ML17279A916

Washington Nuclear Plant 2 Single-Loop Operation Analysis.

ML17279A916
Person / Time
Site:	Columbia
Issue date:	09/30/1987
From:	Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
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ML17279A913	List: ML17279A912 ML17279A916 ML17279A917 ML17279A918 ML17279A919
References
ANF-87-119, NUDOCS 8803300291
Download: ML17279A916 (35)

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ADVANCEDNUCI.EAR FUELS CORPORATION ANf-87-119 Issue Date: 9/ll/87 WNP-2 SINGLE LOOP OPERATION ANALYSIS Prepared By:

J. E. Krajicek BWR Safety Analysis Licensing and Safety Engineering fuel Engineering and Technical Services AFFII.IAIDO AH OF KRAFTWERK VHIOH Q~KiVU

CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con.

cerning the subject matter of this document are those set forth in the Agreement Advanced Nuclear Fuels Corporation and the Customer pursuant to 'etween which this document Is issued. Accordingly, except as otherwise expressly pro-vided In such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor--

mation contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, ap-paratus. method or process disclosed in this document.

The Information contained herein Is for the sole use of Customer.

In order to avoid Impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (ln the patent use of the term) of such Information until so authorized In writing by Advanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless. otherwise expressly provided ln the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this docu-ment.

XN NF.F00.765 (1/87)

ANF-87-119 TABLE OF CONTENTS Section Pacae

1.0 INTRODUCTION

...'.........,.......... ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 2.0

SUMMARY

........... ~ t ~ 2 3.0 TRANSIENT ANALYSES............... ~ ~ ~ ~ ~ ~ ~ ~

3.1 Analysis Bases....................... ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 3.2 Load Rejection Without Bypass (LRWB). ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ 4 3.3 Feedwater Controller Failure (FWCF)..

3.4 Recirculation Pump Trip.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3.5 Recirculation Flow Runup............, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 4.0 PUMP SEIZURE ACCIDENT................ ~ ~ ~ ~ ~ ~ ~ ~ ~ 15 5 0 SAfETY LIMIT............ 18 6.0 STABILITY ANALYSIS....... 20

7.0 REFERENCES

21

ANF-87-119 LIST OF TABLES Table Pacae

2. 1 Summary Of SLO Tr'ansient Analyses................ ~ ~ ~ ~ ~ ~ 3

3. 1 Analysis Conditions For Single Loop Operation .. 7 3.2 Results of Single Loop Operation Plant Transient Analyses ... ~ ~ ~ ~ ~ ~ ~ 8

5. 1 Safety Limit Evaluation Uncertainties........................ ~ ~ ~ ~ ~ ~ ~ 19 LIST OF FIGURES Ficiure Pacae

3. 1 Load Rejection Without Bypass Results , RPT Operable, Normal Scram Speed,.... ~ ~ ~ 9 3.2 Load Rejection Without Bypass Results , RPT Operable, Normal Scram Speed............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ \ ~ \ 10 3.3 Feedwater Controller Failure Results, RPT Operable, Normal Scram Speed....

3.4 Feedwater Controller Failure Results, RPT Operable, Normal Scram Speed............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ a 12 3.5 Recirculation Pump Trip Results...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 13 3.6 Recirculation Pum'p Trip Results...... 14

4. 1 Pump Seizure Accident................ 16 4.2 Pump Seizure Accident. 17

ill ANF-87-119 ACKNOWLEDGMENT Advanced Nuclear Fuels Corporation appreciates the major contribution to the WNP-2 single loop operation analysis made by J. C. Rawlings of ENSA, Inc.

Yi" 0

ANF-87-119

1.0 INTRODUCTION

1 Advanced Nuclear Fuels Corporation (ANF) has performed postulated transient and accident analyses for the Supply System Nuclear Project Number 2 (WNP-2) reactor with a recirculation pump (or loop) out of service. The purpose of the analyses is to demonstrate that two loop operational limits provide protection for the maximum power single loop operation (SLO) condition.

