ML061380035

Revised Redacted Version of Response to Information Request Dated February 15, 2005

ML061380035
Person / Time
Site:	Palo Verde
Issue date:	05/01/2006
From:	Troy Pruett NRC/RGN-IV/DRP/RPB-D
To:	James M. Levine Arizona Public Service Co
References
Download: ML061380035 (558)

Text

UNITED STATES NUCLEAR REGULATORY C O M M I S S I O N REGION I V 61 1 RYAN PLAZA DRIVE, SUITE 400 ARLINGTON, TEXAS 76011-4005 May 1, 2 0 0 6 James M. Levine, Executive Vice President, Generation Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

SUBJECT:

PAL0 VERDE NUCLEAR GENERATING STATION - REVISED REDACTED VERSION OF RESPONSE TO INFORMATION REQUEST DATED FEBRUARY 15,2005

Dear Mr. Levine:

Arizona Public Service (APS) Company's letter (102-05213-DMS/SAB/GAM) and affidavit dated February 15, 2005, submitted your staffs response to an information request in NRC Special Inspection Report 05000528/2004014; 05000529/2004014; 05000530/2004014. In this letter, APS requested that the information in Enclosure 2 and its attachment (except Attachment 2-F) to the letter be withheld from public disclosure pursuant to 10 CFR 2.390. At the request of the NRC staff, APS provided a redacted version of this submittal, dated July 5, 2005, that was suitable for public release. The redacted version of the submittal was subsequently posted on the NRC's public website (ADAMS assession number ML053480465).

We have carefully reviewed both the original February 15, 2005, letter and the redacted version. We have concluded that some of the material that was redacted may be withheld in accordance with 10 CFR 2.390, but that certain other material should be released and placed in the Public Document Room (PDR). The attachment to this letter provides a revised redacted version of the July 5, 2005, submittal which we believe meets the criteria of 10 CFR 2.390(a) for public withholding.

In accordance with 10 CFR 2.390(~)(2),this information was forwarded to you in an NRC letter dated January 24, 2006, (ML060250548) as notice that the information would be placed in the Public Document Room fifteen (15) days from the date of that letter. No response was received from APS within the required fifteen (15) days.

In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at httrx//www.nrc.qov/readinq-rm/adams.html(the Public Electronic Reading Room).

Arizona Public Service Company Should you have any questions concerning this correction, we will be pleased to discuss them with you.

Trby W. Pruett, Chief Project Branch D Division of Reactor Projects Dockets: 50-528 50-529 50-530 Licenses: NPF-41 NPF-51 NPF-74

Attachment:

cc wlattachment: Craig K. Seaman, General Manager Steve Olea Regulatory Affairs and Arizona Corporation Commission Performance Improvement 1200 W. Washington Street Palo Verde Nuclear Generating Station Phoenix, AZ 85007 Mail Station 7636 P.O. Box 52034 Douglas K. Porter, Senior Counsel Phoenix, AZ 85072-2034 Southern California Edison Company Law Department, Generation Resources Hector R. Puente P.O.Box 800 Vice President, Power Generation Rosemead, CA 91770 El Paso Electric Company 310 E. Palm Lane, Suite 310 Chairman Phoenix, AZ 85004 Maricopa County Board of Supervisors 301 W. Jefferson, 10th Floor Jeffrey T. Weikert Phoenix, AZ 85003 Assistant General Counsel El Paso Electric Company Aubrey V. Godwin, Director Mail Location 167 Arizona Radiation Regulatory Agency 123 W. Mills 4814 South 40 Street El Paso, TX 79901 Phoenix, AZ 85040 John W. Schumann Los Angeles Department of Water & Power Southern California Public Power Authority P.O. Box 51111, Room 1255-C Los Angeles, CA 90051-0100

Arizona Public Service Company John Taylor Public Service Company of New Mexico 2401 Aztec NE, MS Z110 Albuquerque, NM 87107-4224 Thomas D. Champ Southern California Edison Company 5000 Pacific Coast Hwy, Bldg. DIB San Clemente, CA 92672 Robert Henry Salt River Project 6504 East Thomas Road Scottsdale, AZ 85251 Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, TX 78701-3326 Karen ORegan Environmental Program Manager City of Phoenix Office of Environmental Programs 200 West Washington Street Phoenix, AZ 85003

Arizona Public Service Company Electronic distribution by RIV:

Regional Administrator (BSMI)

DRP Director (ATH)

DRS Director (DDC)

DRS Deputy Director (RJCI)

Senior Resident Inspector (GXWZ)

Branch Chief, DRPID (TWP)

Senior Project Engineer, DRP/D (GEW)

Team Leader, DRPITSS (RLNI)

RlTS Coordinator (KEG)

Dale Thatcher (DFT)

DRS STA (DAP)

William Maier, RSLO (WAM)

Stephen O'Connor, OED0 RIV Coordinator (TEB)

SUNS1 Review Completed: T W P - ADAMS: J Y e s 0 No Initials: TWP J Publicly Available 0 Non-Publicly Available 0 Sensitive J Non-Sensitive I :vo6 OFFICIAL RECORD COPY IRA/

4/28/06 T=Telepho ne E=E-mail F=Fax

ATTACtiMENT 1 Gregg R. Overbeck Tel (623) 393-5148 Mail Station 7602 Palo Verde Nuclear Senior Vice President Fax (623) 393-6077 PO Box 52034 Generating Station Nuclear e-mail: GOVERBEC@apsc.com Phoenix, Arizona 85072-2034 1 02-05303-GRO/lNW/GAM ATTN: Document Control Desk July 5, 2005 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Dear Sirs

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units I,2 and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Redacted Version of Proprietary Submittal Dated February 15, 2005 Regarding Safety Significance Evaluation of ECCS Containment Sump Voided Piping In letter no. 102-05213, dated February 15,2005, Arizona Public Service Company (APS) submitted to the NRC the safety significance evaluation of emergency core cooling system (ECCS) containment sump voided piping. APS requested that Enclosure 2 and Attachments 2-A, 2-6,2-C, 2-0, and 2-E of that submittal be withheld from public disclosure under 10 CFR 2.390(a)(4) because they contained information considered to be proprietary to APS. Since that time, NRC Region IV personnel have requested that APS submit redacted versions of and Attachments 2-A, 2-B, 2-C, 2-D, and 2-E of the February 15,2005 submittal.

The requested redacted versions of the enclosure and attachments are enclosed.

There are no commitments in this letter. Should you have any questions, please contact Mr.

Thomas N. Weber at (623) 393-5764.

Sincerely, GRO/TNW/GAM/ca

Enclosure:

Redacted Versions of Proprietary Enclosure 2 and Attachments 2-A, 2-B, 2-C, 2-D, and 2-E of APS Letter No. 102-05213, dated February 15, 2005, Regarding Safety Significance Evaluation of ECCS Containment Sump Voided Piping cc: T. W. Pruett NRC Region IV (w/ Enclosure)

B. S. Mallett NRC Region IV Regional Administrator (wlo Enclosure)

M. 6. Fields NRC NRR Project Manager Y G. G. Warnick NRC Senior Resident Inspector for PVNGS u A nember of the STARS ( S t r a t e g i c Teaming and Resource Sharing) A l l i a n c e Callaway Comanche Peak Diablo Canyon Palo Verde South Texas ? r o j e c t Wolf C r e e k

Redacted Versions of Proprietary Enclosure 2 and Attachments 2-A, 2-6, 2-C, 2-0, and 2-E of APS Letter No.

102-05213, dated February 15,2005, Regarding Safety Significance Evaluation of ECCS Containment Sump Voided Piping

REDACTED VERSION ENCLOSURE 2 SAFETY SIGNIFICANCE EVALUATION OF ECCS CONTAINMENT SUMP VOIDED PIPING REDACTED VERSION

REDACTED VERSION Page 7 SIGNIFICANT CRDR 2726509 SAFETY SIGNIFICANCE EVALUATION OF ECCS CONTAINMENT SUMP VOIDED PIPING REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 2 t:

- Executive Summary In July, 2004, Engineering personnel determined that a section of Emergency Core Cooling System (ECCS)piping leading from the containment recirculation sump, in both ECCS trains in each of the three Palo Verde Units, was left in an unfilled condition during normal plant operation. The resultant volume of air could potentially be ingested into the ECCS pumps suction following a Recirculation Actuation Signal (RAS). A review of design basis information determined that this condition was not consistent with the design intent of the ECCS and not consistent with the analyses that demonstrate the ability of the ECCS to perform its design basis safety functions. Condition RepodDisposition Request (CRDR) 2726509 was initiated to document and evaluate the condition.

The purpose of this report is to describe and provide the results of a comprehensive testing and analysis program performed to evaluate the ECCS system response to the voided piping condition. The results of the evaluation are then used in a risk assessment to determine the safety significance of the discovered condition.

Scale model tests were performed at Fauske and Associates which simulated the system response during and following a RAS with the affected section of piping initially unfilled. The scale tests were conducted in phases. The purpose of the first phase (typically referred to as Phase I) was to demonstrate the ability to simulate the transient and measure the important parameters such as void fiaction, pressure, and flow rate. [

Full-scale pump tests were performed at Wyle Labs utilizing a spare Palo 1 a d e High Pressure Safety Injection (HPSI) pump and a representative Containment Spray (CS) pump to determine the impact on pump performance under the projected air ingestion conditions. Tests were performed for a spectrum of flow rates and air ingestion rates based on the results of the scale model test program. Pump Derformance, as defined by developed head and flowrate, was measured as a function o f-

.- Y 1 .. - -

7 1 A series of thermal hydraulic analyses of the Palo Verde Reactor Coolant System and Containment were performed using the Westinghouse CENTS code and the EPRI MAAP code. These analyses established

~

the expected reactor coolant and containment environment conditions that would exist at the time of R4S for a spectrum of Loss of Coolant Accident (LOCA) break sizes. Operator actions, as prescribed in the Palo Verde Emergency Operating Procedures (EOPs), to initiate a cool down and depressun-zethe RCS upon diagnoses of a LOCA wereexplicitly considered in the analyses. B - C REDACTED VERSION.

Safety Significance Determination

REDACTED VERSION Page 3 In addition to the testing program, a computer hydraulic transient analysis of the ECCS voided pipe condition was performed. [

1 Uitimately, the analysis results are compared to the testing program and shown to be complimentary.

Given the resuits of the tests and analyses, the risk significance was determined by making appropriate adjustments to the Palo Verde ProbabilisticRisk Assessment CpRA) model. [

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REDACTED VERSION Page 4 1.1 BackgroundlPurpose of Report In July, 2004, Engineering personnel determined that a section of Emergency Core Cooling System (ECCS) piping leading from the containment recirculation sump, in both ECCS trains in each of the three Palo Verde Units, was left in an unfilled condition during normal plant operation. The resultant volume of air could potentially be ingested into the ECCS pumps suction following a Recirculation Actuation Signal (US). A review of design basis information determined that this condition was not consistent with the design intent of the ECCS and not consistent with the analyses that demonstrate the ability of the ECCS to perform its design basis safety functions. Condition ReportCDisposition Request (CRDR) 2726509 was initiated to document and evaluate the condition.

The purpose of this report is to describe and provide the results of a comprehensive testing and analysis program performed to evaluate the ECCS response to the voided piping condition. The results of the evaluation are then used in a risk assessment to determine the safety significance of the discovered condition.

1.2 Description of Condition The Palo Verde ECCS design employs recirculation from the containment sump after the contents of the Refueling Water Tank (RWT) have been injected into the reactor vessel and containment building. Upon receipt of a RAS, automatic valve actuations result in suction of the ECCS pumps being transferred from the RWT to the containment sumps. Two completely redundant and separated ECCS trains are utilized. Figure 1-1 illustrates a typical ECCS suction piping and component layout.

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REDACTED VERSION Page 5 Emergency Core Cooling and Containment Spray System Suction Piping Train A -

Refueling Water Tank Minimum elev. 94 ft. 4 in.

Containment Recirculation Sum A Minimum elev. 84 f! 6 in.

-Not to scale-Figure 1-1 Typical Palo Verde ECCS Suction Layout As illustrated in Figure 1-1, the containment sump outlet pipe contains an in-board and an out-board containment isolation valve, and a downstream check valve. Engineering personnel determined that this section of the ECCS suction piping, between the two containment isolation valves and between the out-board valve and the downstream check valve, had been routinely left in an unfilled condition during plant operation.

In the unlikely event of a Loss-of-Coolant Accident (LOCA), the contents of the Reactor Coolant System (RCS)will leak into containment and flow into the containment sumps. Automatic ECCS actuation would occur causing the contents of the RWT to be injected into the RCS and the containment building to maintain core cooling and containment pressure and temperature control.

Ultimately the basement of the containment building, including the containment sumps, would become flooded. Once the contents of the RWT are depleted, a RAS would be automatically generated causing both containment sump isolation valves in each train to open, resulting in closure REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 6 of the RWT isolation check valves. The RAS would also cause, by design, the Low Pressure Safety Injection (LPSI) pumps to be turned off. ECCS suction, consisting of a SI pump and a CS pump in each train, would thus be transferred to the containment sump.

With the containment sumps flooded and the section of containment sump piping not filled with water, air would be trapped in the piping. As flow is initiated from the sump, this air could be entrained and/or transported into the ECCS suction piping and potentially into the ECCS p u m p inlets. Industry literature and operating experience indicates that pump performance could be severely degraded, or even result in air binding or pump failure, if the resultant air volume fraction ingested by the pump exceeds the pumps tolerance for air ingestion. Industry literature (Ref. 1 NUREGKR 2792) indicates that a pumps tolerance for air ingestion varies by design and fluid conditions, but at air volume fractions above approximately 3%, pump degradation can be experienced.

Therefore, in order to determine the safety significance of this condition, the air volume fraction that could be ingested by the HPSI and CS pumps would need to be determined. Once the air volume fraction is determined, each pumps tolerance for the projected air ingestion can be assessed, and ultimately the impact on the ECCS safety functions.

1.3 Significance DeterminationApproach The assessment of voided and two-phase fluid behavior is complex. A comprehensive scale model testing program was employed to develop a full understanding of the system response to the void and the resulting aidfluid conditions that would be delivered to the pumps suction inlet. The impact to pump performance was then assessed via full-scale testing, given the projected aidfluid inlet conditions.

The scale model tests were performed at Fauske and Associates, and simulated the system response during and following a RAS with the affected section of piping initially voided. The scaIed tests were conducted in phases. The first phase modeled the RWT and associated piping, and the sump and associated piping down through and including the long vertical run of pipe. The purpose of the first phase (typically referred to as Phase 1) was to demonstrate the ability to simulate the transient and measure the important parameters such as void fraction, pressure, and flow rate. A series of tests were performed to test important scaling parameters to ensure the results of the test could be confidently applied to the hll scale Palo Verde units. A series of phenomenological tests using a larger scale model was incorporated into the test plan to verify that the flow regime in the vertical section of the scaled piping configuration was representative of large pipe behavior.

The second phase extended the scale model to include the individual pump suction piping up to each pump inlet. An extensive series of tests under varying flow and pressure conditions were performed.

t 1 These results established the inlet conditions for the subsequent full-scale pump performance tests.

Full-scale pump performance tests were performed at Wyle Labs utilizing a spare Palo Verde High Pressure Safety Injection (KPSI) pump and a representative Containment Spray (CS) pump to determine the impact on pump performance under the projected air ingestion conditions. The HPSI pump was of the same make and model as those installed at Palo Verde. A spare CS pump of the REDACTED VERSION Safety Significance Defemination

REDACTED VERSION Page 7 Same make and model as the Palo Verde CS pumps was not readily available; therefore a spare CS pump fiom a cancelled WPSS plant was utilized for the testt.This pump is the same make and model as the Palo Verde LPSI pumps and is very similar in design and size to the Palo V&de CS pumps.

The impact on performance for equivalent fluid conditions is expected to be representative. Tests were performed for a spectrum of flow rates and air ingestion rates based on the results of the scale model test program. Pump performance was measured as a function of air volume fraction. A maximum degraded pump performance curve was then constructed using the test results for the tests performed at maximum air volume fractions.

A series of thermal hydraulic analyses of the Palo Verde Reactor Coolant System and Containment were performed using the Westinghouse CENTS code and the EPRI MAAP code. These analyses established the expected reactor coolant and containment environment conditions that would exist at the time of RAS for a spectrum of LOCA break sizes. Operator actions, as prescribed in the Palo Verde Emergency Operating Procedures (EOPs), to initiate a cool down and depressurize the RCS upon diagnoses of a LOCA were explicitly considered in the analyses.1 a.

] For those system conditions in which the required head do not exceed the degraded pump performance capability, continued degraded ECCS delivery (Le. continued pump flow) is assumed until the air inventory available for ingestion into the pump is consumed, at which time restoration of fbll pump performance is assumed.

1 REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 8 2.1 Phase 1 Test Program and Results 2.1.1 Experimental Objectives and Physical Arrangement 7

The objective of the Phase 1 testing was to investigate the potential for the air initially resident in the horizontal piping section from the containment sump to be forced into the vertical downward piping section. Phase 1 tests included the transient effects of switching the supply from the simulated RWT to the simulated containment sump by simultaneously opening the sump suction isolation valves.

Clear piping was used for the horizontal and vertical segments of the simulated suction line to observe and record the flow pattern and the behavior of the initial air filled void. A complete report on the conduct and results of the Phase 1 test program is attached as Attachment 2-A to this report.

