ML20245C450

Revised Proposed Tech Specs Re Low Temp Overpressure Protection Enable Temp

ML20245C450
Person / Time
Site:	North Anna
Issue date:	06/19/1989
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20245C442	List: ML20245C440 ML20245C450
References
NUDOCS 8906260191
Download: ML20245C450 (39)

Text

..

D REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPER ATION 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 450 psig whenever any RCS cold leg temperature is less than or equal to 261 F, and 2) less than or equal to 390 psig l

whenever any RCS cold leg temperature is less than 150 F, or 1

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 l-than or equal to 261 F, except when the reactor vessel head is l removed.

ACTION:

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

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

c. In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted  ;

to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

I NORTH ANNA - UNIT 1 3/4 4-31 8906260191 890619 E PDR ADOCK 0500o338 P PNUt b-_---_-______---.---_--.--

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j REACTOR COOLANT SYSTEM BASES 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 I 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, q

)

The pressure-temperature limit lines shown on Figure 3.4.2 for inservice leak )

and hydrostatic testing have been'provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The minimum temperature for criticality specified in T.S. 3.1.1.5 assures compliance with the criticality limits of 10 CFR 50 Appendix G.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in the UFSAR and WCAP-11777 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. 1 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 with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent opening of greater than 2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 261 F. Either PORV has adequate relieving l 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 {

i 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 261 F. This ]l temperature is the water temperature corresponding to a metal temperature of at least j the limiting RTNDT + 90 F + instrument uncertainty. Above 261 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 88-11. j 1

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NORTH ANNA - UNIT 1 B 3/4 4-8 I i

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_ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ 1

(---- - - - . - . . - . . , . _ _

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1 ATTACHMENT 2 j SAFETY EVALUATION IN SUPPORT OF REVISED HEATUP AND C00LDOWN CURVES VALID TO 10 EFPY

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1 4

i Table of Contents .

List Of Tables . . . . . . . . . . . . . . . . . . . . . . . . 3 i

List Of Illustrations .................... 3 )

1

1.0 INTRODUCTION

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

SUMMARY

OF CAPSULE U ANALYSIS .............. 5 3.0 APPENDIX G CURVE DEVELOPMENT . . . . . . . . . . . . . . . 6 6

3.1 Heat Up Curve Analysis . . . . . . . . . . . . . . . . 7 3.2 Cool Down Curve Analysis . . . . . . . . . . . . . . . 8 3.3 Heat Up And Cooldown Curve Adjustment ........ 9 1

4.0 PORV SETPOINT DETERMINATION . . . . . . . . . . . . . . 11 )

i 4.1 Mass Addition Transient . . . . . . . . . . . . . . 12 l 1

4.2 Heat Up Transient ................13  ;

1 l

4.3 New LTOP Setpoints . . . . . . . . . . . . . . . . . 14 5.0 PTS Evaluation . . . . . . . . . . . . . . . . . . . . . 15

6.0 CONCLUSION

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

8.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . 27 1

Safety Evaluation Page 2  ;

o____-____.-.

O List of Tables 1 Heatup Curve Including Plant Specific Uncertainties . . 21 2' Cooldown Curve Including Plant Specific Uncertainties. 22 3 Cooldown Rates Assumed For Various Temperature Ranges. 23 4 Initial Conditions For The Mass Addition Transient . . . 23 5 Initial Conditions For The Heat Input Transient. . . . . 24 6 North Anna Unit 1 PORV Setpoints Technical Specification 3.4.9.3.a . . . . . . . . . . . . . . . . 24 List of Figures l l

1 Modification of Heatup Curve To Include Plant Specific Uncertainties................................ 25 2 Modification of Cooldown Curve To Include Plant Specific Uncertainties................................. 26 i

4 l

Safety Evaluation Page 3

1.0 INTRODUCTION

The 10CFR50 Appendix G analysis (Reference 3.) performed using the results from Capsule U, removed at the end of Cycle 6, indicates that the heatup and the cooldown curves for North Anna Unit I can be shifted allowing higher operating pressures relative to the current curves (T.S.

Figures 3.4.2 and 3.4.3, Reference 4.). The reason for this curve shift is that the Capsule U analysis shows the reactor vessel has been exposed to a lower fast neutron fluence than previously assumed. The current Technical Specification Curves are based on an assumed design basis neutron fluence through ten effective full power years (EFPY). The measured fast neutron (E >1.0 Mev) fluence data shows that capsule U was exposed to 8.28 x 10 18 n/cm' at 5.90 EFPY (Reference 2.). The measured fluences are significantly lower than the design values due to the

{

i implementation of low leakage fuel management at North Anna. Using the i measured fluence at 5.90 EFPY to estimate the fluence at 10.0 EFPY results in a lower fluence than used to generate the current Technical I

Specification curves. Regulatory Guide 1.99, Revision 2 allows the use j l

of surveillance data to adjust original design values after two or more

]

credible surveillance data sets become available (Reference 6.). Capsule U is the second capsule removed from North Anna Unit 1, Capsule V was removed after 1.13 EFPY.

The heatup and Cooldowri curves provide an upper pressure and

, temperature limit and Reactor Coolant Pump (RCP) operation limits the 1

lower operating pressure. The PORV low temperature overpressurization protection (LTOP) setpoints prevent inadvertent operation at pressures which would exceed the allowable Appendix G curve. This Safety Evaluation Safety Evaluation Page 4 l

4 Report was prepared to summarize the analyses performed to determine i revised heatup and cooldown curves, and LTOP setpoints.

