ML20087D171

Proposed Tech Specs Re Pressure/Temp Limitations & Low Temp Overpressure Protection Sys Setpoints

ML20087D171
Person / Time
Site:	North Anna
Issue date:	12/27/1991
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20087D168	List: ML20087D166 ML20087D171
References
NUDOCS 9201150233
Download: ML20087D171 (128)

Text

. - _.. _ -.

i NORTH ANNA POWER STATION 4

UNITS 1 and 2 TECHNICAL SPECIFICATION. CHANGE REQUEST HEATUP/COOLDOWN LIMITATIONS LOW TEMPERATURE / OVERPRESSURE PROTECTION t

Discussion of Proposed Changes Proposed Changes - Unit 1 Final Copy Proposed Changes - Unit 2 - Final Copy Significant Hazards Considerations Determination (10 CFR 50,92 Evat,..on)

Additional Supporting Inforniation (NA&F Technical Analysis)

(BAW 2146)

(WCAP-12503)

(References List) 9201150233 911227 VIRGINIA EUECTRIC AND POWER COMPANY l

pDR ADOCM 0500

t i

ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATIONS CHANGES l

1 TECHNICAL SPECIFICATION CHANGE REQUEST i

4 DISCUSSION.

NORTH ANNA POWER STATION - UNITS-1 & 2 i

1 l

l VIRGINIA ELECTRIC AND POWER COMPANY l

t..

oock t Nos : 60 338&339 subl No.

91 707 section i Page 2 of 9 DISCUSSION OF PROPOSED CHANGES

[n.ltod u etIon The North Anna Reactor Coolant Systems (RCS), specifically the Reactor Pressuro Vossols (RPV), are protected flom matorial failure during low temperature operations by imposing rostrictions on RCS pressure.

The hoatup and cooldown curvos as well as the Low Temperature /Over-pressuro Protection System (LTOP) sotpoints, provide the rostrictions to bound the area of operation and ensuro RCS protection from non ductilo failure.

The regulatory requirements for providing those rostrictions and roovaluating them, are stipulated in 10 CFR 50, Appendix G.

The curront heatup and cooldown curves and LTOP sotpoints, will not be valid after 10 offectivo full power years (EFPY) cumulative core burnup.

According to our most recent estimatos in September,1991, North Anna Unit 1 id expected to reach 10 EFPY in April,1993 and Unit 2 in Septembor, 1993, in anticipation of the expiration of those curves, Virginia Power has performed a safety evaluation to support implomonting revised curvos and setpoints.

These now curvo values and sotpoints will be valid through 12 EFPY for North Anna Unit 1 and 17 EFPY for North Anna Unit 2.

Backoround The heatup and cooldown curves are required by Appendix G of 10 CFR 50 and have been extrapolated to 12 EFPY and 17 EFPY for North Anna Units 1 and 2, respectively, by including the effects of the incremental radiation exposure on the reactor vossol bottline region.

The results are referenced to the analycos of the North Anna Units 1 and 2 Capsulo U results.

The revised Appondix G curves woro prepared using standard B&W and Westinghouse methodologies including those found in Regulatory Guido 1.99 Rev. 2.

PORV setpoints woro developed to provide bounding heatup and cooldown curve protection for the worst caso mass and heat addition low temperature overpressure transients.

VIRGINIA ELECTRIC AND POWER COMr'ANY L

Dock:t Nos.: 50 338&339 Sul:1 No.

M.707 section 1 Page 3 of 9 l

l

\\

Technleal Specification Chanans All Technical Specification changes described herein apply to North Anna l

Units 1 end 2 with the exception of Specification 3.4.1.2, which is only

[

applicable to Unit 2.

Where possible, the Unit 2 specifications and associated bases have been modified to be consistent in content and format with the Unit 1 speelfications.

These changes are considered to be administrative in nature.

During the discussion of each specification, setpoints that are appilcable to Unit 2 only, will be in parentheses.

i Technical Specification 3.1.2.2 REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING i

An existing footnote to Technical Specification' 3.1.2.2 has been revised to specify that only one boron ilow path is required to be operable whenever the temperature of one or more of the RCS cold legs is less than or equal to 316'F (358'F for Unit 2).

This requirement is.provided to ensure consistency with the requirements of T.S. 3.1.2.4 - (charging pump operability), and to ensure that actual operating conditions' are consistent with those assumed-in the mass addition transient analysis.- The rnass addition transient analysis assumes that only one charging pump will be operable below the temperature defined in T.S. 3.1.2.2.

-The 316'F value (358'F for Unit 2) corresponds to the pressurizer safety' valve lift setpoint of 2485 psig, on the composite lieatup and cooldown curve. Below these temperatures, the anticipated low temperature accidents may be adequately mitigated by the automatic action of the PORV.

In addition, for Unit 2 only, this footnote has been applied to MODE 3 Applicability.

This is necessary because the new Unit 2 setpoint involves MODE 3 operation between 350 F and 358'F.

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 3388339 1

s:rl:t No.

91 707 Section 1 Page 4 of 9 l

Technical Specification 3.1.2.4 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS OPERATING An existing footnote to Technical Specification 3.1.2.4 has boon revised to f

specify that a maximum of one centrifugal charging pump shall be operable whenever the temperature of one or more of the RCS cold legs is l

less than or equal to 316'F (358'F for Unit 2).

This requirement is provided to ensure that actual operating conditions are consistent with those assumed in the mass addition transient arealysis.

The mass addition transient analysis assumos that only one charging pump will be operable below the temperature defined in T.S. 3.1.2.4.

The 316'F value (358'F for Unit 2) corresponds to the pressurizer safety valve lift setpoint of 2485 psig, on the composito heatup and cooldown curve.

In addition, for Unit 2 only, this footnote _has been applied to the MODE 3 Applicability because the now Unit 2 setpoint involves MODE 3 operation between 350*F and 358'F.

Technical Specification 3.4.1.2 REACTOR COOLANT SYSTEM - SHUTDOWN MODE 3 This change applies to Unit 2 only.

A footnote was added to Specification 3.4.1.2 to specify that a reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 358'F unless the secondary water temperature of each -steam generator is less than 50*F above each of the RCS cold leg temperatures. This is necessary because the new setpoint encompasses a small portion of MODE 3 ( >350'F s.358'F).

This requirement is provided to ensure' that actual operating conditions are consistent with those assumed in the heat addition-transient analysis.

The heat addition transient assumes that a 50 degree teroperature differential exists between the secondary and primary sides o' the steam generator when-a reactor coolant pump is started. The 358'F value corresponds to the pressurizer safety valve -lift setpoint of 2485 psig on the composite heatup and cooldown curve.

VIRGINlA ELECTRIC AND POWER COMPNW h, y.

v_

,,r

, ~,

+.,,,.%,'.p.-

y

,'p_.

.,._y.m

.,.,.vr,,.

.m_m_,..

y.,

y.,

.-~.,

.9 p,m_.

Dock:t Nos.: 50 338&339 s: tid No, 91 70'1 section 1 Pago s of 9 Technical Specification 3,4.1.3 REACTOR COOLANT SYSTEM SHUTDOWN MODES 4&5 An existing footnote to Technical Specification 3.4.1.3 has been reviced to specify that a reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 316'F (358'F for Unit 2) unless the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

This requirement is provided to ensure that actual operating conditions are consistent with those assumed in the heat addition transient analysis.

The heat addition transient assumes that a 50 degree temperature differential exists between the secondary and primary sides of the steam generator when a reactor coolant pump is started. The 316'F value (358"F for Unit 2) corresponds to the pressurizer safety valve lift setpoint _ of 2485 psig, on the composite heatup and cooldown curve.

Technical Specification 3.4.9.1 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS T.S. 3.4.9.1 Indicates that T.S. Figures 3.2 2 and 3.4 3 provide pressure and temperature operating limitations at criticality.

The criticality limit line has been eliminated from the heatup curve in Figure 3.4 2 in favor of the more restrictive T.S. 3.1.1.5 minimum temperature for criticality.

The t0 <t of T.S. 3.4.9.1 has been modified to reflect this change.

Technical Specification Figures 3.4-2 and 3.4-3 PRESSURE / TEMPERATURE LIMITS - HEAT UP&COOLDOWN CURVES Revised T.S. Figures 3.4-2 and 3.4-3 have been prepared presenting the revised Unit 1,12 EFPY and Unit 2,17 EFPY heatup and cooldown curves.

The revised curves do not include allowances for temperaNre and measurement uncertainty.

The bases have been modified to reilect this change.

The development of these curves is discussed in greater detall in a later section. The criticality limit line has been excluded from the heatup curve in Figure 3.4 2 in _ favor of the-more restrictive Technical Specification 3.1.1.5, Minimum Temperature For Criticality.

(The Material Property Bases table, presented on these figures, has been added to the Bases.)

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 330&330 S:rlal No.

91 707 section 1 Page 6 of 9 Technical Specification 3.4.9.3 RCS OVERPRESSURE PROTECTION SYSTEMS The LTOP sotpoints have boon changed to provido bounding heatup and cooldown curve protection.

The developmont of those revised sotpoints is discussed in greator detail in a later section.

For Unit 2, LCO 3.9.4.3 (c),

has boon ollminatod.

The maximum pressurizer water volume requiremont has boon eliminated in favor of requiring low temperaturo ovorpressuro protection via automatic operation of the LTOP system below 321*F.

The maximum pressurizer water volume requirement was ollminated from the Unit 1 Specification 3.4.9.3 in a previous amendment (3).

Technical Specification 3.5.2 EMERGENCY CORE COOLING SUBSYSTEMS Tavg 2 350'F ACTION c, has boon rovised to allow the provisions of Specificatien 3.0.4 to be not applicable to ACTIONS a and b for one hour following heatup above 316'F (358'F for Unit 2) or prior to cooldown below 310'F (358 F for Unli 2).

Also, for ' Unit 2 only, a footnoto and a # sign have boon added which are applicable to LCO 3.5.2 (a) and (b).

This added footnoto specifies that a maximum of one centrifugal charging pump shall be operable whenover the temperature of one or more of the RCS cold legs is loss than or equal to 358'F for Unit 2 and is necessary because the now Unit 2 setpoint involves MODE 3 operation between 350 F and 358"F.

Those changes ensure that actual operating conditions are consistent with those assumed in the mass addition transient analysis.

The mass addjuon transient analysis assumos that only one charging pump will be operable below 316 F (358'F for Unit 2).

The 316'F value (358'F for Unit 2) corresponds to the pressurizer safety valvo lift sotpoint of 2485 psig, on ll.a corr.posite heatup and cooldown curve.

VIRGINIA ELECTRIC AND POWER COMPANY

_ - ~

- - - ~. - - - - -

Dod:t Nos.: 50 3384339 S:rlal No.

91 707 Section 1 Page 7 of 9 l

l Technical Specification 3.5.3 EMERGENCY CORE COOLING SUBSYSTEMS Tavg < 350*F The existing footnote in the Specification has been revised such that a maximum of one centrifugal charging pump shall be operable whenever the temperature of one or more of the RCS co:d legs is less than or equal to 316'F (358'F for Unit 2).

These changes ensure that actual operating l

conditions are consistent with those assumed in the mass addition transient analysis.

The mass addition transient analysis assumes that only one charging pump will be operable below 316'F (358'F for Unit 2).

The 316'F value (358'F for Unit 2) corresponds to the pressurizer safety valve lift setpoint of 2485 psig, on the composite heatup and' cooldown curve.

Conclusions The attached Technical Analysis supports the following conclusions:

The heatup and cooldown curves requ! red by Appendix G of 10 CFR 50 have i

been extrapolated to 12 EFPY and 17 EFPY for North Anna Units 1 and 2, respectively, by including the effects of the incremental radiation

+

exposure on the reactor vessel beltline region.

The results are referenced to the analyses of the North Anna Units 1 and 2 Capsule U toaults.

The revised Appendix G curves were prepared using standard B&W and Westinghouse methodologies including Regulatory Guide 199 Rev. 2.

LTOP 1

setpoints were developed to provide bounding heatup and cooldown curve protection for the worst case mass-and heat addition low temperature overpressure transients.

The next Unit 1 reactor vessel surveillance capsule -(Capsule-X) is scheduled to be-removed after the tenth fuel cycle (10'EFPY) which allows sufficient time for analysis prior to exceeding 12 EFPY, The next Unit 2 reactor vessel surveillance capsule (Capsulo W) is scheduled - to be removed after the thirteenth fuel cycle (15 EPPY) which allows sufficient-time for analysis prior to exceeding 17 EFPY.-

The heatup and cooldowr.,urves prepared.by B&W and-Westinghouse were determined in a conventional manner according to Section lll of the ASME code as required by 10_ CFR 50 Appendix G.-

Both-- steady state and transient thermal conditions were considered in order -to bound the possible combinations of pressure (i.e. membrane) and thermal stresses.

i VIRGIN lA El.ECTRIC AND POWER COMPANY

- - ~

Dock t Nos.: 50 3388339 S:rbl No.

91 707 Section 1 Page 8 of 9 The new North Anna Unit 1 low temperature ovorpressure protection system PORV lift settings should be loss than or equal to 450 psig whenever any RCS cold leg temperature is loss than or equal to 270'F, and loss than or equal to 300 psig whenever any RCS cold log temperature is less than 150'F.

The now North Anna Unit 2 low temperature overpressure protection sys' o PORV lift sottings should be less than or equal to 510 psig wha iover any RCS cold leg temperature is less than or equal to 321'F, and less than or equal to 360 psig whenover any RCS cold log temperature is less than 210'F.

Pressurized Thermal Shock (PTS) ovaluations were madn for the limiting boltline locations.

It was demonstrated that: (1) predicted end of license fluences do not result in RTPTS values in excess of the screening criteria when calculated using thu methodology of Regulatory Guide 1.99, Revision 2;

(2) thoro is an excellent comparison-between-exporimentally determined and calculated vo si fluences; and (3) the extrapolated fluences at the burnup limit to which the revised hoatup and cooldown curves are applicable for each unit are r.ignificantly loss than th extrapolated end of Ilconse fluences (which have 'oeen demonstrated to not result in a violation of PTS screening criteria).

On this basis it may be concluded that there is neither a significant change in predicted RTPTs values; nor is there a PTS concern for either unit up to the burnup limit to which the revised heatup and cooldown curves are valid.

VIRGINIA ELECTRIC AND POWER COMPANY v

--..--y

<c, - - - - -, ~

,-w.,

...m.

.e-,,w.,

--~.-w.-

<-n.p,,-

-,e.,,,,-r

-re.--~

m--e-

=--

ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES TECHNICAL SPECIFICATION CHANGE REGUEST FINAL COPY NORTH ANNA POWER STATION - UNIT 1 l

VIRGINIA EL CTRIC AND POWER COMPANY

BEA_CTMTY cot /TROWSIE!aS FLOW PATHS - OPERATit1G LIMITING C_OiJDITION FOR OPERATION 3.1.2.2 Each of the following boron injection flow paths shall bo 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 Cociant System.

AEEUCAD'.LII1:

hiODES '. 2, 3 AND 44.

ACI]Ohl.

a With tho flow path from the boric acid tanks inoperablo, restoro the Inoperable flow path to OPERABLE status whin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or bo in at loact HOT STANDBY and borated to a SHUTDOWN A'ARGIN equivalont to at least 1.77% Ak/k at 200*F within the nort 6 ho **1 rostore the flow path to OPERABLE status within the next 7 days or bo 1. D;OLD 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 phth from the refueling water storage tank inoperab'o, restoro 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 />.

SURVEILLANCE REOUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a At least onco por 7 days by verifying that tN smporaturo of the heat traced portion of the flow path from the boric acid ta'4s is 2115'F.

Only one boron injection flow path is required to bo OPERABLE whenever the temperature of one or more of the RCS cold logs is less ths, or equal to 316*F.

l l

l l

NORTH ANNA UNIT 1 3;4 19 Amendment NoA 6,68,4+7,

i REACTMTYCONTROL SYSTEMS CHARGING PUMPS OPERATING

]

JMITING CONDITION FOR OPERATION 3.1.2.4 At least two charg!ng pumps shall bo OPERABLE.

AEf1[CABILIIY:

MODES 1,2,3 and 4*.

AClEU:

With only one charging pump OPERABLE, restoro a second charging pump 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 hour6.944444e-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 bo 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 of Specification 3.0A are not applicable for one hour following hoatup above 316'F or prior to cooldown below 316"F.

.SURVEll1ANCE REOUIREMENTS.

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

4.1.2.4.2 All chaming pumps, except the above required OPERABLE pump, shall be doraoMtrated inoporable at least onco por 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 rnoro of the RCS cold logs is loss than or equal to 316'F by verifying that l

the switches in the Control Room have boon placed in the pull to lock position.

A maximum of one contrifugal charging pump shall be OPERABLE whenover the temperature of one or more of the RCS cold legs is less than or equal to 316'F.

l i

NORTH ANNA UNIT 1 3/4112 Amendment No.4A4,4+7,

_,,9=..*ges%yu 4

.,..m,.,i,%g, g4._.y.

j,,,,_3

,_3_

g,.

m e

r--.m y.w

! REACTORCOOl## SYSTEM i

SHLITDOWN UMITING CONDITION FOR OPERATION 3.4.1.3 a At least two of the coolant loops listed below shall be OPERADLE:

1. Roactor Coolant Loop A and its associated steam generator and reactor coolant pump,'

2. Roactor Coolant Loop 8 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 Heat Removal Subsystem D."

b. At least one of the above coolant loops shall be In operation."'

APPLICABILITY:

MODES 4 and 5 ACI m a With loss than the above required loops OPERABLE, immediately Initiato correctivo action to return the requlted loops to OPERABLE status as soon as possible; bo in COLO SHUTDOWN within M hours.

b. With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immodlately initiato 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 temperctures less than or equal to 316*F untoss the secondary water temperature of l

each steam generator is less than 50*F above each of the RCS cold log temperatures.

• The olisite or emo,gency power soifree may be inoperable in MODE 5.

• *

• All reactor coolant pumps and residual boat removal pumps may be de energized for up to t hour provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) _ core outlet temperature is maintained at least 10*F below saturation temperature.

l-NORTH ANNA UNIT 1 3/4 4-3 Amendment NoA6,82,4+7,

f1EACTOR COOL >NT SYSTDA T4.4.9 PRESSURE / TEMPERATURE LIMITS REACTORCOOLR4T SWIEM UMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizor) temperature and pressure shall be limited in accordance with the limit linos shown in Figuros 3.4 2 and 3.4 3 during heatup, cooldown, and inservice leak and hydrostatic testing with:

l

a. A maximum heatup of 60'F in any one hour period,

b. A maximum cooldown of 100'F in any one hour porlod.

c. A maximum temperature chango of Icss than or equal to 10'F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLtCABILITY:

At all times.

ACTK)N With any of the above timits exceeded, rostoro the temperaturo and/or pressure to within the limit within 30 minutos; perform an engineering ovaluation to l

determine the offects of the out-of limit condition on the structural integrity of the Reactor Coolant Systorn; dolormine that the React < ' Coolant System romains acceptable for continued operations or be la at leastgpT 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 reduce the RCS Tavg and pressuro,. less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

F SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be dolormined to be l

within the limits at least onco per 30 minutos during system heatup, cooldown and inservice leak and hyd ostatic testing operations.

4.4.9.1.2 The reactor vessel materia! irradiation surveillance specimens shall be removed and examined, to dotormino changos in material proporties, at the intervals required by 10 CFR 50, Appendix H. The results of thoso examinations shall be used to update Figures 3.4 2 and 3.4 3.

i l

NORTH ANNA UNIT 1 3/4426 Amendment No.-86,

Figure 3.4 2 Unit 1 RCS HEATUP P/T Limits Valid to 12 EFPY Heatup Rates: 0 60'F/Hr.

(Margins for Instrument Errors NOT Included) i i

2750 -

' MATERIAL PROPERTY BASf3

?-

9. *. t..

P. <.1.-

t.-.r-. >. -

, Controlling Material:

3 7

f.

, Lower Shell Plate Forging

-,s.

2500 - Copper Content: 0.15 W1.%

i-

..!

• 4-

' Nickel Content: 0.80 Wt %

~-

, Initial RTndt: 38'F

>. Leak Test 4

i.

R

. RTndt After 12 EFPY:

I-Limit

-i

• f* ~ f~-- > '

2250 -- j f4T' 145'F

~

C

.F.

3/4T.122'F 3

. Heatup Rate 560'F/Hr.

.l.

.;... l..

i..}.

i P

2000

.,.1..

r

.y.

3

.y.

.p...:....:...q.

.y.

e 3

S 1750

(.

U

~.

. s..

.s.

. w...u.

r 1500 e

UNACCEPTABLE 4

.b

'r..:-

.s.

