ML20196B918

Proposed Tech Specs,Supporting Operation of Plant W/Revised Heatup & Cooldown Curves Valid to 10 EFPY

ML20196B918
Person / Time
Site:	North Anna
Issue date:	11/30/1988
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20196B903	List: ML20196B900 ML20196B918 ML20196C128
References
NUDOCS 8812070089
Download: ML20196B918 (50)

Text

i Attachment 1 Proposed Technical Specification Changes l

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9812070080 S'31130 PDR ADOCK 050003':::

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REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION _

3.1.2.2 Each of the following boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4# .

ACTION:

a. With the flow path from the boric acid tanks inoperable, restore the inoperable flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN HARGIN equivalent to at least 1.77% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the flow path to OPERABLE status within the next 7 days or be in COLO SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

b. With the flow path from the refueling water storage tank inoperable, restore the flow path to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SVRVEILLANCE RE0VIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is a 11**F.

80nly one boron injection flow path is required to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 324'F.

NORTH ANNA - UNIT 1 J/4 1 9

M REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATIM LIMITING CONDIT10f4 FOR OPERATI0f4 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4 .

ACT10ti:

With only one charging pump OPERABLE, restore a second charging pumo to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.77% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a second charging pump to OPERABLE status within the next 7 days or % in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The provisions o' Specificatico 3.0.4 are not applicable for one hour following heatup above 324*F or prior ' cooldown below 324'F.

SEVElll ANCE REQUIREMENTS _

4.1.2.4.1 The above required charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure of n 2410 psig when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All charging pumps, except the above required GPERABLE pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 324*F by l verifying that the switches in the Control Room have been placed in the pull to lock position.

9 maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 324'F.

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t NORTH ANNA - UNIT 1 3/4 1-12

REACTOR COOLAPIT SYSTEM SHUTDOWri LIMIT!!IG C0fiDITIOPI FOR OPERAT10f1 3.4.1.3 a. At least two of the coolant loops listed below shall be OPERABLE:

1. Reactor Coolant loop A and its associated steam generator and reactor coolant pump,*

2. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,*

3. Reactor Coolant loop C and its associated steam generator and reactor coolant pump,*

4. Residual Heat Removal Subsystem A,**

5. Residual Hot Rcmoval Subsysten' B.**

b. At least one of the above coolant loops shall be in operation.***

APPLICABillTY: MODES 4 and 5.

ACTI0ti:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible: be in COLD SHUTDOWft within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />,

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and iraediately initiate corrective action to return the required coolant loop to operation.

• A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 324'F unless the secondary water l temperature of each steam generator is less than 50'f above each of the RCS I cold leg terperatures.

"The offsite or emergency power source may be inoperable in MODE 5.

• • • All reactor coolant pumps and residual heat removal pumps may be de energized for up to I hour provided 1) no operations v e permitted that would cause dilution of the reactor coolant system boron concentration, and

2) core outlet temperi.ture is maintained at least 10'F below saturation temperature, fiORTH Afif4A - UtilT 1 3/4 4-3

Material Procerv Basis Controlling Material: Circumferential Weld Copper Content: 0.086' WT5 Nickel Content: 0.jlWT5 Initial RT !iDT; 19 F RT

?iOT 1/4T,136.3y 3/4T, 116.1 e Curves Applicable for Service Periods Up To 10 EFPY And t ontain Margins Of 20 0F And 80 nsi for Possible Instrument Errors

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,, 4 100 150 200 250 300 350 Cold leg Temperature (degrees O Figure 3.4.2 Reactor Coolant System Pressure-Temperature Featup Limitations NORTH ANNA - UNIT 1 3/4 4-27

Material Precer9v Casis Controlling Material: Circumferential Wald Cooper Content: 0.036 Wit Nickel Content: 0.11 WT1 Initial RT NOT; 13 F .

AT,,0T 1/47, 135.3 38F 3/4T, ;;6.1 F Curves Acolicable For Service Pericos Uo To l' ".FPY And Contain Mr.rgins Of 20 0 F And 80 pst for Possible Instruaent Errors

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION _.

3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two power operated relief valves (PORVs) with a lift setting of: l

1) less than or equal to 450 psig whenever any RCS cold leg temperature is less than or equal to 247'F, and 2) less than or equal to 390 psig whenever any RCS cold leg temperature is less than 150'F, or

b. A reactor coolant svstem vent of greater than or equal to 2,07 .

square inches. I APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 247'F, except when the reactor vessel l head is removed.

ACTION:

a. With one PORV inoperable, either restore the inoperable FORV to OPERABLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; traintain the RCS ;a a vented condition until both PORVs have been restored to OPERABLE status,

b. With both PORVs inoperable, depressurize and vent the RCS through a

, 2.07 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a l Vented condition until both PORVs have been restored to OPERABLE i status.

1

c. In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (-) on the transient and any corrective action necessary to prevent resurrence,

d. The provisions of Specification 3.0.4 are not applicaLle.

NORTH ANNA - UNIT 1 3/4 4 31

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T 2 350*F U};ITING CONDITION FOR OPERATION 3.5.? Two independent ECCS subsyster6s shall be OPERABLE wit! sach subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One 00ERABLE low head safety injection pump,

c. An OPERABLE flow path capable or transferring fluid to the Reactor Coolant System when taking suction from the refueling water storage tank on a safety injection signal or from the containment sump when suction is transferred during the recirculation phase of operation or from the discharge of the outside recirculation spray pump.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated ac;uation cycles to date.

c. The provisions of Specification 3.0.4 are not applicable to 3.5.2.a and 3.5.2.b for une hour following heatup above 324*F or prior to cooldown below 324*F.

NORTH ANNA UNIT 1 3/4 5-3

EMERGENCY CORE COOLING SYSTEMS f_C.CS SUBSYSTEMS - T > 350*F L1!iLTING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump # ,

b. One OPERABLE low head safety injection pump #, and

c. An OPERABLE flow path capable of automatically trarsferring fluid to the reactor coolant system when taking suction from the refueling water storage tank or from the containment sump when the suction is transferred during the recirculation phase of operation or from the discharge of the outside recirculation spray pump.