Capsule U analysis is summarized in Section two. The revised heatup i l

and cooldown curve analyses are discussed in Section three. Section four discusses the revised PORV setpoints. A PTS evaluation is presented in I Section five. Finally, a 10 CFR 50.59 evaluation is presented to support the proposed Technical Specification changes. j i

2.0

SUMMARY

OF CAPSULE U ANALYSIS j i

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, E > 1.0 Mev, fluence of 8.28 x 1018 n/cm2 . The calculated, based on actual cycle power distributions, fast neutron fluence at the capsule location was 8.85 x 1018 n/cm" which compares favorably with the dosimeter fluence. The peak calculated fluence at the inside surface of the reactor vessel was calculated to be 8.83 x 10 18 n/cm2 which shows + fiat the capsule has been exposed to slightly more neutrons than the vessel (Reference 2.).

The material property testing included Charpy V-riotch 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-lb 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 Safety Evaluation Page 5

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Capsule U were obtained from the same girth weld and forging materials as that used in the Reactor Vessel beltline'(Reference 2.).

I The irradiated specimens test results were compared to unirradiatied specimen test results. The Charpy V-notch impact test results show the 4

irradiation has increased the average 50 ft-lb transition temperature by 80 to 110 'F. depending on the specimen metal. Irradiation has increased the average 30 ft-1b transition temperature by 65 to 300 'F. The upper I

shelf energy, average energy absorption at full shear, results show the l

worst decrease to be 25 ft-lb when comparing irradiated samples to unirradiated samples. The lowest average upper shelf energy was determined to be 92 ft-lb which is greater than the 10 CFR 50 Appendix G low limit of 50 ft-lb (Reference 7.). The Charpy impact test results from Capsule U were also satisfactorily compared to the Capsule V results.

I Tension test results show a slight increase in the ultimate tensile strength and the yield strength due to irradiation. Reference 2 should j be consulted for specific test results. l i

3.0 APPENDIX G CURVE DEVELOPMENT l

4 10 CFR 50 Appendix G, " Fracture Toughness Requirements," establishes  ;

pressure and temperature operational limits for the Reactor Vessel.

Virginia power utilizes two types of graphs to identify plant specific limits. The graphs are know as the Heatup and Cooldown curves. The present heatup curve depicts three curves, the leak test limit, the heatup limit (up to 60 'F/hr) and the criticality limit. The cool down crve depicts a series of curves for typical cooldown rates (0, 20, 40, 60, and 100 'F/hr). The range of the graphs is from the pressure limits h fety Evaluation Page 6

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corresponding to .80 'F to the' Pressurizer' Safety. Valve set' point, 2485

-l psig. The Control Room' 0perators use these Technical Specification graphs.

to ensure RCS pressure _and' temperature are within acceptable values.

3.1 Heat Up Curve Analysis <

l The analysis done to determine heatup curves includes the development' )

I of pressure-temperature relationships for steady-state conditions as'well l as finite heatup rate conditions asseming the presence of a 1/4T defect .

at the inside of the vessel wall. The thermal- gradients during heatup j produce compressive stresses' at the inside of the wall that alleviate the i

tensile stresses produced by internal pressure. Therefore, steady-state conditions can be limiting for the inside wall so both heatup and steady j state must be considered.  ;

i The heatup curve calculations must also~ consider the case of-a 1/4T flaw at the outside surface. The thermal and pressure stresses never' l cancel for this situation. The thermal stresses are dependent. on' both the rate of heatup and the coolant temperature along the heatup ramp.

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The use of a composite curve is required to make sure. that the l limiting condition is always protected.against. For. example protection  :

must be provided if the _ limiting location shifts. from the'inside to the. <

outside surface. Therefore, a composite 'heatup curve is generated by comparing, on.a point-by point basis, the steady-state curve at the inside of the wall- along with the .various heatup- rate .' curves at the. outside  !

surface. Th'us at any given temperature, the allowable pressure is taken.

Safety: Evaluation' Page 7-

f to 'be the lowest of the: values .from each of the curves under.

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consideration.

.i 3.2 Cooldown Curve Analysis During cooldown, the controlling.-location of the flaw'is. always'at' i

the inside . of the ' wall beca'use' the thermal . ' gradients' produce tensile ')

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stresses at the .inside, 'which increase with increasing cooldown rates.  !

Allowable pressure-temperature relations a re' generated for both steady-state and finite cooldown rate situations. A lower bound composite curve from the steady state and cooldown conditions is constructed for each cooldown rate of interest.

The use of a composite curve in the cooldown analysis.is necessary

]

because control of the cooldown procedure is based on measurement of-reactor coolant temperature,' whereas the limiting pressure. is actually dependent on the material temperature at the tip of the assumed flaw. i l

During cooldown the tip is at a higher temperature than the fluid adjacent j to.the vessel inner wall. -This condition,: of course, is not true for the ,

steady-state situation. It follows that at any given reactor coolant j i

temperature, the temperature gradient developed during cooldown results in a higher value.of KIR at:the,1/4T location for finite cooldown. rates i

than for steady-state operation. So, if the temperature changes'such that KIR increases faster than KIT the steady-state can be limiting.