The test facility that was used was comprised of two tanks with water inventories, a centrifbgal pump, piping, valves, and associated instrumentation. The piping and valves used to establish and visualize the flow pattern development from the initial location between the valves and into the downcomer piping were all 4 inch in diameter. Clear plastic piping facilitated observation of the initial air inventory behavior during the opening of the motor operated valves. The vertical segment was also clear plastic piping that allowed for the observation [ ] in the downward vertical flow. [

1 2.1.2 ScalinP Considerations As indicated, 4 inch diameter piping was used to simulate the sump horizontal and vertical downward sections of piping. Since actual Palo Verde piping is 24 inch in diameter, this results in a I/6*

geometric scaling factor. This geometric (lengths and diameters) scaling factor was maintained through out the Phase 1 tests to the extent possible.

Previous tests and experiments described in the literature have demonstrated that maintenance of the Froude number, particularly for horizontal flow regimes, will result in prototypical behavior in scaled experiments. As such, flow rates were scaled in the Phase 1 tests so as to maintain the same dimensionless Froude Number parameter as would exist in the Palo Verde units.

2.1.3 Phase 1 Results and Observations A series of twelve tests were performed with varied [

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REDACTED VERSION Page 9 2.2 Phenomenological T-estingProgram 2.2.1 Experimental Objective and Phvsical Arraneement Design reviews conducted before and after the Phase 1 tests and an independent review [

1 resulted in the identification of several phenomenological investigations that could be pertormed to provide REDACTED VERSION Safety Significance Detennination

REDACTED VERSION Page 10 The test arrangement also provided the opportunity to observe the flow patterns and influence of the HPSI and CS branch connections off the lower header piping.

2.2.2 PhenomenoloPical Testine Results and Observations An extensive series of tests using the [ ] scale test apparatus were performed. Key observations from these tests were REDACTED VERSION Safety Significance Detennination

REDACTED VERSION Page 1 7 c

2.3 Phase 2 Test Program and Results 23.1 Exwrimental Obiectives and Physical Arraneement The test facility for Phase 2 was similar to that of Phase I [

1 iyJ Volume iFA 1-3 Figure 2-1 Phase 2 Test Arrangement.

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REDACTED VERSION Page 72

] In the plant system under accident conditions, air transported through the HPSI line would influence the pump performance and cause a decrease in the flow rate being pumped. Reduced flow rate would cause a corresponding reduction in the rate of air ingestion.

Thus, the air intrusion rate deduced from these scaled experiments provides a conservative representation of the plant response.

The test instrumentation is also illustrated in Figure 2-1. A computer with a CIO-DASOO8 data acquisition card was used to collect the data. Key pieces of instrumentation included 1

Various pressure , level, and flow meters 1

During the Phase 2 tests, the flow rate through the CS pump was again held constant at the maximum predicted flow rate equivalent to 4885 gpm, except for several tests in which CS flow was set to zero to simulate a HPSI flow only scenario. HPSI flow rate was varied ranging from the equivalent to 200 gprn to an equivalent maximum nm-out flow of 13 10 gpm. LPSI start scenarios were also tested for a range of LPSI flow rates.

-23.2 Scalinp Considerations The same I/& geometric scaling used in Phase 1 was used for the Phase 2 experiments. Flow rates were scaled to maintain the same Froude number that would exist at Palo Verde. The Froude number relationship was maintained for both the total flow and the individual flow rates to the simulated WSI, CS,and LPSI pumps.

REDACTED VERSION 1

Safety Signmcance Determination

REDACTED VERSION Page 13 In this horizontal orientation, the principal scaling parameter has been well established previously 1

(References 3 and 4) to be the Froude number which is a ratio of the inertial and buoyancy forces, i.e.

where:

D is the diameter of the horizontal piping, g is the acceleration of gravity, U is the onedimensional velocity of the flow in this line, 0 pe is the air density, and 0 pwis the water density.

Since pw>> pg,this reduces to the familiar form U

NFr =JgD REDACTED VERSION Safety Significance Determination

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REDACTED VERSION Page 75 2.33 Phase 2 Results and Observations A series of twentyeight tests were initially performed with varied flow rates, containment level, and containment pressure conditions. Additional tests were later performed to investigate the air transport process during potential LPSI pump start scenarios. Key observations from the tests were:

Flow Patterns Digital movie cameras were used to record the flow patterns in all the Phase 2 tests. Each test was 1

initiated by simultaneously opening the sump containment isolation valves. As the valves open, water REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 76 is seen to enter the initially voided horizontal piping segment and induce mixing of the water and air.

The air is swept out of the horizontal segment and into the vertical piping segment. [

1 HPSZ Air lngeslion Rofes 1

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These results show that the air flow ingestion rates increase to their maximum value within

' approximately [ 1 seconds for the scald experiments and then subsequently decay towards zero as the air inventory in the horizontal suction header becomes insufficient to enter the HPSI line. Similar evaluations for scaled HPSI flow rates [

1 REDACTED VERSION Safety Significance Detennination

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With a 1/6th h e a r scale, the respective volumes are determined by the cube of this linear scale, i.e.

the scaled up quantities are defined by the volume multiplied by 216. More simply put, the area is scaled by the square of the diameter times the length. Thus six cubed equals 216. Since mass is directly proportional to volume at a given pressure and temperature, mass quantities are also scaled by a factor of 216.

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REDACTED VERSION Page 20 Using the results from the Phase 2 tests, these scale factors are applied and the results illustrated in Figure 2-4 for the &e of a "SI flow rate of I3 10 gpm. As shown, the meaningful delivery period for the air flow is approximately [ 1 REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 21 Since Reference 1, and other pump performance tests described in the literature, indicates that pump performance is typically assessed as a function of air volume fraction, the peak mass flow rate data obtained during the Phase 2 tests was converted to air volume fractions for use in the full-scale pump tests. [

1 REDACTED VERSION Safety SignificanceDetermination

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REDACTED VERSION Page 23 I3 I  :. .

'. Hygyaulic.

I - . Transient Arialygis 3.1 Description of Analysis and Computer Model A hydraulic computer model of a typical Palo Verde ECCS system was developed 1

3.2 Analysis Results REDACTED VERSION Safety SignMcance Determination

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REDACTED VERSION Page 26 3.3 Hydraulic Transient Analvsis Conclusions Safety Significance Determination 1

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REDACTED VERSION Page 37 Pump Performance t With Air Ingestion 4.1 Description of Test Facility The pump performance tests were conducted at Wyle Labs in Huntsville, AL. The test facility consisted of two closed pump loops each drawing suction from, and discharging to, a common 30,000 gallon pressure vessel. One loop was constructed to provide for testing of the spare HPSI pump. Suction and discharge pipe sizes were selected to correspond to the actual pipe sizes at Palo Verde. The specific suction piping configuration leading into the HPSI suction nozzle was explicitly reproduced. The second loop was provided for testing of the representative CS pump.

4.2 Test Conduct A series of tests were conducted at each base case flow rate. The base case flow rates of 600 gpm, 900 gpm, and 13 10 gpm were selected to span the range of flow rates that could be expected at the time of RAS during a postulated LOCA.

For each base case, tests were performed at incrementally increasing air injection mass flow rates.

The resulting air volume fraction, defined as the ratio of volumetric air flow rate to total volumetric air flow rate, was then determined. [

] Figure 4-1 illustrates the final test for the 900 gprn base case. 1 1

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REDACTED VERSION Page 32 Figure 4-1 Air Injection and Air Volume Fraction for Final 900 gpm Series Test During every test, the duration of air injection was specified to assure that the total volume of air

[ ] exceeded the total volume of air predicted by the scale model tests. Pump performance data was taken during each test for subsequent assessment of the air ingestion on pump performance.

Visual observations, and digital camera recordings, were made for all HPSI test cases.

4.3 TestResults Visual observations through the clear spool piece on the "SI suction line confirmed [

] similar in nature to that observed during the scale model Phase 2 tests. The visual observations confirmed the proper scaling of the Phase 2 tests and gives reasonable confidence that the Phase 2 and Phase 3 tests closely approximate the full-scale plant conditions. Pump performance data was taken using a data acquisition system that recorded each data point 10 times per second. The recorded data was then inserted into Excel spreadsheets to facilitate calculation of pump developed I

1 The data represents the calculated developed head (TDH)from the recorded pump inlet and outlet pressure data taken every 0.1 seconds, and the corresponding flow rates as measured on the pump discharge line. The data represents that obtained over a specific time period during which the air injection rate was at its maximum steady state value and the corresponding peak air volume fractions were obtained. The data points, as expected, fall along the test loop system curve.

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REDACTED VERSION Page 35 As illustrated in the preceding three figures, and as would be expected, pump performance progressively degrades as inlet air volume fraction increases. This progressive degradation is consistent with data reported in NUREGKR 2792 (Reference 1). The following figure 4-5 is taken from Reference 32 is cited in the NUREG.

Figui REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 36 A maximum bounding degraded pump curve is then constructed as shown in Figure 4-6. As illustrated, the maximum degraded pump curve conservatively bounds all recorded data for the peak air volume fraction cases tested. The use of this maximum degraded pump curve redults in additional conservatism since the Phase 3 tests conditions in some cases exceeded the specified air volume fraction from the Phase 2 scale model tests.

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REDACTED VERSION Page 37 8 Xn&wae o F ~ ~ ~ w & c T of r l p r ~ Zir.

s Figure 4-7 hfluence of Number of Stages on Performance Degradation (from NUREGKR-2792)

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REDACTED VERSION Page 38 5.1 Thermal Hydraulic Analysis of Spectrum of LOCA Break sizes A series of thermal hydraulic analyses of the Palo Verde ECCS system were performed using the Westinghouse CENTS code and the EPRI MAAP4 code. These analyses established the expected reactor coolant system and containment environment conditions that would exist at the time of RAS for a spectrum of LOCA break sizes. Operator actions as prescribed in the Palo Verde Emergency Operating Procedures (EOPs) to initiate a cool down and depressurization of the RCS upon diagnosis of a LOCA were explicitly considered in the analyses. In this way, best-estimate parameters such as RCS and containment pressures at time of RAS were established. The CENTS and MAAP codes were used to mutually develop the conclusions associated with the LOCA scenarios. Summary descriptions of the two codes are presented, followed by descriptions of application of the HPSI and CS pump test data in the transient results. Detailed descriptions of the codes and their applications and limitations are within References .- These references also provide detailed descriptions of the individual transient results.

5.1.1 MAAPJ Analvsis Code DescriDtion MAAP is a computer code that simulates tight water reactor system response to accident initiation events. The Modular Accident Analysis Program (MAAP), an integral systems analysis computer code for assessing severe accidents, was initially developed during the industry-sponsored IDCOR Program. At the completion of UXOR, ownership of MAAP was transferred to Electric Power Research Institute (EPRI). Subsequently, the code evolved into a major analytical tool (MAAP 3B) for supporting the plant-specific Individual Plant Examinations (PES) requested by NRC Generic Letter 88-20. Furthermore, MAAP 3B was used as the basis to model the Ontario Hydro CANDU designs. As the attention of plant-specific analyses was expanded to include accident management evaluations, the scope of M A P (its design basis) was expanded to include the necessary models for accident management assessments. h4AAP4 is the first archived code that contains a graphical representation of the reactor and containment response. MAAP4, like MAAP 3B, is currently being maintained by Fauske & Associates, LLC (FAI) for EPRI and the MAAP Users Group (MUG).

MAAP4 is an accident analysis code that provides results with confidence in all phases of severe accident studies, including accident management, for current P W R reactorkontainment designs and for ALWRs. MAAP4 includes models for the important accident phenomena that might occur within the primary system, in the containment, and/or in the auxiliarykeactor building. For a specified reactor and containment system, MAAP4 calculates the progression of the postulated accident sequence, including the disposition of the fission products, from a set of initiating events to either a safe, stable state or to an impaired containment condition (by overpressure or over-temperature) and the possible release of fission products to the environment.

Since the beginning of the MAAP code development, the codes have represented all of the important safety systems such as emergency core cooling, containment sprays, residual heat removal, etc.

M A P 4 allows operator interventions and incorporates these in a flexible manner, permitting the user to model the operator response and the availability of the various plant systems in a general way.

REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 39 The user can represent operator actions by specifying a set of values for variables used in the code andor events, which are the operator intervention conditions. There is a large set of actions that the operator can take in response to the intervention conditions.

h4AAP4 has been developed under the FA1 Quality Assurance Program, in conformance with IOCFRSO Appendix B and with the International IS0 9000 Standard. Furthermore, the new software has been subjected to review by a Design Review Committee, comprised of senior members of the nuclear community, in a manner similar to that exercised for MAAP 3B.

MAAP4 has been benchmarked against plant experience and large-scale integral experiments and also against one integral computer code. Most of the plant experience and experiment benchmarks are documented in the M A P 4 Users Manual PPlU, 2003aJ.

The USNRC reviewed and approved MAAP 3.OB for support of probabilistic risk assessment (PRA) activities at licensed power reactors in the U.S., particularly the IPEs that occurred in the late 1980s and early 1990s. While MAAP4 has not undergone a formal review process by the NRC,the code owner, EPRI, Fauske & Associates, and the M A P Users Group previously engaged in MAAP4 familiarization activities with the NRC when MAAP4 was first released. Recently, a M W 4 Information Exchange between these parties has been undertaken in view of the expanding scope of MAAP4 application and MAAP4-supported submittals to the NRC.

MAAP4 has been used previously for safety analyses outside of the risk arena with NRC approval.

For example, an NRC Safety Evaluation Report (SER) was written for the D.C. Cook plant in its assessment of minimum safe sump level in the containment recirculation sump during a small LOCA event. This assessment involved small LOCA scenarios that are similar to those in the present analysis for PVNGS.

The MAAP4 RCS model uses momentum equation selectively for sub-models that demand a momentum equation for model integrity. One of the aspects for which a full-fledged momentum equation is not implemented is water flow. Consequently, MAAP4 cannot void the core by reversing flow from the core 10 the downcomer and loop piping during a large LOCA event. However, small breaks of the size being analyzed for this analysis do not engage in such significant flow reversal, so this limitation is not relevant to this analysis.

The M A P 4 containment model can accommodate most physical phenomena that would occur.

However, since it does not entrain pre-existing liquid and condensate from heat sink surfaces, it does not mechanistically bring suspended water droplets into the containment atmosphere (although the model could accommodate droplets if such liquid entrainment was added). Consequently, it conservatively predicts excess gas-phase superheat and pressurization during the blowdown stage of a large LOCA event. Since small breaks of the size being analyzed for this analysis do not engage in this phenomenon, this limitation is not relevant to this analysis. Documented containment benchmarks are testament to the adequacy of the containment model for predicting short-term and long-term containment pressurization under small and medium LOCA conditions, which is necessary for an accurate depiction of containment spray actuation signal (CSAS) timing in this analysis.

The latest MAAP4 archived revision, MAAP 4.0.5 EPFU, 2003b1, was used with the latest PVNGS-specific plant model (a.k.a., parameter file). [

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REDACTED VERSION Page 40 1

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REDACTED VERSION Page 41 REDACTED VERSION Safety Significance Determination I . . .

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The analyses provide three key results. The first result is the RCS pressure that would exist at the time of RAS for various size breaks. These results are provided in Figure 5-1.

Break ~ RCS Pressure at RCS Pressure at Size RAS (psia) RAS (psia) Suction Discharge Leg Leg Breaks Breaks 1 1386 1384 2 546 438 Table 5-1 RCS Pressure at RAS for Various Break Sizes from CENTS This parameter is used in the following section to 1 assess ECCS performance (i.e. HPSI flow) under the maximum predicted air ingestion conditions.

The second result from these analyses is that break sizes of 2 diameter or smaller [

] alternate method of core cooling is available should the HPSI pump fail due to air ingestion. The current PVNGS Emergency Operating Procedures fully implement this recovery strategy.

1 5.2 Determination of Degraded HPSI Flow REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 43 The resulting HPSI system performance or operating points, given the degraded pump performance and the system resistance curves developed above, can be determined and illustrated graphically as shown in Figure 5-2. The developed head and flow rate of the degraded pump is determined by the intersection of the system curves and the degraded pump curves.

REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 44 As indicated in Figure 5-2, the static head associated with the 1" diameter small break LOCA at the time of RAS is well above the developed head of the degraded HPSi pump under maximum air ingestion.

For break sizes 2" diameter and larger, Figure 5-2 indicates the degraded HPSI pump has sufficient developed head to continue delivering ECCS flow to the RCS for the short time until the volume of air originally resident in the voided piping is exhausted. After the total air volume is ingested, the Phase 3 pump performance tests demonstrated the HPSI pump would recover and return to its normal nondegraded performance. [

1 5.3 HPSI Pump (Emergency Core Cooling) Safety Function Impact Conclusion From the Phase 3 pump performance tests under air ingestion, a bounding degraded HPSI pump performance curve was developed. The bounding degraded performance curve envelopes the maximum predicted air volume fractions ingested by the HPSl pump, based on Phase 2 scale-model testing. This study then compared the resulting degraded pump performance with the calculated system resistance that would exist at the time of RAS,for the spectrum of break sizes. The comparison indicates the degraded "SI pump would develop sufficient discharge head to maintain flow to the RCS for all break sizes except for the smallest breaks less than 2". The degraded flow rate delivered to the RCS would only exist [ ] until the air inventory available to be ingested is exhausted, at which time pump performance can be assumed to return to normal. The analyses performed using the CENTS and MAAP codes determined that for the full spectrum of REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 45 5.4 Containment Spray Safety Function Conclusion Tests were conducted on the representative CS pump by injecting air at rates up to approximately

[ ] air volume fraction. This air volume fraction conservatively bounds the amount of air predicted by scale model testing for all scenarios tested. The pump experienced a reduction in flow during the period of air ingestion, and then returned to normal baseline performance after air injection was suspended. Jt is concluded that the voided pipe condition does not have a significant impact on Containment Spray pump functionality.