.s.

.s.

OPERATION

. ~..

1250 l

l l

p s

.:..q.

.j..p.

. :....p.

g 1000 - r.-

. s.... 4..

.t

..s...

.s.

i

.s...s......

..r......

. s... s..

750

..;. A'.. :..

ACCEPTABLE 7

T..

. ;...+f.... r.T. T-. :".". :.- OPERATION -

500

.s.

.i... :.... :..

250,

80 100-120 140 160-180 200 220 240 260 280 300 320 Minimum RCS Cold Leo Temperature + degrees F 1

NORTH ANNA. UNIT 1 3/4427 Amendment NoA6,64.4+7, ti-w-+

wm ----- - - - - -

-m

.-a 2

Figure 3.4 3 Unit 1 RCS COOLDOWN P/T Limits Valid to 12 EFPY Cooldown Rates: 0100'F/Hr.

(Margins for Instrument Errors NOT included) 2750 MATERIAL PROPERTY BASIS 4

4 Controllin9 Material:

~.

~*-

'e- '"

' + ' '

'T]*. 'c.

Lower Shell Plate Forging 2500 - Copper Content: 0.15 W1.%

. :....:..)...:...).

.i.

Nickel Content: 0.80 Wt.%

..P..' )* *.: * '* *j *. *. :.. '.

• Initial RTndt: 38'F R

2250 - RTndt After 12 EFPY:

.. '..l l

l C

'14T,145*F

.s.

. s....

. i..

.u.

S 5i4T,122*F

~

~-

. - -.~.

.~.

• f' ~ *. ' '! 3 '!' ' 'b

' l' '

2000 i Cooldown Rate: s100'F/Hr.

' l- * *'*

4

.y...;.

. ;...>..y.

.q.

.,s.

.,. <. 3

.q..y.

P

. e....,.

.r

.s......

..,.....s.....

. s..

.s.

r e

1750 e

e.

s

.s

.s.

.s....

U 1 0

.s.

r P

3 3

OPERATION

..:.. :/...:.

e 1250 t

P s

1000 l

.p.

.:..}.

.:...{.

0

.:..y...;.

.}....g 3

.:...q.

3

.:..,.t..

.:..y...:.

.t.

.;...e.

q..

'F/hr

.t. p

.4

.s.

.,s...,..

.s.

}.-

+y-..

500 --

to.

.,c.

4o.

,,o...y +

. t,.

ACCEPTABLE

.e.

s o*

3

.:...t..y.

3 OPERATION

.e 250

.s....

O i' 80 100 120 140.160.180 200 220 240 260 280 300 320 Minimum RCS Cold Leg Temperature degrees F NORTH ANNA. UNIT 1 3/4:4 28 Amendment NoA6,68,44 7,

(

REACTOR COOLAVTSYSTOA OVERPRESSURE PROTECTION SYSTEMS UMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a Two power operated tellof vr.1ves (PORVs) with a lift setting of: 1) less than or equal to 450 prig whenever any RCS cold leg temperaturo is less than or equal to 270'F, and 2) less than or equal to 390 psig whenover any RCS cold l

leg temperaturo is loss than 150*F, or

b. A reactor coolant system vont of greator than or equal to 2.07 square inches.

APPLICABILITY:

When the temperature of one or more of the RCS cold legs is less than or equal to 270'F, except when the reactor vessel head is removed.

l ACTION-a With one PORV inoperable, olther restore the inoperablo PORV to OPERABLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vont(s) within the next 8 houtst maintain the RCS in a vented condition until both PORVs have boon 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 />; maintaln the RCS in a vented condition un'il both PORVs have been restored to OPERABLE status.

c. In the event olther the PORVs or the RCS vent (s) arJ 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 (s) on the transient and any corrective action necessary to provent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

l NORTH ANNA UN!T 1 3/4 4 31 Amendment NoA6,N,44-7, m m-m m

ErAERGENCYCORE COOLN3 SYSTEMS ECCS SUBSYSTEMS Tavo a 350*F UMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. Ono OPERABLE centrifugal charging pump,

b. One OPERABLE low head safety injection pump,

c. An OPERABLE flow path capable of transferring fluid to the Roactor 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 dischargo of the outside recirculation spray pump.

APPLICABILITY:

MODES 1,2 and 3.

ACTION:

a. With one ECCS subsystem inoperablo, 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 HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. In the ovent the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report sf.all 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 actuation cycles to date.

c. The provisions of Specification 3.0.4 ar0 not applicable to 3.5.2.a and 3.5.2.b for one hour following heatup above 316'F or prior to cooldown below 316'F.

l l

NORTH ANNA - UNIT 1 3/4 5 3 Amendment No.0,44,4-N,

EMERGENCYCORECOOLN3 SYSTEMS ECf.S SUBSYSTEMS Tavo < 3507

_ LIMITING CONDITION FOR OPERATION 3.5.3 As a minirnum, one ECCS subsysterr comprised of the following shall be OPERABLE:

a Ono OPERABLE contrifugal charging pump #,

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

c. An OPERABLE flow path capable of automatically transferring fluid to the reactor coolant system when taking suction from the refueling water storage tank or from the containment sump when the suction is transferiod during j

the recirculation phase of operation or from the discharge of the outsido i

recirculation spray pump.

APPL'OABILITY:

MODE 4.

ACTION-a With no ECCS subsystem OPERABLJ becauso of the Inoperability of olther the contrifugal charging pump or the flow path from the refueling water storage tank, restore at least ono ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the 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, restoro at least one ECCS subsystem to OPERABLE status or maletain the Reactor Coolant System Tavg loss than 350'F by use of alternato heat removal methods.

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall bo prepared and submitted to the Commission pursuant to Specification 6.5.2 within 90 days describing.the circumstances of the actuation and the total accumulated actuation cycles to dato.

l l

A maximum of one centrifugal charging pump and one low head safoty injection pump shall bo OPERABLE whenever the temperature of one or more of the RCS cold logs Is-less than or equal to 316'F.

.l NORTH ANNA UNIT 1 3/4 5 6

. Amendment No.0,4-6,84,4+7,

DAEITIJCYCORE COOL 1Y3 SYSTEMS SURVEILLANCE REOUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE por the applicablo Surveillanco 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 inoperablo at loast ohco por 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 oold logs is loss than or equal to 310'F by verifying that the switches in the Control Room ato in the pult l

to lock pos!! ion.

/

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

Amendment NoA6.H-7,

d REACTMTY CONTROL SYSTEMS MSES 3/4.1.2 BORATION SYSTEMG (ContinutJ)

With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injaction systems are provided to ensure single functional capability in the event an assumed fal!ure renders one of the systems inoperable, Allowable out-of-service-periods ensure that minor component repair or corrective action may be completed without undue rlst to overc'l facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from expeded operating conditions of 1.77% Ak/k after xenon decay and cooldown to 200'F.

This expected boration capability requirement ot, curs at EOL from full powcr aquilibrium xenon conditions and requires 6,000 gallons of 12,950 ppm borated water from the boric acid storage tanns 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 excep' t% required OPERABLE pump to be inoperable below 316'F provides assurance that a mass adJNon pressure transient l-can be relieved by the operation of a single PORV.

With the RCS temperature 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 restriotions prohibiting CORE ALTERATIONS and positive reactivity change in-the event the singl ' a Wlon syr, tem becomes inoperable, The boron capabl. -

. aired below 200'F. is sufficient to provide a SHUTDOWN MARGIN of 1.77% ak/h atter zenon decay and cooldown from 200*F-to 140'F. - This condition requires either 1378 gallons of 12,950 ppm borated water from the boric acid e

storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of-discharge line location and other phvsical characteristics. The OPERABILITY of one boron injection system during REFUEL"23 insures that this system is available for reactivity-control while in MODE 6.

NORTH ANNA UNIT t B 3/4 1-3 Amendment No.4.14.G-8,00.4+7, m

3/4.4 REACTMTY CONTROL SYSTEMS N

3L4.A.1 REACTOR COOLANT [f0PS The plant is designed to operate with all reactor coolant loops in ope allon and maintain the DNRR above the design limit during all normal operations and anticipated translerw in MODES 1 and 2 with one reactor coolant loop not in operation, this specification requirea that the plant be in at least HOT STANDBY witnin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal caphMity for removing decay heat; however, single failure considerations require that two loops be OPEMBLE.

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufflctent 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.

After the reactor has shutdown and entered into MODE 3 for. at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow ia'.; of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperatare less than or equal to 140*F. Since the decay heat power production rate decretces with time after reactor shutdown, the requirements for RHR i

system decay heat removal also decrease. Adequate decay heat removalls provided as long as the reactor has been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficleat to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 316*F are provided to prevent RCS pressure transients, caused by energy additions l

from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will-not exceed the limits of Appendix G by restricting staring of the RCPs to when the secondary water temperature of each steam generator is iess than 50'F above each of the RCS cold le' ' mperatures.

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

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

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 Amendment NoA6,42,W,W,W,

REACTIVITYCONTROLSYSTEMS BASES The ACTION statement permitting POWER OPERATION to ec1tinue for limited time periods with the primary coolant's specific activity > 1.0 pCl/ gram DOSE EQUIVALENT l 131, but within the allowable limit shown on Figure 3.41, accommodates possible iodine phenomenon which may occur followi.1g changes in THERMAL POWER.

Reducing Tayg to < 500'F prevents the rel9ase of activity should a steam generator tube rupture 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 lodine 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 LIMITS All c( mponents 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 UFSAR. 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 goveming location.

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. The:>e stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the cuter _ wall of _the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp;.

L 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.

l-NORTH ANNA UNIT 1 B 3/4 4-G Amendment No.-96,44-7,

REACTMTY CONTROL SYSTEMJ BtGES The heatup limit curve, Figure 3.4 2, is a composite curve which was prepared by l

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 l

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 gradl6nts tend to produce tensile stresses while producing compressive stre.sses at the outside wall.

The heatup and cooldown curves are prepared based upon the mNt limiting value of the predicted adjusted reference temperature _ at the end of 12 EFPY, The adjusted referer.ce l-temperature was calculated using results from a capsule removed after the sixth fuel cycle.

The results are documented in Westinghouse Report WCAP 11777, February 1988 and Babcock and Wilcox Report BAW-2146, October,1991.

The reactor vessel materials have been tested to determine their, altlal RTNDT. The results of these tests are shown in the UFSAR and WCAP 11777. Reactor operation and resultant fast neutron (E>1 Mov) Irradiation will-cause an increase in the RTNDT.

Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using US NRC Regulatory Guide 1.98, Revision 2.

The heatup and cooldown limit curves (Figure 3,4 2 and FiQure 3.4-3) inc!ude predicted I

adjustments for this shift in RTNDT at the end of 12 EFPY. The reactor vessel beltline region material prnperties are listed in Table B.3.41.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing end evaluating, in accordance with ASTM E185 70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift 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 ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for inservice leak and l

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-11777 to l

assure compliance with the requirements of Appendix H to 10 CFR Part 50.

l 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 analysis performed in accordance v ith the ASME Code l

requirements.

NORTH ANNA - UNIT 1 B 3/4

,-7 Amendment NoA47,

1REACTMTY CONTROL SYSTEMS BASES 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 270 F. Either PORV has adequate relieving capability to protect the RCS l

from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to TO'F above the RCS ccid leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.

Automatic or passivo low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 270*F. This temperature is the water l

temperature corresponding to a metal temperature of at least the limiting RT DT+C>^F+

N Instrument uncertainty.

Above 270'F administrative control is adequate p;.cction 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.

Table B.3.41 MATERIAL PROPERTY BASIS Controlling Material:

Lower Shell Plate Forging Copper Content:

0.15 Wt.%

Nickel Content:

0.80 Wt.%

initial RTndt:

38'F RTndt After 12 EFPY:

1/4T. 145'F 3/4T, 122*F Cooldown Rate:

5100* F/H r.

Heatup Rate:

560

• F/ H r.

i i

\\

NORTH ANNA - UNIT 1 B 3/448 Amendment No.-74,4+7,

REACTIVITY CONTROL SYSTEMS E%SES ECCS SUBSYSTEMS (ContinuMfi 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 316*F provides assurance that a mass addition pressure transient can bo l

relieved by the operation of a single PORV.

The Saveillance 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 OPERABILITY is maintained.

1/4.5.4 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.

l The limits on injection tank minimum contained volume and boron concentration ensure that l

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 111*F at 15,750 ppm boron.

NORTH ANNA - UNIT 1 B 3/4 5 2 Amendment NoA6.68,4+7,

i ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS CHANGES TECHNICAL SPECIFICATION CHANGE REQUEST FINAL COPY NORTH ANNA POWER STATION - UNIT 2-l VIRGINIA ELECTRIC AND POWER COMPANY

REACTMW cot #ROL SYSTEMS

)

FLOW PATFIS-OPERATING UMITING CON,DITION FOR OPERATION, 3.1.2.2 At least two of the following three 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.

)

b. Two flow paths from the refueling water storage tank via charging pumps l

to the Reactor Coolant System.

APPLICABILITY:

MODES 1,2,3# AND 4.

l ACTION With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System 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% delta k/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 at least two flow paths to OPERABLE status within the next 7 days or be 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 />.

SURVEILLANCE PC.OUIREMENTS 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE:

a At least or,ce per 7 days by verifying that the temperature of the heat traced l

portion of the fbw path from the boric acid tanks is greater than or equal to l

115'F when it is a required water source.

i 1

Only 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 358*F.

l NORTH ANNA - UNIT 2 3/4 1-9 Amendment No.-64,4-29,

REACTMTYCONTROLSYSTEMS CHARGING PUMPS-OPERATING UMITING CONDITION FOR OPERATION 3.1.2,4 At least two charging pumps shal! be OPERABLE.

APPLICABILITY:

MODES - 1, 2, 3# and 4.

ACTION:

With only one charging pump OPERABLE, restore a second charging pump 'o OPERABLE status wi;hin 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 eaulvalent to at least 1.77% delta k/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 />; restor 13 a second charging pump to OPERABLE status within the next 7 days or be 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 of Specification 3.0.4 are not applicable for one hour following heatup above -

358'F or prior to cooldown below 358'F.

_l:

SURVEILLANCE 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 cf greater than or equal 'o 2410 psig when tested pursuant to Specification 4.0.5, 4.1.2.4.2 All charging pumps, except the above required-OPERABLE 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 358*F by verifying that l

the control switch is in the pull to lock position.

1. A maximum of one centrifugal charging pump 'shall be OPERABLE whenever the

-- l temperature of one or more of the RCS cold legs is less than or_ equal to_358*F.

l-L NORTH ANNA - UNIT 2 3/4 1-12 Amendment No.

.~.

~, _

. -. ~.. -.,

4 REACTORCOOLANTSYSTEM HOT STANDBY UMITING COND_ITION FOR OPERATION- - _ - -. _ -

3.4.1.2

a. At least two of the reactor coolant loops listed below shall be OPERABLE:

1. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, l

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

l

3. Reactor Coolant Loop C and its associa'ed steam generator and reactor coolant pump,'

l

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

l APPLICABILITY:

MODE 3 ACTIC'.*

a. With less than the above required loops OPERACLE, restore the required loops to OPERABLE status with in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

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

SURVEILLANCE REOUIREMENTS l

4.4.1.2.1 At least the above required reactor. coolant pumps, if not in operation, shall be i

determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

A reactor coolant pump shall not be started with one or more of the RCS cold leg-l temperatures less than or equal to 358*F unless the secondary water temperature of each l

steam generator is less than 50*F above each of the RCS cold leg temperatures.

All reactor coolant pumps may be de-energized for up to 1 Mur provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 1o'F below saturation temperature.

The requirement to have one coolant loop in operation is exempted during the performance of the boron mixing tests as stipulated in License Condition 2.C(15)(f) and 2.C(2o)(b).

NORTH ANNA UNIT 2 3/4 4-2 Amendment NoA9,

4,4:

s~w e

.4 a__n Lt

~ -

,,4 REACTORCOOLWT SYSTBA SHUTDOWN UMITING CONDITION FOR OPERATION 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 genorator and reactor coolant pump,

2. Reactor Coolant Loop B and its cssociated 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 Heat Removal Subdistem B.**

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

APPLICABILITY:

MODES 4 and 5 ACTION.

a. With less than the above required icops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN 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 immediately 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 358'F unless the secondary water temperature of l

each steam generator is less than 50 F above each of the RCS cold leg temperatures.

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 1 hcur provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

1 NORTH ANNA - UNIT 2 3/4 4-3 Amendment No.

l l

REACTOR COOUNTSYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS i

REACTOR COOUNTSYSTEM UMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown in Figures 3.4-2 and 3.4 3 during heatup, cooldown, and inservice leak and hydrostatic testing with:

l

a. A maximum heatup of 60'F in any one hour period,

b. A maximum cooldown of 100'F in any one hour period.

c. A maximum temperature change of less than or equal to 10*F in any one hour perloo during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the e;fects of the out of limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations 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 reduce the RCS Tavg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be j

witr,In the limits at least once per 30 minutes during system heatup, cooldown and uservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determire changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 nd 3.4-3.

l 1

l NORTH ANNA - UNIT 2 3/4 4-26 Amendment No.-60,

Figure 3.4-2 Unit 2 RCS HEATUP P/T Limits Valid to 17 EFPY.

Heatup Rates: 0 60 F/Hr.

(Margins for Instrument Errors NOT Included)--

2750 MATERIAL PROPERTY BASIS *t*

P

• > ' 'i* t.

' ?. -*t-t.

'. ?

• Controlling Material:

.t.

.t.

....i.

3 Lower Shell Plate For in

.6

4..

.6.

2500 -- Copper Content: 0.13 Wt.%

~

Leak Test

's.

.t.

Nickel Content: 0.83 Wt.%

A Inlllai RTndt: 56'F

..:...k.

R

' RTndt After 17 EFPY:

t- -t P-i- 9

-?- i-

, ' -.?-

2250 -

C 1/4T.196*F

,...g.

7

.g.

,...g.

S 3/4T.172'F

.6 Heatup Rate s60*FIHr.

m-

.,s..;.

6 P

2000 r

.:..y.

3 3

.a.

.t.

.;..~: . ':. '

'. '. '* ;* *' z '..' :'.'* *; : . :"..'

'. ;* *.~. ' ".

O S

1750 S

4 u

....a.

.~.

.,.... ~.

1500 e

4 4

UNACCEPTABLE h..

e 4- -4 e

F -- -4

+

P.

r

.s....

.s....

.s.

OPERATION 1, 4.:. /

. /.. J..:.

.:.. J.

,,3o p

S

.:.. t...:.

3 3..

9 1000

. r.....

.n.

2..

. r..

.....s

. :...;...:...).

. t..

.s.

.s 750

. :...). /('.

... ACCEPTABLE OPERATION

.t.

. '.'..>.,.p

.s

.t.

500

.,..s.

. s...,..

4 250 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 Minimum RCS Cold Leg Temperature - degrees F' l-NORT H ANNA. UNIT 2 3/4427 Amondment No. 60

. 1 1

s - - -....

e e.~

N v e. r e.4 ;r '

ri t

I Figure 3.4-3 Unit 2 RCS COOLDOWN P/T Limits Valid to 17 EFPY. CooldoWn Rates

0-100 F/Hr.

(Margins for Instrument Errors NOT Included) 2750

.~.

MATERIAL PROPERTY BASIS

. Controlling Material:

. ~.

Lower Shell Plate Forging v

r-r-

1" Y

Y r-r-

2500 -

. Copper Content: 0.13 Wt.%

4

.'Nickel Content: 0.83 W1.%

r. -* :...; -.;.

r. - 7..- ;,.-

init. l RTndt, 56,F ia R

2250 - RTndt After 17 EFPY:

C

. 1/4T,196*F

. t....

. 2,.

.s.. t..

.t

- 3/4T,172*F 2000 ; Cooldown Rate: s100'F/Hr.

f- !* *i-D O

' -f- -!-

3

.;..q.

.y.y..t..q.

.,s..).

P

~.

. i,..;.

.t.

.a.

.t.. $,.

.t.

.t.

......j..

r e

1750

$~

.t.

[..

.4 4

..P,

.4

b....

e.t.

.4,.

S u

1500

-UNACCEPTABLE g

.... s..

OPERATION.

.~.

.......c.

e.

....s.

1250 4

4 P

.. -...A s

1000 l

. :..p..;..<...:.

4

.:... :...).

g 3

.{.

4

.:..:..g

.i.

.:._. :... ;...).

.t.

750-3 3.q.

.:. g/

.t..i.

. :... :...).

/..

. 'F/hr

. A./u

.s..4. p p

.s..t..;.

s.

.t..t.

.L.

D.s o.

500- -

20-c-

p 4o.

fI s @...