The analyses considered the following events:

o Load Rejection Without Bypass (LRWB) o Feedwater Controller Failure (FWCF) o Pump Trip o Recirculation Flow Runup o Pump Seizure Accident The ECCS analysis for the SLO condition is reported in Reference 1. The conclusions of these analyses are applicable to future fuel cycles containing ANF/NSSS vendor fuels of the current 8x8 design.

I W4" ANF-87-119 2.0

SUMMARY

The most limiting transient events and the pump seizure accident have been analyzed for the maximum expected power state during single loop operation (SLO) of WNP-2. The analyses were performed using current ANF transient analysis methodology for a core configuration representative of Cycle 3.

The results of the SLO transient analyses are summarized in Table 2. 1. The two loop MCPR operating limits (rated conditions) bound the requirements for single loop operation. Therefore, the single loop transient analyses need not be performed on a cycle by cycle basis and the two loop HCPR operating limits applicable for a cycle are appropriate for single loop conditions.

Operation in the single loop mode results in higher uncertainties for core flow, radial power and axial power. Considering these SLO uncertainties, the

~

HCPR safety limit was determined to increase by 0.01 to 1.07.

~ ~ ~ However, the two loop HCPR limits at SLO flow conditions bound the required MCPR limits for SLO conditions including the higher safety limit HCPR.

A postulated pump seizure accident was evaluated for SLO conditions. The event is less severe than the design basis loss of coolant accident (LOCA).

The radiological consequences of this accident are well within the 10 CFR 100 N

limits.

ANF-87-11 TABLE 2. 1

SUMMARY

OF SLO TRANSIENT ANALYSES Required SLO MCPR Limit 2 Loo MCPR Limits Event GE ANF GE ANF LRWB 1.24 1.21 1.32 1.30 FWCF 1.09 1.09 1.32 1.30 Pump Trip 1.09 1.07 1.32 1.30

ANF-87-119 3.0 TRANSIENT ANALYSES 3.1 Anal sis Bases The WNP-2 single loop transient analyses were performed using ANF methodology (Reference 2) consistent with that applied to the normal reload analyses.

Reference 3 requested that the analyses support plant operation to 75% of rated thermal power. Normally the transient analyses are performed at 104 '%

of rated core thermal power which corresponds to the 105% steam flow condition. For consistency, these analyses were performed at 104,2% of the 75% power state or 2596.9 Hwt. The core flow for analysis purposes was assumed to be 54.0 Hlbm/hr. The steam flow at this power level was taken to be 10.79 Mlbm/hr. A conservative dome pressure of 1020 psia was used in the analyses.

The system conditions at which the SLO transients were evaluated are summarized in Table 3. 1. . The steam flow/feedwater enthalpy characteristic from Reference 4 was used to initialize the plant transient simulation code COTRANSA (Reference 5). A jet pump H-ratio of 3.2 was used for initialization of the COTRANSA model at 54.0 Hlbm/hr core flow. In addition, the following assumptions are made for all analyses performed:

Normal scram speed (NSS).

2. Technical specification scram delay.

3. Integral power multiplier of 110%(5).

3.2 Load Re 'ection Without B ass A t cycle exposures equal to or greater than EOC -2000 HWd/HTU, and rated or increased (106% of rated) core flows, the pressurization transient events such as the load rejection without bypass (LRWB) and the feedwater controller failure (FWCF) usually establish the HCPR operating limits. However, at the reduced power and flow conditions for SLO relative to rated conditions, there

ANF-87-11 is a reduction in the steam flow to the turbine. With the lower steam flow, the pressurization of the reactor vessel is reduced in comparison to rated condit'ions -when the turbine control valve is closed following the generator load rejection signal. The resulting power excursion and associated thermal margin reduction are less than that for 'the full power case.

Figures 3. 1 and 3.2 present the time variance of critical reactor and plant parameters from the analysis of the load rejection without bypass transient at the SLO reduced power and flow c'ondition. The analysis assumes normal. scram speed and recirculation pump trip (RPT). The delta CPR and other peak conditions during the event are shown in Table 3.2. The appropriate delta CPR results from the WNP-2, Cycle 3 plant transient analysis (Reference 6) are also provided for comparison.

