ML20206M979

Requests Approval of Locked Rotor Analysis Submitted W/ 850502 Core Uprating Amend Request,Based on Clad Temp Acceptance Criterion.Safety Evaluation Using Acceptance Criterion in WCAP-11298 Submitted for Approval.Fee Paid

ML20206M979
Person / Time
Site:	North Anna
Issue date:	04/13/1987
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
87-150, NUDOCS 8704200328
Download: ML20206M979 (17)

Text

-_

~

.th .. ]

VIuGINIA ELucTunc AND I'owan COMPANY HicnxoNn,VINGINIA 20261 w.L.sTBWANT Vics Passinant Noctuam Oramarione April 13, 1987 United States Nuclear Regulatory Commission Serial No.87-150 Attention: Document Control Desk E&C/KLB:cdk Washington, D.C. 20555 Docket Nos. 50-338 50-339 i License Nos. NPF-4 l NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY h0RTH ANNA POWER STATION UNITS 1 AND 2 LOCKED ROTOR REANALYSIS l

The current North Anna licensing basis for locked rotor reports the predicted

~

percentage of fuel rods with DNBR less than the acceptance criterion and the associated off-site doses, assuming those rods fail. This analysis was submitted to the NRC (letter from W. L. Stewart to H. R. Denton, Serial No. 85-772A, February 6, 1986) in response to an NRC staff request for additional infonnation relating to our core uprating license amendment request.

This methodology differs from that employed in our original uprating submittal and the FSAR which was based on a clad temperature acceptance criterion. The current licensing basis imposes limitations on fuel rod power distribution which are proving difficult to meet in the design of uprated reload cores.

This letter requests your review and approval of the locked rotor analysis originally submitted in our core uprating amendment request (Serial No.85-077, May 2, 1985), which is based on a clad temperature acceptance criterion.

Westinghouse has recently submitted a topical report, WCAP-11298, for NRC review and approval which provides the basis for the clad temperature acceptance criterion. Therefore, the attached safety evaluation using the clad temperature acceptance criterion identified in WCAP-11298 for locked rotor is submitted for your review and approval.

These changes and the supporting documentation have been approved by the Station Nuclear Safety and Operating Committee and by the Safety Evaluation and Control Staff. It has been determined that this request does not pose any l unreviewed safety question as defined in 10 CFR 50.59 nor does it pose a significant hazards consideration as defined in 10 CFR 50.92.

\

f"A8SE"E8856 PDR O; \

Qee'<l hcIL $/3D'00

We have evaluated this request in accordance with the criteria in 10 CFR 170.12. A check in the amount of $150 is enclosed as an application fee.

b W. L. Stewart Attachments

1. Safety Evaluation

2. Application Fee

, . . - . _ _. - - - - , - - - . ., , , . _ _ . , . , _ . - - , , _ . , . . _ . . . , _ . . _ _ , , . ,,y .- - - -

cc: U. S. Nuclear Regulatory Comission 101 Marietta Street, N.W.

Suite 2900 Atlanta, GA 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector d

North Anna Power Station i

l 1

l 1

c k

a

, _ . . _ ~ _ . . , _ _ . _ _ . __ ..__,._ _ . -

a SAFETY EVALUATION - SINGLE REACTOR COOLANT PUMP LOCKED ROTOR

1.0 INTRODUCTION

The original North Anna FSAR analysis of the locked rotor accident incorporated a clad temperature acceptance criterion, i.e., fuel rods experiencing DNB do not fail. The original analysis included no offsite dose calculations. However, NRC staff would not approve the use of this methodology for the recently issued North Anna core uprating license amendment. A recent Westinghouse report (Reference 1) concludes that rods in DNB will not fail following a locked rotor accident. This evaluation is being submitted to request approval for referencing this report and using a clad temperature acceptance criterion for the North Anna locked rotor 4

accident.

As discussed below, Reference 2 submitted a locked rotor analysis for core uprated conditions which reported percentage of failed rods and offsite doses. This analysis constitutes the current locked rotor license basis. The Reference 2 analysis imposes limitations on fuel rod power distribution which are proving difficult to meet in the design of uprated reload cores. The approval of the original FSAR analysis methodology would relieve these limitations while ensuring that a demonstrated conservative methodology is used.