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Safety Evaluation _ Page 8. l

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3.3 Heatup And Cooldown Curve Adjustment 1

Westinghouse performed the Capsule U analysis and provided Virginia j i

Power with heatup and cooldown curves for North Anna Unit 1 (Reference j l

3.). The Westinghouse curves did not include any instrument uncertainty. )

To ensure that instrument error does not allow the Reactor Vessel to be i adjusted I operated at undesirable pressure-temperature combinations, heatup and cooldown curves were generated. The curves were created by shifting the Westinghouse curves 20 'F to higher temperatures and 80 psi to lower pressures. The 20 'F and 80 psi represent the worst case error [

between the control room wide range temperature and wide range pressure indicators and the actual conditions which may exist at the Reactor Vessel Beltline. The plant specific uncertainties were determined using the same methodology used to determine the instrument uncertainties in the emergency operating procedures (EOP) but adjusting the results for normal operations.

l Figure 1 presents the revised heatup curve, with a maximum heatup 4

rate of 60 'F/hr. As indicated in the figure, the curve is based on an j extrapolated fluence to permit operation to 10 EFPY. The revised heatup l

curve does not contain the 10CFR50, Appendix G criticality limit. This  !

limit is not required since limit.ing condition for operation (LCO) 3.1.1.5 restricts the lowrst operating loop average temperature to ;t 541 'F for modes 1 and 2. This LCO defines a minimum temperature for criticality. I It provides substantially more margin to the heatup curve than the criticality limits required by 10CFR50, Appendix G. l Safety Evaluation Page 9

Table 1 is the same heatup curve data used in Figure 1. The data shown in Table 1 is the heatup curve adjusted for instrument errors associated with the Control Room indicators. The instrument error associated with the automatic operation of the PORV for low temperature overpressurization protection is less than the instrument error associated with the Control Room indicators. The Unit 1 cooldown curve was developed for a range of cooldown rates. Table 2 and Figure 2 show the Unit 1 cooldown curve data af ter incorporating the plant specific instrumentation uncertainties.

Comparing the heatup and cooldown curves in Tables I and 2, shows that the heatup curve is more limiting than the cooldown curve except for low temperatures. Less than 180 'F, the cooldown curves for 40 'F/hr and greater become more limiting. Previous discussion of a composite cooldown curve indicated the Reactor Vessel wall has a temperature gradient dependent on cooldown rate.

Due to this temperature gradient and since these higher cooldown rates are not realistically possible at low RCS temperatures, Table 3 was developed to present the assumed cooldown rate for various temperatures. Comparison of a composite cooldown curve, constructed from the information in Table 2 and Table 3, to the heatup curve shews the heatup curve to be the most limiting. The 60 'F/hr heatup curve will be treated as the representative Appendix G limiting curve for all further discussions concerning PORV setpoint determination.

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

corresponding to the Pressurizer Safety Valve setpoint of 2485 psig is 324 'F. .This point is used to bound all of the low temperature accident' analyses. Below 324 'F.the anticipated low temperature accidents may be adequately mitigated by the automatic action of the 90RV or by allowing sufficient time - for operator response. The mass addition transient assumes only one Charging pump will be operable below 324 'F. The heatup tr-isient assumes whenever a RCP is. started below 324 'F the temperature difference between the primary and' secondary fluids in the St'eam Generator is less than 50 'F.

4.0 PORV SETPOINT DETERMINATION

' 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 these Appendix G heatup and cooldown curves 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 b'etween the Steam Generator secondary fluid temperature and the RCS temperature. .These transients represent the limiting mass addition and heat input tran.sients and are analyzed with the RCS water solid. Only one PORV is required to operate during the transients.

Generic transient analysis was previously .used to determine LTOP setpoints which maintain acceptable pressure-temperature combinations on the Appendix G heatup and cooldown curves. The p~1 ant specific analysis-Safety Evaluation- Page 11

allows actual plant characteristics to be modelled rather than.the use o f. generic assumptions. The generic- assumptions have excessive conservatism to allow a wide range of application. These generic conservative assumptions become operationally burdensome when PORV lift setpoints are' too low to allow normal RCS operation without opening a'-

PORV. A plant specific North Anna two. loop RETRAN model was developed' to analyze possible'setpoints. Since generic analysis found that the' mass addition and the heatup transients'were equally limiting, it was necessary to analyze both transients to decermine how the North Anna model' would respond for these transients at different initial conditions. The .

.following sections describe the North Anna model development ' and . the analysis to determine new p0RV setpoints.

4.1 Mass' Addition Transient i The inadvertent startup of a single charging pump was selected as the design basis mass addition transient based on p'revious UFSAR work

.l l

)

(Reference 5. Section 5.2.2.?). Therefore, the LTOP .setpoints were i determined such that an overshoot allowance exists to prevent the Appendix G curves from being exceeded assuming an inadvertent charging pump startup i

during water solid operation. This overshoot allowance is required because of the valve opening characteristic associated . with the air operated relief valves used on the pressurizer at North Anna. (Reference l

~

10. and Reference 11.). -

I Inadvertent operation of a single Charging Pump was modeled assuming initial conditions as listed in Table 4. The initial RCS temperature and the PORV setpoint were varied to observe the effects of these' parameters. {

-1 Safety Evaluation Page 12  !

4 u-The RCS temperatures used were 100 and ,200' 'F. 'The resulting peak pressure was not significantly dependent on initial RCS temperature. The peak RCS pressure has also been found to be insensitive' to the initial-RCS pressure.

4.2 Heat Input Transient The heat input 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 to the Appendix G ar.ves. (Reference 5. Section 5.2.2.2).

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

the RETRAN model was used with a Westinghouse representation' of the PORV.

Finally, a 1. orth Anna specific model of the PORV ' was used in: the.

normalized RETRAN model. The most restrictive pressure overshoot was from the heatup cases initializing the RCS at high temperatures.

Safety Evaluation Page'13

k 4.3 New LTOP Setpoints The North Anna plant specific PORV.model_was verified and new LTOP' setpoints were determined. The new setpoints were established by.using the new' heatup' curve from Reference 3 adjusted for instrumen.t uncertainties for automatic PORV operation (15.68 'F and 67.2 psi).

Limiting condition mass addition and heatup transient RETRAN computer runs were made to validate the new setpoints. An effort was made' to optimize the PORV setpoints to values which limit the Reactor _ Vessel peak pressure to be less than the pressure allowed by the Appendix G curves and not be . operationally restrictive by having too much safety margin.