.,D. i. ' >. '*?-

. ACCEPTABLE: I..! -

1 60

.y..y..;.

.y.. y.

-OPERATION i

+F,17- - -

, 50

. s..

.u c.

..s....

.s.......

.s.

. m..,.

.~.

0 80 100120140160180 200 220 240 260 280'300 320 340 360 =

~

Minimum RCS Cold-Leg Temperature - degrees F NORTH ANNA-UNIT 2 3/4 4-28 Amendment No.40, I-i

REACTORCOOUNTSYSTEM OVERPRESSURE PROTECTION SYSTEMS UMITING 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: 1) less than or equal to 510 psig whenever any RCS cold leg temperature is less than or equal to 321*F, and 2) less than or equal to 360 psig whenever any RCS cold leg temperature is less than 210'F, or

b. A reactor coolant system vent of greater than or equal to 2.07 square inches.

APPLICABILITY:

When the temperature of one or more of the RCS cold legs is less than or equal to 321'F, except when the reactor vessel head is removed.

l ACTION

a. With one PORV inoperable, either restore the inoperable PORV to OPERALLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />st maintain the RCS in a vented conoition 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 vented condition until both PORVs have been restored to OPERABLE status,

c. In the avent 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 (s) on the transient al.d any corrective action necessary to prevent recurrence,

d. The provisions of Specification 3.0.4 are not applicable.

l JRTH ANNA - UNIT 2 3/4 4-30 Amendment No.40,

EMERGEfCf CORE COOUNG SYSTEMS ECCS SUBSYSTEMS - Tava GREATER THAN 350"F LIMITING CONDlT1ON FOR ODE _ RATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump #,

l

b. One OPERABLE low head safety injection pump #,

l

c. An OPERABLE flow path capable of 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.

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

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 actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

c. The provisions of Specification 3.0.4 are not applicable to Specifications 3.5.2.a and 3.5.2.b for one hour following heatup above 358*F or prior to cooldown below 358'F.

SURVEILLANCE REOUIREMENYS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve operators removed:

A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 358*F.

l NORTH ANNA - UNIT 2 3/4 5-3 Amendment No.

1

' i EMERGENCYCORECOOUNG SYSTEMS ECCS SUBSYSTEMS - Tava LESS THAN 40 E UM! TING 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 transferring flula to the-reactor coolant system when taking suction from the refueling watcr storage tank or from the containment sump when the suction is transferred during the recirculation phase of operation.

APPLICABILITY:

MODE 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 OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the 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 Ruactor Coolant System Tavg less than 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 l.

of the actuation and the total' accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle sha!! be provided in this Special Report whenever its value exceeds 0.70.

A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is -

less than or equal to 358'F.

l 1

NORTH ANNA - UNIT 2 3/4 5 6 Amendment No.,74, 1

.~

r,,

,.,g-.,

y

f2dEEOFJKNCORECOOl1G SYSTEMS SLGVEl' LANCE REOU:REMENTS 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 legs is less than or equal to 358'F by verifying that the control switch is in the pull to lock position.

l l

l 1

l i

l l

l NORTH ANNA-UNIT 2 3/4 5-7 Amendment No.

REACTIVITY CONTROL SYSTEMS IA*ES 3/4.1.2 BORATON 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) on emergency powar supply from OPERABLE diosol generators.

With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficiant to provide a SHUTDOWN MARGIN from expected operation conditions of 1.77%

delta k/k after xenon decay and cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6000 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.

With the RCS temperature bebw 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 prohibitin0 CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and toe Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 358'F provides assurance that a mass addition pressure transient l

can be relieved by the operation of a single PORV.

The boron capability required below 200'F is sufficient to provide a SHUTDOWN MARGIN of 1.77% delta k/k after zenon decay and cooldown.from 200aF to 140'F. This condition requires either 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, i

NORTH ANNA-UNIT 2 B-3/4-1-3 Amendment No.-5458,

3/L4 REACTIVITY CONTROL SYSTEMS IRSES 3/4.4.1 REACTOR OOOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above the design limit during all normal operations and anticipated translonts. In MODES 1 and 2 with one reactor coolant loop not in operation, inis specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in MODE 3, a single reactor coolant loop provides sufficient heat removal capability for rem *ng decay heatt however, single failure considerations require that two loops be OPE

'LE.

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 specif. cation requires two RHR loops to be OPERABLE.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a minimum RHR system flow rate of 2000 gpm in MODE 5 is pumhted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140*F. Since the decay heat power production rate decreases with ilme after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removails provided as long as the reactor hos been shutdown for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 358'F are provided to prevent RCS pressure transients, caused by energy additions l

from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of

(

Appendix G by restricting staring of the RCPe to when the secondary water temperature of each l

steam generator is less thar; 50'F above each of the RCS cold leg temperatures.

The requirement to maintain the boron concentration of an isolal?d 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. Verification of the boron concentration in an idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the i

boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at l

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 boron concentration stratification.

Startup of an idle loop will inject cool water from the loop into the core. The reactivity i

transient resulting from this cool water injection is minimized by delaying isolated loop startup l

until its temperature is within 20*F of the operating loops. Making the reactor subcritical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient.

NORTH ANNA UNIT 2 8 3/4 4 1 Amendment No.-120,+2-2,

i REACTMTY CONTROL SY5'TEMS SE The ACTION statement permitting POWER OPERATION to continue for limited time periods.with the primary coolant's specific activity greater than 1.0 - Cl/ gram DOSE EQUIVALENT l-131, but within the allowable limit shown on Figure 3.41, accommodates possible iodine phenomenon which may occur following changes in THERMAL POWER.

Reducing Tavg to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the till pressure of the atmoppheric steam relief valves.

The surveillance requirements provide adequate assurance that excessivo specific activity levels in the primary coolant will be detected in sufficient time tc take corrective action. Information obtained on lodine 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 LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and precsure changes. These cyc!ic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of losd cycles used for design purposes are provided in Section 5.2 of the UFSAR. 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 tenslie 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.

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 well 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.

L i

NORTH ANNA UNIT 2 8 3/4 4-6 Amendment No.48.

l l

l REACTIVITY CONTROL SYSTEMS EMSES The heatup limit curve, Figure 3.4-2, is a composite curve which was 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 are prepared based upon the most limiting value of the l

predicted adjusted reference temperature at the end of 17 EFPY, The adjusted reference temperature was calculated using results from a capsule removed after the sixth fuel cycle.

The results are documented in Westinghouse Report WCAP 12497, January 1990 and WCAP-12503, March,1990.

The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in the UFSAR and WCAP-12497. Reactor operation and resultant fast neutron (E>1 Mev) Irradiation will cause an increase in the RTNDT.

Therefore, an adjusted refe ce temperature, based upon the fluence and copper content of the material in questior d be predicted using US NRC Regulatory Guide 1.98, Revision 2.

The heatup and coo'

.vn limit curves (Figure 3.4-2 and Figure 3.4 3) include predicted adjustments for this snift in RTNDT at the end of 17 EFPY. The reactor vessel beltline region material properties are listed in Table B.3.41.

The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185 70, reactor vessel material irradiation surveillance specimens installed near the inside wall-of the reactor vessel in the core area. Since the neutron spectra at the irradiation-samples and vessel inside radius are essentially identical, the measured transition shift 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 ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation

exposure, l

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 oi 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 12-497 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 provM'ed to assure that the pressurizer is operated within the design cr!!eria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

NORTH ANNA - UNIT 2 2 3/4 4-7 Amendment No.

l l

-I REACTIVII V CONTROL SYSTEMS iMSES 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 rnore of the RCS cold legs aro less than or equal to 321*F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.

Automatic or passive low temperature overpressure protection (LTOP) is required whenever any RCS cold leg temperature is less than 32t'F. This temperature is the water temperature corresponding to a metal temperature of at least the limiting RTNDT+90'F+

instrument uncertainty.

Above 321*F administrative control is adequate 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 8811.

Table B.3.41 MATERIAL PROPERTY BASIS Controlling Material:

Lower Shell Plato Forging Copper Content:

0.13 WI.%

Nickel Content:

0.83 Wt.%

initial RTndt:

56*F RTndt After 17 EFPY:

1/4T, 196'F 3/4T, 172*F Cooldown Rate:

5100*F/Hr.

Heatup Rate:

s60*F/Hr.

3/4.4.10 STRUCTURALINTEGRITY 3/4.4.10.1 ASME CODE CLASS 1.2 and 3 COMPONENTS The inspection program for ASME Code Class 1, 2 and 3 Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for components is in compliance with Section XI of the ASME Boller and Pressure VesselCode.

NORTH ANNA - UNIT 2 B

3/4 4 8 Amendment No.

f

- =..

=

Pages B 3/4 4 9 thru 8 3/4 416 have beon deleted, 1

l l

l NORTH ANNA - UNIT 2 8 3/44-9 Amendment No.

-. - ~

REACTMTYCONTROL SYsIEMS BASES ECCS SUBSYSTEMS (Continuech l

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 contrifugal 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 inject.on pumps except the required OPERABLE purnp to be inoperable below 358'F provides ascurance that a mass addition pressure transient can be l

relieved by the operation of a single PORV.'

The Surveillance Requirements provided to ense c OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsys' OPERABILITY is maintained. Surveillance requirements for throttle valve _ position stops and flow balance testing provide assurance that proper flow resistance ar.d pressurs drop -

in the piping system to each injection point is necessary to:_ (1) prevent total pump flow from exceeding runout conditions _when the system. is in its-- minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS LOCA analyses, and (3) provide an acceptable level of _ total ECCS flow to all injection points equal to or above that assumed in the ECCS LOCA analyses.

3/4.5.4 BOROW 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 contained water volume I'mit includes an allowance for water not usable because.of' tank discharge line

~

. location or other physical characteristics.

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 111*F at 15,750 ppm boron, NORTH ANNA - UNIT 2 8 3/4 5 2 Amendment No.-64, i

,. a..

-.. -..-. ~,.

ATTACHMENT 4 DETERMINATION OF NO SIGNIFICANT H AZARDS CONSIDERATIONS (10 CFR 50.92 EVALUATION)

TECHNICAL SPECIFICATION CHANGE REQUEST NORTH ANNA UNITS 1 & 2 l

l VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 338&339 snial No.

91 707 section 4 Page 2 of 3 BASIS FOR NO SluNIFICANT HAZARDS DETEBMINATION The North Anna Reactor Coolant Systems (RCS), specifically the Reactor Pressure Vessels (RPV), are protected from material failure during low temperature operations by imposing restrictions on RCS pressure.

The heatup and cooldown curves as well as the Low Temperature /Over-pressure Protection System (LTOP) setpoints, provide the restrictions to bound the area of operation and ensure RCS protection from non-ductile failure.

The regulatory requirements for providing these restrictions and reevaluating them, are stipulated in 10 CFR 50, Appendix G.

The current heatup and cooldown curves and LTOP setpoints, will not be valid after 10 effective full power years (EFPY) cumulative core burnup.

According to our most recent estimates (September,1991) North Anna Unit 1 is expected to reach 10 EFPY in April,1993 and Unit 2 in September, 1993.

In anticipation of the expiration of these curves, Virginia Power has performed a safety evaluation to support implementing revised curves and setpoints.

These revised curves required by Appendix G of 10 CFR 50 have been extrapolated to 12 EFPY and 17 EFPY for North Anna Units 1 and 2, respectively, by including the effects of the incremental radiation exposure on the reactor vessel beltline region.

The revised curves were preparet using standard methodologies and Regulatory Guide 1.99 Rev. 2.

LTOP eetpoints were developed to provide bounding heatup and conidown curve protection for the worst case mass and heat addition low temperature overpressure transients.

The proposed changes to Technical Specifications are required to support implementation of the revised heatup and ooldown pressure temperature limitations on the Reactor Coolant System.and the revised eetpoints for the low temperature - overpressure protection (LTOP) system.

Operation of North Anna Power Station Units 1 & 2 in accordance with this change will not involve a significant hazards consideration as defined in 10 CFR 50.92 because it will not:

1.

result in a significant increase in the probability or consequence of an accident previously evaluated.

The application of the revised pressure-temperature limitations ana the revised LTOP setpoints, results in greater restrictions on.the operation of the units and will insure that the requirements of 10 CFR 50, Appendix G, for fracture toughness of the Reactor Coolant System pressure boundary will continue to be satisfied.

VIRGINtA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 3388339 stri:1 No.

91 707 section 4 Page 3 of 3 2.

create the possibility of a new or different kind of accident from any accident previously identified.

There will be greater restrictions on the operation of North Anna Power Station Units 1 & 2 using the revised pressure-temperature limitations and the revised LTOP setpoints.

These restrictions will insure that the requirements of 10 CFR 50, Appendix G, for fracture toughness of the Reactor Coolant System pressure boundary will continue to be satisfied.

The proposed amendments will not result in other changes in the way the units are operated.

3.

result in a significant reduction in a margin of safety.

.The revised pressure temperature limitations and the revised LTOP setpoints will insure that the requirements of 10 CFR 50, Appendix G, for fracture toughness of the Reactor Coolant System pressure boundary will continue to be satisfied.

The safety factors defined in the ASME Code and the requirements of 10 CFR 50 Appendix G provide the basis for the applicable safety margins.

The plant specific information obtained from the testing of the sample vessel-material and utilized to develop the revised pressure temperature I!mitations and the revised LTOP setpolms will confirm that these safety margins are not reduced.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that this change does not involve a significant hazards consideration.

1 l

VIRGINIA ELECTRIC AND POWER COMPANY

1 ATTACHMENT 5 ADDITIONAL SUPPORTING INFORMATION (NA&F TECHNICAL ANALYSIS)

(B AW 2146)

(WC AP 12503)

(REFERENCE LIST)

TECHNICAL SPECIFICATION CHANGE REQUEST NORTH ANNA POWER STAllON UNITS 1 & 2 VIRGINIA ELECTRIC AND POWER COMPANY

s Dock:t Hosa 50 330&339 S:ri:1 No.

91 707 De" Con 5 Page 2 of DO NA&F TECHNICAL ANALYSIS l

VIRGINIA ELECTRIC AND POWER COMPANY

i Dock:t Nos.: 50.J384339 i

Sulal No.

91 707 l

section 6 Page 3 of 80 l

IECHNICAL ANALYSIS l

Survalliance Cansula Analvsla Results The North Anna Units 1(4) and 2(5) reactor vessel materials surveillance program Capsule U analysis reports were transmitted to the NRC in June 1988(F and March 1990(6), respectively.

These reports provide the basis for the iluence projections and RTNDT values used in the generation l

of revised heatup and cooldown curves.

l North Anna Unit 1 Capsule Analysis Results l

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 cumulative fast neutron, 'd > 1.0 Mev, fluence of 8.28 x 1018 n/cm2,.

The calculated fast neutron fluence based on actual cycle power-distributions at the capsule location was 8.85 x 1018 n/cm2-which comparet favorably (within 7%) with the dosimeter fluence.

The peak fluence at the inside 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 (4).

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 ft Ib and 50 ft lb points, and the decrease in the upper shelf enargy.

The tensile specimens were used to determine ultimate tonslie strength and yield strength.

The vessel specimens within Capsule U were obtained from _ the same girth weld and-forging materials es those used in the reactor vessel beltline (4).

The irradiated specimens test results were compared to unirradiated specimen test results.

The Charpy V-notch impact test results show the irradiation has increased the aveiage 50 ft lb transillon temperature by-80 to 110'F cepending on the specimen. metal.

Irradiation has' increased the ' average 30 ft Ib transition temperature by 65 to =100'F, The upper l

shelf energy.(average. energy absorption at full shear) results _ show the worst decrease to be 25 ft lb _when comparing Irradiated samples to unirradiated samples.

4 VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 3384339 s:f t:1 No.

91 707 Section 5 Page 4 of 80 The lowest average upper shelf energy was determined to be 92 ft lb which is greator than the 10 CFR 50 Appondix G low limit of 50 ft lb.01).

The Charpy impact test results from Capsule U were clso satisfactorily comnared to the Capsulo V results.

Tension test results show a slight 15 uten o in the ultimato tensile strength and the yloid strength due to irradiation.

Reference 4 should be consultod for specific test results, in accordance with the methods proscribed by Regulatory Guide 1.99, Revision 2, the lower shell forging adjusted RTNDT at the end of 12 EFPY for Unit 1 was calculated utilizing the data from the surveillance capsule analysis.

Although the use of surveillanco data for the calculation of RTNDT results in a lower value of RTNDT than is obtained when surveillanco data is not used, the limiting 12 EFPY values of RT DT at the 1/4T and N

3/4T locations were shown to occur in the Unit 1 lower shell forging.

The limiting material in the existing Unit 1, 10 EFPY curves was determined to be the lower shell circumferential well(3).

North Anna Unit 2 Capsule Analysis Results Capsule U was removed from North Anna Unit 2 at the end of the sixth cycle of operation.

The capsu!o dosimotors were evaluated and found to have a cumulative fast neutron, E > 1.0 Mov, fluence of 9.55 x 1018 n/cm2, The calculated fast neutron fluence at the capsulo location was 1.06 x 10 19 n/cm2, which compares favorably (within 11%) with the dosimeter fluence.

The peak calculated fluence at the insido surface of the reactor vessel was calculated to be 8.02 x 101 B n/cm2 which shows that the capsulo has been exposed to slightly more neutrons than the vessel (s),

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 transillon temperature increases at 30 ft Ib and 50 ft 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 Capsulo U were obtained from the same girth weld and forging materials at those used in the reactor vessel beltlino(5).

I VIRGINlA ELECTRIC AND POWER COMPANY

i Dock:t Nos.: 50 338&339 s:rlal No.

91 707 section 5 Page 5 of 80 The irradiated specimens test results were compared to unirradiated specimen test results.

The Charpy V notch impact test results show the irradiation has increased the average 50 ft lb transition temperature betsveen 25'F to 55'F dependi'ig on the specimen metal.

Irradiation has increased the avera0e 30 ft lb transition temperature between 25 and 60'F depending on the specimen metal.

The upper shelf energy (average energy absorption at full shear) results showed no decrease in the amrage upper shelf energy of the forging and wold metals.

Both survemance materials exhibit a more than adequate upper shelf level for continued safe plant operation (5).

T3nsion test results show a slight increase in the ultimate tensile ctrength and the yield strength due to Irradiation.

A comparison of the 30 ft lb transillon temperature increases for the Unit 2 surveillance material with predicted increases using the methods of NRC Regulatory Guide 1.99, Revision 2, demonstrated that forging and weld metal transition temperature increases were less than predicted.

Reference 5 should be consulted for specific test results.

ligatuo and Cooldown Curves Heatup and Cooldown Curve Generation 10 CFR 50 Appendix G ostablishes fracture toughness requirements for the reactor vessel.

Virginia Power utilizes two types of graphs to identify plant specific limits.

The graphs are known as heatup and cooldown curves.

The proposed heatup curve depicts two curves: the leak test limit and the heatup limit (up to 60*F/hr).

The criticality limit curve has been omitted as it is redundant with respect to the existing T.S. 3.1.1.5 criticality limitation.

The cooldown curve depicts a series of curves for a range of assumed cooldown rates (0, 20, 40, 60, and 100*F/hr).

The control roorn operators use these Technical Specification graphs to ensure pressure and temperature are within acceptable values.

The heatup and cooldown curve analysis developed 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 durin0 heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure, Therefore, steedy state conditions can be lirniting for the inside wall so both heatup and steady state must be considered, VIRGINA ELECTRIC AND POWER COMPANY J

_ _ _ ~. __

Dock:t Nos.: s0 338&339 s:rl:1 No.

91 707 sect on 5 Page 6 of 80 i

The heatup curve calculations must also consider the case of a 1/4T flaw at the outsido surface.

The thermal and pressure stressos 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 composito curvo is required to make sure that the limiting condition is always protected against.

For o>amplo, protection must be provided if the limiting location shifts from the inside to the outside surface.

Thorofore, a composito heatup curve is generated by comparing on a point by-point basis the steady stato cervo at the inside of the wall along with the farious heatup rato curves at the outsido surface.

Thus at any given temperature, the allowable pressure is taken to be the most limiting of the values from each of the curves under consideration.

During cooldown, the controlling location of the flaw is always at the insida of the wall because the thermal gradlonts produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure temperature relations are generated for both steady stato and finito cooldown rato situations.

A lower bound composito curve from the steady stato and cooldown conditions is constructed for each cooldown rate of intorest.

The use of a composite curve in the ccoldown analysis is necessary because control of the cooldown proceduro is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown the tip is at a higher temperature than the fluid adjacent to the vossel inner wall.

This condition is not true for the steady stato situation.

It follows that at any given reactor coolant temperature, the temperature gradient developed during cooldown results in a higher value of KIR at the 1/4T location for finito cooldown rates than for steady state operation.