1.1 Identification of Causes and Accident Description The locked rotor incident is analyzed to demonstrate that the peak RCS pressure and clad temperature reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits and compromise the integrity of the primary coolant system.

The accident postulated is an insta'itaneous seizure of a reactor coolant pump rotor. Flow through the affected reactor coolant loop is rapidly reduced, leading to a reactor trip on a low-flow signal.

Following the reactor trip, heat stored in the fuel rods continues to be transferred to the coolant, causing the coolant to expand. At the same time, heat transfer to the secondary side of the steam generators is reduced, first because the reduced flow results in a decreased tube-side film coefficient, and then because the reactor coolant in the tubes cools down while the shell-side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the reactor coolant system. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system, opens the power-operated relief valves, and opens the pressurizer safety valves, in that sequence.

The two power-operated relief valves are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure-reducing effect, as well as the pressure-reducing effect of the spray, is not included in the analysis.

04-KLB-386R-1

2.0 ANALYSIS OF EFFECTS AND CONSEQUENCES 2.1 Method of Analysis The single reactor coolant pump locked rotor incident is analyzed in two parts. First, a peak pressure calculation is performed using j conservative assumptions thet tend to maximize the heat transfer from the fuel to the coolant. This calculation assumes that the fuel rods in the core do not experience departure from nucleate boiling.

Second, the calculation is repeated assuming the limiting fuel rod in the core experiences departure from nucleate boiling, and the fuel rod thermal transient is investigated with respect to peak clad i temperature and zirconium-water reaction. The discussion of the film-boiling coefficient, fuel clad gap coefficient, and zirconium-steam reaction applies only to the second calculation.

l Reference 1 provides a description of the reactor coolant pump l locked rotor event. In Reference 1, parameters calculated as a result of this event are related to mechanistic fuel failure phenomena.

Based on an evaluation of fuel rod test data relating Zircaloy clad temperature transients to fuel failures, no failures are predicted for Westinghouse PWR fuel rods when utilizing a 2700 F peak clad l temperature safety limit for the locked rotor event. Therefore, if l the clad temperature remains below 2700 F during the transient, the calculation of offsite radiological doses need not assume fuel rod failures.

The attached results are those originally contained in the core uprating license submittal of Reference 3. The Reference 3 analysis used the methodology described in the original North Anna FSAR and I consistent with Reference 1, which employed a clad temperature I

acceptance criterion without any explicit offsite dose calculations.

This assumes that fuel rods which experience DNB but remain below the 2700*F peak clad temperature safety limit do not fail. This methodology was used again in two subsequently approved 1icense submittals which contained locked rotor reanalyses (References 4, 5).

In Reference 6, the NRC rejected Virginia Power's use of this

! methodology for the core uprating. Reference 2 submitted an alternate l analysis in which the percentage of rods predicted to be in DNB was determined, along with the corresponding offsite doses.

Twodig)tal-computercodesareusedto analyze this transient.

The LOFTRAN code is used to calculate the resulting loop and core coolant flow following the pump seizure, the time of reactor trip based on the loop flow transients, the nuclear power following reactor trip, and to determine the peak pressure. Thethermalbehaviorofthg fuel located at the core hot spot is investigated using the FACTRAN l code, using the core flow and the nuclear power calculated by LOFTRAN.

The FACTRAN code includes the use of a film-boiling heat transfer coefficient. A constant +6 pcm/*F Moderator Temperature Coefficient j (MTC) was used. This is conservative, since Technical Specifications '

require MTC to be zero or negative when power is greater than 70 ,

l percent. I At the beginning of the postulated locked-rotor accident, i.e.,

at the time the shaft in one of the reactor coolant pumps is assumed to seize, the plant is assumed to be in operation under the most 04-KLB-386R-2

adverse steady-state operating conditions with respect to the margin to departure from nucleate boiling, i.e., maximum steady-state power level (plus an allowance'of 2% for uncertainty), minimum steady-state pressure, and maximum steady-state coolant average temperature.