3.3 Feedwater Controller Failure The other limiting pressur'ization event is the feedwater controller failure (FWCF) to maximum demand. This event results in the maximum amount of subcooled feedwater being introduced into the vessel causing a core power increase followed by a high water level isolation signal and a turbine trip.

Because of the reduced core and recirculation flows at the SLO condition relative to rated conditions, the increased subcooling due to the high feedwater flow takes longer to reach the core and the high water level trip occurs earlier, limiting the power rise prior to the turbine trip.

As with the LRWB event, the feedwater controller failure at the SLO power and flow is less severe than that for the full power and flow condition. The time variation of the pertinent reactor and plant parameters are shown in Figures 3.3 and 3.4. The results are tabulated in Table 3.2.

i

ANF-87-119 3.4 Recircul ation Pum Tri The recirculation pump trip transient is modelled to create the most'apid decrease in pump speed. Figures 3.5 and 3.6 illustrate the time variation of the plant parameters for this event. The peak conditions are tabulated in Table 3.2. As expected, this event is less severe than all the other transients and is bounded by the Cycle 3 reload analysis.

3.5 Recirculation Flow Runu In the single loop configuration, the additional constraint of the reduced flow HCPR operating limits is no longer required. The 2 pump flow runup which would encroach upon the HCPR safety limit is not possible as the pump in the idle loop is not running. An inadvertent start of the idle pump cannot affect flow appreciably as the pump is interlocked to prevent starting unless it' associated flow control valve is at the minimum position.

Operation in single loop is preclude'd above the 80% rod line when the total core flow is less than 39% of rated core flow. This limits the most severe pump runup to a flow increase from 39% to approximately 50% core flow. This flow increase is not of sufficient magnitude to violate the HCPR safety'limit if the transient initiates from the two loop HCPR operating limits. This can be seen from the WNP-2, Cycle 3 plant transient analysis (Reference 6, Figure 5.1).

ANF-87-11 TABLE 3.1 ANALYSIS CONDITIONS FOR SLO OPERATION Reactor Thermal Power (1.042 x 75%) 2596.9 Hwt Core Flow 54.0 Hlbm/hr Core In-Channel Flow 48. 12 Hlbm/hr Core Bypass Flow 5.88 Hlbm/hr Idle Jet Pump Back Flow 11.9 Hlbm/hr Core Inlet Enthalpy 510.8 Btu/ibm Vessel Pressures Steam Dome 1020.0 psia Upper Plenum 1025.0 psia Core Pressure 1029.7 psia Lower Plenum 1035.0 psia Jet Pump H-Ratio 3.2 Recirculation Pump Flow 15.7 Hlbm/hr Turbine Pressure 960.5 psia Feedwater/Steam Flow . 10.79 Mlbm/hr

ANF-87-119 TABLE 3.2 RESULTS OF SLO PLANT TRANSIENT ANALYSES Naximum Maximum Haximum Core Average System Delta CPR Neutron Flux Heat Flux Pressure GE ANF GE ANF Event ~Rd f//Rd ~sicQ SLO SLO CY3 CY3 LRWB 143 82.7 1154 0.17 0.14 0.25 0.23 FWCF 80.6 79.8 1121 0.02 0.02 0.26* 0.24*

Pump Trip 78. 2 78.5 1020 0.02 0.00 NA

• Analyzed at 47% power, 106% flow.