The new North Anna Unit 1 setpoints are s 450 psig when s 261 'F and 5 390 psig when s 150 'F. The mass addition transient was the most limiting transient for both the high and low setpoints.

Automatic low temperature overpressurizatiois protection is required whenever any RCS cold leg temperature is less than 261 'F. This-temperature is the RTNDT + AT + 90'F + instrument uncertainty. The RT NDT temperature is 136.3 'F for 1/4T and 116.1 'F for 3/4T (Reference

3. page 12). The AT is the maximum temperature difference between the water and the metal (i.e.15'F at the 1/4T and 32'F at the 3/4T). The L 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 to require automatic ,

protection during startup and shutdown. Above 261 'F administrative control is adequate protection because of Appendix G fracture criterion.

Safety Evaluation Page 14

The analysis has an increased margin at higher temperatures. In addition,

-1 operation of the RCS above 261 'F 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 (Reference 8.). 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 .j required to be enabled.

1 3

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

5.0 PTS EVALUATION Capsule U was located at the sixty five degree location, twenty five degree azimuthal angle, in the North Anna Unit I reactor vessel. The experimentally determined fluence was 8.28 x 1018n/cm' fast neutron fluence (E > 1 Mev). The calculated fast fluence at the capsule center, an ' azimuthal location of twenty five degrees, was 8.85 x 10 18 n/cm2 .

The close comparison between the calculated results and the actual (i.e.

experimental) results confirmed the analytical model which has been used -

to predict the fluence to the end = of the current operating license (Reference 2).

Safety Evaluation Page 15

.)

1 The V' irginia Power PTS submittal was based on calculated fluence.

' Since this fluence is slightly greater than the experimentally determined.

l- value and the material properties have not' changed, the'RTPTS Parameters i

. on record are slightly conservative (Reference 1.). :This conclusion is readily evident'by observation.of the equations which define the RTPTS l parameter: , )

l l

.]

0 .q RT PTS = I + M + (-10 + 47d(Cu)'+ 350(N1)(Cu)) f .270 l

RT PTS

= I + M + 283 f 0.194:

1

)

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i The only parameter in the above equations:which changed;as a result 1

of the capsule eval'u ation :is the fluence factor, f which -is reduced.

slightly (i.e. 0.885 to 0.828) at the 25' azimuthal location and the capsule center radially). . Based on this experimental data point the fast fluence factors are 6.88 % too high. Applying-these differences to the O' azimuthal angle, the effect upon the limiting RTPTS parameters can be determined based on the Reference-1. results. The Lower Shell forging 1

03 material was found to have the most limiting RTPTS j

RT PTS =38 +48 +(-10 + 470(.15) + 350(.80)(.15))(.883)0.270 a

= 185 'F )

'l, The RTPTS at license expiration is ' expected to be. 235 'F.' Thus,'the  ;

current RTPTS. parameter is not significantly different at the end of the  ;

l operating license. l 3

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6.0 CONCLUSION

S The heatup and cooldown curves required by Appendix G~of-10 CFR 50-have been extrapolated to 10 EFPY by . including the effects of the incremental radiation exposure on the reactor vessel beltline region.

The results are referenced to the analysis of Capsule U from North Anna Unit 1. The revised curves allow PORV setpoints higher than the: current setpoints for the LTOP system'to be implemented. -The revised Appendix G curves were prepared using standard Westinghouse methodology including R.G. 1.99 Rev. 2.

The new Heatup and Cooldown curves are valid until 10 EFPY of operation. The next reactor vessel surveillance capsule, Capsule X, is scheduled to be removed after the ninth fuel cycle which allows sufficient time for analysis prior to exceeding 10 EFPY.

The heatup and cooldown curves prepared -- by Westinghouse were determined.in a conventional manner according to Section III of the ASME code as required by 10 CFR 50 Appendix G. Both steady-state and transient-thermal conditions were considered in order to bound the possible comb: nations of ~ pressure ( i .. e . membrane) and thermal stresses. The Westinghouse curves were revised to include plant specific instrument loop uncertainties.

• The new North Anna Unit 1 Pressurizer PORV low temperature overpressurization lift settings should be less than or equal to 450 psig whenever any RCS cold leg ~ temperature is less than or equal to 261 'F,

' Safety Evaluation Page 17 i 1

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I and less than or equal to 390 psig whenever any RCS cold leg temperature is less than 150 'F.

PTS evaluations were made for the limiting beltline locations. The current value of the limiting RTPTS parameter was. updated using .-the calculated fluences . contained in . the' capsule report. No significant'.

change was obtained.

.I 7.0 10 CFR 50.59 EVALUATION' j 1

The proposed changes have been reviewed against the criteria of.10 CFR 50.59 resulting in the' conclusion that an unreviewed safety question does not exist. This determination was reached based on the following -

specific considerations: I

1. The probability or consequences of.any UFSAR event do not increase.

Accident probability is independent of the initial conditions j maintained by T.S. 3.4.9.1. This Specification deals with heatup and cooldown curves. The criticality limit.line has been removed from Figure 3.4.2. to eliminate confusion with the more restrictive -

criticality limit of T.S. 3.1.1.5. T.S. 3.4.9.3 has been changed to revise the PORV setpoints which are used to protect against violation of the Appendix G curves contained - in T.S. 3.4.9.'1. T.S. 3.1.2.2,.

3.1.2.4, 3.4.1.3, 3.5.2, 3.5.3 and 4.5.3.2'are required to ensure the low temperature accident assumptions are consistent with anaiysis assumptions. T.S. 3.4.9.1'.C has been deleted because ,the  !