APPLICABILITY: iDE 4.

ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERA 0LE status within I hour or be in COLD SHUTDOWN within tta next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />,

b. With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less avg thar, 350'F by use of alternate heat removal methods,

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total eccumulated actuation cycles to date.

• ^ maximum of one centrifugal charging pump and one low head safety injection pup shall be OPERABLE whenever the temperature of one or more of the RCS co d legs is less than or equal to 324'F. l NORTH ANNA - UNIT 1 3/4 5 6 l

EMERGENCY CORE C00LINC SYSTEMS SVRVEILLANCE RE0VIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold 1 cgs is less than or equal to 324*F by verifying that the switches in the Control Room are l in the pull to lock position.

NORTH ANNA - UNIT 1 3/4 5-6a

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE C0 EFFICIENT (MTC) (Continuedl conditions other than those explicitly stated will require extrapolation to those conditions in order to permit r accurate comparison.

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. Th transformedjntothelimitingHTCvalue-4.4x10'jsvalueoftheMDCwasthen ak/k/*F. The MTC value of

-3.3 x 10 - ak/k/'F represents a conservative value (with co rections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentrationangisobtainedbymakingthesecorrectionstothelimitingMTC value -4.4 x 10' ak/k/*F.

Once the equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval. For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibriug full power boron concentration s 60 ppm is less negative than

-4.0 x 10' delta k/k/'F. 4 The difference between this value and the limiting MTC value of -4.4 x 10' delta k/k/'F conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER ccnditions) and the licensed end-of-cycle, including the effect of rods, boron concentration, burnup, and end-of-cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, and 4) compliance with Appendix G to 10 CFR Part 50 (see Bases 3/4.4.9).

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

NORTH ANNA UNIT 1 8 3/4 1-2

REACTIVITY CONTROL SYSTEMJ BASES 3/4.1.2 BORAT10N SYSTEMS (Continuedl With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service perio6 ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MAPGIN from expected operating conditions of 1.777, ak/k after xenon decay and cooldown to 200*F. This expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 6,000 gallons of 12,950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the refueling water storage tank.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 324*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

With the RCS t..nperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity

, condition of the reactor and the additional restrictions prohibiting CORE I

ALTERATIONS and positive reactivity change in the event the single injection i system becomes inoperable.

The boron capability required below 200'F is sufficient to provide a SHUTDOWN MARGIN of 1.777. ok/k after xenon decay and cooldown from 200*F to 140'F. This condition requires 1378 gallons of 12,950 ppm borated water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.

The contained water volume limits in '

.. 'lowance for water not available because of discharge line ",

and other physical characteristics. The OPERABILITY of r . jection system during REFUELING insures that this system is .i .eactivity control while in MODE 6.

NORTH ANNA - UNIT 1 B 3/4 1-3

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least H0T STANDBY within I hour.

In MODE 3, a single reactor coolant loop provio'es sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 324*F are provided to prevent RCS pressure l transients, caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. T'

• RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

The operation of one Reactor Coolant Pump or one RHR pwp provides adequate flow to ensure mixing, prevent stratification, and produce mdual reactivity changes during boron concentration reductions in the keactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated loop. Varification of the boron concentration in an idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to baron concentration stratifications.

Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is NORTH ANNA - UNIT 1 B 3/4 4-1

REACTOR COOLANT SYSTEM BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Reducing T to < 500*F prevents the release of activity should a steam generator tube @ ture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LuiLTJ T

All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2 of the VFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

NORTH ANNA - UNIT 1 B 3/4 4-6

REACTOR COOLANT SYSTEM BASES The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The heatup limit curve, Figure 3.4.2, is a composite curve which was l prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves of Figure 3.4.3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall . The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 10 EFPY. The adjusted reference temperature was calculated using results from a capsule removed after the sixth fuel cycle. The results are documented in Westinghouse Reports WCAP-ll777, February 1988 and WCAP-ll791, May 1988.

The reactor vessel materials have been tested to determine their initial RT The results of these tests are shown in the UFSAR and WCAP-ll777. l Rebo.r operation and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluence andJD[.opper content of the material in question, can be predicted using US NRC Regulatory Guide 1.99, Revision 2. The heatup and l cooldown limit curves (Figure 3.4.2 and Figure 3.4.3) include predicted adjustments for this shift in RT at the end of 10 EFPY, as well as adjustments for possible errors O1T the pressure and temperature sensing instruments.

The actual shift in RT of the vessel material will be established periodically during operatioM0Iyt removing and evaluating, in accordance with ASTM E185-82, reactor vessel material irradiation surveillance specimens l installed near the inside wall of the reactor vessel in the core area. Since '

the neutron spectra at the irradiation samples and NORTH ANNA - UNIT 1 8 3/4 4-7

REACTOR C0OLANT SLSl 3 BASES vessel inside radius are essentially identical, the measured transition shif t for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different from the calbIated ART t4DT f r the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4.2 for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The minimum temperature for criticality specified in T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Appendix G.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in the UFSAR and WCAP-ll777 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue anaip is performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 247'F. Either PORV has

! adequate relieving capability to protect the RCS from overpressurization when i the transient is limited to either (1) the start of an idle RCP with the I secondary water temperature of the steam generator less than or equal to 50*F l above the RCS cold leg temperatures or (2) the start of a r Mrging pump and its injection into a water solid RCS.

l Automatic or passive low temperature overpressurization protection (LTOP) is required whenever any RCS cold leg temperature is less than 247'F. This temperature is the Limitin RT g + 90'F + instrument uncertainty. Above 247'F administrative contro is SIfeguate protection to ensure the limits of the heatup curve (Figure 3.4.2) and the cooldown curve (Figure 3.4.3) are not violated. The concept of requiring automatic LTOP at the lower end, and administrative control at the upper end, of the Appendix G curves is further discussed in NRC Generic Letter 88-11.

NORTH ANNA - UNIT B 3/4 4-8

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EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump and one low head safety injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and low head safety injection pumps except the required OPERABLE pump to be inoperable below 324'F provides assurance that a j mass addition pressure transient can be relieved by the operation of a single PORV.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILilY is maintained.

3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.