REDACTED VERSION Safety Significance Oetennination

REDACTED VERSION Page 46 Other Considerations 6.1 Waterhammer The ECCS voided piping condition did not present any negative impacts stemming from waterhammer. Numerous analyses and experiments (References 12 through 14) have been performed to evaluate the influence of air in a system during a strong hydraulic transient such as a pump start.

As stated by Martin (Ref. 12):

The effect of the presence of entrapped air on transient pressures of a liquid pipeline can either be beneficial or detrimental, depending on the amount of air, the two-phase flow regime of the mixture (whether homogeneous or slug), and the nature and cause of the transient.

Of particular importance are those situations which could be detrimental to the piping system.

Generally these are conditions in which a significant coherent gas volume has formed on the discharge side of the pump. Significant means a volume that is comparable to or larger than the integrated volumetric flow discharged from the pump during the time that it comes up to speed.

Given these conditions the pump can accelerate to essentially runout flow conditions with the only resistance being the frictional forces generated by the moving water column between the pump discharge and the air pocket. Subsequent to this, the moving water column will begin to compress the air volume and the gas pressure will increase dramatically as volume is reduced.

For example, under these conditions, the gas bubble pressure more than doubles when the gas volume is reduced by one half and similarly more than doubles again when it is reduced again by one half, etc. Hence, with a low pressure gas volume on the discharge side of the pump, the compression of the gas bubble will eventually absorb the kinetic energy of the water column. For this to occur, the gas volume pressure can increase to values much greater than the maximum pump discharge pressure.

Conversely, if the air volume is on the suction side of the pump such as in the case of the Palo Verde ECCS voided piping, [

1 REDACTED VERSION Safety Significance Determination

REDACTED VERSION Page 47 ANALY.TIC PREDICTION EXPERIMENTAL DATA 0

1 ,

BUBBLY F L O W EO. ( 8 2 )

0.20 0 40 0.60 V O I D FRACTION, Q 0.80 1.0  :

Figure 6-1: Comparison of the ratio of the two-phase propagation velocities to the water sonic velocity for selected flow patterns (taken from Henry, Grolmes and Fauske, 1971).

As illustrated, for stratified flow the pressure wave propagation velocity was reduced by a factor of four while bubbly mixtures experienced a reduction of as much as two-orders of magnitude.

Consequently, a uniformly distributed gas volume will slow the response to transients, i.e. stabilize the flow. This is consistent with the example calculations provided by Martin (Ref. 12).

In summary, if a large air bubble exists in the pump discharge piping, the pump start transient can experience pressure surges with peak values well in excess of the pump shutoff head. The extent of the pressure increase is determined by the gas volume, pump runout flow, etc. For those conditions with air on the suction side of a pump, the air flow rate will be determined by the pressure difference from the pump header to the pump inlet, the dispersed air flow will have a greatly reduced volume in the discharge piping and will slow (stabilize) the hydraulic response of the piping network.

REDACTED VERSION Safety Significance Detemination

REDACTED VERSION Page 48 6.2 Net Positive Suction Head NUREGKR-2792 (Ref. 1) provides discussion and guidance regarding the affect of pump air ingestion on NPSH considerations. For example, Section 3.2.3 states that the presence of air at the inlet.. ...increases the limiting NPSH required for satisfactory operation. The increased degradation at the pump inlet, as inlet NPSH or pressure is lowered, results from the increased volumetric expansion of air between the pump inlet flange and the impeller inlet. Thus pumps operating with air ingestion will have higher NPSH requirements than those required in single-phase operation.

Section 4.2 goes on to establish an arbitrary relationship for the purpose of minimizing this volumetric expansion that occurs between the inlet and the impeller eye. The relationship is:

Where AF is the air volume fraction in percent. It is noted that this relationship is only intended for use with air volume fi-actions less than 2%

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REDACTED VERSION Page 49 REDACTED VERSION Safety Significance Detennination P

REDACTED VERSION Page 50 isk Assessment Probabilistic Risk Assessment Conclusion From the CENTS thermal-hydraulics analyses and the Phase 3 pump performance tests, modifications to tbe Palo Verde Probabilistic Risk Assessment (PRA) model were made to assess the risk significance of the voided pipe condition. The Palo Verde model contains an event tree for small break LOCAs of 2.3 inch diameter and smaller. The model was revised by inserting a failure of the HPSI pumps at RAS (failing the high pressure recirculation function) for small-break LOCA due to air binding, and modeling the subsequent plant cool down and depressurization and LPSI alignment for low pressure recirculation. Consideration was also given to small LOCA events that are induced through the lifting of a PSV and the subsequent failure to reseat. An estimate of the risk increase due to small LOCAs resulting from seismic events was also calculated. Since the pump performance tests indicate that for breaks 2 inches in diameter and larger failure of the HPSI pump is not likely, medium and large LOCA events were unaffected by the voided condition. Thus the small LOCA event would be the dominant contributor to the risk increase due to the voided pipe condition.

Engineering Study 13-NS-CO74, Revision 0 (Attachment 2-F) calculated the increase in risk associated with the unfilled containment sumps suction lines. The following table shows the overall impact of loss of High Pressure Recirculation (HPSR) for break sizes of two inches or less.

Initiator Delta-CDF (per Y w Small LOCA 4.5E-6 PSV -Internal Events Plus Fire 2.OE-6 Seismic 4.7E-7 Total 7.0M

~~ ~ ~ ~

Table 7-1 Over-all Risk Associated with Loss of HPSR The above described model adjustments were applied to the entire range of small break L E A events (i.e. 2.3 diameter and smaller). The pump testing and analysis program described in the previous sections of this report demonstrate that continued functionality of the HPSI pump for the upper end of the SBLOCA range (those breaks approaching 2 in diameter and larger) would be expected. For the small end of the SBLOCA range of approximately 0.5 in diameter or less, analyses using the CENTS and MAAP code demonstrate that complete depressurization of the RCS to shutdown cooling conditions would be achieved prior to RAS. Therefore, no additional risk is associated with REDACTED VERSION Safety SigniffcanceDetemination

REDACTED VERSION Page 51 the breaks on the small end of the SBLOCA range. Therefore, the above result provided in Table 7-1 is considered to be a conservative estimate of the incremental risk associated with the ECCS voided piping condition.

REDACTED VERSION Safety Significance Determination P

REDACTED VERSION Page 52 A comprehensive testing and analysis program was conducted to conservatively estimate the risk significance of the ECCS voided piping condition. The scale model testing program simulated bounding conditions and parameters to provide high confidence the air ingestions rates obtained from the tests exceeded the air ingestion rates the ECCS pumps would have actually experienced had an accident requiring containment recirculation actually occurred. Subsequent pump performance tests were conducted under conditions considered to be more severe than would have been experienced during an actual emergency. The results of the pump performance tests were then used in a set of thermal hydraulic analyses of the Palo Verde Reactor Coolant System and Containment. The analyses determined that performance of the ECCS and containment and temperature control functions would have been maintained. For most postulated accidents scenarios, the ECCS safety function would have been maintained by the HPSI pumps. For a subset of SBLOCA scenarios, the ECCS function would have been maintained by the use of any available CS or LPSI pump following RCS cooldown and depressurization by the Plant Operators, if the HPSI pumps were to have failed due to air ingestion. Utilizing the results of the testing and analysis program in a conservative manner, the incremental risk associated with the ECCS voided piping condition is estimated to be 7.0 x IO 4.

REDACTED VERSION Safety Significance Determination

.I

REDACTED VERSION Page 53 I . NUREGICR-2792 An Assessment of Residual Heat Removal and Containment Spray Pump Performance Under Air and Debris Ingesting Conditions. Published September 1982.

2. Paranjape, S. S. et al. 2003. Interfacial Structures in Downward Two-Phase Bubbly Flow. 1 l*

international Conference on Nuclear Engineering (ICONE 1 1 ). Tokyo, Japan.

3. Ishii, M. et.al., 1989 The Three-Level Scaling Approach With Application to the Purdue University Multi-Dimensional Integral Test Assembly (PUMA), d e a r Engineering and Design, 186, pp. 177-21 1

4. Wallis, G.B. Conditions for a Pipe to Run Full When Discharging Liquid into a Space Filled With Gas, Transactions for the ASME, Journal of Fluids Engineering, June 1977, pp. 405413 c 3

6. Wallis, G.B., 1969. One Dimensional Two-Phase Flow, McGraw-Hill, New York.

12. Martin, C. S., 1976, Entrapped Air in Pipelines, Second Intl Conf. on Pressure Surges, Sept. 22-24, London, England, pp. F2-15to F2-28.

13.Chaik0, M. A. and Bnnckman, K. W.,2002, Model for Analysis of Waterhammer in Piping with Entrapped Air, Transactions of the ASME, Journal of Fluids Engineering, 124, pp. 194-204.

14. Lee, N. H. and Martin, C. S., 1999, Experimental and Analytical Investigations of Entrapped Air in a Horizontal Pipe, Proceedings of the Third A S W J S M E Joint Fluids Engineering Conference, July 18-23, San Francisco, California.

15. Henry, R. E., Grolmes, M.A. and Fauske, H. K., 1971, Pressure Pulse Propagation in Two-Phase One- and Two-Component Mixtures, Argonne National Laboratory Report, ANL-7792.

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ATTACHMENT 2-A FA1104-65, Revision 0 Test Report for Phase 1 of Experimental Investigation of Post RAS Air Intrusion into ECCS Suction Piping for Palo Verde Nuclear Generating Station REDACTED VERSION

PRBPWETARY FAIIO4-65 Page 1 of 34 REDACTED VERSION Rev. 0 Date: 09117/04 FAUSKE & ASSOCIATES, INC.

CALCULATION NOTE COVER SHEET 3 T I O N TO BE COMPLETED BY AUTHOR(%

Calc-Note Number FAU04-65 Revision Number 0 Title Test Report for Phase 1 of Expenmental lnvestigation of Post RAS Air Intrusion Into ECCS Suction Piping for Pa10 Verde Nuclear Generatinn Station Project Number or Project Arizona Public Service (APS) Shop Order APSO03

Purpose:

This report documents the scaled experiments that were conducted to investigate a past plant operability question regarding the possibility of any of the air initially residing in the horizontal segment of the sump suction line being swept into the vertical downcomer and subsequently into the ECCS pumps. The nature of the two phase flow pattern produced in the vertical segment was also investigated.

Results Summary: A range of containment overpresswe and system flow rates were investigated. The set of conditions that would be expected for a large break LOCA event were found to result in the air being relocated from the horizontal segment into the vertical segment. The two-phase flow pattern in the vertical segment was seen to be liquid continuous with dispersed air bubbles.

References of Resulting reports, Letters, or Memoranda (Optional)

Author(s): Completion Name (Print or Type) Signature Date R. 1. Hammenlev September 17.2004 ECTION TO BE COMPLETED BY VERIFIER(S):

Verifier(s): Completion Name (Print or Type) Signature Date W. E. Bereer SeDtember 24.2004 Independent Review or Method of Verification: Design Review , Alternate Calculations X , Testing Other (specify)

SECTION TO BE COMPLETED BY MANAGER:

Responsible Manager: Approval Name (Print or Type) Signature Dale

iwHwWw FAII04-65 Page 2 of 34 REDACTED VERSION Rev. 0 Date: 09/17/04 CALC NOTE NUMBER FAVO4-65 PAGE 2 CALCULATION NOTE METHODOLOGY CHECKLIST CHECKLIST TO BE COMPLETED BY AUTHOR(S) (CIRCLE APPROPRIATE RESPONSE)

1. Is the subject and/or the purpose of the design analysis clearly stated? .......................................................................................................... NO

2. Are the required inputs and their sources provided? .................................... NO N/A

3. Are the assumptions clearly identified and justified? .................................... NO N/A

4. Are the methods and units clearly identified? ................................................. NO N/A

5. Have the limits of applicability been identified? ............................................. NO N/A (Is the analysis for a 3 or 4 loop plant or for a single application.)

6. Are the results of literature searches, if conducted, or other background data provided? .................................................................. -NO N/A

7. Are all the pages sequentially numbered and identified by the calculation note number?.. .............................................................................. NO

8. Is the project or shop order clearly identified? ................................................ NO

9. Has the required computer calculation information been provided? ..................... YES NO-IO. Were the computer codes used under configuration control? ............................... YES NO

11. Was the computer code(s) used applicable for modeling the physical andor computational problems identified? ........................................................ YES NO (Le., Is the correct computer code being used for the intended purpose.)

12. Are the results and conclusions clearly stated? ............................................... NO

13. Are Open Items properly identified ................................................................... YES NO

14. Were approved Design Control practices followed without exception? .............YES NO (Approved Design Control practices refers to guidance documents within Nuclear Services that state how the work is to be performed, such as how to perform a LOCA analysis.)

15. Have all related contract requirements been met? ........................................... NO N/A NOTE: If NO to any of the above, Page Number containing justification

+RMwEww FAItO4-65 Page 3 of 34 REDACTED VERSION Rev. 0 Date: 09/17/04 FAVO4-6.5 Test Report for Phase 1 of Experimental Investigation of Post U S Air Intrusion Into ECCS Suction Piping for Palo Verde Nuclear Generating Station Prepared For:

Palo Verde Nuclear Generating Station Arizona Public Service Prepared By:

Fauske & Associates, LLC I6 W070 West 83rd Sfreet Burr Ridge, Illinois 60527

-TEL: (630) 323-8750 FAX: (630) 986-5481 September, 2004

-Pwwww FAI104-65 Page 4 of 34 REDACTED VERSION Rev. 0 Date: 0911 7/04 ABSTRACT This report documents the scaled experiments that were conducted to investigatea past plant operability question regarding the possibility of any of the air initially residing in the horizontal segment of the sump suction line being swept into the vertical downcomer and subsequentlyinto the ECCS pumps. The nature of the two phase flow pattern produced in the vertical segment was also investigated.

A range of containment overpressure and system flow rates were investigated. The set of conditions that would be expected for a large break LOCA event were found to result in the air being relocated from the horizontal segment into the vertical segment. The two-phase flow pattern in the vertical segment was seen to be liquid continuous with dispersed air bubbles.

REDACTED VERSION PURPOSE FAIl04-65 Page 5 of 34 Rev. 0 Date: 09/17/04 The purpose of this report is to document the Phase 1 test conditions and results for the APS experimental investigation of the post RAS air intrusion into ECCS suction piping.

RE F.9Y04-65 Page 6 of 34 Rev. 0 Date: 09117/04 INPUT DATA AND ASSUMPTIONS The Phase 1 experiments were configured and conducted per the approved test plan (FAI, 2004). The initial conditions, major components, and key dimensions for these tests are described in the test plan.

-wawEMw '

REDACTED VERSION FAIl04-65 Page I of 34 Rev . 0 Date: 09/ 17/04 TABLE OF CONTENTS Page CALCULATION NOTE COVER SHEET ............................................................................... 1 CALCULATION NOTE METHODOLOGY CHECKLIST .................................................... 2 TITLE PAGE .............................................................................................................................. 3 ABSTRACT .................................................................................................................................. 4 PURPOSE ..................................................................................................................................... 5 INPUT DATA AND ASSUMPTIONS ........................................................................................ 6 TABLE OF CONTENTS .............................................................................................................. 7 LIST OF FIGURES .................................................................................................................... 9 LIST OF TABLES ................................................................................................................... 10 1 .0 PHASE 1 TEST OBJECTIVES...................................................................................... 11 1.1 Technical Issue .................................................................................................... 11 1.2 Experimental Objectives ..................................................................................... 11 2.0 PHASE 1 TEST FACILITY ........................................................................................... 13 2.1 Physical Arrangement ......................................................................................... 13 2.2 Instrumentation ................................................................................................... 15 2.3 Scaling Considerations........................................................................................ 17 3.0 PHASE 1 TEST MATRIX AND TESTING OBSERVATIONS ................................... 21

3. I Initial Conditions and Test Matrix ...................................................................... 21 3.2 Observations During Phase 1 Testing ................................................................. 21 4.0 PHASE 1 TEST RESULTS ............................................................................................ 25 4.1 Key Observations ................................................................................................ 25 4.2 Discussion of Resul ts .......................................................................................... 25

5.0 CONCLUSION

S ............................................................................................................. 33

FAU04-65 Page 8 of 34 Rev. 0 Date: 09117/04

6.0 REFERENCES

............................. ......._.... _._.. .............................. 34 APPENDIX A: Phase 1 Test Data ........................................................................................... A-1

REDACTED VERSION LIST OF FIGURES FAII04-65 Page 9 of34 Rev. 0 Date: 09/17/04 Page Figure 1 Phase 1 test configuration for post R A S air intrusion ......................................... I4 Figure 2 Two-phase flow patterns reported by [ ] for vertical downflow ............................................................................................... 19 Figure 3A Total flow rate (Tests 1-4) .................................................................................. 29 Figure 3B Total flow rate (Tests 5-8) .................................................................................. 30 Figure 3C Total flow rate (Tests 9-12) ................................................................................ 31

REDACTED VERSION LIST OF T U L E S FAV04-65 Page 10 of 34 Rev. 0 Date: 09117/04 Page Table 1 Instrumentation for Phase 1 Test ........................................................................ 16 Table 2 Phase I Test Matrix ............................................................................................ 22 Table 3 Phase 1 Test Results and Observations ............................................................... 26 Table 4 Valve Opening Times ........................................................................................ 32

+-m"w FAI/04-65 Page I 1 of 34 REDACTED VERSION Rev. 0 Date: 09/17/04 1.0 PHASE 1 TEST OBJECTIVES 1.1 Technical Issue The Palo Verde Nuclear Generating Station (PVNGS) has identified a concern that their sump recirculation flow paths to the Emergency Core Cooling System (ECCS) pumps contain a pocket of air trapped between the sump isolation Motor Operated Valves (h4OVs) and check valve that could potentially be forced into the operating pump suction upon an initiation of a Recirculation Actuation Signal (RAS) during a design basis event. PVNGS has requested analysis of this concern to determine:

(1) If any volume of air between the inboard sump isolation valve and the downstream check valve could be forced into the suction of the operating High Pressure Safety Injection (HPSI) and Containment Spray (CS) pumps upon full opening of the sump isolation vaives at the time of RAS.