Applicability Statement temperature change. makes this option- l obsolete. Finally the applicability of 3.4.9.3 has been modified to require automatic and passive overpressurization protection only.

during the the lower portion, below RTNDT+110'F,.of the Appendix G q heatup and cooldown curves. The upper portion'of the the Appendix G i curve limits will be ensured through administrative compliance with  ;

T.S. 3.4.9.1. ,

None of the above changes increase accident probability but rather 1 change the' initial conditions assumed in the safety analysis or are.

administrative in nature.

Accident consequences are not increased by the proposed Technical Specification changes. The proposed changes to T.S. 3.4.9.1 are required by Appendix G to maintain the prescribed margin to the reference stress intensity factor. The design criteria are met for Safety Evaluation Page 18

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J both the trass addition and heat input transients.which form the cold - j overpressure protection design basis. Therefore, the consequences i are not increased by these changes.  !

It is noted 'that plant specific instrument _ uncertainties' have been included in the revised curves which are more conservative than the previously used values. The change proposed for T.S. 3.4.9.3 reflects  ;

the peak pressure' overshoot from the two design transients. 1 i

Changes to the Technical Specifications Basis Sections 3/4.1.1.5,. j 3/4.1.2, 3/4.4.1, 3/4.4.9 and 3/4.5.3 - are necessary to maintain consistency . -with. the Limiting Condition .for Operation and j j Surveillance Requirements Sections. . Table B 3/4 4-1, Figures B i 3/4.4-1 and B 3/4.4-2 were deleted from -the Basis Section' and the text {

appropriately revised. The Table and Figures do not reflect . the j current methodology used to determine the adjusted reference  !

temperature, RTNDT. These Basis Section changes are administrative ,.i and do not increase the probability or consequences , of any UFSAR l event.

2. No new or different accident type is generated as a result of the revised heatup and cooldown curves, T.S. 3.4.9.1. These operational' curves provide restrictions on the pressure and temperature of the  !

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

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

the instrument loop. -;

i The changes to T.S. 3.1.2.2, 3.1.2.4, 3. 4.1. 3, 3. 5. 2, 3.5.3 and 4.5.3.2 are required to ensure the low temperature accident l assumptions are consistent with analysis assumptions. These require only a shift in an existing operational temperature limit and does not involve hardware modifications.

Changes .to the Technical Specifications Basis Sectior s 3/4.1.1.5,  !

3/4 1.2, - 3/4.4.1, 3/4.4.9 'and 3/4.5.3 are necessary to maintain  !

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

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3. The margin of safety is'not reduced. . The proposed changes to T.S. '

3.4.9.1 are required. by::10 CFR 50 Appendix..G on 'a periodic basis.

-The analyses. include margin in the- fluence calculations, the-reference temperature determination and the ? stress . calculations.

These margins. include =the safety factors' required by.the ASME code.

The criticality -limit .line has been : removed from Figure < 3 4.2 - to eliminate confusion with the 'more restrictive criticality limit of'

~T.S. 3.1.1.5.

The changes: to 'T.S. 3 .1'. 2. 2, 3.1.2.4, 3.4.1;3, 3.5.2, 3.5.3 and ,

,4.5.3.2. are required to ensure thei low y temperature: ' accident- )

assumptions are consistent with analysis assumptions. Plant specific  ;

design basis transients were analyzed and .the limiting pressure-overshoot was used to determine ' the . revised PORV setpoint. ' Plant specific instrumentation uncertainties. were . factored ' into the results. T.S. 3.4.9.1.C has' been _ deleted .because the Applicability: .

Statement temperature change makes this option obsolete. Therefore, )

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

Changes to the Technical Specifications Basis Sections 3/4.1.1.5, 3/4.1.2, 3/4.4.1, 3/4.4.9 .and 3/4.5.3 are necessary to maintain consistency with the Limiting Condition for Operation and Surveillance Requirements Sections. Table B 3/4 4-1, Figures B 3/4.4-1 and B 3/4.4-2 were deleted from the Basis.Section and the text-appropriately revised. The Table and Figures do not reflect the current methodology used to determine the adjusted reference ~ j temperature, RTNDT. These Basis'Section changes are administrative and do not reduce the~ margin of safety.

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

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Table 1: HEATUP CURVE INCLUDING PLANT SPECIFIC UNCERTAINTIES  ;

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Indicated Leak Test . Heat Up Rate Criticality -1' Temperature Limit _

To 60 SF/hr Limit

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

466.45 -----

110.0 ----

466.45 ----

120.0 ----- 466.45- --- ~

')

130.0 ----

474.01 ----

1 140.0 ----

488.56 ----

]

145.0 ---- ----

150.0 -----

508.98 '

160.0 ----

534.84- ---- .

i 170.0 ----

566.14 ----

180.0 ----

603.20 ----

190.0 ----

646.59 ----

)

200.0 ----

697.05 ----

)

210.0 ----

755.53 ----

220.0 ----

823.15- ----

l 230.0 ----

901.20 ----

(

240.0

.991.20 ---- I 250.0 ----

1094.88 ---- ,

260.0 ----

1214.17 ----

d 264.0 1920.0 ----

270.0 1351.30 ----

280.0 1508.69 ----

285.0 2405.0 1082.53 288.85 2494.0  ;

290.0 ----

1689.04- 1094.88 l 300.0 ----

1895.56- 1214.17 310.0 ----

2331.32 1351.30 f 320.0 ----

2399.14 1508.69.

323.54 ----

2494.0 330.0 ---- ----- -

1689.04 340.0 ---- -----

1895.56.