The limits on injection tank minimum contained volume and boron concentration ensure that the assumptions used in the steam line break analysis are met.

The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of Ill'F at 15,750 ppm boron.

NORTH ANNA - UNIT 1 8 3/4 5 2

Attachment 2 Safety Evaluation Report In Support Of Revised Heatup and Cooldown Curves >

Valid to 10 EFPY l

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October 28, 1988 l __

Table of Contents List Of Tables . . . . . . . . . . . . . . . . . . . . . . . . 3 List Of Illustrations .................... 3

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . 4 2.0

SUMMARY

OF CAPSULE V ANALYSIS .............. 5 3.0 APPENDIX G CURVE DEVELOPMENT . . . . . . . . . . . . . . . 6 3.1 Heat Up Curve Analysis . . . . . . . . . . . . . . . . 7 3.2 Cool Down Curve Analysis . . . . . . . . . . . . . . . 8 3.3 Heat Up And Cooldown Curve Adjustment ........ 9 4.0 PORV SETPOINT DETERMINATION . . . . . . . . . . . . . . 11 4.1 Mass Addition Transient . . . . . . . . . . . . . . 12 4.2 Heat Up Transient . . . . . . . . . . . . . . . . . 13 4.3 New LTCP Setpoints . . . . . . . . . . . . . . . . . 14 5.0 PTS Evaluation . . . . . . . . . . . . . . . . . . . . 15

6.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . 17 7.0 10 CFR 50.59 EVALUATION . . . . . . . . . . . . . . . . 18

8.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . 27 Safety Evaluation Page 2

List of Tables 1 Heatup Curve Including Plant Specific Uncertainties . 21 2 Cooldown Curve Including Plant Specific Uncertainties. 22 3 Caoldown Rates Assumed For Various Temperature Ranges. 23 4 Initial Conditions For The Mass Addition Transient . . 23 5 Initial Conditions For The Heat Input Transient. . . . 24 6 North Anna Unit 1 PORV Setpoints Technical Speci fication 3.4.9.3.a . . . . . . . . . . . . . . . 24 List of Figures 1 Modification of Heatup Curve To Include Plant S pec i f i c Un ce rtai n ti e s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2 Modification of Cooldown Curve To Include Plant S peci f i c Unce rta i n t i e s. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 l

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Safety Evaluation Page 3

1.0 INTRODUCTION

The 10C'FR50 Appendix G analysis (Reference 3.) performed using the results from Capsule V, removed at the end of Cycle 6, indicates that the heatup and the cooldown curves for North Anna Unit I can be shifted allowing higher oper-ating pressures relative to the current curves (T.S. Figures 3.4.2 and 3.4.3, Reference 4.). The reason for this curve shift is that the Capsule V analysis shows the reactor vessel has been exposed to a lower fast neutron fluence than previously assumed. The current Technical Specification Curves are based on an assumed design basis neutron fluence through ten effective full power years (EFPY). The measured fast neutron (E >1.0 Mev) fluence data shows that capsule U was exposed to 8.28 x 1018 n/cm8 at 5.90 EFPY (Reference 2.). The measured fluences are significantly lower than the design values due to the implementa-tion of low leakage fuel management at North Anna. Using the measured fluence at 5.90 EFPY to estimate the fluence at 10.0 EFPY results in a lower fluence than used to generate the current Technical Specification curves. Regulatory Guide 1.99, Revision 2 allows the use of surveillance date. to adjust original design values af ter two or more credible surveillance data sets become available (Reference 6.). Capsule V is the second capsule removed from North Anna Unit

1. Capsule V was removed after 1.13 EFPY.

The heatup and Cooldown curves provide an upper pressure and temperature limit and Reactor Coolant Pump (RCP) operation ' limits the lower operating pressure. The PORV low temperature overpressurization protection (LTOP) setpoints prevent inadvertent operation at pressures which would exceed the allowable Appendix G curve. This Safety Evaluation Report was prepared to Safety Evaluation Page 4

summarize the analyses performed to determine revised heatup and cooldown curves, and LTOP setpoints.

Capsule U analysis is summarized in Section two. The revised heatup and cooldown curve analyses are discussed in Section three. Section four discusses the revised PORV setpoints. A PTS evaluation is presented in Section five.

Finally, a 10 CFR 50.59 evaluation is presented to support the proposed Tech-nical Specification changes.

2.0 StktiARY OF CAPSULE U ANALYSIS Capsule U was removed from North Anna Unit 1 at the end of the sixth cycle of operation. The capsule dosimeters were evaluated and found to have a cumu-lative fast neutron, E > 1.0 Mev, fluence of 8.28 x 1018 n/cm*. The calculated, based on actual cycle power distributions, fast neutron fluence at the capsule location was 8.85 x 1018 n/cm2 which compares favorably with the dosimeter fluence. The peak calculated fluence at the inride surface of the reactor vessel was calculated to be 8.83 x 1018 n/cm2 which shows that the capsule has

been exposed to slightly more neutrons than the vessel (Reference 2.).

l The material property testing included Charpy V-notch impact testing and tension testing of several specimens located within the surveillance capsule.

The Charpy tests are performed to determine the transition temperature increases, at 30 f t-lb and 50 f t-lb points, and the decrease in the upper shelf energy. The tensile specimens were used to determine ultimate tensile strength and yield strength. The vessel specimens within Capsule U were obtained from Safety Evaluation Page 5

L the same girth weld and forging materials as that used in the Reactor Vessel beltline (Reference 2.).

The irradiated specimens test results were compared to unirradiatied specimen test results. The Charpy V-notch impact test results show the irra-diation has increased the average 50 ft-lb transition temperature by 80 to 110

' F. depending on the specimen metal. Irradiation has increased the average 30 f t-lb transition temperature by 65 to 100 'F. The upper shelf energy, average energy' absorption at full shear, results show the worst decrease to be 25 f t-lb when comparing irradiated samples to unirradiated samples. The lowest average upper shelf energy was determined to be 92 ft-lb which is greater than the 10 CFR 50 Appendix G low limit of 50 f t-lb (Reference 7.). The Charpy impact test results from Capsule U were also satisfactorily compared to the Capsule V results. Tension test results show a slight increase in the ultimate tensile strength and the yield strength due to irradiation. Reference 2 should be consulted far specific test results.