So, if the temperature changes such that KIR increases faster tht,n KIT, the steady-state can be limiting.

Revised heatup and cooldown curves valid to 12 EFPY and 17 EFPY have been generated for North Anna Units 1(9) and 200), respectively.

The reports documenting the development of those curves are presented in Appendicos A and B.

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 60 338&339 s: rial No.

91 707 section 5 Page 7 of 80 Heatup and Cooldown Curve interpretation Babcock and Wilcox (B&W) and Westinghouse have provided Virginia Power with revised heatup and cooldown curves for North Anna 1(9) and 2(10),

respectively.

Figures 1 through 4 present these revised Technical Specification heatup curves and cooldown curves.

Data for these figures is presented in the attached References 9 and 10.

Although the proposed Technical Specifications heatup curves present only the curve generated for a 60'F/hr heatup rate, heatup curves were generated for heatup rates of 20', 40', and 60'F/hr.

The cooldown curves were generated for 20',

40', 60', and 100'F/hr cooldown rates.

The steady state condition was also considered.

As indicated in the figures, the curves are based on extrapolated fluences to perrait operation to the specified cumulative core average burnup.

The proposed revised heatup curve does not contain the 10 CFR 50, Appendix G criticality limit.

The criticality limit is not required since limiting condition for operation (LCO) 3.1.1.5 restricts the lowest operating loop average temperature to 2 541*F for Modes 1 and 2. This LCO defines a minimum temperature for criticality that provides substantially more margin to the heatup curve than the criticality limits required by 10 CFR 50, Appendix G.

A previous discuss!on of the composito cooldown curve indicated that the reactor vessel wall has a temperature gradient dependent on cooldown rate.

Because of this temperature gradient, and because higher cooldown rates are not realistically possible at low RCS temperatures, lower achievable cooldown rates at lower temperatures were assumed.

Table 1 presents the assumed maximum cooldown rate for various temperatures.

These lower achievable cooldown rates permit the elimination of the most restrictive (Iow temperature) portions of the 20, 406, 60', and 100 F/hr cooldown curves for the purpose of establishing LTOP setpoints.

Tchie 2

presents assumed maximum heatup rates for various

' anperatures.

As in the case of the cooldown curves, these lower ssumed heatup rates permit the elimination of the most restrictive p# tions of the 20' and 40'F/hr heatup curves for _ the purpose of establishing LTOP setpoints.

The temperature-dependent heatup and cooldown rates will be incorporated into the Unit 1 and 2 plant operating procedures to ensure that they are not exceeded during normal operation.

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 338&339 S:ti:1 No.

91 707 section 5 Page 8 of 80 l

After modifying the cooldown curve and heatup curve to reflect lower assumed cooldown and heatup rates at lower temperatures, the most limiting points from each curve were selected at each temperature to construct a composite curve for use in the development of revised LTOP PORV setpoints.

Tables 3 and 4 present the composite, modified heatup and cooldown curve for Units 1 and 2, respectively.

Since the design basis transients are defined with operational assumptione related to the pressurizer safoty valve setroints, certain operational restrictions must be enforced to ensure the low temperature accident analysis assumptions are valld.-

The temperature on the composite, modified heatup and cooldown curve _ corresponding to the pressurizer safety valve.setpoint of 2485 psig is 316'F'(358'F for Unit 2),

This point is used to bound all of the. low temperature accident. analyses.

The mass additlori transient assumes only one chargint, pump will be operable below 31G'F (358'F for Unit 2).

The heatup transient assumes whenever a RCP is started below 316*F (358'F for Unit 2) the temperature difference between the primary and secondary fluids in the Steam Generator is less than 50'F.

Because the Unit 2 temperature which corresponds to 2485 psig (358'F) Is i

greater than 350*F, it was necessary -to apply the single charging pump operability requirement to MODE 3.

The implications of this change on postulated events at RCS -temperatures equal-to or less than -358'F must be considered.

Because the-stored energy of the core would be low at conditions where RCS teraperatures are less than or equal to 358'F, the MODE-3 single charging pump operabi_llty requirement-presents no safety

concerns, it Is reasoned that although one of the charging pumps would not be available for automatic initiation, it would still be available for manual initiation.

Therefore, it has been concluded that extension of this requirement to 358 F has no impact of the ability of the operator to mitigate the consequences of postulated events, V'RGINIA ELECTRIC AND POWER COMPANY

l Dock:t Nos.: 50 3388339 s:ri:1 No.

91 707 section 5 Page 9 of 80 P_O.R V SelgnJals

Background

Cold overpressure protection is provided to ensure that the normal operation 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 to prevent violation of the composite, modified heatup and cooldown curve should a RCS pressure transient occur.

The limits have been set by two design basis accidents:

the inadvertent start of a charging pump and the startup of a reactor coolant pump in an RCS loop with a 50 F difference between the steam generator secondary fluid temperature and the RCS temperature.

These translents represent the limiting mass addition and heat input transients and are analyzed with the RCS water solid.

Only one PORV is required to operate during the transients.

Generic transient analyses have been used previously to determine LTOP setpoints 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 generic assumptions have excessive conservatism to allow a w!de range of application.

These generic conservative assumptions becomo operationally burdensome when PORV lift setpoints are too low to allow normal RCS operation without opening a PORV. A plant specific North Anna two loop RETRAN02/ MOD 03(15) model was developed to analyze possible setpoints.

(Information supporting the use of the RETRAN Code version RETRAN02/ MOD 03 is presented in to Reference 2.)

The analysis revealed that the mass addition transient produces the most limiting results.

The following sections describe the North Anna model development and the analysis to determine new PORV setpoints.

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 3308339 sort:1 No.

91 707 secti:n 5 Page 10 of 80 Mass Addition Transient The inadvertent startup of a single charging pump was selected as the design basis mass addition transient based on previous UFSAR work (Reference 12, Scction 5.2.2.2).

The LTOP setpoints were determined such that a pressure overshoot allowance exists to prevent the composite, modified heatup and cooldown curve from being exceeded assuming an inadvertent charging pump startup during water solid operation.

This overshoot allowance is required because of the valve opening characteristic associated with the air operated rollef valves used on the pressurizer at North Anna 03Mid).

loadvertent operation of a single charging pump was modeled assuming initial conditions as listed in Table 5.

The initial RCS temperature, pressuro, and PORV setpoint were varied to observe the effects of changes in these parameters. A range of RCS temperatures between 100 and 325'F were examined, as well as a range of initial pressures, The results revealed a gradually decreasing PORV setpoint pressure overshoot with increasing initial RCS temperature and PORV setpoint.

The peak RCS pressure was found to be relatively insensitive to the initial RCS pressure.

Heat Addition Transient The heat addition transient assumes the Technical Specification limit of a 50'F temperature 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 addition transient for LTOP setpoint determination relative 'o the composite, modified heatup and cooldown curve (Reference 12, Section 5.2.2.2).

The heat addition transient was modeled assuming the initial conditions listed in Table 6.

The secondary to primary heat transfer modelling included a very conservative evaluation of the local secondary side convection heat transfer coefficient and an assumed constant bulk i

secondary side temperature (no credit taken for decreasing temperature due to secondary to primary heat. transfer).

The pump startup flow characteristic was also modelled in a conservative fashion.

The analysis revealed that the results of the heat addition transient are easily bounded by those of the mass addition transient.

VIRGINtA ELECTRIC AND POWER COMPANY

Dock:1 Nos.: 50 338&339 serlil No.

91 707 section 5 Page 11 of 80 Revised LTOP Setpoints Using the PORV setpoint overshoot results from the analysis described above performed with the North Anna plant specific RETRAN(15) model, revised LTOP setpoints were determined.

The revised setpoints were established using the composite, modified heatup and cooldown curves in Tables 3 and 4 which take credit for the assumed cooldown and heatup rates presented in Tables 1 and 2, and exclude measaroment uncertainty.

I It is reasoned that measurement uncertaintles may be excluded from consideration b the development of LTOP setpoints on the basis that these uncertainties are insignificant when compared to the margin terms included in the ASME Section lil Appendix G methods.

Specifically, the pressure stress !s multiplied by a factor of two, resulting in conservative stress intensity values.

(As an example, a pressure / temperature limit which shows an allowable internal pressure of 400 psi is actually based upon a stress associated with an internal pressure of 800 psi.)

In addition, the use of a lower bound allowable stress Intensity, KiR, which is shifted in accordance with Regulatory Guide 1.99 Revision 2 methods (i.e., 2c margin on mean predicted shift) ensures a conservative measure of allowable stress intensity as a function of temperature in the heatup and cooldown curve calculations, instrumentation uncertainties have been excluded from consideration in other utility submittals on the basis of their insignificance relative to the connrvatisms of stress intensity factors (19),(20),

Temperature measurement uncertainty was considered in the development of the minimum LTOP enabling temperatures.

Automatic low temperature overpressurization protection is required whenever any RCS cold leg temperature is less than 270 F.

(321 F for Unit 2).

These temperatures are the RTNDT+AT+90 F+1nstrument uncertainty.

For Unit 1,

the 12 EFPY RTNDT temperature is 145'F for 1/4T and 122*F for 3/4T(9). For Unit 2, the 17 EFPY RT DT temperature is 196 F for 1/4T and 172'F for N

3/4T(10).

The AT is the maximum temperature difference between the water and metal (i.e.,15 F at the 1/4T; 32'F at the 3/4T).

The instrument uncertainty added was 20 F.

The 90'F addition is considered to be a reasonable range to require the automatic low temperature overpressurization protection.

This is sufficient for automatic protection during startup and shutdown.

Above 270*F (321'F for Unit 2),

administrative control is adequate protection because of Appendix G fracture criteria.

The analysis has an increased margin at higher l

l VIRGINIA ELECTRIC AND POWER COMPANY

Dock *t Nos.: 50 3388339 Smlil No.

91 707 Sect!on 5 Page 12 of 80 temperatures.

In addition, operation of the RCS above 270'F (321*F for Unit 2) decreases the effects of the two design basis transients.

The concept of requiring automatic LTOP 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(16).

The Generic Letter states that due to the impact of Implementation of Revision 2 of Regulatory Guide 1.99, Standard Review Plan Section 5.2.2 and Branch Technical Position RSB 5 2 will be revised to define the temperature where automatic protection is required to be enabled.

The PORV setpoints were optimized to values which limit the reactor vessel peak pressure during postulated overpressure transients to vaiues less than those represented by the composite, modified heatup and cooldown curve, and which are not operationally restrictive because of too much safety margin, The proposed North Annn t' nit i setpoints are s 450 psig when the RCS temperature is s 270'F, aria s 390 psig when the RCS temperature is s 150 F.

The proposed North Anna Unit 2 setpoints are s 510 psig when the RCS temperature is s 321 F, and s 360 psig when the RCS temperature is s 210 F.

Table 7 presents the current and proposed LTOP PORV setpoints.

The setpoints are not dramatically changed, reflecting the not impact of excluding instrumentation uncertainties in the heatup and cooldown curves and the extrapolation of RTNDT to a higher applicable burnup.

RTPTs Evaluation The results of the Capsule U analyses for both North Anna Unit 1(4) and Unit 2(s) reveal an excellent comparison between the experimentally determined fast neutron fluence in the surveillance capsule and the calculated fluence at the capsule center.

These results confirm the analytical model used to predict vessel fluence to the end of the current operating license.

Reference 17 presented estimated fluences for North Anna Unit 1 and Unit 2 as a function of burnup.

These values were utilized to estimate the end-of-license RTpTs for the controlling materials in the North Anna 1 and 2 reactor vesse' beltlines.

Utilizing the method of Reg Guide 1.99, Revision 2 as specified in 10 CFR 50.61(18), the following results-were obtained:

North Anna Unit 1 - Limiting Forging (Lower Shell) - (Estimated EOL Fluence - 6.79x1019 n/cm2, inner Surface)

EOL RTpts - 240 F (Screening Criterion - 270 F)

VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nes.: f,0 30ftt,339 S:rlal No.

91407 Section s Pye 13 cf 80 North Anna Unit 1 Limiting Wold (Circumferential Weld).

(Estimated EO!. Fluence - 6.79x1019 n/cm2 - Inner Surfaco)

EOL RTPTS - 148'F (Screening Criterion - 300'F)

North Anna Unit 2 - Limiting Forging (Lower Sheli) ' (E!siirnated EOL Fluence - 6.96x1019 n/cm2. Inner Surface)

EOL RTpis - 230*F (Screening Criterion - 270*P North Anna Unit 2 - Limiting Weld (Circumferential Wold) -

(Estimated EOL Fluence - 6.96x1019 n/cm2. Inner Surface)

EOL RTPTs - 60'F (Screoning Criterion - 3005F)

The extrapolated fluence at tho above locations at 12 EFPY (Unit 1) is estimated to be 1.64 x 10 9 nk;m2 H).

This is well within the fluences 1

demonstrated above to not resuit in a PTS concern for Unit 1.

The extrapolated fluence at the above locations at 17 EFPY (Unit 2) is es'! mated to be 2.35 x 10 9 n/cm2 (s).

Again, this is well within the 1

fluences demonstrated to not result in a PTS cor.cern for Unit 2.

Because of the excellent comparison betweeq experimentally determined and calculated fluences, and because there are no expected changes in operating conditions that would significantly impact vessel fluence estimates, the EOL RIPTS values presented above (based on previous fluence estimates) are still considered valid.

Measured ARTNOT values from the capsule analyses and predicted values as calculated by Regulatory Guide 1.99 Revision 2 compare well, with the R.G.

1.99 Rev. 2 values generally being more limiting.

This reinforces the conclusion that there is no PTS concern for North Anna Units 1 and 2 for burnups up to the respective applicable burnups for each unit.

Conclusions The heatup and cooldown curves required by Appendix G of 10 CFR 50 have been extrapolated to 12 EFPY and 17 EFPY for North Anna Units 1 and 2, respectively, by including the effects of the incremental radiation exposure on the reactor vessel beltline region.

The results are referencad to the analyses of the North Anna Units 1 and 2 Capsule U results.

The reviseo Appendix G curves were prepared using standard B&W and Westinghouse methodologies methodology including Regulatory Guide 1.99 Rev. 2.

PORV setpoints were developed to provide bounding heatup and cooldown curve protection for the worst case mass and heat addition low temperature overpressure transients.

VIRGINtA ELECTRIC AND POWER COMPANY

= _ _ _

Dock:1 N:s.: 50 3388339 seri:1 N3.

91 707 secti:n 5 P:ge 14 of 80 The next Unit 1 reactor vessel surveillance capsulo (Capsule X) is scheduled to be removed after the tonth fuel cycle (10 EFPY) which allows sufficient time for analysis prior to exceeding 12 EFPY.

The next Unit 2 reactor vessel surveillance capsule (Capsule W) is scheduled to be removed after the thirteenth fuel cycle (15 EFPY) which allows sufficient time for analysis prior to exceeding 17 EFPY, The heatup and cooldown curves prepared by B&W and Westinghouse were determined in a conventional manner according to Section lil of the ASME code as required by 10 CFR 50 Appendix G.

Both steady state and transient thermal conditions were considered in order to bound the possible combinations of pressure (l.o. membrane) and thermal stresses.

The new North Anna Unit 1 low temperature overpressure protection system PORV lift settings should be less than or equal to 450 psig whenever any RCS cold Icg temperature is less than or equal to 270'F, and less than or equal to 300 psig whenever any RCS cold leg temperature is less than 150'F.

The new North Anna Unit 2 low temperature overpressure protection system PORV lift settings should be less than or equal to 510 psig whenever any RCS cold leg temperature is less than or equal to 321*F, and less than or equal to 360 psig whenever any RCS cold leg temperature is less than 210'F.

PTS ovaluations were made for the limiting belt!ine locations.

It was demonstrated that (a) predicted end of license fluences do not result in RTPTS values in excess of the screening criteria when calculated using the methodology of Regulatory Guide 1.99, Revision 2; (b) there is an excellent comparison between experimentally determined and calculated vessel fluences; and (c) the extrapolated fluences at the burnup limit to which the revised heatup and cooldown curves are applicable for each unit are significantly less than the extrapolated end of license fluences (which have been demonstrated to not result in a violation of PTS screening criteria).

On this basis it may be concluded that there is neither a significant change in predicted RTPTS values; nor is there a PTS concern for either unit up to the burnup limit to which the revised heatup and cooldown curves are valid.

.i j

l i

VIRGINlA ELECTRIC AND POWER COMPANY

=

Dock:t Nos.: 50 338&330 s: rial No.

91 707 section 5 P:ge is of 80 List of Tables 1.

Cooldown Rates Assumed for Various Temperature Ranges 2.

Heatup Rates Assumed for Various Temperature Ranges 3.

Unit 1 Composite, Modified Heatup and Cooldown Curve Data for PORV Sotpoint Development 4.

Unit 2 Con.pos.ite, Modified Heatup and Cooldown Curve Data for PORV Setpoint Development 5.

Initial Conditions for the Mass Additions Transient 6.

Initial Conditions for the Heat Additions Transient 7.

Current and Proposed LTOPS PORV Setpoints l

)

VIRGINtA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 50 330&339 Seri:1 N2.

91 707 Section 5 Page 16 of 80 Table 1:

COOLDOWN RATES ASSUMED FOR VARIOUS TEMPERATURE RANGES Temperature Rango 'F Cooldown Rate 'F/hr.

T>200'F 100'F/hr.

180'F<Ts200'F 60'F/hr.

150'F<Ts180'F 40'F/hr.

120*F<Ts150*F 20'F/hr.

Ts120'F O'F/hr.

Table 2:

HEATUP RATES ASSUMED FOR VARIOUS TEMPERATURE RAf GES Temperature Range 'F Heatup Rate 'F/hr.

Tc150'F 20'F/hr.

170'F>T> 150*F 4 0'F/h r.

T2170'F 00'F/hr.

i VIRGINIA ELECTRIC AND POWER COMPANY

Doclot Nos,: 50 3364339 Serial No.

91 707 Secti3n 5 Pa9e 17 ef 80 Table 3 COMPOSITE, MODIFIED HEATUP AND COOLDOWN CURVE NORTH ANNA UNIT 1 (Data for LTOPS Setpoint Development)

Indicated Tcmperature Composite Pressure Limit

('F)

(Oslo).

75 550 80 554 85 559 90 565 95 572 100 579 105 587 110 595 115 604 120 614 125 597 130 610 135 624 140 639 145 655 150 673 155 659 160 681 165 705 170 715 175 737 180 760 185 782 190 805 195 829 200 855 t

l 205 883 210 914 215 946 220 981 225 1019 230 1059 I

4 VIRGIN!A ELECTRIC AND POWER COMPN #

Dock *.1 Nos.: 50 338&&39 S:iial No.

91 707 l

Secti:n 5 Pa9e 18 cf t,0 Table 3. (continued)

Indicated femperature Composite Pressure Limit

(* F)

(pslo).

235 1103 240 1149 245 1199 250 1253 255 1311 260 1373 265 1440 270 1512 275 1590 280 1673 2 8 T.

1762 i

290 1858 295 1962 300 2073 305 2192 310 2320 315 2458 320 2605 VIRGINIA ELECTRIC AND POWER COMP /NY

Docket Nos.: 50 338&339 Seri:1 No.

91 707 Section 6 Page 19 of 80 Table 4 COMPOSITE, MODIFIED HEATUP AND COOLDOWN CURVE NORTH AtNA UNIT 2 (Dats for LTOPS Setpoint Development)

Indicated Temperaturo Composite Pressure Limit

('F)

(pslo).

85 513 90 514 95 518 100 522 105 527 110 532 115 538 120 545 125 525 130 53; 135 537 140 545 145 552 150 561 155 534 160 544 165 555-170 539 175 550 180 561 185 574 190 586 195 601 200 616 205 596 210 622 215 649 220 679 225 711 230 737 VIRGINIA ELECTRIC AND POWER COMPANY

~

-. - ~ -. - - ~ _

F Dock:t Nos.: 50 3384339' Seri:.1 Ns.

91 707 Secti:n 5 Page 20 of 80 l

t i

Table 4. (continued) t Indicated Temperature Composite Pressure Limit

('F)

(D$l0).