Consequently, the initial power level for this incident was assumed to be 102% of nominal (2910 MWt), with three loops operating.

When the peak pressure is evaluated, the initial conservatively estimated as 30 psi above nominal pressure (pressure is 2250 psia) to allow for errors in the pressurizer pressure measurement and control channels. This is done to obtain the highest possible rise in the coolant pressure during the transient. To obtain the maximum pressure in the primary side, conservatively high-loop pressure drops are added to the calculated pressurizer pressure. The pressure response shown in Figure 1 is the response at the point in the reactor coolant system having the maximum pressure.

2.2 Evaluation of the Pressure Transient After pump seizure, the neutron flux is rapidly reduced by control rod insertion. Rod motion is assumed to begin 1 second after the flow in the affected loop reaches 87% of nominal flow. No credit

' is taken for the pressure-reducing effect of the pressurizer relief valves, pressurizer spray, steam dump, or controlled feedwater flow after plant trip. Although these operations are expected to occur, and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effect.

The pressurizer safety valves are actuated at 2575 psia; their capacity for steam relief is 39.2 ft3/second. Due to the very conservative method of analysis, the peak surge rate is high enough to cause the reactor coolant pressure to exceed the pressurizer safety valve actuation pressure. However, this condition exists only for a few seconds; consequently, the pressurizer water volume does not change significantly (less than 150 ft3). Therefore, the transient is not sensitive to the initial pressurizer level, and the programed value is used.

2.3 Evaluation of the Effect of Departure from Nucleate Boiling in the Core During the Accident For this accident, departure from nucleate boiling is assumed to occur in the core and, therefore, an evaluation of the consequences with respect to fuel rod thermal transients was performed. Results obtained from analysis of this hot-spot condition represent the upper limit, with respect to clad temperature and zirconium-water reaction.

In the evaluation, the rod power at the hot spot is conservatively assumed to be three times the average rod power (i.e., F0 = 3.0) at the initial core power level.

2.4 Film-Boiling Coefficient The film-boiling coefficient is calculated in the FACTRAN code using the Bishop-Sandberg-Tong film-boiling correlation. The fluid proper *.ies are evaluated at film temperature (average between wall and bulk temperatures). The program calculates the film coefficient at every time step based upon the actual heat transfer conditions at the 04-KLB-386R-3 L

. j time. The neutron flux, system pressure, bulk density, and mass flow rate as a function of time are used as program input, i

l For this analysis, the initial values of the pressure and the i

bulk density are used throughout the transient, since they are the

most conservative with respect to clad temperature response. For

conservatism, departure from nucleate boiling was assumed to start at .

the beginning of the accident. ,

2.5 Fuel-Clad Gap Coefficient i

l The magnitude and time dependence of the heat transfer coefficient between fuel and cladding (gap coefficient) has a 4 pronounced influence on the thermal results. The larger the value of the gap coefficient, the more heat is transferred between pellet and t

clad. Based on investigations into the effect of the gap coefficient

, upon the maximum clad temperature during the transient, the gap

' coefficient was assumed to increase from a steady-state value

consistent with the initial fuel temperature to 10,000 Btu /hr-ftr_op at the initiation of the transient. This assumption causes energy stored in the fuel to be released to the clad at the initiation of the transient, and maximizes the clad temperature during the transient.

2.6 Zirconium-Steam Reaction Yhe zirconium-steam reaction can become significant for clad temperatures above 1800'F. The Baker-Just ~ parabolic rate equation shown below is used to define the rate of the zirconium-steam j reaction:

d(wn ) = 33.3 x 106 exp -45,500  ;

dt 1.986T i

j where w = amount reacted, mg/cm2

)-

t = time, sec j T = temperature, *F i

j The reaction heat is 1510 cal /gm.

i 2.7 Locked-Rotor Results  ;

Transient values of reactor vessel flow coastdown, neutron flux, i and hot channel heat flux following a locked rotor are shown in i Figures 2 and 3 for three-loop operation.