3.0 APPEh0!X G CURVE DEVELOPMENT 10 CFR 50 Appendix G, "Fracture Toughness Requirements," establishes pressure and temperature operational limits for the Reactor Vessel. Virginia Power utilizes two types of graphs to identify plant specific limits. The graphs are know as the Heatup and Cooldown curves. The present heatup curve depicts three curves, the leak test limit, tha heatup limit (up to 60 'F/hr) and the criticality limit. The cool down curve depicts a series of curves for typical cooldown rates (0, 20, 40, 60, and 100 'F/hr). The range of the graphs Safety Evaluation Page 6

is from the pressure limits corresponding to 80 'F to the Pressurizer Safety Valve setpoint, 2485 psig. The Control Room Operators use these Technical Specification graphs to ensure RCS pressure and temperature are within accept-able values.

3.1 Heat Up Curve Analysis The analysis done to determine heatup curves includes the development of pressure-temperature relationships for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. Therefore, steady-state conditions can be limiting for the inside wall so both heatup and steady state must be considered.

The heatup curve calculations must also consider the case of a 1/4T flaw at the outside surface. The thermal and pressure stresses never cancel for this situation. The thermal stresses are dependent on both the rate of heatup and the coolant temperature along the heatup ramp.

The use of a composite curve is required to make sure that the limiting condition is always protected against. For example protection must be provided if the limiting location shifts from the inside to the outside surface.

Therefore, a composite heatup curve is generated by comparing, on a point-by-point basis, the steady-state curve at the inside of the wall along with the Safety Evaluation Page 7

T various heatup rate curves at the outside surface. Thus at any given temper-ature, the allowable pressure is taken to be the lowest of the values from each of the curves under consideration.

3.2 Cooldown Curve Analysis During cooldown, the controlling location of the flaw is always at the nside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pres-sure-temperature relations are generated for both steady-state and finite cooldown rate situations. A lower bound composite curve from the steady state and cooldown conditions is constructed for each cooldown rate of interest.

The use of a composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant [

temperature, whereas the limiting pressure is actually dependent on the material te.nperature at the tip of the assumed flaw. During cooldown the tip is at a higher temperature than the fluid adjacent to the vessel inner wall. This  ;

condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the temperature gradient devel- l i

oped during cooldown results in a higher value of KIR at the 1/4T location for j finite cooldown rates than for steady-state operation. So, if the temperature changes such that KIR increases faster than KIT the steady-state can be limiting, f l

t Safety Evaluation Page 8 t

,_m_- - , . - - - - . - - - - - , - - - - - - - - - - - -'

4 3.3 Heatup And Cooldown Curve Adjustment .

l Westinghouse performed the Capsule U analysis and provided Virginia Power with heatup and cooldown curves for North Anna Unit 1 (Reference 3.). The i Westinghouse curves did not include any instrument uncertainty. To ensure that ,

instrument error does not allow the Reactor Vessel to be operated at undesirable pressure-temperature combinations, adjusted heatup and cooldown curves were '

generated. The curves were created by shifting the Westinghouse curves 20 'F to higher tecoeratures and 80 psi to lower pressures. The 20 'F and 80 psi represent the worst case error between the control room wide range temperature and wide range pressure indicators and the actuti conditions which may exist  ;

at the Reactor Vessel Beltline. The plant specific uncertainties were deter- l I

mined using the same methodology used to determine the instrument uncertainties l

in the emergency operating procedures (EOP) but adjusting the results for normal  !

operations.

Figure 1 presents the revised heatup curve, with a maximum heatup rate of 60 'F/hr. As indicated in the figure, the curve is based on an extrapolated  ;

fluence to permit operation to 10 EFPY. The revised heatup curve does not l

contain the 10CFR50, Appendix G criticality limit. This limit is not required i

since limiting condition for operation (LCO) 3.1.1.5 restricts the lowest  :

operating loop average tempe,ature to 2 541 'F for modes 1 and 2. This LCO l

defines a minimum temperature for criticality. It provides substantially more margin to the heatup curve than the criticality limits required by 10CFR50, Appendix G.

(

Safety Evaluation Page 9 P

4 Table 1 is the same heatup curve data used in Figure 1. The data shown in Table 1 is the heatup curve adjusted for instrument errors associated with '

the Control Room indicators. The instrument error associated with the automatic operation of the PORV for low temperature overpressurization protection is less than the instrument error associated with'the Control Room indicators. The Unit 1 cooldown curve was developed for a range of cooldown rates. Table 2 and Figure 2 show the Unit 1 cooldown curve data after incorporating the plant specific instrumentation uncertainties.

Comparing the heatup and cooldown curves in Tables 1 and 2, shows that the heatup curve is more limiting than the cooldown curve except for low temper-atures. Less than 180 'F, the cooldown curves for 40 'F/hr and greater become more limiting. Previous discussion of a composite cooldown curve indicated the Reactor Vessel wall has a temperature gradient dependent on cooldown rate. Due to this temperature gradient and since these higher cooldown rates are not realistically possible at low RCS temperatures, Table 3 was developed to present the assumed cooldown rate for various temperatures. Comparison of a composite cooldown curve, constructed from the information in Table 2 and Table 3, to the heatup curve shows the heatup curve to be the most limiting. The 60 'F/hr heatup curve will be treated as the representative Appendix G limiting curve for all further discussions concerning p0RV setpoint determination.

Since the design basis transients are defined with operational assumptions related to the Pressurizer Safety Valves setpoints, certain operational restrictions must be enforced to ensure the low temperature accident analysis assumptions are valid. The Appendix G Curve temperature corresponding to the Safety Evaluation Page 10

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i Pressurizer Safety Valve setpoint of 2485 psig is 324 'F. This point is used to bound all of the low temperature accident analyses. Below 324 'F the anticipated' low temperature accidents may be adequately mitigated by the auto- f matic action of the PORV or by allowi:1g sufficient time for operator response.

