. PRESUlu2ER CONTROL 2550 2450 -

P RCS 1

S 2350 U

I R

1 I

i k 2250 -

R P

RE 2150 -

SS URE 2O50 PS 1950 A

1 1850 O

4O Sb i20 ISO 200 TIME i N SECONOS wiwswww resoSACM 47

T a

9 I

e F10UllE 4.3.3 NA TURBINE TRIP /PORY CORI ATG TEMPEIATUllI F MDmfUM FEEDB&cK. PREBEURIZER CONT 110L SCO 590 -

C O

R

[

E A

V 580 -

G T

EM PE R 570 -

A T

U R

E V SSO 4

550 O

40 SO 120 18O 200 TIME IN SECONOS i

l wewiwww regeeAcu 48

r

.r e

5,0 CONCLUSIONS Results of the analyses show that a total loss of external electrical load without a direct or immediate reactor trip below 30% of Rated Thermal Power presents no hazards to the integrity of the reactor coolant system or the main steam system. Pressure relieving devices incorporated in the two systems are adequate to keep the maximum pressure within the design limits.

The analysis demonstrates that for a complete loss of forced reactor coolant flow initiated from the most adverse preconditions of a turbine trip, the integrity of the core is maintained by operation of the reactor protection

system, i.e.,

the ONBR will be maintained above the design limit value.

Thus, there will be no cladding damage and no release of fission products to the reactor coolant system.

The analysis also demonstrates that on a better estimate basis, the pressurizer PORVs are not challenged at any time during the transient, although conservative assumptions were used to demonstrate the pressurizer response.

The proposed changes will not have a significant impact on the frequency of a small break 1.0CA caused by a stuck-open PORV.

I

\\

49

+

5,1 10 CFR 50,59 EVALUATION i

P From the analyses presented in this report, it is concluded that the results i

of this analyses, i.e.,

minimum DNBR and peak pressurizer pressure are bounded by the design limit values and consequently that no unreviewed safety questions as defined in 10CFR50.59 exists as a result of the proposed change.

The results of this evaluation can be stated as follows:

1.

No increase in the probability of occurrence or consequences of an accident analyzed in the UFSAR will result from elimination of reactor trip on turbine trip below 30% of Rated Thermal Power (RTP).

The analyses results shows that the DNBR does not decrease below the design limit at any time.

The analysis also shows that, except under conservative assumptions, the pressurizer PORVs are not challenged

[

during the transient. Pressure relieving devices incorporated in the primary and the secondary systems are adequafe to hep the maximum pressure within the design limit. Since the predicted results are within the range of existing safety analysis values, it is concluded that operation with the proposed ~ Technical Sp~ecificaticn changes will neither increase the probability of occurrence nor the consequences of initiating events for any known accident.

2.

The possibility of a new or dif ferent accident type not previously i

considered in the UFSAR is not created by this proposed change. The complete loss of unit load without a direct reactor trip on turbine trip is a de<*gn event and is addressed in Section 15.2.7 of the UFSAR.

The results for a loss of flow due to fast bus transfer failure after a turbine trip are bounded by the results for a complete loss of flow from full power, which is discussed in Section 15.3.4 of the UFSAR.

Thus, the results of all the relevant accident analyses show that operation l

with this modification does not create the possibility of an accident of a different type than any evaluated previously in the UFSAR.

3.

The margin of safety is not reduced.

The proposed Technical Specifications changes have been incorporated in the safety analyses.

These analyses have demonstrated that calculated results meet all design acceptance criteria as stated in the UFSAR.

i 50

c o

6,0 REFERENCES 1.

Letter from B.

R. Sylvia (Vepco) to H.

R. Denton (NRC),"Response to NUREG-0737 TMI Action Items", December 15, 1980.

2.

WCAP-9804,"Probabilistic Analysis and Operational Data in Response to NUREG-0737, Item II.K.3.2 for Westinghouse NSSS Plants", Feb. 1981.

l 3.

Smith, N. A.,"VEPCO Reactor System Transient Analysis using the RETRAN Computer Code", VEP-FRD-41A, May, 1985.

l t

4.

Sliz, F. W. and Basehore, K. L.,"VEPCO Reactor Core Thermal-Hydraulic r

Analysis Using the COBRA IIIC/MIT Computer Code", VEP-FRD-33-A, October, 1983.

l 5.

Anderson, R.

C.

and Wolfhope, N.

P.,"Qualification of WRB-1 CHF i

l Correlation in the Virginia Power COBRA Code", VEP-NE-3, Nov.,1986.

6.

Anderson,"Statistical DNBR Evaluation Methodology", VEP-NE-2A, June, 1987.

I t

I l

l l

I I

I i

l l

1 51

f ATTACHMENT 4 10 CFR 50.92 Evaluation 1

1

7 w'

10 CFR 50.92 EVALUATION Virginia Power is proposing the following Technical Specification Changes which support blocking of reactor trip on turbine trip below 30% of rated thermal power. The following three items have been' considered from a safety analysis standpoint and are addiessed in this evaluation:

The results of i

this avaluation can be stated as follows:

)

1.

No significant increase in the probability of occurrence or consequences of an accident analyzed in the UFSAR will result from elimination of reactor trip on turbine trip below 30% of Rated Thermal Power (RTP).

The analyses results show that the DN8R does not decrease below the design limit at any time. The analysis also shows that, except under conservative assumptions, the pre s st.ri zer PORVs are not challenged during the transient. Pressure relieving devices incorporated in the primary and the secondary systems are adequate to keep the maximum pressure within the design limit. Since the predicted results are within the range of existing safety analysis vrlues, it is concluded that operation with the proposed Technical Specification changes will neither significantly increase the probability of occurrence nor the consequences of initiating events for any known accident.

2.

No new or different accident type not previously considered in the UFSAR is created by this proposed change. The complete loss of unit load without a direct reactor trip on turbine trip is a design event and is addressed in Section 15.2.7 of the UFSAR. The results for a loss of flow due to fast bus transfer failure after a turbine trip are bounded by the results for a complete loss of flow from full power, which is discussed in Section 15.3.4 of the UFSAR.

Thus, the results of all the relevant accident analyses show that operation with this modification does not create a new or dif ferent accident type than any evaluated previously in the UFSAR.

3.

The margin of safety is not reduced. The proposed Technical Specification changes have been incorporated in the safety analyses.

These analyses have demonstrated that calculated results treet all design acceptance criteria as stated in the UFSAR.

s 6