(2) The impact on pump performance if any amount of air from the sump suction piping is injected into the operating pumps.

1.2 ExDerirnental Obiectives An experimental investigation has been initiated to address this technical issue and investigate the two-phase flow patterns for the scaled horizontal and downward vertical flow segments. The objective of the Phase 1 testing was to investigate the potential for the air initially resident in the horizontal sump suction line to be forced into the vertical downward piping section.

1

-PRwwww FAIJ04-65 Page 12 of 34 Rev. 0 Date: 09/17/04 REDACTED VERSION

[ 1 The Phase 1 tests were configured and performed in accordance to the approved test plan (FAI, 2004).

FAI/04-65 Page 13 of 34 Rev. 0 Date: 09/17/04 2.0 PHASE I TEST FACILITY 2.1 Physical Arraneernent The test facility that was used for the Phase I testing was composed of two tanks with water inventories, a centrifugal pump, piping, valves, and associated instrumentation as indicated in Figure

1. The piping and valves used to establish and visualize the flow pattern development from the initial location between the valves and into the downcomer piping were all 4 inch in diameter. The horizontal segment [

1 Thevertical [

1 1

REDACTED VERSION FAI104-65 Page 14 of 34 Rev. 0 Date: 09/17/04 Figure 1 Phase 1 test configuration for post RAS air intrusion.

REDACTED VERSION FAl/04-65 Page 15 of 34 Rev. 0 Date: 09/17/04 2.2 Instrumentation The test instrumentation is indicated in Figure 1 and listed in Table 1. A personal computer (PC>[ ] was used to collect data during [

] . Each data channel was sampled at a rate of [ ] The data that was recorded for each test included:

I ] General observations as visible through the clear pipe sections were made and noted by the test engineers.

These observations were used to characterize the air behavior and flow patterns.

REDACTED VERSION Table 1 FAIl04-65 Page I6 of 34 Rev. 0 Date: 0911 7/04 Following the first four tests in the test matrix the test data was reduced and plotted. The results were inspected for internal consistency as well as confirmation of the proper functioning of the instrumentation. The data collected on instrument P4 appeared to be contaminated with excessive noise. [

1 Thus,in addition to relocating the P4 pressure transducer it was reoriented such that instead of being at the

[

1

FAI/04-65 Page 17 of 34 REDACTED VERSION Rev. 0 Date: 0911 7/04 23 Scalinp Considerations

REDACTED VERSION FAIl04-65 Page I8 of 34 Rev. 0 Date: 09117/04

REDACTED VERSION FAU04-65 Page I9 of 34

-mommRF Rev. 0 Date: 0911 7/04

-pRwwww FAIJ04-65 Page 20 of 34 REDACTED VERSION Rev. 0 Date: 09/17/04

3.1 REDACTED VERSION 3.0 PHASE I TEST MA TRCX AND TESTING OBSER VATIONS Initial Conditions and Test Matrix FAI104-65 Page 2 I of 34 Rev. 0 Date. 09117/04 The initial conditions were as follows:

Relative to the elevation o the center line of t , e horizontal segment of t,e pump suction line.

The test matrix as provided in the approved Test Plan was modified based on observations during the Phase I tests by the Westinghouse project team and the APS representatives [

] who were observing the tests. The revised test matrix executed in the Phase 1 testing is provided in Table 2. The key observations for each test included [

3.2 Observations During Phase 1 Testing During the execution of the Phase 1 test matrix several general observations were made in addition to the key object [

1

-?mmEMw REDACTED VERSION FA1104-65 Page 23 of 34 Rev. 0 Date: 09/17/04

REDACTED VERSION F.41104-65 Page 24 of 34 Rev. 0 Date: 09/17/04

wtomrmw FAIf04-65 Page 25 of 34 REDACTED VERSION Rev. 0 Date: 0911 7/04 4.0 PHASE I TESTRESULTS 4.1 Key Observations The key observations for the Phase 1 air intrusion test relate to the specific test objectives.

The objectives are to observe the behavior of the air in the initially voided horizontal segment and the nature of the flow pattern produced in the vertical downcomer segment. The observations fortbe 12 tests performed in the Phase 1 testing regarding these objectives are as follows:

the air initially resident in the voided horizontal segment is removed from the horizontal segment during the initial transient phase, 0 the two-phase flow pattern produced in the vertical segment is found to be liquid continuous with the air dispersed as a bubbly flow.

4.2 Discussion of Results The test data and movies for each of the twelve Phase 1 tests were reviewed. Table 3 summarizes the results of this review. Table 3 includes [

1

REDACTED VERSION 1

+4RwMMw REDACTED VERSION FA110465 Page 27 of 34 Rev. 0 Date: 09117/04

FAY04-65 Page 29 of 34 REDACTED Rev. 0 Date: 09/17/04 Figure 3A: Total flow rate (Tests 1-4).

FAIi04-65 Page 30 of 34 Rev. 0 Date: 09/17/04 Figure 3B: Total flow rate (Tests 5-8).

FA1/04-65Page 3 1 of 34 REDACTED VERSION Rev. 0 Date: 09/17/04 Figure 3C: Total flow rate (Tests 9-12).

REDACTED FAI/04-65 Page 32 of 34 Rev. 0 Date: 09117/04 Table 4

5.0 CONCLUSION

S The Phase 1 tests results lead to the conclusion that the air void initially contained in the horizontal sump suction piping can be swept down and through the vertical piping in the suction line.

FAI/04-65 Page 34 of 34 REDACTED VERSIO Rev. 0 Date: 09117/04

6.0 REFERENCES

FAI, 2004, FAU04-61, "Test Plan for Experimental Investigation of Post I U S Air Intrusion Into ECCS Suction Piping for Palo Verde Nuclear Generating Station," September.

ATTACHMENT 2-B FA1104-86, Revision 0 Test Report for Phase 2 of Experimental Investigation of Post RAS Air Intrusion Into ECCS Suction Piping for Palo Verde Nuclear Generating Station REDACTED VERSION

-REDACTED VERSION Westinghouse Non-Proprietary Class 3 FAIIO4-86 Page 1 of 106 Rev. 0 Date: 0 2 1 1 !05 FAUSKE & ASSOCIATES, LNC.

CALCULATION NOTE COVER SHEET E D I O N TO BE COMPLETED BY AUTHOR(S):

Calc-Note Number FAl/Q4-86 Revision Number 0 Title Test Report for Phase 2 of Experimental lnvestieatron of Post-RAS Air Intrusion Into ECCS Suction P i ~ t for n ~Palo Verde Nuclear Generatinz Station Project Number or Project Arizona Public Service IMS) Shop Order APSOOS

Purpose:

This report documents the scaled integral experiments (Phase 2) that were conducted to investigate a past plant operability question regarding the possibility of air initially residing in the horizontal segment of the sump suction line being swept into the pump suction header and ECCS pumps. The nature of the two phase flow patterns in the ECCS suction piping was also investigated. .

Results Summary:

References of Resulting reports, Letters, or Memoranda (Optional)

Authods): Completion Name (Print or Type) Signature Date Robert J. Hammerslev February I 1.2005 Robert E. Henry Februarv 1 1.2005 ECTION TO BE COMPLETED BY VERIFIER(S):

Verifieds): Completion Name (Print or Tme) Sirmature Date William E. Bemer February I I . 2005 Independent Review or Method of Verification: Design Review , Alternate Calculations X , Testing Other (specify)

SECTION TO BE COMPLETED BY MANAGER:

Responsible Manager: Approval Name (Print or Type) Signature Date R. E. H e w Februarv 1 1.2005 Q 2005 Fauske & Associates, Inc.

All Rights Reserved

REDACTED VERSION w FAI/04-86 Rev. 0 Page 2 of 106 Date: 0211 1/05 CALC NOTE NUMBER FAU04-86 PAGE 2 CALCULATION NOTE METHODOLOGY CHECKLIST CHECKLIST TO BE COMPLETED BY AUTHOR(S) (CIRCLE APPROPRIATE RESPONSE)

I. Is the subject andlor the purpose of the design analysis clearly stated? ................................................................................................. NO

2. M e the required inputs and their sources provided? .................................... NO N/A

3. Are the assumptions clcarly identified and justified? .................................... NO N/A

4. &e the methods and units clearly identified? ................................................ NO NIA

5. Have the limits ofapplicability been identified? ............................................. NO N/A (Is the analysis for a 3 or 4 loop plant or for a single application.)

6. Are the results of literature searches, if conducted, or other background data provided? .................................................................. -NO NIA

7. Are all the pages sequentially numbered and identified by the calculation note number?.. ........................................................................... 0..YES NO

8. Is the project or shop order clearly identified? ................................................ NO

9. Has the required computer calculation information been provided? ..................... YES NO-lo. Were the computer codes used under configuration control? ............................... YES NO 1I. Was the computer code(s) used applicable for modeling the physical andor computational problems identified? ........................................................

(i.e., Is the correct computer code being used for the intended purpose.)

YES NO a

12. Are the results and conclusions clearly stated? ....................................... ..... NO

13. Are Open Items properly identified ................................................................... YES NO

14. Were approved Design Control practices followed without exception? ............. YES NO (Approved Design Control practices refers to guidance documents within Nuclear Services that state how the work is to be performed, such as how to perform a LOCA analysis.)

15. Have all related contract requirements been met? .................................. ..... NO NIA NOTE: If N O to any of the above, Page Number containingjustification

REDACTED VERSION FAI/04-86 Rev. 0 Page 3 of 106 Date: 0 2 I 1/05 FAVO4-86 Test Report for Phase 2 of Experimental Investigation of Post-RAS Air Intrusion Into ECCS Suction Piping f o r Palo Verde Nuclear Generating Station Prepared For:

, Arizona Public Service Prepared By:

Fauske & Associates, LLC I 6 WO 70 Wesf 83rd Street Burr Ridge, Itlinois 60527

-TEL: (630) 323-8750 F U : (630) 986-5481 November, 2004

v F.41/04-86 Page 4 of 106 REDACTED VERSION Rev. 0 Date: 02/11/05 A B S T M CT This report documents the Phase 2 scaled experiments that were conducted to investigate a past operability question for the Palo Verde plants regarding the possibility of the air initially residing in the horizontal segment of the sump suction line being swept into the vertical downcomer and subsequently into the ECCS and Containment Spray (CS) pumps. Phase 1 tests pM, 2004a) addressed the behavior of the vertical downcomer. The nature of the two phase flow pattern produced in the pump suction piping for the High Pressure Safety Injection (HPSI), Low Pressure Safety Jnjection (LPSl), and CS systems was investigated in these Phase 2 tests.

A range of containment overpressure and system flow rates were studied. [

1 Test cases were also included with the HPSI and CS pumps running at the time of U S with the Low Pressure Safetyhjection (LPST) started later. Ln general these tests demonstrated that most of the air was pulled through the WSI suction line before the LPSI pump was started. For most of these tests the HPSI pump was assumed to fail and was shutdown when the flow decreased to one-half of the initial value. Some tests were performed to address the possible operator action of keeping the CS pump on one train and shutting down the CS pump on the other train in favor of the LPSI pump if HPSI were to fail on both trains. With this event sequence, stopping the CS pump enabled the air in the lower header to rise up through the downcomer, pass backward through the check valve and be discharged into the sump thus eventually rising to the containment atmosphere.

Consequently, there was no air in the header when the LPSI pump was started.

REDACTED VERSION PURPOSE FA1/04-86 Page 5 of 106 Rev 0 Date: 02/11/05 This report documents the scaled integral experiments (Phase 2) that were conducted to investigate a past operability question regarding the possibility of air initially residing in the horizontal segment of the sump suction line being swept into the pump suction header and ECCS pumps. The nature of the t w o phase flow patterns in the ECCS suction piping was also investigated

w REDACTED VERSION FAV04-86 Page 6 of 106 Rev 0 Date: OUI 1 /05 INPUT DA TA AND ASSUMPTIONS The Phase 2 experiments were configured and conducted per the approved test plan (FAI, 2004b). The initial conditions, major components, and key dimensions for these tests are described in the test plan

REDACTED VERSION TABLE OF CONTENTS FAV04-86 Rev . 0 Page 7 of 106 Dale: OUl1/05 Pao,e CALCULATION NOTE COVER SHEET ............................................................................... 1 CALCULATION NOTE METHODOLOGY CHECKLIST .................................................... 2 TTTLE PAGE ............................................................................................................................. 3 ABSTRACT .............. ............................................................................................................. 4 PURPOSE ..................................................................................................................................... 5 INPUT DATA AND ASSUMPTIONS ........................................................................................ 6 TABLE OF CONTENTS .............................................................................................................. 7 LIST OF FIGURES .................................................................................................................... 9 LIST OF TABLES ................................................................................................................... 12 1 .0 PHASE 2 TEST OBJECTIVES.................................... .............................................. 13 1.1 Technical Issue .................................................................................................... 13 1.2 Experimental Objectives ..................................................................................... 13 2.0 PHASE 2 TEST FACILITY ........................................................................................... 15 2.1 Physical Arrangement ......................................................................................... 15 2.1.1 Configuration 2A ......................... ........................................................ 14 2.1.2 Configuration 2B ................................... .............................................. 18 2.1 -3 Configuration 2C ..................................................................................... 21 2.2 Instrumentation ................................................................................................... 26 2.3 Scaling Considerations........................................................................................ 28 2.3.1 Two-Phase Flow Pattern Considerations ..........................

2.3.2 Vertical Scaling of Two-Phase Downward Flow ..................................... 29

REDACTED VERSION 2.3.3 Scaling of the Initial Air Volume and the Isolation Valves .....................

FAI104-86 Rev . 0 Page 8 of 106 Date: 021 I 1/05 33 2.3.4 Materials ........................................................................................

3.0 PHASE 2 INITLZL CONDITIONS AND TEST M A T m .......................................

4.0 PHASE 2 TEST RESULTS ............................................................................................ 43 4.1 Configuration 2A ................................................................................................ 43 4.1 . 1 Key Observations ..................................................................................... 43 4.1.2 Discussion of Results ....................... ................................ .......... 49 4.2 Configuration 2B ................................................................................................ 94 4.2.1 Key Observations ..................................................................................... 94 4.2.2 Discussion of Results ................................................. ..........................94 4.3 Configuration ZC ................................................................................................ 96 4.3.1 Key Observations ..................................................................................... 96

5.0 CONCLUSION

S ............................ ............................................................................... 101

6.0 REFERENCES

............................................................................................................. 105 APPENDIX A: Phase 2 Configuration 2A Test Results......................................................... A-1 APPENDIX B: Phase 2 Configuration 2B Test Results ......................................................... B-1 APPENDIX C: Phase 2 Configuration 2C Test Results ......................................................... C-1 APPENDED: [

1 ........................................................................................... D-I APPENDIXE:

] ........................................................................................................ E-1

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LIST OF TABLES Table 1 Test Dimensions.................................................................................. ...._... . _...I 7 Table 2 Instrumentation for Phase 2 Test ......................... ........................ 27 Table 3 Phase 2 Test Matrix for Configuration 2A ......................................................... 40 Table 4 Phase 2 Test Matrix for Configuration 2B 41 Table 5 Phase 2 Test Matrix for Configuration 2C .......................................................... 42 Table 6 Phase 2 Configuration 2A Test Results and Observations 44 Table 7 Phase 2 Configuration 2B Test Results and Observations Table 8 Phase 2 Configuration 2C Test Results and Observations _...._...__._._..._._____.._.___.. 47 Table 9 Summary of Tests for Reproducibility ................................................................ 90

REDACTED VERSION I.0 P m S E 2 TEST OBJECTIVES FAIl04-86 Rev. 0 Page I3 of 106 Date: 0211 I105 1.1 Technical Issue The Palo Verde Nuclear Generating Station (PVNGS) has identified a concern. Specifically, all three units have sump recirculation flow paths to the Emergency Core Cooling System (ECCS) pumps which contain a pocket of air trapped between the sump isoIation Motor Operated Valves (MOVs) (butterfly valves) and check valve that could potentially be forced into the operating pump suction upon an initiation of a Recirculation Actuation Signal ( U S ) during a design basis event.