235 763 i

240 791 245 820 l

250 852 255 886 260 923 265 963 l

270 1005 275 1051 280 1099.

j 285 1152 290 1208 295

'1268 300 1333 305 1402 310 1477 315 1556

..t 320 1642 t

325 1733 330 1830 335 1935 340 2046 345 2186 350 2283 355 2402 i

i l

a l

r VIRGINIA ELECTRIC AND POWER COMPANY A

^

.m -

E

.-.m.-.

at.. -

1 Docket Nos.: 50 33P,$339 S: rial No.

91 707 l

Section 5 P ge 21 of 80 f

Table 5:

INITIAL CONDITIONS FOR THE MASS ADDITIONS TRANSIENT i

Reactor Coolant Temperature 100, 150, 200, i

('F) 250, 300, 325 Reactor Coolant Pressure 200, 250, 300 (psl0) 340, 380, 400 Maximum Charging Pump Flowrate 705 Opm (design head / flow curve)

Pressurizer Steam Volume 0 ft3 Pressurizer Water Volume '

-1400 ft3 Reactor Coolant System Flow 10%

PORV OPEN Sepoint Varlabte l

PORV CLOSED Setpoint OPEN.15 psi

.t i

4 5

1 I'

l l

VIRGlhlA ELECTRIC AND POWER COMPANY- '

l i

l

[

+a--

b-Aa--

,A s,

-.-4

.h-rem

-4as..

'6+L A -

x-

--a m

-Aa

-,A-naa

-X---

- i.---

da>

Dock:t Nos.: 50 338&339 Seri:1 Nr.

91 707 Secti:n 5 P:ge 22 cf 80 Table 6:

INITIAL CONDmONS FOR THE HEAT ADDITIONS TRANSIENT Reactor Coolant Temperature 100'F Reactor Coolant Pressure 280, 340(psig)

RCS/SG AT 50*F Pressurize. Steam Volume O ft3 Pressurizer' Water Volume 1400 ft3 RCP Spoods in Affected Loop startup 10%

100 %

in Unaffected loop, coastdown 10 %

0 %

PORV OPEN Sotpoint Variable PORV CLOSED Setpoint

_ OPEN 15 pol T'Ne 7:

CURRENT AND PROPOSED LTOPS PORV SETPOINTS t

JORTH ANNA UNIT 1 Current:

5450 psig for Cold Leg Ts261*F s390 psig for Cold Leg Ts150'F Proposed:

5450 psig for Cold Log Ts270*F 5390 psig for Cold Leg T4150'F tt3')gt, JNA UNIT 2 Current:

s520 psig for Cold Leg Ts340*F s375 psig for Cold Leg Ts190*F Proposed:

s510 psig for Cold Leg Ts321'F 5360 psig for Cold Leg Ts210'F

.+

- VIRGINIA ELECTRIC AND POWUR COMPANY l

Docket Nos.: 50 338&339 Seri-J No.

91 707 Secti:n 5 P:ge 23 of 80 B AW-2146 I

i 1

. VIRGINIA ELECTRIC AND POWER COMPANY

f)ocket Nos.: 50 3384339 Sort) No.

91 707 Sectsn 5 Pgj4gs0 October 1991 i

l 4

h0RTH ANNA UNIT 1 PRES $URE TEMPr.RATURE LIMITS FOR 12 EFPY AND NORTH ANNA UNIT 2 PRES $URE TEMPER /TURE LIMITS FOR 12 AND 15 EFPY 4

VIRGINIA ELECTRIC AND POWER COMPANY by A. D. Nana M. J. Devan l

B&W Document No. 77 2146 00 (See Section 3 for Document Signatures) l B&W NUCLEAR SERVICE COMPM f t

Engineering and Plant Services Division P. O. Box 10935 Lynchburg, Virginia 24506 0935 VIRGINIA ELECTRIC AND POWER COMPANY

Dock:t Nos.: 5o 3388339 S: rial No.

91 707 Section 5 Pa9e 25 of Bo 1.

INTRODUCTION This report presents the North Anna l'ait I and Unit 2 reactor vessel beltline region pressure temperature operating limits applicable,to 12 EFPY.

Also included in this report are the North Anna Unit 2 reactor vessel beltline region pressure temperature operating limits for 15 EFPY. The data used to develop the 12 EFPY limitations for North Anna Unit 1 are based on the analysis of North Anna Unit 1 Reactor Vessel Surveillance Capsule U as reported in WCAP ll777,' and the data used to develop the 12 and 15 EFPY limitations for North Anna Unit 2 are based on the analysis of North Anna Unit 2 Reactor Vessel Surveillance Capsule U reported in WCAP 12497.8 The-report contains data which support the development of the pressure temperature limits for ncnsal operation, both heatup and cooldown, insarvice leak and hydrostatic tests and reactor core operation applicable to 12 EcPY for North Anna Unit I and 2 and 15 EFPY for North Anna Unit 2.

These limits are adequate for current operations and are justified by the data obtained from the surveillance capsules as presented in WCAP ll777 for North Anna Unit I and WCAP 12497 for North Anna Unit 2.

t L

11 VIRGINIA ELECTRIC AND POWER COMP /NY

~

1 3

1 Dock:t Nos.: So 33Pa339 S:tlal No.

91 7o/

-l Section 5 Page 26 of 80 l

l I

t 2.

DETERMINATION OF REACIOR VESSEL BELTLINE I

j REGION PRES $URE-TEMPERATURE LIMITS t

I The pressure temperature limits of the reactor vessel baltline region of North -

Anna Unit 1 and Unit 2 are established in accordance with the requirements of 10CFR50 Appendix G.*

The objective of these limits is to prevent nonductile.

failure during any normal operating condition, including anticipated operational occurrences _and system hydrostatic-tests.- The loading conditions of interest

(

include the following 1.

Normal operations including heatup and cooldown.

2.

Inservice leak and hydrostatic _ tests.

3.

Reactor core operation.

The beltline regions of the North Anna Unit I and Unit 2-reactor vessels have been analyzed in accordance with 10CFR50, Appendix G. For the service period for L

which the limit curves are established, the maximum-allowable pressure as a function of fluid temperature is obtained through a point by point comparison of the limits isposed by the beltline region.. The maximum allowable pressure is taken to be the lowest of the allowable pressures at the one quarter T and three-quarter T vessel wall locations (T = wall. thickness measured from the inside l

surface)andduetosteadystateconditions.

North Anna Unit 1 The limit curves for North Anna Unit.1 are based on the predicted values of.the adjusted reference temperatures of all the beltline region' materials at the end :

of the twelfth EFPY, LThe twelfth EFPY was selected for this analysis to extend the applicability of the Unit-1 curves to provide sufficient time to perform the 10 EFPY surveillar.ce capsule analysis. and evaluation.

i l

21 l

VIRGINIA ELECTRIC AND POWER COMPANY :

i

,,,-,+-,,_,nJ.+.,.,.,-,

, n-r

.c.

e w.,,-,,.,,nn..,.-~,-+.,.n.,,

.,,-,-~.,

-n

,n n.-~.,---,,.-,~

t Dock:t Nos.: So.330a339 Seri:.1 Ns.

91 707 l

Section 5 Page 27 cf 80 l

Unirradiated Charpy impact properties were determined for surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H.'

For beltline region materials for which the measured properties are not available, unirradiat-ed Charpy impact properties and residual, element compositions, as originally determined, are listed in Table 2-1.

The~ adjusted reference temperatures are calculated by adding the predicted radiation induced shifts in RTuor and the initial RTuor. The radiatio'n induceo RT, values were predicted as a function of the material's copper and nickel content and exposure to neutron fluence in accordance with the guidelines' presented in Regulatory Guide 1.gf. Revision 2.*

The neutron fluence and adjusted RTuor values of the beltline region materials at the end of the twelfth full power year are listed in Table 2-1. The adjusted RTuor values are given for the one quarter T and three quarter T vessel wall locations.

Figures 2-1 through 2-3 show reactor vessel's pressure temperature limit cerves for normal heatup and inservice leak and hydrostatic tests. These figures also show the core criticality limits as required by 10CFR50, Appendix G.

Figure 2-4 shows the vessel's pressure temperature limit curves for normal cooldown.

The above pressure temperature limit curves are applicable through 12 EFPV.

Protection against nonductile failure for the North Anna Unit I reactor beltline region is ensured by maintaining the reactor vessel downconer pressure below the upper limits of the pressure-temperature limit curves. The. acceptable pressure and temperature combination for reactor vessel operation are below and to the right of the limit curves. The reactor is not permitted to go critical until the pressure temperature combinations are to the right of the criticality limit curve.

To establish the pressure temperature limits for protection against nonductile failure, the limits pre?.ented in Figures 21 through 2-4 must ~ be adjusted by the pressure-differential between the; point of system pressure measurement and the pressure on the reactor vessel' controlling the limit curves

(

as well as accounting for possible instrument errors.

E 22-t VIRGINIA ELECTRIC AND POWER CCMPANY -

s.

u

,_2_,__..

_ _ _. _,. _._.m a. -_-

e t Docket Nos.i 50 338&339 SerH N3.

91 707 Section 6 Page 28 of 80

)

In addition to the North Anna Unit I revised heatup and cooldown curves, the original Babcock and Wilcox Report ' BAW-2146 documented revised curves for North-Anna Unit 2 applicable to burnups of 12 and 15 EFPY.

these curves as they are not being submitted for review a y

this time.

J l

t VIRGINIA ELECTRIC AND POWER COMPANY -

a

T g~ {*

4u uUO T

[

4 7 6 S S A

6

)

/

5 I S F

( E zo eon 7

T W

RF 3

L I 9 !

1 E

l 1

d l

moa m ni 9 DO n

eF s'

tI

(

r ef jo 4

2 8 s e 4

3 5

]

d

/

8 3 el

/

2 4 4

Ad 7

I l I q

1 n

I 1 1 E

1

(

T 4

6 4 4 9 A

8 F a

/

2 3 3 4 M,

i 3-W 3 I gre 4-4 4 4 2 A

2 F o

l

/

3 3 3 S f

h t

ly

$ I T

t ro de K

cfeJ T

5 S 7 9 A

9 F d

r a d'F 4

2 G 8 T e

n

/

W 3 S F

o l.

3-E.

f n

f etp t et sY t

t 4

eP e 7

/

2 F t t A

2 8 l

vF dT1 T

4 S i e 4

V 1 9 9

rE eR l

u R

I C2 l

t e T.,

ih l

mc t

5 F 8 t 9

9 0 9

8 n T.

i I 3

i. u 1 3 9

I o IR -

1 9

r 1

eh y

r 6

y rT t r 0

a 9

M u

se 1

lt 2 o l 0 4

2 9 3

t t e ec 2 t l 3 S

5 8 9

s 2

al ea 1

rb hf l 1 u

n t

ea g

e u

l pc A

a l

mi f

n l

t et F e

el eo t t s

I l

l l l l

l le Wj e e e e l

I Tp l o t t t t A

e p

e e m

1 at 6 2 f 6 t f e

4c/ o /n 7 1

l. F.

W rf t

9 t

eA e

/

i.

l.

1 9

s r

c 3l 2 4 4 l n

i u-e 8 4 4

v 1

u l n F t f F R

o e

s -

l l e s

f ej 9 f t l I l t

d el W t "a 6

e + e l

1 l t

t t t 8

s a c 4 4 4 4

/

4 s

A t E t

re e

1 E

A 4 c Ps

/o4 S. 9 8 9 M

9 M.

P.

y 9

1 T

l 5 1 3 6 s

r f e f

3 I i

o m

oY

- e-t l

S o

4 2 0 e 9

l 9 d

e n

l h-7 2 0

l.

1 I.

8 1

a ll e nr 1

c a e 1{,

n p

i 8 9 f J 8

S 8 8

lo oo i c lf rtl n it e s s /e f

t c 6 r t ee 8

8 e ohp r

8 9 p ca ntu e

6 7 6

9 1

ee o

p 4 !*

6 0 5 0

d tR C

p 1 1 1

0 1

0 y

s t t

e o

o 0 6 4 **

/

0 9 0

M gl t-a e a r1 P

e u A c t

d 1

d d

ri g

l l

9

. a l

n g n o

e w s I

l e e e

1 on e

n i g W W

er er I.

3 n s I

c fU s

t i g n T.

t jr 9 o e

e p 9 t

e i

ee o g r r

s'.R r np np A

1 s

aa nc wf e

o i

r i l

i e T

oC C

dt dt N y e e h e s

i C

a e

ed C

l e

t n lt l

3 n t

f l l r e a ea r e R r e n

l el t DA ee e h l n

a e be be2 l

e sd sd e

c t

e a s t

u a a

r ', T.

tl h S e I l

e p

g S

l j

b 1 R 7 d e

h

)

t l2 t

s e 3 t e t e.

F a

. S o5 o) of sv f n F 9 d n

s s t4 tS t ipe sv o

1 e 6 1

i r r r e e 9

.S hor iat l

e heI t 7 h l

l T f

r t m r0 r0 t

. s c.

s l

s 2

o n e o6 e1 e

,e9 c

i F e e.

t p I t N( h( l s 1 sj f

,e9 ol d s o

.c 9 T. c 9, p

l. r a s e

1 n m.

n m

e RaI ReI o p pd n e l

e s

.._ 2 2 2 dl e dl e t s a c l b

l t

l cA nk u c

s c eid eld t

a 1 l t r

r r t el t el e

ed h v ee l

i T

tp C.

t. C. A A

A svo aee

- c rl et l

e i a t y 8 s0 rG rC fT e e0 e.

, lee.lue.

e p r ul a l

e e c g

n e

l S l $ h h

h l gl g

h e ps cs e

t A A 1 e o

e at e afe c t

,d el n

S '

S S

S CeR Cof a T. i n s l e

d l

s l e o I

s d R n v r 5 M9 3

9 3

l o

l t

4 e el i r n A

8 f

c e t e e ao g g gS 5

4 g 4

e e t l

iN n n n S 9

9 n 0

r nt e t C r

i i i c

et g g g d d

d g d

H 7 I R R

(

i t

o i d a )

t e r r r l e1 n a e ee e o o e e

e o e

l l

r MM F f F e l

W

• J f

W

• m *' * "'

f

$EgyS fbQ" @pO gO$TD O li D

1 4

I

~

a Dock:t Nos.: 50 338433g-S:rld N1 91 707 Section 5 Page 30 of 80 ll In addition to the North Anna Unit 1-revised heatup and cooldown curves, the original Babcock and Wilcox _ Report BAW-2146 documented revised curves for North Anna Unit 2 applicable to burnups of-12 and 15 EFPY.

these curv; 5 Virginia Power has removed the pages from BAW-2146 which document this tiA?

as they are not being submitted for review and approval at-9 9

s!

VIRGIN!A ELECTRIC AND POWER ",OMPANY 4

I Figure 2-1.

North Anna Unit 1 Reactor Vessel Pressure-Temperature Limit Curves for Normal Doeration - Heatuo. Aeolicable f6r First 12 EFPY Up to 26*F/hr

[

d l

2500 i

i Leek Test Limil 4.

. O_

e I

E h

o' 2000 f

The acceptable pressere-temperature combleetless are below and

-[

m to the right of she llelt cerves. the llett cerves do not.

m

- faclude tne pressere differeettel between the pelet of systee

}

og -

b e

pressere seeseresset and the pressere en the reacter vossa!

4 Q

- 0 too controllleg the liett cerve, ser de they loclase say w

1500

[

3?

E

[

o e a

"U c) h O

i O

g 1000 3

o 98 i

2ll:

- e 4

}-

• --CrHicality Limit (273 F) 500 ee a

- u

.O.

Assumed Adjusted RTer, I_..E.,- t e

o 5-L l

8eltilneRegion1/4T.14#

e c

p Beltilne Region 3/41 122 3

l 1

1 1

I I

I I

I I

?*

C 0

1 0

100 200 300 400 4

o Reactor Vessel Coolant Temperature,. F j

g 4

i I

t 4

I l

l Figure 2-2.

North Anna Unit 1 Reactor Vessel Pressure-Temperature Limit Curves for t

Normal Doeration - Heatuo. Applicable for First 12 EFPY Up to 40*F/hr i

1 i

i 2500 j

Leak Test Limit n.

x e'

2000 t

F The acceptable pressore-temperatore cediestless are below and -

I m

g to the rlyht of the llett cerves. The llett cerves de set I

5

.m

_ lectede tes pressore itfforestIel betuose the pelet of systee o

.g pressere messerement and the pressere se the reactor vessel l

reelee contreilleg the llelt cerve, nor de they lectede say

-4 m-E/

Q.

edditteesi eergte of safety for possible lastrusset error..

0 1500 C

y E

o S

U 1000 o

8

==

m ACriticeHty Limit (273 F) 4 go o

00 AssumedAdjustedRT.,,Fy [h gs e

e 8eltline Region 1/4T I49a >F e

Beltilne Region 3/4T 122' E

I I

I I

I I

I I

I O

=

0' 100 200 300 400 h

i E

W s

Reactor Vessel Coolant Temperature, F 1

W*

s 3

i Figure 2-3.

North Anna Unit 1 Reactor Vessel Pressure-Temperature Limit Curves for l

Normal Operation - Heatup. Applicable for First 12 EFPY Up to 60*F/hr 2500 l

Leak Test Limit cn 3

i t

o.

559 e'

2000 f

The acceptable pressere-teeperatore combleetless are below and to the rfght of the llelt cerves. The llelt cerves de met m

. m

_ laclede the pres ~ere differeettel between the point of systes b'

pressere messorense.d and the pressere en the reacter vessel r

g reglen contreilla "the llelt eveve, nor de they loclude any s

w m

g, additional eeryIn of safety for possible lastrement error.

r 6

1500 L

E on m

[

~

o g

o O.

g 1000

.O g

O

=

e Criticality Limit (273 r) g T

'500 c

_o Assumed Adjusted RT...FR,3, 8 n

~

t m

Beltilne Region 1/4T 14$" f a

z e

Beltilne Region 3/4T 122 ~

.i tr l

I I

I I

I I

I I

I 2*f 0

m O

100 200 300 400 O

Reactor Vessel Coolant Temperaturt F

sa e

+

4

=

~ -

Figure 2-4.

North Anna tinit 1 Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation - Cooldown. Apolicable for First 12 EFPY I

2500 Assumed Adjusted RT,,,,

F O

Beltline Region 1/4T 145 55 m

Beltline Region 3/4T 122 o

c.

2 2000

.y m

?

E 3

Q m

.x5 8

1,500 3 '.*

1 i

tE E

o a

E O

1000 o

o O

Cooldown Rates 3

• F/Hr The acceptable pressure-tagerature combinettens are below and O

to the right of the llelt cerves. The Italt cerves de m

500 2j inciade tne pressere differeattel between the point of syjg M W

II" pressere nessarement and the pressere en the reacter reesg a g 60 regioncontrollingthelleltcerve,merdetheyincludeanyz$

too a6dittenal margin of safety for possible lastrument errorm P

?

j g

i 1

i i

i i

ye; 0

100 200 300 400

• h 600

.~g Reactor Coolant Temperature, F

$5 o

=

4 t

e

^

Dockit Nos.: 50 338&S39 S:rl'.I No.

91 707 Section 5 Page 35 of 80 l

In t.ddition to the North Anna Unit I revised heatup and cooldown curves, the original Babcock and Wilcox Report BAW2146 documented revised curves for North Anna Unit 2 applicable to burnups of 12 and 15 EFPY. Virginia Power has removed the pages from BAW-2146 which document these cur ves as they are not being submitted for review and approval at this time, f

i VIRGIN!A ELECTRIC AND POWER COMPANY

DockIt N:s.: 50 3388339 l

+

S: rial No.

91 707 Section S Pa0e 36 of 80 1

3.

CERTIFICATION The pressure temperature operating. limits for North Anna Units 1 and 2 reactor pressure vessel were calculated using approved procedures and established methods and techniques in accordance with the requirements of 10CFR50, Appendix G.

&%w Io //ohl A. D. NTfir -

'Date Materials and Structural Analysis MO eW 10 llo } 4 l--

M. J. 6evan Date Materials and Structural Analysis This report has been reviewed for t ical-content and-accuracy.

c15 W I&f/91 L.? B'. Grcss. P.E. (Materials Analysik)-

Date

~

Materials and Structural Analysis nk

/c 6 E/-

K. K. Yoon, P. E4 (Fractu're Analysis)-- 'Da'te Materials and Structural Analysis -

Verification of independent review.

JDYJ f

J

'K. E. Moore, Manager:

'Date Materials and Structural Analysis This report has been approved for release.

/DYfl9/

An' T. L. Baldwin -

Date Program Manager

[

3-1 VIRGINIA ELECTRIC AND POWER COMPANY r

r

,.-y y

.. ~. _.

Dock:t Nos.:- 50 338&339 Sul:1 No.

91 70?-

Section 5 Page 37 of 80 i

4.

REFERENCES 1.

S. E. Yanichko, et al'., Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance J

Program, WCAP-11777, Westinghouse Electric - Corporation, Pittsburgh, Pennsylvania, February.1988, 2.

E. Terek, et al., Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 2 Reactor. Vessel Radiation Surveillance Program; EAP-12497, Westinghouse - Electric Corporation, Pittsburgh, Pennsylvania, January 1990.