~

The mass addition transient assumes only one Charging pump will be operable )

t below 324 'F. The heatup transient assumes whenever a RCP is started below 324 -

'F the temperature difference between the primary and secondary fluids its the l

Steam Generator is less than 50 'F. l 4.0 PORV SETPOINT DETERMINATION Cold, overpressure protection is provided to ensure that the normal oper-t i ation heatup and cooldown curves are not violated during operation with a water solid system. The PORVs on the pressurizer are set at a pressure low enough I to prevent violation of these Appendix G heatup and cooldown curves should a RCS pressure transient occur. The limits have been set by two design basis [

l accidents: the inadvertent start of a charging pump and the startup of a reactor [

coolant pump in an RCS 'oop with a 50 'F difference between the Steam Generator l secondary fluid temperature and the RCS temper 4ture. These transients represent [

j  !

the limiting mass addition and heat input transients and are analyzed with the l

4

RCS water solid. Only one PORV is required to operate during the transients.

i Generic transient analysis was previously used to determine LTOP setpoints l

which maintain acceptable pressure-temperature combinations on the Appendix G heatup and cooldown curves, The plant specific analysis allows actual plant characteristics to be modelled rather than the use of generic assumptions. The Safety Evaluation Page 11

• .q l

l generic assumptions have excessive conservatism to allow a wide range of application. These generic conservative assumptions become operationally burdensome when PORV lift setpoints are too low to allow normal RCS operation l without opening a PORV. A plant specific North Anna two loop RETRAN model was developed to analyze possible setpoints. 'Since generic analysis found that the .

I mass addition and the heatup transients were equally limiting, it was necessary '

to analyze both transients to determine how the North Anna model would respond I for these transients at different initial conditions. The following sections }

describe the North Anna model development and the analysis to determine new PORY setpoints.

4.1 Mass Addition Transient l The inadvertent startup of a single charging pump was selected as the design i

basis mass addition transient based on previous UFSAR work (Reference 5. Section ,

5.2.2.2). Therefore, the LTOP setpoints were determined such that an overshoot allowance exists to prevent the Appendix G curves from being exceeded assuming I an inadvertent charging pump startup during water solid operation. This over-shoot allowance is required because of the valve opening characteristic associated with the air operated relief valves used on the pressurizer at North i Anna. (Reference 10. and Reference 11.).

l Inadvertent operation of a single Charging Pump was modeled assuming initial conditions as listed in Table 4. The initial RCS temperature and the PORV setpoint were varied to observe the effects of these parameters. The RCS temperatures used were 100 and 200 'F. The resulting peak pressure was not Safety Evaluation Page 12

, 4  :

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significantly dependent on initial RCS temperature. The peak RCS pressure has j i

also been found to be insensitive to the initial RCS p-essure.

t 4.2 Heat Input Transient t b

The heat input transient assumes the Technical Specification limit of a 50 'F temperatura difference between the Steam Generators and the RCS. A Reactor Coolant Pump startup in one loop is also assumed to maximize the heat transfer. This scenario has been determined to be the design basis heat addi-  ;

tion transient for LTOP setroint determination relative to the Appendix G f

curves. (Reference 5. Section 5.2.2.2). (

i The heat addition transient was modeled assuming the initial conditions l listed in Table 5. The RETRAN model was normalized to produce results equal to the LOFTRAN model used for the generic analysis. Next, the RETRAN model was  !

used with a Westinghouse representation of the PORV. Finally, a North Anna f I

specific model of the PORV was used in the normalized RETRAN model. The most {

restrictive pressure overshoot was from the heatup cases initializing the RCS f

at high temperatures.  !

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4 4.3 New LTOP Setpoints The North Anna plant specific PORV model was verified and new LTOP setpoints were determine'd. The new setpo!nts were established by using the new heatup curve from Reference 3 adjusted for instrument uncertainties for automatic PORV operation (15.68 'F and 67,2 psi). Limiting condition mass addition and heatup transient RETRAN computer runs were made to validate the new setpoints. An effort was made to optimize the PORV setpoints to values which limit the' Reactor Vessel peak pressure to be less than the pressure allowed by the Appenda J curves and not be operationally restrictive by having too much safety margin.

The new North Anna Unit 1 setpoints are s 450 psig when s 247 'F and s 390 psig when s 150 'F. The mass addition transient was the most limiting transient for both the high and low setpoints.

Automatic low temperature overpressurization protection is r., quired when-ever any RCS cold leg temperature is less than 247 'F. This temperature is the RTNOT + 90'F + instrument uncertainty. The RTNDT temperature is 136.3 'F for 1/4T and 116.1 'F for 3/4T (Reference 3. page 12). The instrument uncer- g tainty added was 20 'F. The 90 'F addition is considered to be a reasonable i range to require the automatic low temperature overpressurization protection.

This is sufficient to require au:omatic protection during startup and shutdown.  !

Above 247 'F administrative control is adequata protection because of Appendix G fracture criterion. The analysis has an increased margin at higher temper-atures. In addition, operation of the RCS above 247 'F decreases the effects of the two design basis transients. The concept of requiring automatic LTOP Safety Evaluation Page 14

. y at the lower end, and administrative control at the upper end, of the Appendix G Pressure - Temperature limit curve is further discussed in NRC Generic Letter 88-11 (Reference 8.). The Generic Letter states that due to the impact of implementation re Revision 2 of Regulatory Guide 1.99, Standard Review Plan Section 5.24. and Branch Technical Positinn RSB 5-2, will be revised to define the temperat.re where automatic protection is required to be enabled.

Table 6 compares the current setpoints to the new setpoints. This comparison shows that the current setpoints cause the PORVs to lift at lower pressure for all temperatures.

5.0 PTS EVALUATION Capsule U was located at the sixty five degree location, twenty five degree azimuthal angle, in the North Anna Unit I reactor vessel. The experimentally determined fluence was 8.28 x 10 18n jem fast neutron fluence (E > 1 Mev). The calculated fast fluence at the capsule center, an azimuthal location of twenty five degroes, was 8.85 x 1018 n/cm .8 The close comparison between the calcu-lated results and the actual (i.e. experimental) results confirmed the analytical model which has been used to predict the fluence to the end of the current operating license (Reference 2).