PVNGS has requested analysis of this concern to determine:

(1) If any air volume between the inboard sump isolation valve and the downstream check valve could be forced into the suction of the operating High Pressure Safety Injection (HPSI) and Containment Spray ( C S ) pumps upon opening of the sump isolation valves at the time of RAS.

(2) The impact on pump performance if any amount of air from the sump suction piping is injected into the operating pumps.

1.2 Experimental Objectives Phase 1 testing (FAI, 2004a) demonstrated that the flow demand on the containment sump pump suction line following RAS was suficient [

] Therefore, Phase 2 experimental investigation was initiated at FAI to investigate the two-phase flow patterns [

] The objectives of the Phase 2 testing were to investigate the extent of air transport to the HPSI and CS pumps as well as the LPSI pump for those accident sequences where this could be started. Full scaling testing of the pump performance for the resulting air intrusion will be performed in a Phase 3 test facility at Wyle Laboratories in Huntsville, Alabama. In the Phase 2 testing, the nature of the flow pattern (dispersed bubbly flow, plug flow, slug flow, etc.) at the pump suctions

REDACTED O

-N V FAU04-86 Page I4 of IO6 Rev. 0 Date, 0211 1/05 will be observed including the transient effects of switching the water supply fiom the simulated Reactor Water Tank (RWT) to the containment sump while simultaneously opening the sump suction isolation (butterfly) valves. Transparent piping was used for the horizontal and vertical segments of the simulated pump suction line to observe and record the flow pattern and the behavior b - 7 of the initial air filled volume The Phase 2 tests were configured and performed in accordance with the approved test plan (FAI, 2004b).

FM04-86 Page I5 of 106 REDACTED VERSION Rev. 0 Date 0211 1/05 2.0 PHASE 2 TEST FACILITY 2.1 Physical Arrangement 2.1.1 Confieuration 2A

[

The use of 4 inch diameter (Schedule 40) pipe to represent the 24 inch diameter (Schedule 20 and 30) pipe in the plant defmed a linear scale ratio of approximately 1/6 (FAI, 2004b). Thus, the balance of the suction line pipe lengths and valve locations also used a 1/6th scale unless there were other considerations [

1

-mawww REDACTED VERSION FAU04-86 Page 16 of 106 Rev. 0 Date: 0211 1/05 Figure 1: Phase 2 Test Configuration 2A for Post-FUS Air Intrusion.

REDACTED VERSION FAI/04-86 Page 17 of I06 Rev 0 Date: 02/11/05

] Both the HPSI and CS pumps are single stage centrifugal pumps in the test apparatus. For the plants, the HPSI umps are eight stage centrifugal designs.

Table 1 Test Dimensions

+RmmwF F ~ I l 0 4 - 8 6 Page 18 of 106 REDACTED VERSION Rev. 0 Date: 021 I 1\05 1

2.1.2 Confieuration 2B

v FA1/04-86 Page 19 of 106 REDACTED VERSION Rev. 0 Date 0 2 1 1/05 Figure 2:

[ I

REDACTED VERSION FA1/04-86 Rev 0 Page 20 of 106 Date 0 2 1 1/05 Figure 3: Phase 2 Test Configuration 2B for Post-RAS Air Intrusion.

REDACTED VERSION FA1/04-86 Page 2 1 of 106 Rev. 0 Date: 02/ 1 1/05 2.13 Confieuration 2C

REDACTED VERSION FA1104-86 Rev. 0 Page 22 of 106 Date: 0211 1/05 Figure 4: Phase 2 Test Configuration 2C for Post-R4S Air Intrusion.

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-+wwwww FAV04-86 Page 26 of 106 REDACTED VERSION Rev. 0 Date: 021 1105 2.2 Instrumentation The test instrumentation is similar for all three test configurations and is indicated in Figures 1, 2 and 3 and listed in Table 2. A personal computer (PC) [

] was used to collect data during the transient following the opening of the isolation valves as well as the subsequent steady state recirculation flow that followed. Each data channel was sampled at a rate of once per [ ]which is much faster than the hydraulic transient which takes tens of seconds. ]

] Each experiment had the following data recorded:

REDACTED VERSION u

m

REDACTED VERSION FAI/04-86 Rev. 0 Page 28 of 106 Dale: 0211 l l O S Digital movie cameras were used to record the flow patterns in the clear piping sections. General observations in the clear pipe sections were made and noted by the test engineers. These observations were particularly important to characterize the water-air flow patterns in the various suction pipes.

2.3 Scaling Considerations The test plan (FAI, 2004b) presented the scaling assessment for the Phase 2 tests. T h e scaling assessment addressed 1

] The scaling considerations are discussed below.

2.3.1 Two-Phase Flow Pattern Considerations

REDACTED VERSION FAI104-86 Rev 0 Page 29 of 106 Date 0211 I /OS 2.3.2 [ 1

REDACTED VERSION FA110446 Page 30 of 106 Rev. 0 Date: 02/11/05 Figure 6: Terminal velocity of air bubbles in filtered or distilled water as function of bubble size reported by Haberman and Morton and shown in Wallis (1969).

6ov 50 I I I I L l l i

1. I I I I I I I l l I 0.01 0.02 0.04 0.06 0.1 0.2 0.4 0.6 0.8 1.0 2.0 4 .Q Equivalent mdius, Rbr cm

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REDACTED VERSION FA1/04-86 Rev. 0 Page 37 of 106 Date: 02/11/05 23.4 Materials Like the postulated accident, water and air are the fluids used in the void behavior and flow pattern observation experiments. Visual observations of the air-watertwo-phase flow patterns in the plastic piping provided the insights needed for the Phase 3 testing program on full scale pumps. For accident conditions, the plant sump water temperature would be elevated and the sump water could also contain chemicals such as boric acid and trisodium phosphate (TSP) due to the sources ofwater that accumulate in the containment and the sump pH control. Prototypic concentrationsofboric acid and TSP were investigated in separate phenomenological tests and found to ,be- i

.tA REDACTED VERSION 1

FAV04-86 Rev. 0 r

Page 38 of 106 Date: 02/11/05 The approach to address sump water temperature is discussed is Section 4, Phase 2 Test Results.

+wwwFMY REDACTED VERSION FAV04-86 Page 39 of 106 Rev. 0 Date: 02/ 1 1/05 3.0 PHASE 2 INITIAL CONDITIONS AND TEST MATRCY The range of initial conditions were as follows.

The test matrices for Configurations 2A, 2B and 2C as provided in the Phase 2 Test Plan C A I ,

2004b) are reproduced in Tables 3 , 4 and 4 respectively. With the observations of the initial tests, the test matrix was expanded during the testing program to investigate specific phenomena as well as demonstrate reproducibility of the results. The expanded test matrix executed in the Phase 2 testing is presented in Section 4.0, Phase 2 Tests Results. A cross reference is provided between the expanded test matrix and the test matrix from the test plan. Key observations for each test include the two-phase flow pattern [ ] Other observations include [

I Relative to elevation of center line of the lower horizontal header for the HPSI, CS and LPSI pump suction lines.

REDACTED VERSION

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m

REDACTED VERSION 4.0 PHASE 2 TEST RESULTS FAI104-86 Rev. 0 Page 4 3 of 106 Date 0211 1/05 The test data and digital movies of the Phase 2, Configurations 2A were reviewed for the tests specified for this configuration in the Phase 2 Test Plan (FAX, 2004b). After reviewing and discussing the results from the original twelve tests for Configuration 2A with APS and Westinghouse personnel, it was decided to expand the Configuration 2A test matrix to 29 tests.

Table 6 summarizes the results for all 29 of the Configuration 2A tests [

Upon the completion of the expanded set of Configuration 2A tests, review of the experimental data and thedigital video recordsas well as other supportingplant analyses (Phase 4 of the overall program), it was decided to investigate two other pump combinations. The results for the Configuration 2B and 2C experiments are summarized in Tables 7 and 8. Observations and insights gained from these configurations are discussed after those resulting from Configuration 2A.

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$2 da

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W

REDACTED VERSION 4.1 4.1.1 Configuration 2A Key Observations REDACTED VERSION FAI/04-66 Rev. 0 Page 49 of 106 Date: 02/11/05 The key observations for the Phase 2 air intrusion tests relate to the specific test objectives, i.e. (1) to investigate the air delivery rates to the HPSI and CS pump suctions and (2) document the associated two-phase flow patterns. Observations from the 29 tests performed in Configuration 2A of the Phase 2 testing are as follows:

4.1.2 Discussion of Results 4.1.2.1 General Comments

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REDACTED VERSION FAI/04-86 Rev. 0 Page 75 of 106 Date: 02/1 1/05 where:

Av is the cross-sectional flow area for the 8 inch pipe, 0 hl is the initial water height, and h2 is the measured water height at a later time.

Knowing the volume change means that the mass can be obtained by multiplying this with the air density which can be calculated from the perfect gas law, i.e.

where:

0 P,k is the total pressure for the air in the gas separator, 0 MW,, is the molecular weight of air (29.2),

0 R, is the universal gas constant, and 0 T, is the absolute temperature of the gas.

(To avoid confusion in units, these parameters are evaluated in the international system of units and then converted to British units once the flow rate is determined.) Hence, the collected air mass is Differentiating this with respect to time produces the air m a s flow rate into the separator.

Figures 25 and 26 r' . .. . _.

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me---- - -.-._

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1

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REDACTED VERSION FAIl04-86 Rev. 0 Page 95 of 106 Date: 021 1/05 4.2 Confieuration 2B 4.2.1 Key Observations 4.2.2 Discussion of Results 4.2.2.1 General Comments These scoping tests included 8 experiments with the principal difference being the predetermined LPSI flow rate. Table 7 summarizes the results for all tests including the as-tested flow rates for both pumps and the corresponding Froude numbers in the different piping segments.

1

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5.0 CONCLUSION

S FAI/04-86 Rev. 0 Page 102 of 106 Date: 02111105 The following conclusions were derived from the three Phase 2 configurations for the integral 4 inch diameter scaled experiments representing the Palo Verde sump suction line behavior.

I. ConfiEuration 2A

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11. Configuration 2B 1

III. Configuration 2C

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6.0 REFERENCES

FAI. 2004a, "FAU04-65,"Test Report for Phase I of Experimental -1ves1igation of Post R S Air Lntrusion Lnto ECCS Suction Piping for Palo Verde Nuclear Generating Station, September.

FAI, 2004b. FAU04-74, "Test Plan (Phase 2) for Experimental Investigation of Post-RAS Air Intrusion Lnto ECCS Suction Piping for Palo Verde Nuclear Generation Station," December.

FAI, 2005, FAY04-79, "Test Report: Phenomenological Studies for the APS Containment Suction Line," January.

NRC, 1982, "An Assessment of Residual Heat Removal and Containment Spray Pump Performance Under Air and Debris Ingesting Conditions," NUREGKR-2792, September.

Ricou, F. P. and Spalding, D. B., 1961, "Measurements of Entrainment by Axisymmetrical Turbulent Jets," Jr. of Fluid Mechanics, Vol. 1 1, pp. 21 -32.

Wallis, G. B., 1969, One-Dimensional Two-Phase Flow, McGraw-Hill, New York.

Wallis, G. B. et al., 1977, "Conditions for a Pipe to Run Full When Discharging Liquid Into a Space Filled With Gas," Trans. ASME, Jr. of Fluids Engineering, June 1977, pp. 405-413.

ATTACHMENT 2-C Test Report 10530R01 Test Report for Testing of a CA Pump and WDF Pump with a Void Fraction Inlet Fluid Condition REDACTED VERSION

REPORT NO. 10530R01,Revision 0 WYLE JOB NO. 10530 CUSTOMER P.O.NO. 500281 122 TEST REPORT PAGE 1 OF 137 PAGE REPORT DATE 01/18/05 SPEClFICATION(S) WLTP 10530TP,Revision 0 TEST REPORT FOR TESTING OF A CA PUMP AND A WDF PUMP WITH A VOID FRACTION INLET FLUID CONDITION Arizona Public Service (APS)

Test Report 1053ORO1 Page ii REDACTED VERSION I THIS PAGE INTENTIONALLY LEFT BLANK I Wyfe Laboratories Huntsvilie Facllity

Test Report 10530ROl Page iii TABLE OF CONTENTS REDACTED VERSION Section Page 1.0 CUSTOMER ............................................................................................................................. 1 2.0 TEST SPECIMEN..................................................................................................................... 1 3.0 MANUFACTURER ................................................................................................................... 1 4.0

SUMMARY

.................................................... 1.......................................................................... 1 5.0 EQUIPMENT DESCRIPTION................................................................................................... 1 6.0 TEST SEQUENCE ................................................................................................................... 2 7.0 PROCEDURES ........................................................................................................................ 3 8.0 RESULTS................................................................................................................................. 3 9.0 QUALITY ASSURANCE........................................................................................................... 3 10.0 TEST EQUIPMENT AND INSTRUMENTATION...................................................................... 3 11.0 APPLICABLE DOCUMENTS AND REFERENCES ................................................................. 3 Attachments AITACHMENT A1 RECEIVING INSPECTION...........................................................................A1-1 ATTACHMENT A2 TEST FACILITY DESCRIPTION. EQUIPMENT SET UP FOR TESTING AND INSTRUMENTATION................................................................................... A2-1 AITACHMENT A3 CHECKOUT TESTING ................................................................................. A3-1 ATTACHMENT A4 PERFORMANCETESTING ......................................................................... A4-1 AlTACHMENT A5 EQUIPMENT INSPECTION AND PREPARATION FOR SHIPPING ...........A5-1 AlTACHMENT A6 PHOTOGRAPHS ......................................................................................... A6-1 ATTACHMENT A7 VIBRATION TEST REPORT ........................................................................ A7-1 Wyle hboratorier Huntsville Facility

Test Report 10530R01 Page iv REDACTED VERSION I THIS PAGE INfENTlONALLY LEFT BLANK I Wyle Laboratories Huntsville Faclllty

1.o CUSTOMER ADDRESS Test Report 10530ROi Page 1 Arizona Public Service Co.

REDACTED VERSION Palo Verde Nuclear Generating Station Tonpah, AR 85354.

2.0 TESTSPECIMEN The equipment to be tested consists of two ESF (Emergency Safety Feature) pump / motor assemblies; the CA pump and motor and the WDF pump and motor.

3.0 MANUFACTURER The pumps were manufacturedby Ingersoll-Rand. The motors were manufactured by Westinghouse.

4.0

SUMMARY

This document has been prepared by Wyle Laboratories to document the results of a test program on the CA and WDF pumps and motors to determine the performancewith a void fraction inlet fluid condition.

This testing was performed in accordance with Wyle Laboratones Test Procedure 10530TP, 'Test Procedure for Testing of a CA pump and a WDF pump with a Void Fraction Inlet Fluid Condition'. The testing meets the requirements of the APS Purchase Order 500281122.

5.0 EQUIPMENT DESCRIPTION The equipment to be tested consists of two pump / motor assemblies; the CA pump and motor, and the WDF pump and motor.

Description:

The equipment description is as follows:

CA Motor (CA):

Westinghouse Electric Frame 5810H Class 1E Rated at 1000 HP, %Phase, 60 Hz,4000 Volts Speed: 3553 rpm Weight 4,800 Ibs Motor Identification Number: 17535LN01 Pump (CA):

4x11CA-8 Nameplate Head = 2850 ft Horizontal shaft Nameplate Rated flow = 900 gpm Weight: 4.400 Ibs Suction diameter: 10" sch 40 Discharge diameter 4" sch 80 Wyle Laboratories Huntsvllle Faclllty

Test Report 1053OROq I .

Attachment A I m f i r r o # qn t h Page A I -2 REDACTED VERSION RECEIVING INSPECTION DATA SHEET PUMP DATA TAG NO.: 087634 SERIAL NO: 087634 MANUFACTURER: Inaersoll-Rand RATED FLOW: A300 wrn NOMINAL SIZE: 8 x 20 WDF SHUT OFF 335 ft HEAD:

END CONNECTION: 14" 30W inlei 18" outlet MOTOR DATA MANUFACTURER: Westinqhouse FRAME: 55010-P39

~

MODEL#: (ID) VSWF SERIAL #: 1s-78 INS. CLASS: B VOLTAGE: 4000 CURRENT @RATED 62 SPEED: 4 776 VOLTAGE FREQUENCY: 60 Hz DESCRIBE CONDITION OF RECEIVED ITEM:

Motor received on metal pallet marked WB. PumD c a s h on pallet #1 B.

Received bbx containina diffuser. DumD seal DiDina. struts. electric box and Wyh Laboratorler Huntsville Facllity

Test Report 10530ROl Attachment A1 Page A 1 3 RECEIVING INSPECTION DATA SHEET PUMP DATA TAG NO.: N/A SERIAL NO: 1178141547 MANUFACTURER: Inqersoll-Rand RATED FLOW: 900 qpm NOMINAL SIZE: 4x11 CAx8 SHUTOFF b.-

2850 ft HEAD:

END CONNECTION: . 10' 300# inl,et

/4" 150#

outlet MOTOR DATA MANUFACTURER: Westinqhouse FRAME: 5810H MODEL #: (ID) HSW2 SERIAL #: 17535LN01 INS. CLASS: F VOLTAGE: 4000 CURRENT @RATED 123 SPEED: 3553 VOLTAGE FREQUENCY: 60 Hz DESCRIBE CONDITION OF RECEIVED ITEM:

PiDe on bottom of pump appears to be bent. Miscellaneous parts with Dump include dates and all thread. Seals. aaskets. and spare bearinqs included. SS shaft is s/n 557. Coupler sleeve and coupler both have taqs with 62013784.