3. - Code of Federal Regulations, Title 10, Part 50 Domestic Licensing of Production and Utilization Facilities. Appendix G, FractureL Toughness Requirements, November 30, 1986.

4.

U.S. Nuclear Regulatory Commission, Radiation Embrittlement --of Reactor Vessel Materials, Egoulktory ruide 1.99. Revision 2, May 1988.=

5.

J. C. Schmertz, 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 i

Operation, WCAP-11791,: Westinghousei-Electric Corporation,' Pittsburgh, Pennsylvania, May 1988.

6.

A; L. Lowe, Jr., Reactor. Pressure Vessel and Surveillance Program Materials.

Licensing Information for North Anna Units 1 and' 2, BAW-1911. Revision :1,-

Babcock & Wilcox, Lynchburg, Virginia, August 1988.

4-1 I

l VIRGINIA ELECTRIC AND POWER COMPANY

~.

Docket Nos.f 50 330&J39.

Sulil No.

91 707 Section 5-Page 08 of 80 7.

N. K. Ray, and J. M. Chicots North Anna Unit 2 Reactor Vess61 Heatup and Cooldown Limit Curves for Normal Operation (Capsule U), MCAP_12101, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, Merch 1990.

8.

Code of Federal Regulations. Title -lo, Part 50, ~0omestic l,icensing of ProductQn and Utilization Facilities, Appendix H. Reactor Vessel 14aterial

)

Surveillance Program Requirements, November 30, Ig%.

9.

VEPC0 North Anna Power Station Units 1 and 2 Final Safety Analysis Report, Part B, Volume III-A, Paragraph 5.4 Figure 5.4-1, dated January 3,1973 USNRC Docket Nos. 50 338 and 5-339.

4 4-2 VIRGINIA ELECTRIC AND POWER COMPANY

Docket Nos.: 50 338&339 Stritt No.

91 707 Section 5 Page 39 of 80

(

APPENDIX A.

North Anna Unit-1 Data Points for Heatup and Cooldown Applicable to 12 EFPY 4

A-1 VIRGINIA ELECTRIC AND POWER COMPANY.

l:

Dock:t Not.: 50 338&339 Sarlal No.

91 707 H

Section S Page 40 of 80 merth eres Bettiene uncorrected P/T Limits unit 1 12 (FPV m0TE: The P/T Lleits provided:

a) ao not account for location & instrument channet umertainties i

M es not incluse Cleoure head Limits.

'J

......................................... M i n P /T L i e l t s "... " " " " ". " " """ CD " ".

sheLL

.......... NU......a...

Shelt Fluid 20

- 40 60 35 FLuis 20 40 60

-100-Ta F/hr F/hr F/hr

. T/4 ton F

F e

up

- eeeeeeeeeeeeeeeeeeeeee...e..

. /hr/hr-F/hr F/ht seasone:__

rea m esseeseeeeeeeeeeeeeeeese -

75 150

-549 548 552 570 3532 3532 3532 3532 to 554 552 552 558 545 3511.

3512 3514 3516 85 559 557 555 565 560 3495 3491 3491' M94 to Se5 542 540-5 72 555 3485 34 73 3470 3471 95 572 -

M7

$45 SM 550 3478 3458' 3451

-3449 100 SM 5 74 570 587 545-3474 3447 M35 3429 105 587 580 576 596 540 M71 -

3437 Mit Met 110 595 588 578 605 535 34e9 3430 MOD 3392 til 404 596 581 615 530 3448 3424 3399 3376 120 614 605 585 626 525 3448 3419 3390 3M1 125 625 615 591-637 520 3467 3415 3382

-3347 130 636 625 599 650 515 3u?

3413 3375 3335 135 MS 636 608 M3 510 347 M10 3370 3323 140-462.

648'

- alt 677 505 3467-

-3409 3345 '

3313 145 -

6 76 M1 G1 M2 500 3467 E07 3M1 -

3304 150 M1 675 M5 709 495.

3448 34M 3358 3295 155 M7 M0 640 727 490 3448 3406 3355 -

3287 160 725 706 677 74 485 3448 -

3405L

-3352-

.?200 165 143 723 M5

- FM -'

480 3448 3405 3350 3274:

170 7M 742 715 188 475 348-

-3405 3349 3268 175 185 762 73 7 812 470 3448--

Moe -

3347 32 6 -

100 809 783' 760 837-45 3468 3404 3346 3258 185 834 806 782 M5 440 3448 Mes 3345 3254 190 861 831 805 894 455 3448 '

3404 3M4 3250-195 000 857 829 6M 450 -.3448 3404 3M3 32M 200 921 886 855 9e0 445 3468 3404 -

3M3

~ 3M3 205 954 917 883 997 440 3448 3404 3342 3240 -

210 900 950 914 10M 435 3448 3404 3342 3237 215 1029 985 94 1018-430 3448 3404 3341-3235-220 1971 1023 981 1124 425 ~-

3448 3404 3341 3232 225 1115

- 10M 1019 1173 420 3449 '

5404 3341' 3230--

230 1163-1108 1059 1225-415 349-3404 3341 3229-235 1215 1156 1103 1282 410 3449__

3405 3341 3227 240 1271 1206 1149 1343 405 -

34M 3405 3341.

- 32M >

245 ~

1331 1261

• • 99-MOS 400 34M -

3405 3341 3225 250 -

1395 1320

253 1478 395 349 3405 3341 3224-255

%M 1383 1311 1554 390 34M 3405 '

3341 3223:

260 1539 1451

.1373 105 385 3449 3406 3442 3222 M5 1619 1524 1440 1723 380

. 3449 --

3406 -

3342 3221-270 1705 1603 1512 1817 -

370

~3469 34M

. 3342:

3220 375 34M - - 3406 3342 3221 275 1797 1M8 1590 1917 280 1896 17M 1673 20M 345 3449 -

3406 3342 3220 285 2003~

1876-1762 2143 360 3449 -

3406 3342 3219-290 2117 1981 1858 2268 355' 34M 3406 3342-3219 295 2M1 2994 1962 2403 350 3469 3406 3342-3219 300 2373 2215 2073 2548 345

- 34M 3406 3M2-

-3219 305-2515-2345 2192 2704 340 34M 3406 3342 3218 310 2468 2485 2320 2872 335 3449 3406 3342 3218 315 2832 2435-2458 3052-330

- 34 70 3406 3342

'3218 320 3008 2796 2605 3244 325 3455 3406 3342 3217 4

325 3198 2970 2FM -

3455 320 32 4 -

32 4

'3246 3217 330 3402 3154 2934 3532 315.

3052-3052 3052 3052.

335 3532 33 %

3117 3532 310 2872

' 2872

-2872 2872.

. 340 3532 3532 3314-3532 305 2704 2704 2704 2704 345 3532 3532-3519 3532 300 2548.

2548 2548 2548 350 3532 3532 3532 3532 295 2403 2403~

-2403 2403 355 3532 3532 3532 3332 290 2268 2248 2268 2264 -

340 3532 3532 3532 3532 285 2143 2143 2143 2143 345 3532 3532 3532 3532 280 2026

- 20M 2026 20M 370 3532 3532 3532 3532 2 75 1917--

1917 1917 1918.

375 3532

-3532 3532 3532 270 1817 1817 1817

-1817 380 3532 3532 3532 3532 265 1723 1723 1723 1725 i

385 '

3532-3532 3532-3532 260 105 105 1635 105 390.

3532 3532 3532 3532 255 1554 1554 1554 1554 VIRGINIA ELECTRICyp POWER COMPANY l

- - - -.. ~ -

~ - - - -. -.

Dock:t Nos.:.50 338&339 Striil No.

91 707-Section S Pace 41 of 80 395 3532 3532-3532 3532 250 1678 1478 1673 1473-400 3532 3532 3532 3532 245 1608 140d 140s 140s 405 3512 3532-3532 3532 240 1343 1M3 1M3 1M3 410 3532 3532 3532 3532

-235 1282 1282 1242 1282 415 3532 3532 3532 3532 130 1225 12 5 1225 1225 420 3532 3532 3532 3532 2M -

1173 1173 1173 1173 45 3532 3532 3532 3532 220 1124 1124

'1124 1124 410 3532 3532 3532 3532 215 1074 1975 107B 1975 -

435 3532 3532 3532 3532 210 1028 1025 1027 10M 440 3532 3532 3532 3532 445 3532 3512 3532 3532

- 205 905 975 976 987 200 946 935 928 930 450 3532 3532 3532 3532 1M 908 SM 884 877 455 3532 3512 3532 3532 190 4 74 857 El 428 460 3532 3532 3532 3532 185 M2 422 805 732 445 3532 3532 3532

• 3532 180 812 789 769 739 470 3532 3532 3532 3532 175 735 759 736 700 475 3532 3532-3532 3532 170 759 731 705 663 480 3532 3532 3532 3532 165 735 705 67F 629 445 3532 3532 3532 3532 160

- 713 681 651 598 490 3532 3532 3532 353Z 155 692 659 627 564 495 3532 1532 3532 3532-150 673 634 604 Mt 500 3532 3532 3532 3532 145 655 619 583 516 505 bill 3532 3532 3532 140 639 491 564 492 510 3532 3512 3532*

3532 135 634 See M6 471 515 3532 3532 3532 3532 130 610-569 529 451 520

'J532 3532 3512 3532 15 597 555 5 14.

433 55 3532 3532 3532 3532 120 SM

$42 4M-

- 416 530 3532-3532 3532 3532 115-5F3 530 -

436 400 535 3532 3532 3532 3532 110 M2

$19 474 386 540 3532 3532 1532 3532 105 553 508 463-3F3 545 3532 3532 3532 3532 100 M4 499 453 360 550 3532 3532 3532 3532 95 535 490 444 M9 555 1532 3332-3532 3532 90 527 482-435 339 MO 3532 3532 3532 3532 tb 520 4 74 -

427 330 M5 3532 3532 3532 3532 30

$13-447 419 321 75 SST 461 412 313 Leak-Test Data.

(12 EFPY)-

i Pressure Temperature-l fosia)

(F) 1971 250.0 2297 265.0 2488 272.5 9

- VIRGINIA ELECTRidAkD POWER COMPANY

Dock _t Nos.: 50 3388339 Sirial No.

91 707 Section 5 Page 42 of 80 1

In addition to the North Anna Unit I revised heatup and cooldown curves, the original Babcock and Wilcox Report BAW-2146 documented revised curves for North Anna Unit 2 applicable to burnups of 12 and 15 EFPY.

Virginia Power has removed the pages from BAW-2146 which cocument these curves as they are not being submitted for review and approval at this time, s

VIRGINIA ELECTRIC AND POWER COMPANY

Dock;t Nos.: 50 338&339 S:rkl No.

91 707 Section 5 Fage 43 of 80 WC AP-12503 l

l l

VIRGINIA ELECTRIC AND POWER COMPANY

WESTINGHOUSE CLASS 3 Docket Nos.: 50 3388332' S: rill No.-

91 707:

1 Section 5 -

Page 44 cf 80 WCAP-12503

~

i s

C NGRTH ANHA: UNIT 2 REACTOR VESSEL HEATUP AND-C00LDOWN LIMIT CURVES

.FOR NORMAL OPERATION (CAPSULE U);-

t N.-K. Ray--

J. N. Chicots Work Performed for. Virginia-Power Company March 1990 Approved by:

M b ft' T. A. Meyer. M(nager-..

Structural Materialsland Reliability; Technology Work Performed Under Shop Order'VJGP-139 WESTINGHOUSE ELECTRIC-CORPORATION

' Nuclear and Advanced = Technology Division 1 4

= P.O.- Box 2728.

Pittsburgh, Pennsylvania 15230-2728 ei 1990'Wastinghouse Electric Corp.

1 VIRGINIA ELECTRIC AND POWER COMPANY-

- w w-3

-r,--re-.-w

+-,e.

--,m

.w--

~

v,*.--e e-i+--

Dock:t Nos.: 50 338&339 Sert:1 No.-

91 707-Section 5-Page 45 of 80-TABLE OF CONTENTS F

Section Title?

Page TABLE OF CONTENTS.

i LIST OF TABLES ii LIST OF FIGURES iii

1.0 INTRODUCTION

1 2.0 FRACTURE TOUGHNESS PROPERTIES

.1 4

3.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 2

4.0 HEATUP AND C00LDOWN LINIT CURVES 5

5.0 ADJUSTED REFERENCE TEMPERATURE 6

6.0 REFERENCES

-19 APPENDIX A:' DATA POINTS FOR HEATUP AND C00LDOWN CURVES A,

4 VIRGINlA ELECTRIC ANg POWER COMPW

.m.ma m

-,..w..-

..a-

Dock t Nos.: 50,338&339 4

S:ri:1 No.

91 707 Secti:n 5 Page 46 of 80 LIST OF lABLES Jable 7,i tl e.-

Pai*

1 North Anna Unit 2. Reactor Vossal Bsitline Region Material Properties 8-2 Calculation of Chemistry Factors Using North Anna Unit 2 Surveill,ance Capsule Data-9 3

Vessel Neutron Exposure Values (E>1.0 MeV) for use 10,

in Generation of Heatup/Cooldown Curves 4'

Calculation of Adjusted Reference Temperatures for North Anna Unit 2 Reactor Vessel Material -

Lower Shell 11 5

Calculation of Adjusted Reference Temperatures for North Anna-Unit 2 Reactor Vessel Material -

Intermediate Shell 12 6

Calculation of Adjusted' Reference _ Temperatures for North Anna Unit 2 Reactor Vessel Material -

i Welds 13 7

Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Location 14 VIRGINIA ELECTRIC ANDPOWER COMPANY

-=

m Dock:t Nos.: 50 338&339 Seri:1 No.

91 707 Section 5 Page 47 of 80 LIST OF FIGURES Figure Title.

Page 1

North Anna Unit 2. Reactor Coolant System Heatup Limitations Applicable for the First 17 EFPY (Without Margins for Instrumentation Errors)

Up to 20'F/hr 15 2

North Anna Unit '2 Reactor holant System Heatup Limitations Applicable for the First 17 EFPY (Without Margins for Instrumentation Errors)

Up to 40'F/hr 16 3

North Anna Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 17 EFPY (Without Margins for Instrumentation Errors) Up to 60'F/hr 17 4

North Anna Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 17 EFPY (Without Wa,rgins for Instrumentation Errors) 18 VIRGINIA ELECTRIC A$ POWER COMPANY

.~

Dock:t Nos.: 50338&339 S; rill N2.

91 7o7 Secti:n 5 Page 48 cf 90 HEATUP AND CODLDOWN LIMIT CURVES FOR NORMAL OPERATION

1.0 INTRODUCTION

Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel.

The most limiting RT f the material in the core region of the reactor vessel NDT is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ART RT is NDT.

HDT designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RT increases as the material is exposed to fast-neutron radiatien.

NDT Therefore, to find the most limiting RTNDT at any time period in th.

reactor'slife,ARJNDT due to the radiation exposure associated with that,

time period must be added to the original unieradiated RT The extent of NDT.

the shift in RT is enhanced by certain chemical elements (such as copper NDT and nickel) present in reactor vessel steels.

The Nuclear Regulatory Commission (NRC) has published a method for predictina radiation embrittlement in Regulatory Guide 1.s9 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)Ill. The fluence values used in this report are from Surveillance Capsule U report (2).

l 2.0 FRACTURE TOUGHNESS PROPERTIES i

The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plani 33 The pertinent chemical and mechanical properties of the beltline region plate and weld materials of the North Anna Unit 2 reactor vessel are given in table 1.

VIRGINIA ELECTRIC AND POWER COMPANY

_. _ ~

.. ~ _ -. -

E e

.Docktt Nos.: So.3384339 Seri:J No.

91 7o7-l Section 5 Pcg3 49 of 80 The chemistry factors and " margin" (M) tcres th'at are also shown in table.1 were determined in accordance with the Regulatory _ Guide _1.99 Revision 2.

1 1

Chemistry factor and margin values in table 1 that were= based upon cred!ble surveillancemeasurementsarealsogivenfoEthebeltlineregionmateriais j

where this data was available.

Table 2 gives further information on the determination of the chemistr'y factors from this data in accordance with the applicable procedure from Revision 2 to Regulatory Guide 1.99.

For.those reactor vessel materials ~where credible surveillance data exists, the respective chemistry factor and margin terms are lower than the values based-on material chemistry measurements.

However, credible surveillance data is not available for lower shall plate which is.the-limiting material.

3.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating.the allowable limit curves for.various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup g

-- cooldown cannot be greater than the reference stress intensity factor,_

KIR, f r the metal temperature at that time.

K is obtained-from the-IR reference fracture toughness curve, defined in-Appendix G to the ASME Codein.

The K curve is given by the following equation:

IR KIR=26.78+1.223-exp[0.0145(T-RTNDT + 150)]'

(1) i where KIR = reference stress intensity factor as a function of_t'e metal h

temperature T and the metal reference nil-ductility temperature RT NOT Therefore, the governing equation for the'heatup cooldown analysis is defined in' Appendix G of the'ASME CodeI43 as foll m -

CKgg + KIT

• EIR (2) l l

VIRGli41A ELECTRICfD POWER CCMPANY

.,,,, _,7

e Dock:t N:s.: 5o 3388330=

S:ri:1 Ns.

91 707-Section 5-Page So of 8o 1

.nare K.H astressintensityfactorcausedbymembrane(pressure) stress f

KIT = stress intensity factor caused b'y the thermal. gradients a

KIR = function of temperature relative to the RTNOT of the matir'al C

= 2.0 for Level A and Level B service limits C

= 1.5 for-hydrostatic-and leak test conditions during which the reactor'

, core is not critical

- At any time during the heatup or cooldown transient, K is determined by.

gg thi metal temperature at the tip of the postulated flaw, the appropriate valu(.

- for RTNDT, and the reference fracture toughness curve.- The thermal stresses-resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, f r the reference flaw are computed.

From equation 2.-the pressure-stress intensity factors are obtained and, from these,-the allowable pressures are calculated.

For the calculation of the allowable pressure versus~ coolant temperature during cooldown, the-reference flaw of Appendix G to the-ASME Code is assumed to exist at the inside of the vessel wall

'During cooldown, the controlling location of the flaw is'always at the inside of the wallibecause the thermal-I gradients produce tensile stresses at the11nside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady state and finite cooldown rate-situations. From these relations, composite limit curves are constructed for each cocidown rate of interest.

The use of the composite curve in the cooldown analysis is necessary 'because control of the cooldown procedure is based on the measurement of reactor' coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

'ui - ie VIRGINIA ELECTRIC AND POWER COMPANY A - a

. =

m g,

,%s,

,,mg%

-q w yh w-py

.g,m n,g-+

y 9m.lf-y9.-

g rp $g

-9 99w.-

Dock:t Nos.: 50 338&339 Serial N3, 91 707 Sectbn 5 Page 51 of Bo During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vessel 10.

This condition, of course, is not true for the steady-state situation.

It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state J

operation. Furthermore, if conditions exist so that the increase in K IR exceeds KIT, the calculated allowable pressure during cooldown will be greater than the steady-state value, The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temocrature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore, the Kgg for the 1/4 T crack during heatup is lower than the K f r the 1/4 T crack during steady-state conditions at the same IR time coolant temperature.

During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR s do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates w'</;n the 1/4 T flaw is considered.

Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep outside

~

surface flaw is assumed.

Unlike the situation et the vessel inside surface.

VIRGINIA ELECTRIC 4WD POWER COMPANY

• ' " = ' '

Dock:t Nos.: 5o 338&339 S: rid No.

91 7o7-S:ction 5 Page 52 of 80 the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present.

These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatap camp'.

Since the thermal stresses at the outside are tensile and increase with increasing heatup ratas, each ieatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-ooint comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to se the lesser of the three values taken from the curves under consideration.

The use of_the composite curve is necessary to set conservative heatuo limitations because it is possible for conditions-to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressura l b.it must at all times be based on analysis of the most. critical criterion.

9 Finally, the 1983 Amendment to 10CFR50(5L has a rule which addresses the metal temperature of the closure head flange and vessel flange regions.

This rule states that the metal temperature of the closure flange regions must exceed the material RT by at least 120'F for normal operation when the NDT pressure exceeds 20 percent of the preservice hydrostatic test pressure.

Table 1 indicates that the limiting RTNDT of -22*F occurs in the closure head flange of North Anna Unit 2, so the minimum allowable ter.perature of this region is 98'F. These limits are shown on-Figures 1 through 4, whenever applicable, 4.0 HEATUP M'D C00LDOWN LIMIT CURVES l

l Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section 3.0.

i Figures 1 through 4 are applicable for the first 17 EFPY.

No instrumentation error margins are considered in developing the heatup and cooldown curves.

VIRGINIA ELECTRIC 6ND POWER COMPANY

• m *'c2 = 5

.J

__ __..___ ~ ____._ _.

Dock:t Nos.: 50 338&339'

~

Seri:1 No.