The Virginia Power PTS submittal was based on calculated fluence. $1nce this fluence is slightly greater than the experimentally determined value and the material properties have not changed, the RTPTS parameters on record are Safety Evaluation Page 15 i

., 4

.~

~q slightly conservative (Reference 1.). This conclusion is readily evident by observation of the equations which defirl t) 7 73 parameter:

RTPTS = I + M + (-10 + 4;C.*Cu) + 350(NL)(Cu)) f0.270 RT yps = I + M + 283 f 0.194 The only parameter in the above equations which changed as a result of the capsule evaluation is the fluence factor, f which is reduced slightly (i.e.

H 0.885 to 0.828) at the 25' azimuthal location and the capsule center radially).

Based on this experimental data point the fast fluence factors are 6.88 % too

! high. Applying these diffarences to the O' azimuthal angle, the effect upon the limiting RTp73 parameters can be determined based on the Reference 1.

results, The Lower Shell forging 03 material was found to have the most limiting RTPTS RT PTS f =38 +48 +(-10 + 470(.15) '. 350(.80)(.15))(.883)0.270 1

= 185 'F The RTPTS .t license expiration is expceted to be 235 'F. Thus, the current RTPTS parameter is not significantly different at the end of the operating license.

Safety Evaluation Page 16

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6.0 CONCLUSION

S The heatup and cooldown c'Jrves required by Appendix G of 10 CFR 50 have been extraf.olated to 11 EFPY by including the effects of the incremental radi-L ation exposure on the eeactor vessel belttine region. The results are i

referenced to 'he analysis of Capsule U from North Anna Unit 1. The revised i

curves allow PORV setpoints higher than the current setpoints for the LTOP system to be implemented. The revised Appendix G curves were prepared using standard Westinghouse methodology including R.G. 1.99 Rev. 2. .

I i '

The new Heatup and Cooldown curves are valid until 10 EFPY of operation.

{ The next reactor vessel surveillance capsule, Capsule X, is scheduled to be o

removed after the ninth fuel cycle which allows sufficient time for analysis f

prior to exceeding 10 EFPY.

The heatup and cooldown c'Jrves prepared by Westinghouse were determined l in a conventional manner according to Section III of the ASME code as required by 10 CFR 50 Appendix G. Both steady-state and transient thermal conditions ,

J i were considered in order to bound the possible combinations of pressure (i.e.

l membrane) and thermal stresses. The Westinghouse curves were revised to include plant specific instrument loop uncertainties.

The new N' th Anna Unit 1 Pressurizer PORV low temperature overpressuri-l zation lift set ings should be less than or equal to 450 psig whenever any RCS f f ,

i cold leg temperature is less than or equal to 247 'F, and less than or equal [

l I to 390 psir .h <ever any RCS cold leg temperature is less than 150 'F. 5 L

j Safety Evaloation Page 17 i

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4 PTS evaluations were made for thi limiting beltline locations. The current value of the limiting RTPTS parameter was updated using the calculated fluences contained in the capsule report. No significant change was obtained, s

7.0 10 CFR 50.59 EVALUATION The proposed changes have been reviewed against the criteria of 10 CFR 50.59 resulting in the conclusion that an unreviewed safety question does not exist.

This determination was reached based on the fo. owing specific considerations:  !

1. The probability or consequences of any UFSAR event do not increase. Acci-dent probability is independent of the initial conditions maintained by T.S.

3.4.9.1. This Specification deals with heatup and cooldown curves. The criticality limit line has bwen removed from Figure 3.4.2 to elimtWate i confusion with the more restrictive criticality limit of T.S. 3.1.1.5. T.S. l 3.4.9.3 has been changed to revise the PORV setpoints which are used to  :

protect against violation of the Appendix G curves contained in T.S.

3.4.9.1. T.S. 3.1. 2. 2, 3.1. 2. 4, 3. 4.1. 3, 3. 5. 2, 3.5.3 and 4.5.3.2 are required to ensure the low temperature accident assumptions are consistent with analysis assumptions. T.S. 3.4.9.1.C has been deleted because the i 1

4 Applicability Statement temperature change makes this option obsolete.

Finally the appit: ability of 3.4.9.3 has been modified to require automatic  !

and passive overpressurization protection only during the the lower '

portion, below RTNDT+110'F, of the Appendix G heatup and cooldown curves.

The upper portion oT the the Appendix G curve limits will be ensured through administrative compliance with T.S. 3.4.9.1.

None of the above changes increase accident probability but rather change ,

, the initial conditions assumed in the safety analysis or are administrative i

) in nature.  !

3 Accident consequences are not increased by the proposed Technical Specift- i cation changes. The proposed changes to T.S. 3.4.9.1 are r* quired by l Appendix G to maintain the prescrib(d margin to the reference stress intensity factor. The design criteria are met for both the mass addition ,

and heat input transients which form the cold overpressure protection design basis. Therefore, the consequences are not increased by these changes, j It is noted that plant specific instrument uncertainties have been included t in the revised curves which are more conservative than the previously used  ;

values. The change proposed for T.S. 3.4.9.3 reflects the peak pressure l l

overshoot from the two design transients, i I

Safety Evaluation Page 18 1

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• e Changes to the Technical Specifications Basis Sections 3/4.1.1.5, 3/4.1.2, 3/4.4.1, 3/4.4.9 and 3/4.5.3 are necessary to maintain consistency with the Limiting Condition for Operation and Surveillance Requirements Sections.

Table B 3/4 4-1, Figures B 3/4.4-1 and B 3/4.4-2 were deleted from the Basis Section and tha text appropriately revised. The Table and Figures do not reflect the current methodology used to determine the adjusted reference temperature, RTNOT. These Basis Section changes are administrative and do not increase the probability or consequences of any UFSAR event.

2. No new or different accident type is generated as a result of the revised heatup and cooldown curves, T.S. 3.4.9.'. These operational curves provide restrictions on the pressure and temperature of the reactor coolant system.

In other words, T.S. 3.4.9.1 requires operation in a manner which prevents flaw extension even if a 1/4T flaw existed. The criticality limit line has been removed from Figure 3.4.2 to eliminate confusion with the more restrictive criticality limit of T.S. 3.1.1.5. T.S. 3.4.9.1.C has been deleted because the Applicability Statement temperature change makes this option obsolete. Therefore, the proposed changes do not involve alterations to the physical plant which introduce any new or unique operational modes or accident precursors.