150

2. HEA FLUX

3. REC RCULATI N FLOW

4. VES EL STEA FLOW 125 100 CI cc75 5 p 5 C)

$ 50 25

28. 0 0.4 0.8 1.2 1..6 2.0 2.4 2.8 3.2 3.6 4.0 TIME, SEC FIGURE 3.1 LOAD REJECTION WITHOUT WSS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED

f40 S EL HATE LEVEL (IN)

J 't ~

i20 100 80 60 2

40 20 1

0 O.O 0.4 '0.8 1.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 TIME, SEC FIGURE 3.2 LOAD REJECTION WITHOUT BYPASS RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED

175 150 125

1. NEUTRON FLUX'I EVEL n 2. HEAT FLUX

3. RECIRCULATION .FLOW Fr100 VESSEL STEAN FLOW n 5, FEEDWATER FLOW j75 50 25 0 n 0 ~

10 12 16 18 20 I TIME,'EC CO V

I lO FIGURE 3.3 FEEDMATER CONTROLLER FAILURE ULTS, RPT OPERABLE, NORMAL SCRAM SPEED

140

2. YES EL HATE LEYEL 120 100 80 60 40 20

'0 10 12 18 20 n I

CO TIME, SEC I

FIGURE 3.4'EEDWATER CONTROLLER FAILURE RESULTS, RPT OPERABLE, NORMAL SCRAM SPEED

100

2. HEA FLUX

3. REC RCULATI N FLOW

4. VES EL STEA FLOW 90 80 4 5 2 5 n

cc70 C3

$ 60 50 3

40 3'0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4,.0 4.5 TIME. SEC FIGURE 3.5- RECIRCULATION PUMP TRIP RESULTS

120

2. VES EL WATE LEVEL

~ 4 r ~

100 80 60 40 20 28 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4,5 5.0 TIME, SEC FIGURE 3.6 ..RECIRCULATION PUMP TRIP RESULTS

Pg 15 ANF-87-119 4.0 PUMP SEIZURE ACCIDENT The seizure of a recirculation pump is considered as a design basis accident event'. It is a very mild accident relative to other design basis accidents such as the loss of coolant accident (LOCA). The pump seizure event is a postulated accident in which the recirculation pump impeller speed is rapidly reduced to zero (in 0. 1 seconds). This causes a rapid decrease in core flow and a decrease in the heat removal rate from the fuel rods. Although the vessel water level increases, a high level trip did not occur in the analysis.

However, even without a scram the power decreases consistent with the core flow decrease until natural circulation conditions occur.

A pump seizure accident event was analyzed for WNP-2 to confirm the insignificance of this event relative to the design basis LOCA. The plant response to the pump seizure accident is shown in Figure 4. 1. The core remains covered and natural circulation conditions are approached within three seconds. Any fuel rods which experience boiling transition would be expected to be in the film boiling mode for a short period. In addition, the film boiling would be limited to small, localized areas in the affected fuel assemblies. Because of this short duration, fuel failures due to overheating or clad strain would not be expected as a result of this accident. Thus, the consequences of this event are bounded by the LOCA where fuel failures are assumed to be extensive.

80 4 5 70 60 n 1. NEUTRON FLUX LEVEL

2. HEAT FLUX cc50 3. RECIRCULATION FLOW 4, VESSEL STEAM FLOW

5. FEEDWATER FLOW

$ 40 30 20 100 p 0.4 0.8 1.2 i.6 2.0 2.4 2.8 3.2 3.6 4.0 TIME. SEC

i20

2. VES EL WATE LEVEL (IN) 100 80 60 40 20

.0 0.4 0.8 i.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 I CO TIME, SEC I lD FIGURE 4.2 PUMP SEIZURE ACCIDENT

'I 18 ANF-87-119 5.0 SAFETY LIMIT The HCPR fuel cladding integrity safety limit for single loop operation was calculated using the methodology described in Reference 7.

In this methodology, a Honte Carlo procedure is used to evaluate the impact on the safety limit of plant measurement and power prediction uncertainties. At the reduced flow state with .single loop operation, the plant measurement and prediction uncertainties of core flow, radial power distribution, and axial power distribution increase. The uncertainties used in the SLO safety limit evaluation are shown in Table 5. 1. The XN-3 correlation, Reference 8, is then used to predict the critical heat flux phenomena. 'on-parametric tolerance limits, Reference 9, are used to determine the expected number of rods in boiling transition.