350.0 ---- -----

2131'.32 360.0 ---- -----

~2399.14 363.54 ---- -----

2494.0 i

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Safety Evaluation Page 21

i Table 2: C00LDOWN CURVE INCLUDING PLANT SPECIFIC UNCERTAINTIES Indicated Steady 20 'F/hr 40 'F/hr 60 'F/hr 100 'F/hr 3 Temperature ~ State Cooldown Cooldown Cooldown Cooldown l

('F) (psig) (psig)- (psig) (psig)- (psig) 105.0 500.77 468.95 436.70 403.99 337.13

.110.0 508.64 477.11 445.18 412.84. 346.82-120.0 526.19; 495.37 464.23 432.74 368.72 130.0 546.47 516.54 486.36 455.94 394.36 "

140 0- 569.90 541.06 .512.07. 482.94. 424.34 150.0 596.99 569.45- 541.90 514.34 459.35.

160.0 628.28 602.33 576.51 1550.84 500.17 170.0 664.43 640.38 616.63 593.22' 547.71 180.0 706.19- 684.41 663.12 642.40 603.05 190.0- 754.42 735.33 716.97 699.45 667.38 200.0 810.11 794.21 779.32 765.57 742.11 210.0 874.41 862.27 851.46 842.15 828.84 220.0 948.62 940.89 934.88 930.81 929.41 230.0 1034.23' 1031.68 1031.30 1033.35 240.0 1132.95 j 250.0 1246.70 260.0 1377.66 270.0 1528.27 280.0 1701.21 ,

290.0 1899.71 300.0 2126.88 310.0 2385.61 314.19 2494.00 Safety Evaluation Page 22

8 Table 3: C00LDOWN RATES ASSUMED.FOR.VARIOUS TEMPERATURE RANGES Temperature Cooldown-Range Rate

'F 'F/hr Above 185- 100 From 185.to 155 60 From 155 to 125 40 From 125 to 110 20

'Below 110 0 i

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Table 4: INITIAL CONDITIONS FOR THE MASS ADDITION TRANSIENT

.l Reactor Coolant Temperature 100/200 'F Reactor Coolant Pressure 50 psig- .

Maximum Charging Pump Flowrate 770 gpm Pressurizer Steam Volume 0 ft* ,

Pressurizer Water Volume 1400 ft'. y Reactor Coolant 5'ystem Flow 10 %

PORV Open Setpoint. Variable PORV Closed Setpoint Open - 15 psi Safety Evaluation Page 123

Table 5: INITIAL CONDITIONS FOR THE HEAT INPUT. TRANSIENT Reactor' Coolant Temperature 100/140/180 'F Reactor Coolant Pressure' 50 psig

'RCS/SG AT- 50 'F Pressurizer Steam' Volume 0 ft' Pressurizer Water Volume 1400'ft' RCP Speeds In Affected Loop,.startup 10 to 100 %

In Unaffected Loop, coastdown 10 to 0 %

PORV Open Setpoint Variable PORV Close Setpoint- Open .15 1

Table 6: NORTH' ANNA UNIT 1 PORV SETPOINTS TECHNICAL SPECIFICATION 3.4.9.3.a.

Current New Setpoints Setpoints. ,

s 375 F- 5 261 F s 420 psig s 450 psig

< 185 F < 150 F s 350 psig s 390 psig I

Safety Evaluation Page.24

Material Propery Basis

, Controlling Material: Circumferential Weld Copper Content: 0.086 WT% <

Nickel Content: 0.glWT%

Initial RT NDT: 19 F  ;

RT NDT 1/4T, 136.3 gF 3/4T, 116.1 F I

i Curves Applicable For Service Periods Up To 10 EFPY And Contain Margins Of 20 0F And 80 psi For Possible Instrument Errors

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Figure 1: Modification Of Heatup Curve To Include Plant Specific Uncertainties Safety Evaluation. Page 25

., Material ~ Property Basis

' Controlling Material: Circumferential Wald

,. Copper Content: 0.086 WT%

Nickel Content: 0.11 WT% I Initial-RT NDT; 19 F RT NDT' 1/4T,136.3p 3/4T, 116.1 F Curves Applicable For Service Periods Up To 10 EFPY And Contain Margins Of 20 0 F' And 80 psi' For Possible Instrument Errors 1,In .

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Figure 2: Modification Of Cooldown Curve To Include Plant Specific Uncertainties l I

Safety Evaluation Page 26

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8.0 REFERENCES

1

)

1. North Anna Units 1 Reactor Vessel Fluence and RTPTS. Evaluations,"

WCAP-11016 Revision 3, Heinecke, C.C., et. al. January 1988.

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

WCAP-11777. S. E Yanichko, L. Albertin, E. P. Lippincott, February j

1988.

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

Normal Operation."'WCAP-11791. J. C. Schmertz. May 1988.

i

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

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

1 Units 1 & 2, Virginia Electric and Power Company. -!

6. " Radiation Embrittlement Of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, May 1988.

7. Code of Federal Regulations; Title 10, " Energy;" Part 50, " Domestic Licensing of Production and Utilization. Facilities;" Appendix G, .;

" Fracture Toughness." Published Janauary 1,_1988 by the Office 'of '

the Federal Register National Archives and Records Administration, j

8. "NRC Position On Radiation Embrittlement Of Reactor Vessel Materials And Its Impact. On Plant Operations (Generic Letter .88-11)," U.S.

Nuclear Regulatory Commission, July 12, 1988. ,

9. " Pressure-Temperature Limits,"-Section 5.3.2 US NRC Standard Review ,

Plan (NUREG-75/087), Revision 0, November 24, 1975. I

~

10. "EPRI PWR Safety and Relief Valve : Test . Program, -Safety and Relief  ;

. Valve Test Report," EPRI, NP-2628-SR,' December 1982

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

December 1985. j i

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l ATTACHMENT 3 RESPONSE TO NRC REQUEST FOR INFORMATION ON THE NAPS l UNIT 1 HEATUP AND C00LDOWN PACKAGE .j l

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Table of' Contents List of Illustrations ....................... 3  :

1.0 INTRODUCTION

......................... 4 2.0 INFORMATION ON THE MASS ADDITION TRANSIENT ...........4 .j 3.0. INFORMATION ON THE USE OF RETRAN-02 ............. 6

4.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . 7-  !