91 7o7 Section 5 Pag) 53 of Bo Allowable combinations of temperature and pressure for specific twoeretmo change rates are below and to the right of the limit lines shown in figures 1 through 4.

This is in addition to other criteria which must be met before the reactor is made critical.

The leak liinit curve shown in figures 1, 2, and 3 represent the minimum temperature requirement at the leak test pressure specified by applicable codes (3,4)

The leak test limit curve was determined by methods of references 3 and 5.

Figures 1 through 4 define limits for ensuring prevention.of nonductile failure for the North Anna Unit 2 Primary Reactor Coolant System.

5.

ADJUSTED REFERENCE TEMPERATURE FroT Regulatory Guide 1.99 Rev. 2 I13 the adjusted reference t6mperature (RT) for each material in the beltline is given by the following expression.

ART = Initial RTNDT + ARTNDT + Margin (3)_

Initial RT is the reference temperature for the unirradiated material as NDT defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code.

If measured values of initial RT for the material in-NOT L

question are net available, generic mean values-for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class..

ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

g = (CF]f(0.28-0.10 log f)

(4)

ART To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following-formula must first be used to. attenuate the fluence at the specific depth, f

I*

I f(depth X) " Isurface (5)

VIRGINIA ELECTRICfAND POWER COMPANY

,~,

...~. -..

m Dock;t Nos.: 50 3384339 -

Serhl No.

91 707 Section' 5 Page 54 cf 80 -

where x (in inches)'is the depth into the vessel wall measured from the vessel z

inner (wetted) surface.- The resultant fluence is then put into equation-(4) to calculate ARTNDT at the specific depth.

~

CF ('F) is the chemistry factor, obtained from reference 1. -The fluence values from reference 2 are reproduced in table 3.

The peak fluence value is used to project cumulative fluence up to 17 effective full power years. The calculation of ART for the beltline region are materials shown in tables' 4, 5, and 6.

Table 7 summarizes the results of RTNDT at 1/4T and 3/4T locations.

t i

VIRGINtA ELECTRICAND POWER COMPANY L

Dock t N:s.:.50 338&339 Srt:1No.91-707 Section 5 PIge 55 cf 80 TABLE 1-NORTH ANNA UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES i

Ni,

1(*)

M(b) in.%)

(Wt.5)

CF

(*F)

(*F)

Intermediate Shell Plate

.09

.83 58 75

.34 (35.1)

(17)

Lower Shell Plate

.13

.83 96 56 34 Weld

.07

.05 38 56 (10.4)

(28)

Closure Head Flange (d)

-31 Vessel Flange (d)

-22 (a)

The initial RTNDT (I) values for the platss and welds are measured values.

.(b) Margin (M) as per Reg. Guide-1.90, rev. 2; the standard deviation for the initial RTNOT margin term is assumed to be zero since'the initial RTNOT values were obtained from conservative (i.e., " upper bound") test results.

(c) Numbers in ( ~ ) corresponds to surveillance capsule data.

(d)

Initial RTNOT values for closure head-flange and vessel flange will be considered for the adjustment of heatup/cooldown curves..

.VIRGINm ELECTRIQPND PO'NER COMPANY.

• m - is

Deck:t Nos.: 50 338&339 S:rt:1 No.

91 707' Section S Page 56 of 80 TABLE 2 CALCULATION OF CHEMISTRY FACTOR USING NORTH ANNA UNIT 2 SURVEILLANCE CAPSULE DATA-

.~

Material Capsule f

ff ART ff x ART ff NDT NDT 4

19 2

i (10 n/cm )

(.p)

Int. Shell V

.241

.615 9

5.535

.3782 (Tangential)

U

.956

.987 25 24.675

.9742 Int. Shell V-

.241

.615 9

5.535

.3782, (Axial)

U

.956

.987 60 59.22

.9742 I=

94.97 2.7048-o Chemistry Factor-(plate) =

I(ff x aRTNOT) 94.97 35*1-

=

2 2.7048 g(ff )

Weld Metal V

.d41-

.615 2

1.23-

.3782 U

.956

.987 13

-12.831-

.9742 I=

14.061 1.3524 Chemistry Factor (weld metal) =

14.061 o

10.4

=

.1.3524 Notes:

f = fluence in 1019 n/cm2 ff = flaence factor Capsule V information are from reference 6.

VIRGINIA ELECTRICRND POWER COMPANY I

A

u Dock:1 Nos.: 60 3384339 S:ti:1 f 91 707 Section 5 Page $7 of 80 TABLE 3 VESSEL NEUTRON EXPOSURE VALUES (E > 1.0 MeV) FOR USE IN GENERATION OF HEATUP/C00LOOWN CLRVES (2) 15 EFPY 32 EFPY 2

2 (n/cm )

(n/cm )

19 19 0'(a) 2.06 x 10 4.47 a 10 19 19 15' 1.03 x 10 2.23 x 10 19 19 30' 5.99 i 10 1.30 x 10 18 18 45' 4.19 x 10 9.11 x 10 (a) Maximum ooint on the pressure vessel s

VIRGifRA ELECTRIC gD POWER COMPANY 4:n. Sue o se

i

~.

Docket Nos.: 50 338&339 8: rial No.

91 707 Section 5 Page 50 of 80 TABLE 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR NORTH ANNA UNIT 2 REACTOR VESSEL MATERIAL -

LOWER SHELL y W

Regulatory Guide 1.99 - Revision 2 17 EfPT 1/4 T 3/4 I Chemistry Factor, CF 96 96 19 n/cm)(a) 1.483

.590 2

Fluence, f (10 Fluence Factor, ff 1.109

.852 eee.............n.*****ne.....u.......*n.,*,,,*............******.***.....u..

ARTN07 = CF x ff ('F) 106.4 81.8-Initial RTHDT,I('F) 56 56 Margin,M('F)(b) 34 34 e n. n *,n.......,***.... n.... n....... n.

n. n,**

,* n. n n u.... n......,...

Revision 2 to Regulatory Guide 1.99 Adjusted Reference lenprature, 196 172 ART = Initial RTNOT + ARTNDT + N"F91"

.n....

n.....

.n

. n... n.n n.. ** n u ne........ n..,n n. n.. n..,* n...n..

9 (a) Fluence, f, is based upon f surf I.10 nice, E>l Nev) = 2.35 at 17 EFPY.

The North Anna Unit 2 reactor vessel wall thickness is 7.677 inches at the beltline region.

(b) Margin is eticulated as, W = 2 (og2

,20.5 The standard deviation 3

for the initial RTNDT margin term (og) is assumed to be O'F since the initial RT is a measure value.

NDT YlRGINtA E0ECTRIC AND POWER CJAPANY

\\\\

4 Docket Nos.: 50 338a339 S:rlal N3.

91 7o7 Section 6 Pc9e 59 of 80 TABLE 5 i

CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR NORTH ANNA UNIT 2 REACTOR VESSEL MATERIAL -

INTERMEDIATE SHEtt Regulatory Guide 1.99 - Revision 2 17 U PT

~

1/4 T 3/4 I ChemistryFactor,CF('F) 58(35.1) 58(35.1) n/cm)(*)

1.483

.590 2

19 Fluence,f(10 Fluence Factor ff 1.109

.852

..n.... u. n... n.n n n n n n en u n,n... n

....****,n*,*******.,,,,,n *,*,,,n

. n t.RTNDT = CF x ff ('F) 64.3(38.9)

- 49.4(29.9)

Initial RTNDT.1('F) 75 75 Margin,M('F)(b) 34(37)

>34(37) n eas.nu nnesenne. sun.nw.,,asenn.aensaecus, vees.nwa*,,see wwwn.as ena.wassen Revision 2 to Regulatory Guide 1.99 AdjustedReferenceTemperature.

173(131) 158(122)-

ART

• Initial RTH0T + ARTNOT + N'"9i" l

. n a r. n n. n.. n.,* n. n.. n n, n... n.... n...... **,*. n.......... n **. n *..... n **,

19 2

(a)

Fluence, f, is based upon.fsurf(10 n/cm. E>l Mov) = 2.35 at 17 EFPY.

The North Anna Unit 2 reactor vessel wall thickness is 7.677 inches at the beltline

region, t

(b) Margin is calculated as, M = 2 (og2

,,2 0.5 3

The standard deviation for the initial RT NDT margin term (o;) is assumed to be O'F since the initial l-RT is a measured value.

The standard oeviation for ARTNOT,(ca)is17'F NDT for the plate. o is8.5'Ffortheplate(cutintohalf)whensurveillancedata g

is used.

( ) numbers in parenthesis were calculated using surveillance capsule data.

VIRGINIA QlfECTRIC AND POWER COMPANY

.m 4ue.o io

Dock:t Nos.: 50 3384339 Sort:I No.

91 707 Secti:n $

Page 60 of 80 TABLE 6 CALCULATION OF A0 JUSTED REFERENCE TEMPERATURES FOR NORTH ANNA UNIT 2 REACTOR VESSEL MATERIAL -

WELDS

?

Regulatory Guide 1.99 - Revision 2 17 EFPT 1/4 I 3/4 I Chemistry Facter, CF (*F) 38(10.4) 38(10.4) n/cm)(a) 1.483

.590 19 2

Fluence,f(10 Fluence Factor, ff 1.109

.852

.n n e..... n n.. n n u. n... n..n n... n.n....... n. n...w..... n.. n. n. n..nn...

ARTNDT=CFxff('F) 42.0(11.5) 32.0 (8.91 Initial RTNDT, I ('F)

-48

-48 Margin, M ('F) (b) 42(11.5) 32(9)

..e... *

• n.... n.. n

• n......n...........n..... u n.....u n. n n... n n.. n, n n. n Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 36(25) 16(-30)

ART = Initial RTNDT + ARTNDT + M*F91"

..n.. n... n... n. n n. n. *

• n.* * *... n..... n. n... n... u....n....... n.. n.n n e n,

19 2

(a)

Fluence, f, is based upon fsurf (10 n/cm, E>l Nev) = 2.35 at 17 EFPY, Tie North Anna Unit 2 reactor vessel wall thickness is 7.677 inches at the beltline region.

(b)

Margin is calculated as, M = 2 (og2

, 2)0.5 The standard deviation for the initial RTNDT margin term (og) is assumed to be O'F since the initial RT is a measured value.

The standard deviation for ARTNDT,(ca)is28'F NDT for the weld, e is 14'F for the weld (cut into half) when surveillance data is 3

used.

Also o need not exceed 1/2 ART g

NDT'

( ) numbers in parenthesis were calculated using surveillance capsule data.

• u ' ** u VIRGINIA ELESTRIC AND POWER COMPANY

-.. ~ - -. - -

(

Dock:t Nos.: 60 3384339 S:rt:1 No.

91 707-i Sectlin 5 Page 61 cf 80 TABLE 7 i

~~

i

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES (ART) AT 1/4T AND 3/4T LOCATION 17 EFPY I

i Component 1/4T 3/4T

('f)

(*F) l Intermediate Shell 173 158 (131)

(122)

Lower Shell 196.0*

172*

Weld 35 15

(-25)

(-30) i Numberswithin(

) are using chemistry factor based on surveillance capsule, Adjusted Reference Temperatures-of 196'F at 1/4T and 172'F at 3/4T used to generate hoatup/cooldown curves applicable up to 17 EFPY.

i-t VIRGINIA ELECTRIC QD POWER COMPANY

.m...

=

Dock:t Noa.: 50 3384339 Sortl N3.

91 707 MATERIAL PROPERTY BASIS CONTROLLING MATEP!AL:

LOWER SHELL PLATE INITIAL RTNDT:

56'F RT AFTER 17 EFPY:

1/4T, 196'f NOT 3/4T, 172*F i

CURVES APPLICABLE FOR HEATUP RAli$ it? R Wi r R FOR THE SERVICE PERIOD UP TO 17 EFFY.

CONTAINS NO MARGIN FOR P D 6Li ',STRUMENT ERRORS.

2500

.u vre i

1-i i i TT i i

i i ai i

' i i i

llll;;ll l

l l

l i

Leak Test Limit -

II It I

I i

6 I

l' I

I -

$150 M

6 II1 I

i i i 1

e i,

i l

I i !

i i i I

I

.1 T

!l i

I if I

1 a

2000 l

?

l i

i i

i i

e f

I It s

e s

{

]

<f ~

1750 l!

ll

/

l/

l l!

t,

I

!I

< i I

T F

f I

G 1500 Unacceptable Operstion

/

/

l l ] I j j{

f f

("

i I

/:

i

,i 7

i,

, i i 1

'1,,,

i w 1250 ll

/

l

- Heatup Rates up to 20'F/hr I

Acceptable Operation g

i.

m ll lj E 1000

'.,l 7

i

~

a

.i e

"-Criticality Limit nased on w

/

O

/

Intervice Hyerostatic Test u

750

,,l

/

Temperature 324'F for the j

f Service period up to 17 ErpY i

I I !

/

l

. ?

l_ i i

I I

i i,

ii i

i i l 1

I I I i

i I

i } ! 1 l 1

i i

4

i i

i 6 i.

i l

9 0

50 100 150 200

.250.

300 350 400 450 500' iwo scATED TEMPERATU9c (DEfr.F)

Figure A-2, North. Anna Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 17-EFPY (Without Margins for -

Instrumentation Errors) Up to 20'F/hr VIRGINIA ELECTRKl&ND POWER COMPANY l

Dock:t Nos.: 50 330A339 S: rial No.

91 707 Section 5 Page 63 of 80 1ATERIALPROPERTYBASIS CONTROLLING MATERIAL:

LOWER SHELL PLATE

!NtT!AL RTNOT:

56*F j

RT AFTER 17 EFPY:

1/4T. 196'T NDT t

3/4T, 172'F i

CV4yES APPLIDBLE FOR HEATUP RATES UP TO 40'F/HR FOR THE SERVICE PERIOD UP TO 17 EFPY.

CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS.

• bbhi M,2IEI'O !

! +

1 1 Ti 4

I e

iiiil l i I r

'l I

i i i

I j

r r

1 !

i 1 Leak Test Limit i

J J

i

\\_/

/

[

i 2250l i i 1 i

! i 1 r

r J

i I

I nI 2000 l

[

Y

,;,lj i

7,

1 i

r l

l

!?$0

[

/

l, l

i I

I i.

i

< i r

r ii!

j j

g gge Unacceptable Operation I

i j

i 1

iii i,

j j

, i i j

i 4

ie e i e ii i I

Acceptable Operation Z

$ gggol,

u

"~

A, I

ttIttt!

Heatup Rates Up to 40'F/hr

/

' i w

N i

1. !

i m

E 1000 N/

l _

I i '

1 i

/

i.iiiii o

W l

a

+-Criticality Limit Based on

/

u 750

/

--. Intervice Hydrostatic Test 5

li Temper $ture 324'F for the Service period up to 17 [FPY 5

. ;r 4

4 4 i ii is 500 !I W'

l ll i i i

i ! i,,

250

,i i

)

10

C0 li3 i;0 2i3 300 150 dC0 m

-100 INDIC ATCD TCMPERATVRE (DEG.F)

Figure A-3 North Anna Unit 2 Reactor Coolant System Heatup Limitations l

Applicable for the First 17 EFPY (Without Margins for InstrumentationErrors)Upto40'F/hr i

VIRGINIA ELECTRigfD POWER COMPANY ov.4m i,

Docket Nos,: 50 338&339 Sutt No.

91 707 Sects:n 5 Page 04 cf 80 WATERIAL PROPERTY BA$l$

CONTROLLING MATERIAL:

LOWER SHELL PLATE INITIAL RTNDT:

56'F RT AFTER 17 EFPY:

1/4T, 196'T t

NDT 3/4i, 172'F CURVE $ APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 17 EFPY.

CONTAINS NO MARGIN FOR POS$1BLE INSTRUMENT ERRORS.

1400 ! cs;>'ra i

i i e

i I

! f f

i i

i i

I il i

I Leak fest Limit I

r 1

i.i, lit 0 V N!

i l

I 5

l l

l l',

1 i

I r

r

j 2000 l

l i

I 1

I I

i I?to j

l l

/

j /

l

,1 i

i r

r

+

i ii i

i I

i!

7 1500

/

l:;

{

Unacceptable Operation I

r f

7-

/

/

=

1 t ii i

i i

I I

AcceptableOperationl w 1250 h

/

i I

4 i

W w:

I J

i i

i E 1000 l

,l

/

r o

Heatup Aates.Up to 60'F/hrq

/

'i N

• -Criticality Limit based on S

?!0

}/

Intervice Hydrostatic fest f

Teaperature 324*F for tha i

g f

l Service period up to 17 EFpY l

i '

500 l

M '

{j ',

~

l 4

250 l

l 1

l li l

l,

-m 50 i00 450 s00 L i 's 300 150 00 450 il0 INolCATCD TCWPCRATuRC (oCC r) i Figure A-4, North Anna Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 17 EFPY (Without Margins for Instrumentation Errors) Up to 60'F/hr VIRGINIA ELECTRIC k 4D POWER COMPANY

~

Docket Nos.: 50 3384339 S:rlal No.

91 707 Secti:n 5 Page 65 of 80 MATERIAL PROPERTY BA515 CONTROLLING MATERIAL:

LOWER SHELL PLATE

!NITIAL RTNDT:

56'T RT AMER 17 EFm 1/4T, 196*F ND7 i

3/4T. 172'F i

CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 17 EFPY. CONTAINS NO MARGIN FOR POSSIBLE INSTRUMENT ERRORS.

2500 u, m,r

,i 4

3 i!

t i

i j

i i

i i

i

! I I iy 2250 l

l

/

, i i

t i

i i 1

ii i

I 1

i 2000

/

,l 1

I r

1 a

I thiacceptable Operation

-f I

G 1500

/

l f

E

,F

/

l1250 w

7 i

i i i i ;

O i

Acceptable Operstion E 1000 a

s w

I r

i

' l 2

i i

u 750 jj [

C8'D!

nates-jg7 l

j ww

/

i 0

f~

540 20 -

-2f f

i i

~

i

-e

/

i i

i i

l iv +

l l

i 250

,, 100 l

l

,i3 3

-m d;0 450 0')

0

5) -

100 i?0

&OO E50

!?>

no i

INDIO&lCD TCWFCRATORc (C(6,f) l Figure A-5.

North Anna Unit 2 Reactor Cool. int System Cooldown Limitations Applicable for the first 17 EFPY-(Without Margins for instrumentation Errors)

VIRGINIA ELECTRIQ4ND POWER COMPANY

I Dock t hos.: 5o.33o8339 i

seri:1 NL 91 7o7 i

Secti:n 5 Page 66 cf 8e

6.0 REFERENCES

t (1) Regulatory Guide 1.99, Revision 2, " Radiation Erbrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission May, 1988.

(2) Terek, E., Anderson, S. L., Albertin, L., WCAP 12497, January 1990, l

" Analysis of Capsule U from the Virginia Electric Power Company North Anna 2 Reactor Vessel Radiation Surveillance Program".

(3)

" Fracture Toughness Requirements.* Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Otatiderd moview Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. LWR Edition, NUREG-800,1981.

(4) ASME Boiler and Pressure Vessel Code, Section !!!, Division 1 -

Appendixes, " Pules for Construction of Nuclear Power Plant Components.

Arspendix G. Protection Against Nonductile failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York, 1986.

(5)-CodeofFederalRegulations,10CFR50,'AppendixG,"FractureToughness Requirements," U.S. Nuclear Regulatory Commission. Washington, D.C.,

Federal Reg 4 ster, Vol. 48 No. 104, May 27, 1983.

[6]

" Analysis of Capsule V, Virginia Electric Power Company North Anna Unit 2. Reactor Vessel Materials Surveillance Program", BAW-1794, Lowe, A. L. Jr. et al. October,1983.

VIRGINIA ELECTRIC Af% POWER COMPANY

Dock.t Nos.: 50 338&339 S:ri:t No.

91 707 Secti:n 5 P:gs 67 cf80 APPENDIX A Data Points for Heatup and Cooldown Curves 4

l VIRGINtA ELECTRIC AN6 POWER COMPANY

,.4

l

[

VG8 M2AfuP CURVE REG. GUIDE 1.93, KEV.2 01/24/90 l

l THE FOLLOWING DATA WERE CALCULATEDFGR THE INSERVICE HvDROSTATIC LEAM IEST.

I MINiseUM JNSERVICE LEAK TEST T E sePE Ra f uRE ( 17.000 EF.*v )

PRES $URE (PSil TEsePERATURE (DEG.F 3 2000 302 2485 324

$x9z 5

m PRES $uRE PRESSURE STRESS t. 5 M ise g

(PSI)

(PSI)

(PSI SQ.RT.IN.)