Motor received on Dallet.

wyle Laboratories Huntsville Facility

Test Report 10530ROl Page 4 REDACTED VERSION ITHIS PAGE INTENTIONALLYLEFT BLANK I Wyre Laboratories Huntsville Facility

Test Report 10530R01 Attachment A l REDACTED VERSION Page A I -1 Attachment A I Receiving Inspection RESULTS Receiving inspections were performed on November 22.2004 for both the CA and WDF motor and pump assemblies upon receipt at Wyle Laboratories in accordance with section 3.1 of Wyle LaboratoriesTest Procedure No. 10530. Revision 0.

The CA pump and motor arrived as two individual pieces. The coupling and miscellaneous spare parts were supplied with the pump.

The WDF pump and motor amved as three pieces; the inlet piping and pump casing assembly, the motor assembly and a box of miscellaneous parts including the seal piping and impeller.

The nameplate data and results of the inspection were recorded on the attached Receiving Inspection Data Sheet The specimen pump and motor assemblies were as described in paragraph 5.0 of this report.

Wyl. labomtories Huntsville Facllfty

Test Report 10530R01 REDACTED VERSION Attachment A2 Attachment A2 Test Facility Description, Equipment Setup for testing and Instrumentation RESULTS Test Facility Description

-The test facility is a two closed loop system consisting of a 30,000 gallon pressure vessel with one loop for each test specimen pump. One loop is the test loop for the CA pump motor and the piping and control valves are sized based on the supplied pump curve. The second loop is the test loop for the WDF pump / motor and the piping and control valves are sized based on the supplied pump curve.

No provision is provided for fluid cooling or heating. The pressure vessel also has the ability to be pressurized to a specified pump net positive suction head. This pressure vessel pressure can be adjusted and controlled.

The test medium is de-ionized water under ambient conditions.

The overall test facility is illustrated in Figure 1 in Attachment A6. [

1 Wyle Laboratories Huntsville Facility

Test Report 10530R01 Attachment A2 Page A2-2 REDACTED VERSION 1

CA Pumphdotor Equipment Setup for Testing The CA pump and motor were setup in accordance with section 3.2 of Wyle Laboratories Test Procedure No. 10530TP, Revision 0.

From December 01 to 09,the following activities were completed:

The pump intemals and visible adjacent inlet piping were inspected to ensure cleanliness and no visual damage.

The pump and motor skid to the lo" 300# RF ANSI inlet flange and the 4" 1500# RF ANSI outlet flange was installed.

Correct connection of seal flush piping was verified.

The pump casing was filled and vented with water.

Oil for the pump and motor was installed and verified by inspection of the site glass at the pump, inboard motor and outboard motor locations.

The coupling installation was performed under the guidance of a representative.

The alignment and coupling of the motor to the pump was be performed under the direction of q-!personnel.

0 The instrumentation as listed below was installed.

Wyle Laboratories Huntsville Facility

Test Report 10530ROl Attachment A2 Page A 2 3 REDACTED VERSION Instrumentation Following the CA test specimen pump and motor installation and alignment, the instrumentation was installed.

The following table summarizes the instrumentation used for the test program and the identification numbers (TAG) used by Wyle Laboratories:

CA Pump Loop Instrumentation:

.n; Wyle Laboratorlet Huntsville Facility

Test Report 1053ORO1 Attachment A2 Page A24 REDACTED VERSION

Test Report 1053OROI Attachment A2 Page A24 REDACTED VERSION All Wyle Laboratories' test equipment is calibrated on a periodic basis with the calibration interval displayed on a decal. This decal is affixed to the equipment indicating the last calibration date, the next calibration due date, accuracy, and by whom calibrated. The instrumentation equipment sheet for all the instrumentation is presented in this attachment.

In addition to individual component calibration, prior to and immediately following the test series, an end-bend system calibration was performed on the pressure transducers.

WDF PumplMotor Equipment Setup for Testing The WDF pump and motor were setup in accordance with section 3.3 of Wyle Laboratories Test Procedure No. 10530TP, Revision 0.

From December 09 to 13, the following activities were completed:

The pump intemals and visible adjacent inlet piping were inspected to ensure cleanliness and no visual damage.

The pump casing was installed to the 14' 300# RF ANSI inlet flange and the 8" 30W RF ANSI outlet flange.

The casing and casing studs and gasket surfaces were inspected for cleanliness and no visual damage.

The rnotor/stuffing box/rotating element was installed into the casing under the direction -of personnel.

The pump casing was filled and vented With water.

Oil for the motor was installed at the proper level.

Correct connection of seal flush piping was verified.

The instrumentation as listed below was installed.

wyle Laboratories Huntmdlle Faclllty

Test Report 10530R01 1 *-@I J & a b = -hhLI Attachment Az Page A26 REDACTED VERSION Instrumentation Following the WDF test specimen pump and motor installation, the instrumentationwas installed.

wyle taboratories Huntsville Facility

Test Report 10530R01 Attachment A2 Page A2-7 REDACTED VERSION All Wyle Laboratories test equipment is calibrated on a periodic basis with the calibration interval displayed on a decal. This decal is affixed to the equipment indicating the last calibration date, the next calibration due date, accuracy, and by whom calibrated. The instrumentation equipment sheet for all the instrumentation is presented in this attachment.

In addition to individual component calibration, prior to and immediately following the test series, as. end-to-end system calibration was performed on the pressure transducers.

Wyle Laboratories Huntsville Facility

Key of Attachments:

Test Report 10530ROl Attachment A2 Page A24 REDACTED VERSION Instrumentation Sheet for Test Program. (3 pages)

Calibration Data for the Turbine Flow Meters. (2 pages) wyle Laboratories Huntsville Facility

Test Report 10530R01 Attachment A2 Page A2-9 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 1053OROl

-MeMEww Attachment A2 REDACTED VERSION Page A2-10 Wyie Laboratories Huntsville Facility

Test Report 1053OROl Attachment A2 Page A2-11 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 10530ROl Attachment A2 REDACTED VERSION Page A2-12

-pRBpRcHARy Wyle Laboratodes Huntsvllle Facllity

Test Report 10530R01 Attachment A2 Page A2-13 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 1053OROl Attachment A3 Page A34 REDACTED VERSION Attachment A3 Checkout Testing RESULTS Prior to the actual testing, a test facility and test specimen check out was performed to v e m facility capabilities, test specimen operation and instrumentation functionality for the two test loops.

Du&g this checkout test program, the data channels were acquired at@zc) samples per w m p l e frequency. A Test Log datasheet waslped to record t&i run descriptions, a s m l l as test data and time information and thwmbient temperature, pressure and flow'wnditions. The test log datasheets obtained during the check out testing are presented in this attachment.

Note that throughout testing, two successive starts from ambient temperature are permissible provided the motor is allowed to fully coast down between starts.

After two successive starts, the motor shall be idle for 30 minutes between addMona1starts.

Initially, the motor was bumped to check motor rotation for both pumps.

Prior to the motor/pump check out testing, a check list was used to v em that the test facility, test specimen and instrumentation were correctly configured to begin the test A copy of the Check list for Start Up is presented in this Attachment as an illustration.

A total of five shakedown runs were performed on the pump/motor specimens as documented in the attached test log from 12/11/04 to 12/13/04. These runs were performed to verify proper facility operation, instrumentation functionality aBd test specimen performance. These tests were recorded as data files as follows:

wylo Laboratories Huntsville Fadlity

Test Report 10530RO1 Attachment A3 Page A3-2 w REDACTED VERSION Run Test Date Data File Name Notes

~ ~ _ _

1 IpsicheckOl First motor/pump test on the WDF pump.

2 12/11/04 Ipsicheck02 Longer duration test run on the WDF pump.

3 12111/04 hpsicheckout01 Short Motor bump test on the CA pump.

4 12/11/04 hpsicheckout02 Long duration test to ensure the required pump curve range is achievable for the CA pump.

5 12/13/04 hpsicheckout04 Used to adjust manual valve position for pump run out protection and to check out air injection system for the CA pump-IpsicheckoutOl Checkout test prior to actual performance testing on the WDF Pump-In all cases, the data files have been supplied to APS separately.

Bote that in checkout runs 1 -4 above, a 14 and 10 strainer were installed in the WDF and CA pump test loops respectively to ensure debris removal in the water inventory.

The data taken during run 4 (hpischeckout02) served to provide a CA performance pump curve prior to the air injection test program. This data is evaluated and compared to the pump performance curve after air injection in Attachment A 5 An instmmentation equipment sheet for the testing is presented in attachment A2.

Wyre Laboratories Huntsville Facility

Test Report 10530R01 REDACTED VERSION Attachment A3 Page A3-3 Key of Attachments:

Test Log Sheets for the Check Out Testing (4 pages)

Start up Check list (CA pump test) (2 pages)

Wyle Laboratories Huntsville Facllity

Test Report 10530R01 Attachment A3 Page A34 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 10530R01 Attachment A3 Page A3-5 REDACTED VERSION Wyle Laboratories Huntsville Facllity

Test Report 10530ROl Attachment A3 Page A3-6 REDACTED VERSION Wyle Laboratories Huntsvllle Facility

Test Report 10530R01 Attachment A3 Page A3-7 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 10530R01 Attachment A3 Page A34 Wyle Laboratories Huntsville Facility

Test Report 10530R01 Attachment A3 Page A3-9 REDACTED VERSION 1

we Laboratories Huntsville Facility

Test Report 10530R01 Attachment A4 Page A4-1 REDACTED VERSION Attachment A4 Performance Testing RESULTS The intent of the testing was to determine if temporary performance degradation occurs during the ingestion of a void fraction, and to identify any permanent degradation of performance after un-voided inventory returns to the pump.

A summarized test matrix for both pumps is presented in this Attachment.

puring the test program, t t p data channels described in Attachment A3 were acquired kt ten samples per second4by the Wyte Laboratories data acquisition system. A Test l o g datasheet was used to record test run descriptions, as well as test data and time information and the ambient temperature, pressure and flow conditions. The log is presented in Attachment A3.

Note that throughout testing, two successive starts from ambient temperature are permissible provided the motor is allowed to fully coast down between starts.

After two succewive starts, the motor shall be idle for 30 minutes between I -

additional starts. -

The instrumentationkquipment sheet for this testing is presented in Attachment A2.

&nor to the motor/pump performance testing, a check list was used to verify that the test

' racility, test specimen and instrumentationwere correctly configured to begin the test. .

Actual Test Matrix Throughout the test program, the required data described in Attachment A2 was recorded. This data covers the complete test proqpm. Note that the test matrix presented here represents the target data for testing. ' Actual durations and peakmass flow rates were evaluated separately by APS and are not presented in this report.

The actual test data files consisting &deos of the voided fluid at the sight glass during each test, digital data for the instrumdation listing and vibration data were transmitted to APS, as documented in Wyle Transmittal No. 1053OW-03 dated 1/06/05 for the complete test program.

• Wyle Laboratories Huntsville Facility

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-wwRmww REDACTED VERSION

Test Report 10530R01 Attachment A4 REDACTED VERSION Page A 4 4 A summary of the actual test data plots is presented here for the following test cases; 1D rerun, 2E, 3C and 4B.

Wyle Laboratories Huntsville Facility

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REDACTED VERSION Test Report 10530R01 Attachment A5 REDACTED VERSION Page A54 Attachment A5 Equipment Inspection and Shipment RESULTS As noted in the Test Matrix in Attachment A4, a post test performance test was performed on the CA pump.

The pump head cuwes were developed for both the checkout02 test and the post test results were the pump was operated over the required range of flow rates.

The pump head curves are attached.

Based on the results of this performance test, not CA pump degradation was observed.

Based on the results of the air injection test on the WDF pump, no inspection was required.

Therefore no inspection of the CA or WDF pump was performed.

The motor of the CA pump was removed from the test loop and is currently in storage.

Wyle Laboratories Huntsville Facliity

Test Report 10530ROl Attachment A5 REDACTED VERSION Page A5-2 1

Wyle Laboratories Huntsville Faclilty

Test Report 10530R01 Attachment A6 Page A6-2 REDACTED VERSION Figure I- Overview of the Test Facility with the 30,000 gallon pressure vessel and Enclosure containing the two Test Loops and two Test Specimens.

Wyle Laboratories Huntsville Facility

REDACTED VERSION

Test Report 10530ROl Attachment AB Page A 6 4

-REDACTED VERSION Figure 3 - Photograph showing the Installation in the Test Loop for the WDF Pump and Motor Test Specimen Wyle Laboratofies Huntsville Faclllty

REDACTED VEWSiQN Test Report 10530R01 Attachment A6 Page A 6 4 REDACTED VERSION 1

Wyle Lsboratories Huntsville Facility

ir-REDACTED VERSIOM

Test Report 10530R01 Attachment A6 Page A6-8 REDACTED VERSION Wyle Laboratories Huntsville Facility

Test Report 10530R01 Page A640A6 Attachment -

REDACTED VERSION Figure 9 - Photograph showing the orifice plate assemblv and differential pressure transducer for water flow rate instrumentation. .. ,

WyreLaboratories Huntsvllle Facility

Test Report 10530ROl Attachment A6 Page A6-12

-REDACTED VERSION Figure 11 - Photograph showing the location of the Tqaxial accelerometer for Bearing 2 on the CA pump outboard thrust bearing.

we Laboratorh Huntsville Facility

REDACTED VERSIUN Test Report 10530ROl Attachment A6 Page AB-14 w

REDACTED VERSION Figure 13 - Photograph showing the location of the Triaxial accelerometer for Bearing 4 on the CA pump motor outboard bearing.

wyre Laboratorler, Huntsville Facility

Test Report 1053OROl Attachment A7 Page A74 REDACTED VERSION Attachment A7 wyle Laboratorler Huntsville Facility

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w REDACTED VERSION

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Test RepoFt 10530R01 mtM%ww Attachment A7 REDACTED VERSION Page A74 1 Wyle Laboratories Huntsville Facllity

Test Report 10530ROl Attachment A7 Page A7-13 REDACTED VERSION Wyle Laboratories Huntsville Facility

?

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Test Report 10530R01 -?e&-

Attachment A?

Page A7-45 REDACTED VERSION wyie Laboratories Huntsville Facltlty

Test Report 1053OROl Attachment A7 Page A747 REDACTED VERSiON Wyle Laboratories Huntsville Faclllty

REDACTED VERSION Test Report 1053oROl v Attachment A? REDACTED VERSION Page A749 Wyle Laboratories Huntsvllle Facility

Test Report 10530R01 Attachment A7 Page A7-21 REDACTED VERSION Wyle Laboratories Huntsville Facillty

ATTACHMENT 2-D FA1105-06, Revision 0 Summary Report of MAAP4 LOCA Analysis in Support of Past Operability Assessment of Degraded HPSl Performance During Containment Recirculation at Palo Verde REDACTED VERSION

Westinghouse Non-Pfopriefary Class 3 FAUSKE & ASSOCIATES, LLC CALCULATION NOTE COVER SHEET REDACTED VERSION SECTlON TO BE COMPLETED BY AUTHOR(S):

Calc-Note Number FAy05-06 Revision Number Title Summarv Report of W 4 LOCA Analysis in Support of Past Owrability Assessment .

of Demded "SI Performance Durinn Containment Recirculation at Palo Verde Project Number or Arizona Public Service Shop Order APSO07

Purpose:

See Section 1.0.

Results Summary: See Section 5.0.

Completion Name (Pnnt or Type) Signature Date ChristoDher E. Henrv Februarv 11.2005 .

U I

SECTION TO BE COMPLETED BY VERIFIER(S):

G.Thomas Elicson Februarv 11.2005 .

Independent Review or SECTION TO BE COMPLETED BY MANAGER:

Responsible Manager:

Name (Pnnt or Type) Signature Approval Date I I Christopher E. H e m W

Februarv 1 I . 2005 .

0 2005 Fauske & Associates, LLC - All Rights Reserved FAI/05-06. Rev. 0 I February 2005

C A W NOTE NUMBER FAuO5-06 CALCULATION NOTE METHODOLOGY CHECKLIST REDACTED VERSION PAGE ii CHECKLlST TO BE COMPLETED BY AUTHOR(S) (CIRCLE APPROPRIATE RESPONSE)

I. Is the subject a n d o r the purpose of the design analysis clearly staled? ............................................................................................... a YES NO

2. Are the required inputs and their sources provided?................. ..... NO NIA

3. Are the assumptions clearly identified and justified? ......................... NO NfA

4. Are the methods' and units clearly identified? .......................... ..... NO N/A

5. Have the limits of applicability been identified? ................................ NO NIA (Is the analysis for a 3 or 4 loop plant or for a single application.)

6. Are the results of literature searches, if conducted, or other background data provided? ............................................................. NfA

7. Are all the pages sequentially numbered and identified by the calculation note number? ................................................................... NO

8. Is the project or shop order clearly identified?. .................................. 0 YES No

9. Has the required computer calculation information been provided? ...a NO N/A

10. Were the computer codes used under configuration control? ............ NO N/A

11. Was the computer code(s) used applicable for modeling the physical and/or computational problems identified? ................................

(ie., Is the correct computer code being used for the intended purpose.)