5.0 REFERENCES

.......................... 10 SETR719 Page 2

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1 List of Illustrations Figure 1. Mass Addition Transient Initial Temperature Dependence .8 )

2. Low Head Safety Injection' Pump Sensitivity Study ...

Figure .9 1

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1.0 INTRODUCTION

l Virginia Electric and Power Company (Virginia Power) has submitted a 1 l request for a licensing amendment to revise the heatup and cooldown curves i

! contained in the North Anna Unit 1 Technical Specifications . The curves 2 i contained in this amendment request are less restrictive than the current  !

curves so approval is required prior to implementation. Virginia Power l has verbally requested approval of the amendment before restart of Unit j'

1. The NRC has responded promptly by assigning a reviewer who has verbally requested further information. The purpose of this report is to provide a response to that request. j 2.0 INFORMATION ON THE MASS ADDITION TRANSIENT l I

Question l The reviewer asked why the mass addition transient was only analyzed at 100'F and 200'F? Additionally, why was the analysis not performed for j temperatures as high as 324'F which is the full pressurization l temperature? )

l Answer i l

The safety evaluation states that the analysis is not sensitive to initial temperature. This conclusion was reached after doing the analysis at both j initial temperatures and finding that the peak pressure was not '

significantly different. Two pressure transients were run with a PORV I lift setpoint of 365 psia. One had an initial temperature of 100'F and the other had an initial temperature of 200'F.~ The overshoot past the setpoint was 109.1 psi for the first case and 107.3 psi for the second case or a difference of about 2 psi. Similarly, two cases were run with c lift setpoint of 435 psia. One had an initial temperature of 100'F and the other had an initial temperature of 200'F. The overshoot for the first case was 96.6 psi and for the second case it was 99.4 psi or a difference of about 3 psi. Pressure plots are provided in Figure 1 for the lower lift setpoint cases.

It was not necessary to perform analyses at temperatures greater than '

200'F for three reasons. First, the work described above showed that initial temperature was not a significant variable. Second, the heatup and cooldown curves begin to increase sharply at temperatures above 200'F. So, any increase in overshoot due to different thermodynamics is offset by the increased space between the setpoint, determined to prevent reaching the Appendix G curve at a lower temperature, and the Appendix G curve. So, the increasing derivative of the Appendix G curve allows more overshoot at temperatures above about 200'F. Finally, credit was taken for Revision I to Branch Technical Position RSB 5-22 which requires that the low temperature overpressure protection system be enabled below a temperature defined as the water temperature corresponding to a metal temperature of at least RTt4DT+ 90'F at the beltline location (1/4t or 3/4t) that is controlling in the Appendix G limit calculations SETR719 Page 4

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. Instrument uncertainty was. included in this calculation. ';

1 Question .

i The mass addition transient was initiatsd from' a steady' state condition j at 50 psig. Why did the analysis exclude the . inadvertent startup. of a  ;

low head safety injection pump in . addition to the charging. pump? It  ;

should be noted that during MODE 4 operation.T.S.3 3.5.3 permits both aL l

low head . safety injection. pump and a centrifugal charging pump to - be i operable.  !

i Answer The mass addition transient is initiated from a steady ' state condition at 50 psig. Once the transient is initiated the pressure ' rises rapidly because of the limited compressibilitylof water. .Therefore"the. shutoff head of the low head injection system would be reached in such a.short time that the contribution of one low' head pump would be negligible. j The design basis for the low temperature overpressure protection system-

~

mass addition transient is the startup of a single L safety ' injection pump'. The selection of the safety injection pump performance curve to use in the generic analysis was based on a consideration of only high head; '

1 and intermediate head pumps as indicated in the following quote from the above reference: i

.I From inspection of FigLre 2.3.2, it is evident .that the system represented by Curve C is the worst case in that'the system delivery into the reactor coolant system is the greatest of all the systems shown over the reactor coolant range of 400 to.600 psig, the range  !

of interest for the transient analyses. Therefore, the system j delivery described by Curve C was used in the study and is referred to as the referense safety injection pump startup case.

Curve C is from an intermediate head SI pump which is not used in the ,

North Anna design. Hence in the plant specific analysis for North Anna  ;

only the centrifugal charging pump was considered.  !

i In order to verify the correctness of . ignoring the low head pump in the design basis generic analysis and in the North Anna plant . specific  !

analysis it was decided to include.the low head pump in' the RETRAN model j as a sensitivity study. The RETRAN model was changed to include the mass  ;

addition from the low head pump along with the charging pump. Figure 2 1 shows the pressure plot for the base case and the case with the low head contribution included. The results of the analysis indicate that the low -i head pump shutoff head is reached at 1.4 seconds into' the transient. The  ;

resultant pressure is about 3.2 psi higher than with the charging pump i alone and the valve opens 1.2 seconds earlier. The results show that even- l with the low head contribution the peak pressure is below the limiting :i Appendix G curve by about 6 psi. Therefore the impact of the low head pump is negligible.  !