The change to T.S. 3.4.9.3 involves changing a setpoint on existing instrumentation. No alterations have been made to the components in the instrurrent loop.

The changes to T.S. 3.1. 2. 2, 3.1. 2. 4, 3. 4.1. 3, 3. 5. 2, 3. 5. 3 a nd 4. 5. 3. 2 a re required to ensure the low temperature accident assumptions are consistent with analysis assumptions. These require only a shift in an existing operational temperature limit and does not involve hardware modifications.

Changes to the Technical Specifications Basis Sections 3/4.1.1.5, 3/4.1.2, 3/4.4.1, 3/4.4.9 and 3/4.5.3 are necessary to maintain consistency with the Limiting Condition for Operation and Surveillance Requirements Sections.

Table B 3/4 4-1, Figures B 3/4.4-1 and B 3/4.4-2 were deleted from the Basis Section and the text appropriately revised. The Table and Figures do not reflect the current methodology used to determine the adjusted reference temperature, RTNDT. These Basis Section changes are administrative and do not introduce any new or different type accident.

3, The margin of safety is not r' educed. The proposed changes to T.S. 3.4.9.1 are required by 10 CFR 50 Appendix G on a periodic basis. The analyses include margin in the fluence calculations, the reference temperature determination and the stress calculations. These margins include the safety factors required by the ASME code. The criticality limit line has been removed from Figure 3.4.2 to eliminate confusion with the more restrictive criticality limit of T.S. 3.1.1.5.

The changes to T.S. 3.1. 2. 2, 3.1. 2. 4. 3. 4.1. 3, 3. 5. 2, 3. 5. 3 a nd 4. 5. 3. 2 a re required to ensure the low temperature accident assumptions are consistent Safety Evaluation Page 19

• 4 1

with analysis assumptions. Plant specific design basis transients were analyzed and the limiting pressure overshoot was used to determine the revised PORV setpoint. Plant specific instrumentation uncertainties were factored into the results. T.S. 3.4.9.1.C has been deleted because the Applicability Statement temperature change makes this option obsolete.

Therefore, it can be concluded that no reduction in the margin of safety results from these Technical Specification changes.

Changes to the Technical Specifications Basis Sections 3/4.1.1.5, 3/4.1.2, 3/4.4.1, 3/4.4.9 and 3/4.5.3 are necessary to maintain consistency with the Limiting Condition for Operation and Surveillance Requirements Sections.

Table B 3/4 4-1, Figures B 3/4.4-1 and B 3/4.4-2 were deleted from the Basis Section and the text appropriately revised. The Table and Figures do not reflect the current mu; odology used to determine the adjusted reference j temperature, RTNOT. These Basis Section changes are administrative and do

not reduce the margin of safety.

The capsule results were used to determine if the PTS projections subtaitted previously (Reference 1) had changed significantly as required by 10 CFR 50.61. The changes were shown to be insignificant.

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• o Table 1: HEATUP CURVE INCLUDING PLANT SPECIFIC UNCERTAINTIES Indicated Leak Test Heat Up Rate Criticality Temperature Limit To 60 'F/hr Limit

('F) Pressure (psig) Pressure (psig) Pressure (psig) 105.0 ----

- 466.45 ----

110.0 ----

466.45 ----

120.0 ----

466.45 ----

130.0 ----

474.01 ----

140.0 ----

488.56 ----

145.0 ---- ----

150.0 ----

508.98 ----

160.0 ----

534.84 ---- i 170.0 ----

566.14 ----

180.0 ----

603.20 ----

190.0 ----

646.59 ----

200.0 ----

697.05 ----

210.0 ----

755.53 ----

220.0 ----

823.15 ----

230.0 ----

901.20 ----

240.0 ----

991.20 ----

250.0 ----

1094.88 ----

260.0 ----

1214.17 ----

264.0 1920.0 ----

{

270.0 1351.30 ----

280.0 1508.69 ----

285.0 2405.0 1082.53 288.85 2494.0 290.0 ----

1689.04 1094.88 L 300.0 ----

1895.56 1214.17 l 310.0 ----

2131.32 1351.30 320.0 ----

2399.14 1508.69 323.54 ----

2494.0 330.0 ---- -----

1689.04 340.0 ---- -----

1895.56 350.0 ---- -----

2131.32 360.0 ---- -----

2399.14 i 363.54 ---- -----

2494.0 [

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Safety Evaluation Page 21 l

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Table 2: C00LDOWN CURVE INCLUDING PLANT SPECIFIC UNCERTAINTIES Indicated Steady 20 'F/hr 40 'F/hr 60 'F/hr 100 'F/hr Temperature State Cooldown Cooldown Cooldown Cooldown

('F) (psig) (psig) (psig) (psig) (psig) 105.0 500,77 468.95 436.70 403.99 337.13 110.0 508.64 477.11 445.18 412.84 346.82 120.0 526.19 495.37 464.23 432.74 368.72 130.0 546.47 516.54 486.36 455.04 394.36 140.0 569.90 541.06 512.07 482.94 424.34 150.0 596,99 569.45 541.90 514.34 459.35 160.0 628.23 602.33 576.51 550.84 500.17 170.0 664.43 640.38 616.63 593.22 547.71 180.0 706.19 684.41 663.12 642.40 603.05 190.0 754.42 735.33 716.97 699.45 667.38 200.0 810.11 794.21 779.32 765.57 742.11 210.0 874.41 862.27 851.46 842.15 828.84 220.0 948.62 940.89 934.88 930.81 929.41 230.0 1034,23 1031.68 1031.30 1033.35 240.0 1132.95 ,

250.0 1246.70 260.0 1377.66 270.0 1528.27 280.0 1701.21 290.0 1899.71 300.0 2126.88 310.0 2385.61 314.19 2494.00 Safety Evaluation Page 22

• o e-Table 3: COOLDOWN RATES ASSLNED FOR VARIOUS TEMPERATURE RANGES Temperature Cooldown Range Rate