..... NO N/A

12. Are the results and conclusions clearly stated? ................................. NO

13. h e Open Items properly identified ...................................................... YES NO a

14. Were approved Design Control practices followed without exception?.YES NO (Approved Design Control practices refers to guidance documents within Nuclear Services that state how the work is to be performed, such as how to perform a LOCA analysis.)

1s. Have all related contract requirements been met? ..................... ..... NO N/A NOTE: If N O to any of the above, Page Number containing justification FAI/05-06. Rev. 0 February 2005

REDACTED VERSION FAVOS-06

SUMMARY

REPORT OF MAAP4 LOCA ANALYSIS IN SUPPORT OF PAST OPERABILITY ASSESSMENT OF DEGRADED msr PERFORMANCE DURING CONTAINMENT RECIRCULATION AT P A L 0 VERDE Submitted To:

Arizona Public Service Phoenix, Aizona Prepared By:

Christopher E. Henry Fauske & Associates, LLC 16W070 West 831d Street Burr Ridge, Illinois 60527 m: (630) 323-8750 (630) 986-5481 February 2005 FAU0.5-06. Rev. 0 iii February 2005

TABLE OF CONTEXTS REDACTED VERSION LlST OF FIGURES ................................................................................................................vi LIST OF TABLES ................................................................................................................ VIII EXECUTrVE

SUMMARY

...................................................................................................... ix 1 .0 INTRODUCTION ............................................................................................ 1-1 O f 1-4 1.1 Background .......................................................................................................

1.2 Post-RAS ECCS a d CSS Status ................................................................... 1-2 1.3 Initiating Event Selection ................................................................................ 1-3 1.4 Break Size and Location Selection ................................................................. 1-3 2.0 M W CODE DESCWTTON .................................................................. 2-1 Of 2- IO 2.1 What is MAAP? .............................................................................................. 2-1 2.2 MAAP History ................................................................................................. 2-1 2.3 Summary of Relevant Benchmarks ................................................................. 2-5 2.3.1 RCS Response to Small LOCA ....................................................... 2-5 2.3.2 Containment Response to LOCA ..................................................... 2-6 2.3.3 RCS Response to Steam Generator Tube Heat Transfer ............... 2-6 2.4 Regulatory Understanding of MAAP ............................................................. 2-7 2.5 MAAP4 Limitations ......................................................................................... 2-8 2.5.1 MAAP4 RCS Model ........................................................................ 2-8 2.5.2 MAAP4 Containment Model ........................................................... 2-8 2.6 Refinements to the MAAP4 Code Revision .................................................. 2-9 3.0 DESIGN INPUT AND ASSUMPTIONS ................................................ 3-1 Of 3-10 3.1 Design Input ..................................................................................................... 3-1 3.1.1 Base Code Revision and Plant Model ............................................. 3-1 3.1.2 Analysis-Specific Plant Model Parametric Input Data ................... 3-1 FAI.05-06, Rev. 0 iv Febmary ZOOS

TABLE OF CONTENTS (concluded)

REDACTED VERSION Page 3.1.3 Analysis-Specific Assumptions of Plant and Operator Response ............................................................................................ 3-4 3.1.3.1 RCS Void Fraction for Phase Disengagement ............ 3-6 3.1.3.2 Post-LOCA Cooldown Methodology ............................. 3-8 3.1.3.3 Post-RAS HPSI S&us .................................................... 3-9 3.1.3.4 Post M S CSS Status ..................................................... 3-9 3.1.3.5 Post-RAS LPSI Status ................................................ 3-10 4.0 MAAP CASES .................................................................. .................. 4-1 Of 4-29 4.1 Series 1 ....................................................................

4.1.1 Detailed Profile of the 3-Inch Case ........ ................................ 4-5 4.2 Series 2 ....................................................................................................... 4-5 4.2.1 Detailed Profile of the 3-Inch Case ................................................. 4-6 4.3 Series 3 ............. ......................................................................................... 4-15 4.4 Series 4 ....................................................................... .............................. 4-15 4.5 Series 5 ....................................................................... .................. 4-19 5.0 MPLAP ANALYSIS

SUMMARY

AND CONCLUSIONS ...................... 5-1 of 5-4 5.1 RCS Thermal-Hydraulic Performance ............................................................ 5-1 5.2 Containment Thermal-Hydraulic Performance ............................................... 5-3 6.0 NOMENCLATURE ......................................................................................... 6-1 Of 6-2

7.0 REFERENCES

................................................................................................. 7-1 of 7-2 FH/05.06 . Rev. 0 V February 2005

LIST OF FIGURES REDACTED VERSION FAI/05-06. Rev. 0 vi February 2005

-Memwmw REDACTED VERSION LIST OF FIGURES (concluded)

FAI/OS-06,Rev. 0 vii February 2005

LIST OF TABLES REDACTED VERSION Table 2- 1 History of MAAP Code Development ................................................................... 2-2 3-1 Analysis-Specific Plant Model Parametric Input Data .......................................... 3-1 3-2 Analysis-Specific Assumptions of Plant and Operator Response ......................... 3-4 5- 1 Series 1 Case Summary .......................................................................................... 5-1 5-2 Series 2 Case Summary .......................................................................................... 5-2 FAU05.06. Rev. 0 viii February 2005

EXECUTIVE

SUMMARY

REDACTED VERSION This report documents MAAP4 calcuJations of Palo Verde Nuclear Generating Station (PVNGS) core, reactor coolant system (RCS), and containment thermal-hydraulic response to a small-to-medium loss of coolant accident (LOCA) in which the high-pressure safety injection (HPSI) and containment spray system (CSS) become degraded. Potential failure of HPSI is also considered. Degradation and potential failure are presumed to occur when the emergency core cooling system (ECCS) and CSS transition between suction from the refueling water tank (RWT) to suction from the containment recirculation sump in response to the recirculation acquisition signal ( R A S ) . This scenario is intended to support a justification for past operations (JPO) assessment regarding degradation and possible failure of the HPSI system due to ingestion of air that actually existed between two valves in the ECCSICSS suction lines during past operation of the plant.

Specifically, a spectrum of break sizes and locations was evaluated to determine the case(s) that could challenge core coverage, long-term core cooling, and long-term containment heat removal. The medium break diameters in the range of roughly 3 to 6 inches were determined to be the most challenging. However, in all cases, MAAP4 predicted that the core would remain completely covered, due almost entirely to the cold leg injection of the safety injection tanks (SIT) (a.k.a., accumulators) during the post-RAS time period. Even when outright post-RAS failure of the HPSI was postulated, SIT injection maintained core coverage until post-LOCA cooldown and depressurization of the RCS below the low-pressure safety injection (LPSI) shutoff head enabled sufficient LPSI flow to provide continued core coverage and long-term core cooling.

FAU05-06, Rev. 0 February 2005

-fwwwMw I.0 INTRODUCTION REDACTED VERSION 1 .I Background On September 28,2004, PVNGS staff [PVNGS, 2004a) submitted a licensee event report (LER) to the Nuclear Regulatory Commission (NRC) that reported a condition in Units 1, 2, and 3 in which air voids in the recirculation sump suction piping (serving both the ECCS and the C S S ) may have prevented the filfillment of the system safety function to removal residual heat and mitigate the consequences of a loss of coolant accident. (Reference [Westinghouse, 20041 provides some additional details that are relevant to all Westinghouse and CE designs.)

PVNGS, in conjunction with Westinghouse and its Fauske and Associates (FAI) subsidiary, investigated this condition with an approach that involved both experiment and analytical elements. Phases 1 through 3 of the investigation were predominantly experimental separate effects testing of HPSYCSS availability and are not considered here. Phase 4 was the integral plant analysis with independent evaluations provided by the MAAP4 and CENTS codes.

This report is confined to MAAP4 analysis portion of Phase 4.

Phase 4 participants from PVNGS, Westinghouse (Windsor, Connecticut office), and FM were charged with considering the core, RCS, and containment response to post-RAS degradation and potential failure of the HPSI and C S S . Furthermore, this circumstance couId result from any of the full spectrum of initiating events (LOCA, transient, station blackout, ...)

that would challenge core coverage, long-term core cooling and, long-term containment heat removal (and by extension long-term containment integrity). Since the outcome of challenges could involve core overheat and damage, the M A P 4 code was selected as a conbibutor to the analysis in view of its ability to model degraded core progression and its influence on the RCS and containment.

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REDACTED VERSION 1.2 P o s t - U S ECCS and CSS Status It has been established that the HPSI system within the ECCS and the CSS could become degraded or even unavailable during p o s t - U S operation due to ingestion of pre-existing air within the suction lines. Elaboration on some key details is instructive.

At the time of RAS, the PVNGS units are designed for automatic switchover of the HPSI and CSS systems. Specifically, these systems are stopped, realigned to the recirculation sump, and then restarted during the automatic switchover. The LPSI system is stopped as part of this process, but it is not automatically restarted. It must be manually restarted by the operator (if necessary) after completion of switchover. Furthermore, the HPSI suction line is the first system to draw from the suction header, This is followed by the CSS suction line and finally the LPSI suction line. Also, the specific configuration of the HPSI suction line makes HPSI more susceptible to air ingestion than the other systems.

Indeed, the noted Phase 1 and Phase 2 experiments, which were responsible for characterizing the two-phase flow through the suction header and individual ECCSICSS suction lines, demonstrated that most air ingestion would occur in the HPSI system with only a relatively small ingestion by the CSS system.

Phase 3 experiments were responsible for evaluating an actual "SI pump with air ingestion boundary conditions dictated by Phase 1 and Phase 2 experiments. These experiments demonstrated that the HPSI system would continue to operate but at a degraded flow condition, with increasing degradation (decreasing flow) at higher system pressure.

Therefore, Phase 4 analyzed both degraded and failed conditions for HPSI, a prescribed degraded condition for CSS and full availability of LPSI in the post-RAS operation. Specific details will be provided in Section 3.

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1.3 Initiating Event Selection REDACTED VERSION As stated above, all initiating events were considered which would challenge core coverage, long-term core cooling, and long-term containment heat removal Furthermore, the Level I1 containment event trees [PVNGS, 19923 for these initiating events were inspected to determine the most challenging set of conditions for high-pressure recircdation degradation or failure. Note, evaluation of the event trees did not entail loss of additional components concurrent with the HPSI degradation or failure. Since this was a deterniinistic (as opposed to probabilistic) analysis that was intended to support justification for past operation, all other systems were assumed to be available, particularly the safety injection ldnks (SIT) and the operator action of post-LOCA steam generator cooldown and depressurization of the RCS via the steam generators.

1.4 Break Size and Location Selection With these ground rules in place, it was determined that a small to medium LOCA (roughly 3 to 6 incbes in diameter) initiating event is most challenging since i t is responsible for significant coolant loss, but the RCS remains at elevate pressure because the break alone is not sufficient to remove decay power. [

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REDACTED VERSION FAU05-06. Rev. 0 1 - 4 FebruaT 2005

2.0 MAAP CODE DESCRIPTION 2.1 What is MAAP?

MAAP is a computer code that simulates light water reactor system response to accident initiation events. The Modular Accident Analysis Bogram (MAAP), an integral systems analysis computer code for assessing severe accidents, was initially developed during the industry-sponsored IDCOR Program. At the completion of IDCOR, ownership of MAAP was transferred to EPRI. Subsequently, the code evolved into a major analyhcal tool ( M A A P 3B) for supporting the plant-specific Individual Plant Examinations (PES) requested by NRC Generic Letter 88-20. Furthermore, MAAP 3B was used as the basis to model the Ontario Hydro CANDU .designs. As the attention of plant-specific analyses was expanded to include accident management evaluations, the scope of MAAP (its design basis) was expanded to include the necessary models for accident management assessments. Through support by the U S .

Department of Energy (DOE), the M A P 4 design basis was further extended to include the Advanced Light Water Reactor (ALWR) designs currently being developed by the reactor vendors. MAAP4 has also been expanded to represent the W E R designs used in Finland and central Europe.

2.2 MAAP History Table 2-1 summarizes the history of MAAP development in terms of the major code versions and the major advancements represented by each version. Two types of Nuclear Steam Supply Systems (NSSS) are modeled in the MAAP4 code: the Boiling Water Reactor (BWR) and the Pressurized Water Reactor (PWR). In addition, MAAP4 is the first archived code that contains a graphical representation of the reactor and containment response (MAAP4GRAAPH).

MAAP4, like MAAP 3B, is currently being maintained by Fauske & Associates, LLC (FAI) for the Electric Power Research Institute (EPRI) and the MAAP User's Group WUG).

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REDACTED VERSION I Table 2-1: History of MAAP Code Development.

Major Advancement MAAP development initiated for BWRs and PWRs.

Primary system and containment thermal-hydraulic models.

Fission product release, transport and deposition models added; local Hz burning (igniters).

Zircaloy-tellurium binding.

In-vessel natural circulation, advanced models for aerosol growth and deposition, suppression pool scrubbing, gas natural circulation in steam generation, C h e x a h y n a n correlation for BWR core power model.

Auxiliary buildinglreactor building model, improved suppression pool scrubbing model, increased RCS nodalization, RCS natural circulation.

CANDU-specific models for the horizontal fuel bundle and pressure tubes, moderator tank, shield tank,multi-unit containment, and vacuum building.

Fuel cans for the PWR core, horizontal steam generator, he1 movement as part of the shutdown mechanism.

Accident management and ALWR models, advanced core melt progression and material creep models, in-vessel cooling, external cooling of the RPV, detailed modeling of the lower head I penetrations, generalized containment, interactive graphical interface, on-site and off-site radiation I dose models.

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REDACTED VERSION The purpose of MAAP4 is to provide an accident analysis code that can be used with confidence by the nuclear industry in all phases of severe accident studies, including accident management, for current reactor/containment designs and for ALWRs. MAAP4 includes models for the important accident phenomena that might occur within the primary system, in the containment, and/or in the auxiliaryheactor building. For a specified reactor and Containment system, MAAP4 calculates the progression of the postulated accident sequence, including the disposition of the fission products, from a set of initiating events to either a safe, stable state or to an impaired containment condition (by overpressure or over-temperature) and the possible release of fission products to the environment.

Severe accident analyses can be divided into four phases: (1) prevention of core damage; (2) recovery prior to reactor pressure vessel breach; (3) recovery after vessel breach, but prior to containment failure; and (4) mitigation of releases of fission products reaching reactor/auxiliary buildings. The previous archived version, MAAP 3B, can analyze phases 1, 3, and 4 for existing reactors, which is sufficient to support the Individual Plant Examination (IPE) studies, the intended purpose of that major MAAP version. However, M M P 3B does not have the ability to treat phase 2, recovery prior to vessel breach but after severe core damage. It bas been estimated that the interval between the onset of severe core damage and the time of vessel breach could vary from 30 minutes to many hours or, as in the TMI-2 accident, vessel integrity can be maintained throughout the accident. Recovery during this interval could obviously reduce, and perhaps eliminate, the likelihood of reactor pressure vessel failure and thereby greatly limit the extent of the accident.

In evaluating the effectiveness of proposed accident management strategies, there is a need to evaluate the integral system response to the proposed actions. Because of the numerous phenomena involved the evaluation is complex, and for many severe accident phenomena, the experimental database is sparse. However, with the extensive "MI-2 data, along with the results of integral experiments such as the LOFT and CORA tests, the major characteristics of the melt progression, primary system thermal-hydraulic response, and core debris-concrete interaction have been demonstrated. Also, with EPRI-sponsored experiments, more data have become available on key phenomena, for example, the mode of vessel breach and the conditions which could prevent vessel failure. The results from these experiments have been included in the MAAP4 FA1/05-06, REV. 0 2 - 3 Februav 2005

REDACTED VERSION modeling enhancements and have resulted in major insights with respect to the effectiveness of accident management actions, particularly for maintaining the integnty of the reactor vessel.

One area where only limited experimental data are available is quenching of overheated debris prior to vessel breach. This of course, is of key interest in recovering from an accident slate and was a major part of the TM1-2 accident. MAAP4 includes models for in-vessel cooling and external cooling of the RPV to evaluate whether a safe, stable state can evolve following water addition to the RCS and/or the containment if the core debris can be retained within the reactor pressure vessel.

MAAP4 also addresses the new and unique features, many of which are passive, included in ALWR designs. These are:

passive beat removal system, such as an in-containment isolation condenser or a passive RHR system, gravity-fed water injection systems, external heat removal from the containment shell, a generalized nodalization scheme for the containment to accommodate the ALWR designs including an incontainment RWST, and the capability to analyze flow through large safety valves, such as an automatic depressurization system for PWR designs.

Since the beginning of the MAAP code development, the codes have represented all of the important safety systems such as emergency core cooling, containment sprays, residual heat removal, etc. MAAP4 allows operator interventions and incorporates these in a flexible manner, permitting the user to model the operator response and the availability of the various plant systems in a general way. The user can represent operator actions by specifjmg a set of values for variables used in the code and/or events, which are the operator intervention conditions.

There is a large set of actions that the operator can take in response to the intervention conditions.

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with W

REDACTED VERSION 4 has been developed under tbe FAI Quality Assurance Program, in conformance IOCFRSO Appendix B and with the lnternational I S 0 9000 Standard. Furthermore, the new software bas been subjected to review by a Design Review Committee, comprised of senior members of the nuclear community, in a manner similar to that exercised for MAAP 3B.