During sustained SLO operation at a HCPR of 1.07 with the design basis power

~

distribution described below, at least 99.9% ~ of the fuel rods in the core are expected to avoid boiling transition at a confidence level of 95%. This supports a safety limit of 1.07, an increase of 0.01 over that for the normal operating 'state for all fuel types.

The design basis power distribution used in this analysis for single loop operation was based on the predicted power distributions which were determined to be the most severe or conservative with respect to the number of rods subject to boiling transition considerations.

19 ANF-87-11 5.1 SAFETY LIMIT EVALUATION UNCERTAINTIES Standard Parameter" Deviation*

Feedwater Flow Rate .0176 Feedwater Temperature .0076 Core Pressure .0050 Total Core Flow Rate .0600 Core Inlet Enthalpy .0024 XN-3 Critical Power Correlation .0411 Assembly Flow Rate .0280 Power Distribution Radial Peaking Factor .0551 Local Peaking Factor .0246

• Fraction of Nominal Value.

20 ANF-87-119 6.0 STABILITY ANALYSIS Single loop stability analyses have been performed with COTRAN for Cycle 3.

The calculations were performed using the two loop APRN rod block equation as specified in Reference 3. The decay ratio calculations were performed with the COTRAN code at the limiting Cycle 3 power and flow conditions. The most limiting'ycle 3 single loop decay ratios are identical to the Cycle 3 decay ratios for two loop operation. The limiting single loop decay ratios and the corresponding power/flow conditions are as follows:

Power Flow

% Rated  % Rated Oeca Ratio 65.0 45.0 0.49*

48.0 27.6 0 84**

Since the Cycle 3 single loop decay ratios are no larger than the Cycle 3 two loop decay ratios, the cycle dependent two loop stability analysis performed for future. cycles will bound single loop operation.

• At right hand boundary of Oetect and Suppress region.

4'

+A

21 ANF-87-119

7.0 REFERENCES

Krajicek, J. E., "WNP-2 LOCA Analysis For Single Loop Operation," XN-NF-87-118, Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1987.

Krysinski, T. L., and Chandler, J. C., "Exxon Nuclear Methodology For 2.

P Ip Inc,. Richland, X~F-WA

-I, Boiling Water Reactors; THERMEX Thermal Limits Methodology; Summary t,n 99352, FE September P, R 11 1986.

I, 1 II I C

3. Vopalensky, R. A., Washington Public Power Supply. System, Letter WPANF-2B-87-0023 to J. B. Edgar, Advanced Nuclear Fuels Corporation, "Supply System Data Package No. 44," dated March 2, 1987.

4. "251 BWR/5 Transient Safety Analysis Design Report," GEX-6413, General Electric Company, Hay 1977.

Kelley, Wt R,n R.

~X-F-Nuclear Co., Inc., Richland,

-1, H., "Exxon Nuclear Plant Transient Methodology For Boiling WA R 11 X (

99352, November 1981.

pp1 d), E

'. Krajicek, J. E., "WNP-2 Cycle 3 Plant Transient Analysis," XN-NF-87-24, Advanced Nuclear Fuels Corporation, Richland, WA 99352, March 1987.

Reactors,"

"Exxon Critical Power Methodology For Boiling Water XN-NF-

~524 A, Revision l, Exxon Nuclear Co., inc., Richland, WA 99352, November 8.

1983.

WIRE CEti I PC Nuclear Co., Inc., Richland, WA Iti,"~XIPEF-IXII,R 99352, March 1981.

11 1,1

9. Sommerville, Paul N., "Tables For Obtaining Non-Parametric Tolerance Limits," Annals of Mathematical Statistics, Vol. 29, No. 2, June 1958, pp 599-601,

0 ANF-87-119 Issue Date: 9/11/87 WNP-2 SINGLE LOOP OPERATION ANALYSIS Distribution D. A. Adki sson R. E. Collingham J. G. Ingham S; E. Jensen T. H. Keheley D. C. Kilian J. E. Krajicek J. L. Haryott G. L. Ritter H. E. Williamson J. B. Edgar/WPPSS (50)

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