SETR719 Page 54 l

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3.0 INFORMATION ON THE USE OF RETRAN Question j The reviewer requested information to. support the use .of the version of. j the RETRAN Code identified as RETRAN02/ Modo 3 since the .SER* issued to j Virginia Power for its RETRAN mcdels identified RETRAN01/ Modo 3. l Answer On. September 4,1984, the NRC issued an SER which' addressed the use of

.both RETRAN01/ Modo 3 and 'RETRAN02/ Modo 2 to the Utility Group for ,

Regulatory Action (UGRA)5 The Virginia Power topical report *, which was y based on RETRAN01/ Modo 3, was submitted to the NRC on April' 14, 1981. .The 1 NRC issued an SER for this report on September 4,1984. .During the review of this topical report, Virginia - Power continued to update . the RETRAN' models being used in-house. This consisted mainly of ' converting the >

RETRAN models from RETRAN01 to RETRAN02. RETRAN02 does have a' number of, enhanced models that are not in RETRAN01 that- address the use of-one-dimensional kinetics, dynamic. slip, vector momentum, steam separators, and an auxiliary neutron void-fraction model for ' analyses . ,

mainly of BWR transients. In addition, a revised pressurizer solution,  !

a local conditions heat transfer model and other modifications are  !

provided to support ATWS. (Anticipated Transients Without Scram) I calculations. These changes extend RETRAN's ability to- analyze  :

additional BWR transients, small break LOCAs and ATWS transients as well  !

as providing better balance,of plant modeling. I The NRC SER required that errors noted during the review of the UGRA i RETRAN02 submittal be corrected in ' the approved versions of the code.

These corrections were made and included in the release of RETRAN02/ Modo 3.

l Virginia Power's RETRAN models were upgraded and checked with that code ,

l version to demonstrate that this version of the code ~ produced results l

consistent with the changes made to the code and with past transient ,

results when the saine model options were used. '

i Virginia Power's desire to have its RETRAN topical extended to RETRAN 02 l was discussed with the NRC reviewers of -VEP-FRD-41 as well as the NRC i Project Manager for the review prior to the approval of VEP-FRD .41. Based on those discussions, Virginia Power subsequently submitted an additional informational package on November 19, 1985', that provided' comparisons  ;

between RETRAN 01 and RETRAN 02 for.a series of transients. This package  ;

demonstrated that the two versions of the code produced results that are '

very nearly identical for the Virginia Power analytical models except'for. '

the changes caused by the ncnequ,ilibrium pressurizer model in RETRAN 02.  ;

It was noted in this package that there were several additional licensing '

[ analyses being prepared for Virginia Power's nuclear plants .that were I based on'RETRAN 02 and that they would be submitted to the NRC soon after this package was provided to the NRC. In that letter, the Company requested NRC approval of use of RETRAN 02 by February 1986 in ' order to 1 support those upcoming submittals. Those submittals were madeL and-subsequently approved by the NRC.

The model used for this submittal was based on the RETRAN model;in the Virginia Power Topical Report VEP-FRD-41-A. This analysis requires that SETR719 Page 6  ;

I e-the pressurizer be assumed . completely' filled at. the .beginning of' the 'j transient; Since the impact of the transient is to force subcooled fluid: "

from the rest of the ~ RCS into the pressurizer, the non equilibrium pressurizer models' have no impact on the solution. for this, transient.

Thus, _ the physical ' representation of the RCS presented . in VEP-FRD-41-A is not affected by this transient. Changes to the model did include' ,

transient specific items such as matching the initial conditions to match the lower initial RCS pressures and temperatures and modeling of the PORV actuation for this mode'of operation.

4.0 CONCLUSION

S-  !

1 Information concerning the mass addition transient' has been prepared for ~ -q presentation to NRC for their use.in the review of the North Anna, Unit i 1 heatup and cooldown curve amendment request. A' discussion of the reason j for doing the. mass addition transient. at only 100'F and 200'F was provided - ]

to show that 'there is little. dependence on initial RCS temperature .and i that the low temperature region is most limiting. A sensitivity study I was' performed .on the mass addition transient. by including one low head safety injection pump, .in addition to the charging pump, in a RETRAN analysis. The results of the sensitivity study show that the pump rises to the shutoff head quickly and results in a peak pressure only slightly-higher than that with only the charging pump The peak pressure is lower )

than the most restrictive Appendix G pressure limit. Additionally, a j discussion of the differences between RETRAN01 and RETRAN02 .for this application was presented.

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5.0 REFERENCES

l

l. '1. Letter from W.R. Cartwright (VEPCo) To USNRC, " Proposed Technical L Specifications Change," Serial No. 88-202A, November 30,1988. l

2. Branch Technical Position RSB 5-2,. "0verpressurization Protection Of- l Pressurized Water Reactors While Operating At Low . Temperatures," l Revision 1, 1988, Attached To SRP 5.2.2, " Overpressure Protection,", j Revision 2, 1988.

3. " Westinghouse Report On RCS Solid Water Overpressurization," July 20,. l 1977

4. Letter from'C.' O. Thomas'(NRC) to W. L. Stewart'(VEPCo), " Acceptance for Referencing of Licensing Topical Report VEP-FRD-41,

'VEPC0 Reactor System' Transient Analysis .Using RETRAN Computer Code'", April-11, 1985. 1

-i

5. Letter from C. O. Thomas-(NRC) to T. W. Schnatz (UGRA), " Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5,. 'RETRAN-A i Program for One Dimensional Transient Thermal Hydraulic Analysis of 2 Chaplex Fluid Flow Systems,' and EPRI NP-1850-CCM, RETRAN-02-A-Pt ogram for Transient Thermal-Hydraulic Analysis of Complex Fluid F ow Systems'" Systems'", September'4,-1984.

C. dReactor System Transient Analyses Using the RETRAN Computer Code,"

VEP-FRD-41, March, 1981.

7. - Letter from W. L. Stewart (VEPCo) to H. R. Denton (NRC), " Reactor System Transient Analyses", Serial No.85-753, November 19, 1985.

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