S y

2000 2 toss e4522

-O

$"7 24a5 26156 105s'23 O

7 O

bn 8,

h En (f) O

$ b. 8

&e:

=zC

• 9$*

Op *O I

246 I

Ow

,NC g,

98 E

CJ U

8lD O

C 4

Docket Nos.: 50 338&339 S:rtl N).

91 707 Section 5 Page 69 of 80 i

a 5

!s: ;;;P:;;i";*.8999;".;

s.. n.......--......

~"t 80tN 2200SER* 82 ~

9 o

g

---

anan a

om

!!Ealtaaaaaalatalaala ji.isessigiggseninis.ii nnnannn nnnnnnn n e

5

-....a,n,..,.,n,.,.,S.an,

.o ann.,

3 o

T-3:5R"* K'2:E035*RE 33 JW-WgdenW.4 J E

j01 3:03n:RERO.

g4.J6 x2283 1

t o%

ESS.$$$$$$$E$$E$$$$!$

!$5$

n 918 0:2g.gfoWo.o.o.o.g od w

4

- nns NN : anan t

g.

~:

a 5

i s

%. }%

.o-an...n..o-an...

-nnunnnnnnnnnnnnnn t*

5a i

9:98";;::.899;*;:;

i 6: 2,g:

e d

.5

. _0 2.:.2 0 : 2 2 3." 3 E..:.: g C A A 3* g3 3

~'

w g-

"s; 5:%?a 253.llfallallalasalaa l

k

'N CE 9te gggfo.odo.osp.jWo e

- :: cec:: e ee I

a S.

~:

8

$ E3 U mm

"""'*****9:UCI**t*

A-3

~

VIRGINIA ELECTRIC AND POWER COMPANY

VG8 40HEATUP C1:RVE REG. GUIDE 1.99 EEW.2 09/24/90 I~

COMPOSITE CDRVE PLOTTED FOR HEAIUP PROFILE 2 HEATUP RATE (5) ( OEG. F /Hr. )

40.0

=

IRRADIATION PERIOD

$7.000 EFF VEARS FLAW DEPTH = (t-AOWIN)7 f

INDICATED INOICATED INDICATED INDICATED INDICATED INDICATED TEMPERATURE PRESSURE TEMPERATURE PRES $URE T E tFE R A TURE PRESSURE t

(DES.F)

(PSI)

(OEG.F)

(PSI)

(Ote. 7 )

(PSI) 1 85.000 466.004 20 180.000 613.92 38 270.000 st22.83

-2 90.000 484,4tj N M 29 185.000 528.27 39 275.000 1174.99 3

95.000 483.06 22 190.000 843.85 40 380.000 1224.82 4

100.000

'493.27 23 195.000 640.12 41 285.000 1275.10 5.

105.000 495.06 24 200.000 677.35 42 200.000 1330.13 j

G 110.000 497.85 25 205.000 694.97 43 295.000 1387.15 7

195.000 50s.67 -

26 210.000 717.54 44 300.000 1447.41 8-

$20.000 506.10 27 215.000 739.72 45 305.000 1512.06 9

125.000 511.40 28 220.000 743.3e 46 310.000 1541.68 i

3 to 130.030 517.28 29 225,000 788.95 47 3t5.000 1655.78 9

ft 835.000 523.06 30 230.000 514.32 48 320.000 17;S.73 Z

12 140.000 531.00 31 235.000 845.67 49 325.000 1820.99 L

5 13 145.000 538.83 32 240.000 877.42 50 330.000 1812.30 I

ITT 14'

.150.000 547.16 33 245.000 991.38 51 335.000 2C09.92 g

IS 155.000 556.35 34 250.000 947.79 52 340.000 2114.38 16 160.000 566.21 35 255.000 996.92 53 345.000 2225.88 h

17 165.000 576.Se 36 260.000 1029.t7 54 350.000 2345.12 m

88 170.000-

'588.36 37 265.000 1074.33 55 355.000 2472.24 g

ts

'175.000 600.63 b

O>

"U E o

t' n

s 4

I$b f

& E if

=z; on O &

- 2 "D 9 Ut I

Y

. ~w g

you r

oN* 98 9

o LD O

O y

a

t-i-

l whts sostAtup Cued W4 NEG GUIDL 9, 99.NE V 2 O t/24 /SMJ f

l -.

C000POSItE CuuvE.PLDITED FOR HEAIUP PROFILE 2 HEAluP RATE (5) (DEG t/HRB =

60.0 IRRADIATION PERIOO 17.000 EFP VEARS

=

FLAW OEPTH = (t-AOWINST INDICATED 1pelCATED Ife1CATED INDICATED IWICAIED INDICATED T E stPE R ATURE ' PRE 55ueE T E ASPER A TURE PRE 55uRE T E asPE R A TURE PRESSURE (DES.F)

(PSI)

.(DEG.F)

(PSI)

(DES.F )

(PSE) 1 85.000 ese-+f 20 180.000 549.29 38 270.000 2005.04 2

90.000 4eu 21 185.000

'573 63 39 275.000 1050.55 dN 22 190.000 544.93 40 280.000 1099.*2

'3 95.000-m M 4

-100.000 N

23 195.000 601.16 49 205.000 t151.82 5

105.000

• • &ef a 24 200.000 stG.G2 42 290.000 120s.04 6

t10.000 473.41 23 205.000 G33.29 43 295,000 1268.40 7

185.000-474.31 JG 210.000 659.06 44 300.000 1332.98 4

8 120.000 476.to 27 215.'000 G70.37 45 305.000 1402.35 e

125.000 479.10 28 223.000 691.09 46 310.000 147s.62 3

90

$30.000 482.79 29 225.000 713.28 47 395.000-1556.29 9'

31' 136.000 487.34 30 230,000 737.25 48 320.000 1641.53 l

Z.

12' 140.000 492.59 31 235 000 762.99 49 325.000 1732.78 5

13

.145.000 498.61-32 240.000 790.59 50 330.000 1430.42 14 150.000 505.22 33 245.000 820.26 51 335.000 1334.85 ITI 55 155.000

'512.67 34 250.000 352.06 52 340.000 2046.38

/ p (. >

fii ts 150.000 520.s2 35 255.000 ses 40 53 345.000 21s5.s0 O

17 165.000 529.76 36 260.000 923.17 54 380.000 2282.78 i(9 1 - 1, y

to 170.000-533.4s 37 265.000 9s2 se 55 355.000 2402.2e g

IS, 175.000 549.90 EI" I<

l O "'

]

R 1

1 8 $. 8

& e. ir

=z-eoZ

~ 3 g

(s TOG

= -o

(

I~

Q 1 L <*,

I.

/) v you

. ~4 e So E

AJ 9

[

Q l.

,i;[Ik^

I

,r Iil>i

}:[ ;(

,i: *' *

??

!,! I

• F; I

[.

e

. Z2:. mobwe9ww#

p

.%l 2o-eabOn i

.8&2 m m.3 yu E,O p

,t e

~

0 9/

42

/

9 0

O EE 34220177207863087 TR) 22383083981594699 AU1 C5$ 27562254761'182318 IS?

02728407428080135 DE(

09922344567880123 NR 11181181111112222 IP 3

E DR n

EU) 00000000000000000 wO T T F.

00000000000000000 AA 00000000000000000 O

CRS L

IEE 80908050808050508 O

DPD 47708990019223344 O

9e(

22222223333333333 4I C

1E T

t ia f

7800t234567880t23 5

3334a444444445S5S f

O A

n E

T S

D

(

Ef 458938469347318897 n

Tt) 590248537512840675 M

A I

t CSS 571S06433594102510 4

ISP 457801357914703604 DE(

S686777777S8899900 E

NR 11 L

IP I

F O

E R

DR P

Ets) 000000000000000000 N

T f F.

000000000000000000 AA 000000000000000000 W

CRG O

1EE 505050505050505050 m

. 2 D

OPD 764990011223344596 L

N9(

111112222222222222 E

V O

IE E

O 5 T

C.

C R a

3 R

E 9

O T

9012345678901234S6 F

1322222222233333S3 P

8 O

F E

E E

D f

.I f

0 U

O 0

. G L

0 P

O 7

EE 4884868163446078e4 G

E 1

TR)

E R

AUI 5 8 4 3 4 9 4 8 3 3 7 6 0 9 4 6.S.2 R

E CSS 693715051730454334-W

• I ISP 223344556679890123 5

DE(

555555555555556666 L

A DN NR

- V.

T OI

- IP R

A IW U

D RO C

EA

- E G P DR N

N

=

EU) 000000000000000000.

W I

N T T F.

000000000000000000 O

W OH 000000000000000000, AA D

O IT CRG L

L IP IEE 505050505060505050 O

L aE DPD 899001822334455667 O

O lO Ns(

11 1111111111918 a

C P

O IE AW T

8 t RA s

G WL e

V t

IF 1234567s9ot2345678 ti1891311

<3Qg 1Thg43 O a

)

T sE f t

.I a

j!.I!

i

![i!',1!Ii'

!i L!IlIl,it

n VG8 COO 1DOWN CURVE S REG. GUIDE 1.99 REv.2 01/24/90 THE. f0LLOw1NG DAIA WERE PLOffEO FOR COOLDOWN PROFILE 2

-(20 DEG-F / HR COOLDOW4 )

IRRADIATION PERIDO +

17.000 EFP VEARS

, FLAW DEPTH = AOWIN Y l>OICATEE INDICATED INDICATED INDICATED IMO!CATED INDICATED i

T E 94PE R A T URE PRESSURE T E 94PE R AluRE PRE 55URE' TEsePERATURE PRES 5URE (DES.F )

(PSI)

( DE G.F )

(PSI).

(DES.F)

(PSI) i-

^ 'SS.OGO

-488.96 -

14 150.000 560.43 27 216.000 747.64 1-l.

2 90.000 492.32 85 155.000 569.39 28 220.000 770.89 3

es.OOO 4es.s7 tG 160.000 s7s.cf 2s 225.000 7es.70 4

100.000 499.89 17 165.000 589.39 30 230.000 822.61 5

106.000 504.04 18 170.000 600.41 31 236.000 851.33 6

110.000.

506.60 19 175.000 612.44 32 240.000 832.39 7

115.000 513.54 20 180.000 525.37-33 245.000 315.70

.s.

120.000 Sts.e4 21 3s5.000 s39.29 34 250.000 951.41 9

125.000

-524.57 22 190.000 E54.13 34 2S5.000 999.83 g

10 130.000 530.72 23 195.000 670.26 3G -

260.000 1031.31 E

'S1 136.000 537.3e 24 200.000 587.59-37 265.000 1075.78 92 '

140.000 544.53 25 205.000 706.12 38 270.000 t123.49 IT1 33 145.000 552.14 26 210.000 726.2t 38' 275.000 1974.s3 b'

O 3g W

oy fy

,e,

'.. ~

4m.

-2 i

02 05 O I

$ !. 8 -

& *- if

= 2-ce o 2

- E i

• Q @ tR t

eeo l

lg' kb ou

[

y w we SD C

w w

eo i

i

h 1

Docket Nos.: 50 3384339 Serial No.

91 707 Sect'cn 5 Page 74 ef 80 4

i i

i 8

-o O

n!" 23:3223'334R AJ A.:ai i p ::n :al::ii.:A s.s.e.

_g U$g !$!$!$$$$$$$

a ll_I,.2.R E. R R.9.t R g S

• e g"

. ~ x.

I 22:2222:2X22 t

3 O

o hj:r:kEE.ETEE8EUN"E e.

d gg.... 2 ens.:.en :

e

,a O!E$l$$$$$Ell$$$

E l

2"! 6Wgeodgs..gggi;o a

9ts :. set.

m

.oo 3

8e

~

4

![

• • C**REN MX-g 8

9$

h: IIEEEEEE8EI.II U

ej:es,::s,s,ee.::,:gge S

g r..

+

yg-e

- g-

-=

,".22' w

og j :3 INkhkk$$!!kkk$$

25" *2288.:*:

t a

• R*

8 8 as its

• 02.c::

8 - - 2.

f 25

.nn,.....o.y:

A-8 VIRGINIA Et.ECTRIC AND POYltR COMPANY

==

css Z3:. eOb=#eHw#

n WUEE zo-e*koM

,O R,O W$T::i:w e

n PS eg 09

/

4 2

/

9 0

DEE 79888404S3O3 0 2 3 7 3 4 3 0.S.9.

0 TR3 AUI CSS 7319497818 4

ISP 8258148271 2

Wi8 6777888980u1 R

1t1 IP

)

E DR N

Eu3 000000000000 W

T t F.

000000000000 O

AA 000000000000 D

CRS S0508080608.70 L

IEE O

DPD S223384596 O

NM( 222222323222 C

IE T

RH

/

78901234667e 222333333333 F

G ED 0

6 D

365150s9031

(

EE t94926e310.04 TR3 a70925952o 39 AuT 02 css 4

IsP a901 34579ta57 DE(

e455555556c66 E

NR L

IP I

FO E

R DR P

EU) 0000000000000 N

T T F.

0000000000000 AA 0000000000000 e

CRG O

8EE 05050505050*0 2

D WPDM(

1111111111222 556677089900 L

W O

IE E

O S T

R.

C R A

9 R E 9

O T

4k67tG0123u56 F

1111t t2222 22 P

D F

E E

E D

?

I f

0 U

D 0 0

. G L

, G.

P D

2209531647 7

EE 5 9 7 7,.2 t 3 9 0 6.

4 64,. 0 E

1 TR)

E R

AUI R

E CSS t482727295203 W

• l' ISP 5112233445 77 S

- DN NR DE(

9444444444444 E

A V

T Ot IP

. R A

Ib U

D RO C

EA

E G P DR.

N N

=

EU) 0000000000000' N

W I

T T F.

0000000000000 O

W OM 00000000000 AA 5050505050 00 D

O II CRG L

L 7P IEE 05

. O L AE DPD 3990011223344

', O O ID NM(

11 1918t - 11 C

F D

IE Aw T

8 E Ra m G H Rl V

T IF t2345G7a90123 1115

>r l II 4 39z5 m Oe3d bo 1

s 2

+

L i'

I.

l

'iL

.l

,l Lf1 Il(

Dock:t Nos.: 50 3384339 S:rt:.1 Ns.

91 707 Section 5 Page 76 of 80 4

8 S

o o

g-2X888823238 S,E_d.is.sidd.[""id J.

snee l

i:B2 $.$.$.$.$.$.$.$.$.$.$.$-

c, ya l

_e D

f XERRRENEXR2R in 3

8

'?

a I _r : : x e g e. s..s.s e.x A-g it;$4,4$$$!.$44$Q 1

2s.u...e..,,su..lE A

.',t 3.t t.

e U

p$

ga

~'

.se*t!!2s R:n o i 8

5 a..

I:

S-

!,!:RaEnnaiesu 2

gra:::sgx,,, gag

=

r...

~ *e:

e e.,,,,

l

o -Sg ls-j et $

:B

=

g

-8 *g 9t.; $$$$$$$$$$$$$

Er 5-ggggg:s.6.ggg.

.. :20...

- sc 5,

.t..B. *,

-nn,..

. g;gg 9

VIRGINtA ELECTRIC dkWER COMPANY

i Dock:t Nos.: 50 3388339

~

Seri:1 No.

91 707 Secti:n 5 Pago 77 of 80 l

In additior, io the North Anna Unit 2 revised heatup and cooldown curves, the original 't H.t'inghouse Report WCAP-12503 documents the results of a study prearett b,y V!estinghouse for Virginia Power on the effects of accelerated cea'/ downs on heatup and cooldown limits.

Virginia Power has removed the N,ges which document these results as they are not being submitted for review arid approval at this time.

i 4

1 l

O l VIRGINIA ELECTRIC AND POWER COMPANY

Dock:1 Nos.: 50 338&339 S:ri:1 No.

91 + 707 Section S Page 78 of 80 i

l REFEriENCE Lists VIRGINIA ELECTRIC AND POWER COMPANY -

l l

Docht Nos.: 50 330&339 S:rlit No.

91 707 Boction 5 Page 79 of 80 REFERENCES (1)

Lottor from W. R. Cartwright to USNRC, " Virginia Electric and Power Company, North Anna Power Station Unit 1 Proposed Technical Specification Change," NRC Lottor Sorlai No. 88 202A, dated November 30,1988.

(2)

Lottor from W. L. Stewart to USNRC, " Virginia Electric and Power Company, North Anna Power Station Unit 1, Proposed Technical Specification Chango Supptomont," NRC Lottor Sorial No. 88 2028, dated June 19,1989.

(3)

Letter from USNRC to W. R. Cartwright,

• North Anna Unit 1, Issuance of Amendment Re: Heatup and Cooldown Curves (TAC No. 72060)," NRC Letter Sorial No. 88 499, dated June 30,1989.

(4)

S. E. Yanichko and L. Albertin: " Analysis of Capsulo U from the Virginia Electric and Power Company North Anna Unit 1 Renctor Vessel Radiation Survoillance Program," WCAP 11777, dated Fobruary,1990.

(5)

E. Terek and L. Albertin:" Analysis of Capsuto U from the Virginia Electrio and Power Company North Anna Unit 2 Roactor Vessel Radiation Surveillanco Program," WCAP 12497, dated January,1990.

(6)

Letter from W. L. Stewart to USNRC, "V:rginia Electric and Power

Company, North Anna Power Station Unit 2, Reactor Vossol Materials Surveillanco Program," NRC Letter Serial No._ 90 097, dated March 8,1990.

(7)

Letter from W. L. Stewart to USNRC, " Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Heatup and Cooldown Curves," NRC Letter Serial No. 91067, dated February 14,1991.

(8)

Lotter from W. L. Stewart to USNRC, " Virginia Electric and Power Con pany, North Anna Power Station Unit 1, Reactor Vossol Materials Surveillance Program," NRC Letter Sorial No. 88 202, dated June 1,1989.

(9)

A. D. Nana, et al.:" North Anna Unit 1 Pressure Temperaturo Limits for 12 ErPY and North Anna Unit 2 Pressure Temperaturo Limits for 12 and 15 EFPY," BAW 2146, dated October,1991.

(10)

N. K. Ray, et al.:" North Anna Unit 2 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation (Capsule U)," WCAP 12503, dated March,1990.

(11)

Code of Federal Regulations, Title 10 Part 50, Appendix G, " Energy; Domestic Licensing of Production and Utilization Facilities; Fracture Toughness Requirements," published January 1,1988 by the Office of the l

Federal Register, National Archivos and Records Administration, VIRGINIA ELECTRIC AND POWER COMPANY

f

- Dockst Nos.: 60 3381339 S:ri:1 No.

01 707 Section 5 Page 80 of 80 (12)

Updated Final Safety Analysis Report," North Anna Power Station, Units 1 and 2 Virginia Electric and Power Company.

(13)

• EPRI PWR Safety and Rollef Valve Test Program, Safety and Relief Valve Test Report," EPRI, NP 2628 SR, December,1982, t

(14)

' Safety and Rollef Valves in Ught Water Reactors,' EPRI, NP 4306 SR, December,1985.

(15)

" Reactor System Transient Analyses Using the RETRAN Computer Code,"

i VEP FRD 41, March,1981; as supplemented by letter from W. L. Stewart to USNRC, " Virginia Electric and Power Company, Surry and North Anna Power Stations Reactor System Translent Analysis" NRC Letter Serial No.

85 753, dated November 19,1985.

(16, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its impact on Plant Operations (Generlo Letter.8811)," U.S. Nuclear Regulatory Commission, July 12,1988.

3 (17)

C. C. Heinecke, et al.:

• North Anna Units 1 and 2 Reactor Vessel Fluence and RTPTS Evaluations," WCAP 11016. Rev. 3, dated January,1988.

r (18)

Code of Federal Regulations Title 10, Par 150.61, Federal Register Volurne 58, No. 94, dated Wednesday May 15,1991 (incorporation -of the Methodology of Regulatory Guide 1.99, Revision 2 into the 10 CFR 50.61 PTE 'ule).

(19)

Letter from J. H. Goldberg (Florida Power and Light) to USNRC, St. Lucie t

Unit 1 Docket No. 50 335, Proposed License Amendment, P T Limits and LTOP Analysis, dated December 5,1989.

-[

(20)

Letter from USNRC to J. H. Goldberg (Florida Power and Light), St. Lucie j

Unit 1 Issuance of Amendment Re: Pressure / Temperature (P/T) Limits and Low Temperature Overpressure Protection (LTOP) Analysis (TAC No.

75386), Docket No. 50 335, dated June 11,1990.

l l

f t

i VIRGINV ELECTRIC AND POWER COMPANY.--- -

.. _ _ _ _. _, _ ~.

_ _. _ _ -,. _.,.. _. _,. - _ ~ _ _ _, _

-.