Maximum reactor coolant system pressure, maximum clad temperature, and amount of zirconium-water reaction are contained in Table 1. Figure 4 shows the clad temperature for the transient.

i i

i l

l 04-KLB-386R-4 L -

- -.- . _ _ - - . = . ._ . _ . .-. -.-

3.0 CONCLUSION

S

1. Since the peak reactor coolant system pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system is not endangered.

2. Since the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2700 F and 1

' the amount of zirconium-water reaction is small, the core will remain

! in place and intact with no consequential loss of core cooling

capability or release of fission product inventory.

3. A positive moderator temperature coefficient does not adversely affect the consequences of a locked rotor at full power with three loops

operating. The integrity of the reactor coolant system is not j endangered as peak pressure during the transient is 2722 psia.

i i

i

}

i I

1, i

i l

i i

04-KLB-386R-5

(

1 .

10 CFR 50.92 No Significant Hazards Determination  ;

The reanalysis of the North Anna locked rotor accident does not involve a significant hazards consideration because operation of North Anna Units 1 and 2 in accordance with this new basis would not:

1. involve any significant increase in the probability or consequences of any accident previously evaluated. The reanalysis of the locked rotor event shows that fuel integrity will be maintained with no release of fission product inventory. Thus the predicted consequences of the locked rotor accident are less severe than the current licensing basis. The consequences of other UFSAR accidents remain unchanged.

The probability of occurrence remains unchanged since the reanalysis only involves a change in analysis methodology and does not impact system design or operation.

2. create the possibility of a new or different kind of accident from any accident previously identified. The new locked rotor basis does not affect any of the physical components in any of the plant systems and j therefore does not produce any new or unique accident precursors.

3. involve a significant reduction in a margin of safety. The locked rotor accident is classified as a Condition IV event in the North Anna UFSAR. Therefore, the limiting safety criterion is based upon the 10 CFR 100 values for offsite dose. Since the reanalysis shows that the fuel remains intact with no release of fission product inventory, the limiting safety criterion continues to remain bounding and the margin 4 of safety is not reduced.

04-KLB-386R-6

. . . . _ - . _ . .-.. - .~ . - _ . . _ . . - . . . - .

f 10 CFR 50.59 Safety Evaluation 1 The reanalysis of the North Anna locked rotor accident using the fuel failure criteria from Reference 1 has established a new analysis basis.

{ Design of future reload cycles to this new basis does not pose an unreviewed safety question as defined in 10 CFR 50.59. The basis for this ,

1 determination is as follows:

1. The probability of occurrence or the consequence of any accidents or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The reanalysis of the locked rotor event shows that fuel integrity will be maintained with no release of fission product inventory. Thus the predicted consequences of the locked rotor accident are less severe than the current licensing basis. The consequences of other UFSAR accidents remain unchanged. The probability of occurrence remains unchanged since the reanalysis only involves a change in analysis methodology and does not impact system design or operation.

2. The possibility for an accident or malfunction of- equipment of a different type than previously evaluated in the safety analysis report i is not created. The new locked rotor basis does not affect any of the 3

physical components in any of the plant systems and therefore does not ,

i produce any new or unique accident precursors.

i

! 3. The margin of safety as defined in the basis for any Technical Specification is not reduced by the proposed change. The locked rotor i accident is classified as a Condition IV event in the North Anna l UFSAR. Therefore, the limiting safety criterion is based upon the 10  ;

, CFR 100 values for offsite dose. Since the reanalysis shows that the i

fuel remains intact with no release of fission product inventory, the

limiting safety criterion continues to remain bounding and the margin l of safety is not reduced.

l l

l 04-KLB-386R-7 i

J .'*

4 i References

1. Skaritaka, J. et al., " Integrity of Fuel Rods During a Locked Rotor / Shaft Break Accident," WCAP-11298, October 1986.

1 1

2. Letter from W. L. Stewart (Virginia Electric and Power Company) to Parold R. Denton (NRC), Serial No. 85-772A, February 6, 1986, Response to Request for Additional Information on Core Uprate.