'F 'F/hr Above 185 100 From 185 to 155 60 From 155 to 125 40 l From 125 to 110 20 Below 110 0 Table 4: INITIAL CO WITIONS FOR THE MASS ADDITION TRANSIENT Reactor Coolant Temperature 100/200 'F Reactor Coolant Pressure 50 psig Maximum Charging Pump Flowrate 770 gpm Pressuri;er Steam Volume 0 ft8

> Pressurizer Water Volume 1400 ft' Reactor Coolant System Flow 10 % l PORV Open Setpoint Variable f PORV Closed Setpoint Open - 15 psi Safety Evaluation Page 23 h

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1 Table 5: INITIAL C0 WITIONS FOR THE HEAT INPUT TRANSIENT Reactor Coolant Temperature 100/140/180 'F Reactor Coolant Pressure . 50 psig RCS/SG AT 50 'F Pressurizer Steam ','nlume O ft8 Pressurizer Water Volume 1400 ft8 RCP Speeds In Affected Loop, startup 10 to 100 %

In Unaffected Loop, coastdown 10 to 0 %

PORV Open Setpoint Variable PORY Close Setpoint Open - 15 Table 6: NORTH MNA UNIT 1 PORV SETPOINTS TEONICAL SPECIFICATION 3.4,9.3.a Current New Setpoints Setpoints s 375 F s 247 F s 420 psig s 450 psig

' < 185 F < 150 F s 350 psig s 390 psig Safety Evaluation Page 24

Material Propery Basis controlling Material: 'Circumferential Weld Copper Content: 0.086 WT1 Nickel Content: 0.glWT1 19 F Initial RTNDT:

RT 1/47, 136.3 F NOT 3/4T, 116.1 F Curves Appitcable For Service Periods Up To 10 EFPY And Contain Margins Of 20 0F And 80 psi For Possible Instrument Errors I,lM t.mb t.4# - lini, Deris y ,

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ColdlegTemperature(degrees 0 Figure 1: Modification Of Heatup Curve To include Plant Specific Uncertainties Safety Evaluation Page 25

• o Material Procertv Basis Controlling Material: Circumfer.tntial Beld Copper Content: 0.086 1T1 Nickel Content: 0.11 WT%

Initial RTHOT: 19 F RT,4DT

'

• O I 3/4T, '16.'1 o

. F Curves Aoplicable For Service Periods Up To 10 EFPY And Contain Margins Of 20 0 F And 80 psi For Possible bstrument Errors

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50 100 150 200 250 300 350 ColdLegTemperature(degreesF)

Figure 2: Modification Of Cooldown Curve To include Plant Specific Uncertaintles Safety Evaluation Page 26

8.0 REFERENCES

1. "North' Anna Units 1 Reactor Vessel WCAP-11016 Revision 3, Heinecke, C.C.,Fluence and RipN88. Evaluations,"

et. al. January

2. "Analysis of Capsule U From The Virginia Electric And Power Company, North Anna Unit 1, Reactor Vessel Radiation Surveillance Program." WCAP-11777.

S. E Yanichko, L. Albertin, E. P. Lippincott, February 1988.

3. "Analysis of Capsule U From The Virginia Electric And Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program, North Anna Unit 1 Reactor Vessel Heatup and Cooldown Limit Curves For Normal Operation."

WCAP-11791. J. C. Schmertz May 1988.

4 North Anna Unit 1 Technical Specifications through Amendment # 104, 6-20-88.

5. "Updated Final Safety Analysis Report," North Anna Power Station, Units 1

& 2. Virginia Electric and Power Company.

, 6. "Radiation Embrittlement Of Reactor Vessel Material s," U. s. Nuclear

Regulatory Commission Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, May 1988.

7. Code of Federal Regulations; Title 10 "Ene rgy; " Part 50, "Domestic Licensing of Production and Utilization Facilities;" Appendix G, "Fracture Toughness." Published Janauary 1, 1988 by the Office of the Federal Register National Archives and Records Administration.

8. "NRC Position On Radiation Embrittlement Of Reactor Vessel Materials And Its Impact On Plant Operations (Generic Letter 88-11)," U.S. Nuclear Regu-latory Commission, July 12, 1988.

9. "Pressure-Temperature Limits " Section 5.3.2 US NRC Standard Review Plan (NUREG-75/087), Revision 0, November 24, 1975.

10. "EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report," EPRI, NP-2628-SR, December 1982.

11. "Safety and Relief Valves in Light Water Reactors," EPRI, NP-4306-SR, December 1985, 1

Safety Evaluation Page 27

4 5

Attachment 3 10 CFR 50.92 Evaluation i

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We have reviewed the proposed changes in accordance with the criteria in 10 CFR 50.92 for determining whether no significant hazards considerations exists. Each criterion is addressed below:

i 1. The proposed amendment does not involve an increase in the probability or consequence of an accident previousl'y evaluated. The proposed change revises the operating restrictions presently in place to prevent non-duc-tile pressurization or overstressing the reactor vessel based on the most recent surveillance capsule and specific instrument uncertainties. The probability of non-ductile vessel failure has not been increased by virture of the refined analysis techniques that were implemented based on specific unit information. The consequences of violating an Appendix G curve or actuating the LTOP system are not increased.

2. The proposed amendment does not create the possibility of a new or different kind of accident and was evaluated in light of the unit specific information that has been derived from the recent evaluations and analyses.

3. The proposed amendment does not involve a significant reduction in the margin of safety. The safety factors defined in the ASME Code and the 1

requirements of 10 CFR 50 Appendix G provide the basis for the safety margins utilized. The plant specific information utilized to evaluate the proposed change provide more explicit information and confirm that the safety margins are not reduced.

Based on this review, we conclude that the proposed change involves no significant hazards consideration.

I The revisions to the format of the existing North Anna Unit 1 Technical Specifications were made to provide more explicit and pertinent 3

instructions to Operations Personnel. These changes are administrative in nature, are consistent with the Westinghouse Standard Technical Specifica-tions (Revision 4), and were included in the significant hazards consideration.

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I Attachment 4 Four Copies of WCAP-11791 i North Anna Unit 1 Reactor Vessel Heatup and Cooldown  !

Limit Curves For Normal Operation f i

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