3. Letter from W. L. Stewart (Virginia Electric and Power Company) to Harold R. Denton (NRC), Serial No.85-077, May 2, 1985 Forwarding License Amendment Request For Rated Thermal Power of 2893 MWt.

ll 4

4. Letter from W. L. Stewart (Virginia Electric and Power Company) to Harold R. Denton (NRC), Serial No. 726, December 30, 1982, License Amendment Request for 587.8*F RCS Average Temperature.

] '

I 5. Letter from W. L. Stewart (Virginia Electric. and Power Company) to

) Harold R. Denton (NRC), Serial No. 666, February 7, 1985, License j Amendment Request For +6 pcm/*F Moderator Temperature Coefficient.

i I; 6. Letter from Edward J. Butcher (NRC) to W. L. Stewart (Virginia Electric and Power Company), Octcber 23, 1985, Request For Additional

Information - Core Uprate. .

7. Burnett, T. W. T., et al., LOFTRAN Code Description, WCAP-7907-P-A, j April 1984.  ;

i j 8. C. Hunin, FACTRAN, A FORTRAN IV Ccde For Thermal Transients In A

i U02 Fuel Rod, WCAP-7908, June 1972.

4 f

r l

I i

l l

i i

! 04-KLB-386R-8 I

i TABLE 1 i LOCKED ROTOR RESULTS 2910 MWt' Uprating Initial Power, percent 102 Moderator Temperature Coefficient, pcm/'F +6 Maximum Reactor Coolant Pressure, psia 2722 Maximum Clad Average Temperature. *F 2203 Amount of Zr-Water Reacted at Core Hot Spot, percent weight 1.1 i

I f

04-KLD-386R-9 ,

s

i .'..

TABLE 1 LOCKED ROTOR RESULTS 2910 MWt Uprating Initial Power, percent 102 Moderator Tenperature Coefficient, pcm/*F +6 i Maximum Reactor Coolant Pressure, psia 2722

Maximum Clad Average Temperature "F 2203 4

j Amount of Zr-Water Reacted at Core Hot

Spot, percent weight 1.1 l

l 4

I

.)

J I

I i

l I

i l l

I i

i k

l i

! 04-KLB-386R-7 l

2800

^

= -

- 2700 -

E g 2600 -

c I 2500 -

E 2400 -

g 2300 -

- I I I I I I I l 2200 I

O 1 2 3 4 5 6 7 3 g 10 I

l TIME (SECONDS) 1 I

i FIGURE 1

ALL LOOPS OPERATING. ONE LOCKED ROTOR 1 PRESSURE VERSUS TIME l

40-GLD-22190-10

~

3.2000  : 1 3.0000< -

1

' g 1

.80000< -

5

.60000< -

E e

.60000< -

.30000- -

9.0 0  ? ?

0 2 4 6 I 10 t

TIMI(SECONDS) i

! l 4

\ l I

l 3

i l

i 1

j FIGURE 2 t

ALL LOOPS OPERATING. ONE LOCKED ROTOR l

CORE FLOW VERSUS TIME i

a 40-GLD-22198-11

]

L_ _ . . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ -

r 1 I

l 1.2000; 4 4 0 0 1

i I 1.0000- -

2 5 .8000< * *

• k + Heat Flux ,

I

.6000< )

g - *

• b 2

w .4000- - *

+ Nuclear Flux

.2000< . *

• 0.0 : i 0 0 0 0 2 4 6 8 10  ;

TIME (seconds)

FIGURE 3 ALL LOOPS OPERATING, ONE LOCKED ROTOR FLUX TRANSIENTS 40-GLD-22190-12

( ,. -

l A #

. NdS <- T

(>

i 2000.O o

(>

f, 1750.0 " <>

l 3 1500.0 o <-

E g 1250.0 o (.

1000.00 - . >

$b0.00" . ,

g u a . -

TIMt 85tti FIGURE 4 ALL LOOPS OPERATING. ONE LOCKED ROTOR CLAD TEMPERATURE VERSUS TIME 40-GLO-2219B-13 m