2.3 Summary of Relevant Benchmarks Tbe following subsections provide a summary of relevant MAAP4 benchmarks against plant experience and large-scale integral experiments and also against one integral computer code. Plant experience and experiment benchmarks are documented in Volume 3 of the MAAP4 Users Manual [EPRI, 2003al. (The MB-2 benchmark is awaiting incorporation into the manual in the next MAAP4 revision cycle this year.)

2.3.1 RCS Response to Small LOCA Since RCS thermal-hydraulic performance under a small LOCA condition is essential to the analysis, some relevant benchmarks are cited here.

MAAP4 RCS thermal-hydraulics has been benchmarked against the Three Mile Island Unit 2 (TMI-2)plant experience, particularly the small LOCA phase of the accident when the pressurizer relief valve was stuck open. MAAP4 RCS thermal-hydraulics has also benchmarked against a similar stuck open pressurizer relief valve event at Crystal River Unit 3. Both benchmarks show reasonable good agreement with the plant data. While these benchmarks are for RCS bot side LOCAs in the pressurizer, they are still relevant to cold side LOCAs since the LOCA modeling in the MAAP pressurizer model is essentially the same as that used for LOCA modeling in RCS loop piping.

As part of the recent Beaver Valley atmospheric containment conversion project, MAAP4 was benchmarked against the Westinghouse small LOCA code, NOTRUMP.

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+wwRmMF REDACTED VERSION 2.3.2 Containment Response to LOCA Since containment response is an important aspect of RAS timing, it is important to insure the integrity of the MAAP4 containment model. MAAp4 has been benchmarked against numerous containment experiments, both separate effects tests and large-scale integral effects tests. Herein, the containment was benchmarked as a stand-alone model with break mass and energy rates from the experiment, specified as a boundary condition to the model. This type of stand-alone benchmark can be performed within the normal MAAP4 code framework via the M A N 4 dynamic benchmarking feature, thereby exercising the exact same containment model that is used in conventional MAAP4 applications that exercise the full code.

Two benchmarks of note are the small LOCA experiment El 1.2 and the medium LOCA experiment T3 1.5 performed at the HDR test facility in Germany, which was a reactor-scale containment that contained a decommissioned low-power reactor. MAAP4 compares well to both short-term and long-term containment pressurization in both experiments.

2.3.3 RCS Response to Steam Generator Tube Heat Transfer Since post-LOCA cooldown and depressurization is an important operator action in this analysis, it is important to insure the integrity of the RCS response to steam generator tube heat transfer.

MAAP4 has been benchmarked the Crystal River Unit 3 plant transient, noted above.

Herein, steam generators temporarily boiled dry during the transient prior to receiving auxiliary feedwater. Also, in a similar event, the Davis-Besse Unit 1 plant transient resulted in the stearn generators boiling dry for a brief period until auxiliary feedwater could be provided. The MAAP4 RCS model, in particular the primary system average temperature, compares well during both the initial steam generator heat transfer and subsequent primary system heatup in the presence of dry steam generators.

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REDACTED VERSION The MAAP4 steam generator model has been compared against an integral steam generator experiment known as the Westinghouse Model Boiler 2 (MB-2). Herein, the steam generator is treated as a stand-alone model with primary system boundary conditions from the experiment provided via user input. Again, like the stand-alone containment benchmark, a stand-alone steam generator benchmark can be performed within the normal MAAP4 code framework via the MAAP4 dynamic benchmarking feature, thereby exercising the exact same steam generator model that is used in conventional MAAP4 applications that exercise the full code. Revision MAAP 4.0.5, which is the code revision used for this analysis, was successfully benchmarked against loss of feedwater tests (both simulated full power and decay power transients) performed at MB-2.

2.4 Reeulatory Understanding of MAAP The U.S. Nuclear Regulatory Commission (NRC) reviewed and approved MAAP 3.OB for support of probabilistic risk assessment (PRA) activities at licensed power reactors in the U.S., particularly the individual plant examinations (IPEs) that occurred in the late 1980s and early 1990s.

While MAAP4 has not undergone a formal review process by the NRC, the code owner, the Electric Power Research Institute (EPRI), Fauske and Associates (FAT), and the MAAP Users Group (MUG) previously engaged in MAAP4 familiarization activities with the NRC when MAAP4 was first released. Recently, a MAAP4 Information Exchange between these parties has been undertaken in view of the expanding scope of MAAP4 application and MAAPCsupported submittals to the NRC.

MAAP4 has been used previously for safety analyses outside of the risk arena with NRC approval. For example, an NRC Safety Evaluation Report (SER) was written for the D.C. Cook plant in its assessment of minimum safe sump level in tbe containment recirculation sump during a small LOCA event. This assessment involved small LOCA scenarios that are similar to those in the present analysis for PVNGS.

FM/05-06.Rev. 0 2 - 7 February 2005

2.5 2.5.1 U P 4 Limitations MAAP4 RCS Model REDACTED VERSION The MAAP4 RCS model uses momentum equation selectively for sub-models that demand a momentum equation for model adequacy. One of the aspects for which a full-fledged momentum equation is not implemented is water flow. Consequently, MAAP4 cannot void the core by reversing flow from the core to the downcomer and loop piping during a large LOCA event. However, small breaks of the size being analyzed for this analysis do not engage in such significant flow reversal, so this limitation is not relevant to this analysis.

2.5.2 MAAP4 Containment Model The MAAP4 containment model can accommodate most physical pbenomena that would occur. However, since it does not entrain pre-existing liquid and condensate from heat sink surfaces, it does not mechanistically bring suspended water droplets into the containment atmosphere (although the model could accommodate droplets if such liquid entrainment was added). Consequently, it is conservatively predicts excess gas-phase superheat and pressurization during the blowdown stage of a large LOCA event.

Again, small breaks of the size being analyzed for this analysis do not promote significant gas superheat, so this limitation is not relevant to this analysis. Furthermore, superheat and excess pressurization are conservative for this analysis since they would lead to earlier RAS timing. As noted previously, the HDR T3 1.5 and El 1.2 containment benchmarks are testament to the adequacy of the containment model for predicting short-term and long-term containment pressurization under small and medium LOCA conditions, which is necessary for an accurate depiction of containment spray actuation signal (CSAS) timing in this analysis.

FAIlO5-06, Rev. 0 2 - 8 February 2005

REDACTED VERSION 2.6 Refinements to the MAAP4 Code Revision The latest MAAP4 archived revision, MAAP 4.0.5 [EPRI, 2003b1, was used with the latest PVNGS-specific plant model (a.k.a., parameter file). [

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REDACTED VERSION FAVO.5-06, RCV. 0 2 - IU February 2005

REDACTED VERSION 3.0 DESIGN INPUT AND ASSUMPTIONS 3.1 Design Input 3.1.1 Base Code Revision and Plant Model The base code revision is the latest MAAP4 archived revision, MAAP 4.0.5

[EPRI, 2003bj. In addition, a revision to the archived subroutine WFLOW was included in this anaIysis to address a finding made during the analysis, as discussed in detail in Section 2.

The base plant model is the latest PVNGS-specific plant model, or parameter file,

[PVNGS,20011 for W 4 .

3.1.2 Analysis-Specific Plant Model Parametric Input Data Table 3-1 summarizes the analysis-specific plant model parametric input data that is most influential to the analysis. Some values are taken directly from the PVNGS base plant model.

Others are analysis-specific changes. (Parameter input of secondary importance is not discussed here, and their values are taken from the base plant model without alternation.) [

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- REDACTED VERSION 3.1.3 Analysis-Specific Assumptions of Plant and Operator Response In addition to plant model parametric input data, there are analysis-specific modeling assumptions of plant and operator response, which area summarized in Table 3-2. As with the parametric input data, assumptions are primarily best-estimate, but some key assumptions, which have a large bearing on RCS and containment response, are biased in a conservative manner.

These are discussed here.

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REDACTED VERSION FAUO5-06,Rev. 0 3-5 February 2005

?Rfmwmw REDACTED VERSION 3.1.3.1 RCS Void Fraction for Phase Disengagement The MAAP RCS model tracks a global primary system average void fraction. When the void fraction exceeds the value of a user input model parameter VFSEP, the gas- and liquid-phases will disengage (or separate). The phases can re-engage if the void fraction is reduced below user input model parameter VFCIRC. Phase disengagement is an important consideration because it has a substantial influence on the rate at which the RCS can depressurize.

Specifically, while the phases are engaged and under natural circulation through the coolant loops, gas and liquid are essentially in thermodynamic equilibrium. The net effect of this condition is that the break discharges at a higher mass and energy rate, which leads to a larger depressurization rate. While the phases are disengaged, gas and liquid are in thermodynamic non-equilibrium. If the phases are disengaged (but all other conditions remain the same), the break discharges at a lower mass and energy rate, which leads to a smaller depressurization rate.

The FLECHT-SEASET was a scaled integral experiment, which studied two-phase natural circulation through the RCS, including phase disengagement. For RCS configurations with inverted U-tube steam generators, phase disengagement occurred at a best-estimate void value of roughly 50%. However, there is significant uncertainty in this quantity. Sensitivity studies of MAAP with the PVNGS plant model showed that a value of VFSEP = 0.10 would disengage the phase early relative to the noted best-estimate value, leading to the noted slower depressurization rate, which is conservative for this analysis. This is demonstrated for the 3-inch LOCA in Figure 3-1. (Values below 0.10 did not result in significantly early disengagement.) Therefore, this value is used as a conservative bound, and it is paired with a corresponding value of VFCIRC = 0.05 for possible re-engagement, although re-engagement does not occur during this analysis.

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REDACTED VERSION FAi/05-06. Rev. 0 3-7 February 2005

3 1.3.2 Post-LOCA Cooldown Methodolorn REDACTED VERSION The post-LOCA cooldown delay time and rate are roughly based upon a representative PVNGS simulator run [PVNGS, 2004b). Herein, the delay time between LOCA initiation and cooldown initiation was roughly 720 seconds (12 minutes). A conservative value of 1500 seconds (25 minutes) was used in the analysis to maintain the RCS at an elevated pressure for a longer period. The cooldown rate in the simulator exercise was roughly 90 F h r .

However, there is not explicit guidance in the EOPs for a nominal cooldown rate, aside from the caveat to not exceed 100 F/hr. For standard industry practice encompassing both normal and emergency operations, a typical range is30-100 F h . Given the 90 F h r used during emergency operation on the simulator, a conservative value of 75 F/hr is appropriate for this an a1ysi s.

Another significant assumption within the cooldown methodology is the entry condition for the cooldown since this can influence the overall timing of the cooldown progression. The typical operator practice in post-LOCA cooldown is that, if any excess overpressure exists within the steam generators, the operator opens the turbine bypass (SDBCS) system to rapidly diminish the generators to a saturation pressure corresponding to the current core exit temperature. This removes excess energy from the steam generators, which may have been acting.as a heat source to the primary system (depending upon the size of the break), and it readies the generators to act as a heat sink. At this point, the operator controls the SDBCS system to provide the core exit temperature with the desired cooldown rate noted above. The operator monitors and updates (if necessary) tbe SDBCS roughly every 10 minutes. (In the current MAAP analysis, this update is presumed to occur in a stepwise manner. If indeed the cooldown is determined to be more of a linear profile rather than a stepwise profile, then this can be easily changed, but ultimately this is a cosmetic consideration that has no bearing on the integral result.)

If during the update of the cooldown, the operator finds that the cooldown is occurring at a rate that is faster than the target rate due to the primary system fluid acting as a heat sink on the steam generator rather than a heat source (which can occur in some of the larger medium LOCAs), then it is assumed that the operator will not chase the primary system cooldown with the steam generator cooldown. Instead, it is assumed that the operator will scale back the FAUOS-06.Rev. 0 3 - 8 February 2005

REDACTED VERSION SDBCS in an attempt to slow the steam generator cooldown rate and attain the target rate. This assumption is consistent with operator training to maintain a rate of less than 100 F/hr to protect the primary system structure components (particularly the vessel) from rapid overcooling even if the primary fluid is cooling itself at a higher rate due to a medium-to-large break size. This methodology is also conservative for this analysis since it slows the primary system depressurization.

3.1.3.3 Post-R4S HPSI status I.

3.1.3.4 Post-IUS CSS Status FAI/O5-06, Rev. 0 3-9 February 2005

REDACTED VERSION 3.1.3.5 Post-RAS LPSI Status As discussed in the background in Section 1, it is virtually impossible for LPSI to experience post-RAS degradation since post-RAS restart of LPSI is not automatic and must be done by remote operator action, which cames a substantial delay relative to the automatic switchover performed by HPSI and CSS.

Therefore, it is assumed that LPSI is available in post-RAS for RCS injection and, if necessary, containment spray and long-term containment heat removal through the containment spray heat exchangers. Even though both LPSI trains are available during post-RAS operation, it is conservatively assumed for this analysis that only 1 train is aligned for post-RAS injection, and no LPSI trains are used to assist contain spray and heat removal.

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REDACTED VERSION 4.0 MAAP CASES This section of the MAAP analysis report (and the corresponding section of the CENTS analysis report) is organized i n terms of several case series, with each series devoted to a particular combination of major boundary conditions (break location, ECCS trains, HPSI availability, etc.). (The full scope of boundary conditions is provided in Section 3 . ) Specific results associated with a series are discussed as part of its presentation below.

An overall summation of the analysis highlights will be conducted in Section 5 .

4.1 Series 1 This senes is Lzfined by the following boundary conditions:

0 Break location: Cold leg discharge 0 Break size: Break diameters of Yz,1, 2, 3, 4, 5, 6, 7, 8, 9, and 10 inches 0 At SIAS: 2 HPSI; 2 CSS; and 2 LPSI trains available 0 At U S : No HPSI; 2 CSS trains degraded to 25% of non-degraded flow; 1 LPSI to RCS; and 1 LPSI in reserve.

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REDACTED VERSION 4.1.1 Detailed Profile of the $Inch Case A detailed profile is being provided for the 3-inch case in Series 2 since its break location is lower and therefore potentiaIly more challenging than Series 1. A dedicated profile for the 3-inch case in Series 1 is not necessary since the same generic insights can be obtained from the profile in Series 2.

4.2 Series 2 FAI/05-06,Rev. 0 4 - 5 February 2005

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4.3 Series 3 This series is defined by the following boundary conditions:

REDACTED VERSION Core coverage and long-term core cooling are never vulnerable, which is expected since the corresponding HPSI failure cases showed no core uncovery.

4.4 Series 4 This series is defined by the following boundary conditions:

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fvmwEMw REDACTED VERSION Core coverage and long-term core cooling are never vulnerable, which is expected since the corresponding HPSI failure cases showed no core uncovery.

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5.0 MAAP ANALYSIS

SUMMARY

AND CONCLUSIONS

- REDACTED VERSION 5.1 RCS Thermal-Hydraulic Performance Key figures-of-merit are summarized for Series I cases in Table 5-1 and Series 2 cases in Table 5-2. The fundamental conclusion illustrated in these tables and discussed in detail in Section 4 is that core coverage is maintained without the use of HPSI for an extensive period between the time of RAS and the time of significant post-RAS LPSI flow, which provides long-term cooling. This is true for even the most challenging break sizes and conservative assumptions for key boundary conditions, particularly early RCS steam-water phase disengagement and a post-LOCA cooldown rate that is substantially less than the maximum aIIowabIe by emergency operating procedures.

FAI/O5-06, Rev. 0 5 - 1 February 2005

+wwwww-REDACTED VERSION FAI/05-06, Rev. 0 5-2 February 2005

REDACTED VERSION 5.2 Containment Thermal-Hydraulic Performance The MAAP containment analysis in Section 4 demonstrated that the 3-inch case is generally the most challenging break size since [

1 FAUO5-06, Rev. 0 5-3 February 2005

REDACTED VERSION As shown in Section 4, this results in a post-RAS pressure peak in containment that is largest for the 3-inch case. However, this peak is well within the containment design basis strength.

Thus, i t can be concluded that, even for the overly conservative assumption of substantial CSS degradation, post-RAS long-term containment heat removal can be achieved.

FAI/05-06. RCV. 0 5 - 4 February 2005

6.0 NOMENCLATURE REDACTED VERSION ADV Atmospheric Dump Valves BAF Bottom of Active Fuel CENTS Combustion Engineering Nuclear Transient Simulation Code CSAS Containment Spray Actuation Signal CSS (or CS) Containment Spray System ECCS Emergency Core Cooling System EOP Emergency Operating Procedures EPRI Electric Power Research Institute FA1 Fauske & Associates, LLC HLI Hot Leg Injection HPSI High-pressure Safety Injection JPO Justification for Past Operations LOCA Loss of Coolant Accident LPSl Low-Pressure Safety Injection MAAP Modular Accident Analysis Program MUG MAAP Users Group PVNGS Palo Verde Nuclear Generating Station RAS Recirculation Actuation Signal RCP Reactor Coolant Pump RCS Reactor Coolant System RWT Reheling Water Tank SDBCS Steam Dump and Bypass Control System SZAS Safety Injection Actuation Signal FAYOS-06,Rev. 0 6 - I February ZOOS

REDACTED VERSION SIT Safety Injection Tank TAF Top of Active Fuel TMI-2 Three Mile Island Unit 2 FAIOS-06. REV. 0 6-2 February 2005

7.0 REFERENCES

REDACTED VERSION EPRI, 2003a, MAAP4: Module Accident Analysis Program for LWR Power Plants, Computer Code Users Manual, May 1994 - January 2003 (Updated to MAAP 4.0.5 specification in January 2003).

EPFU, 2003b. W 4 : Module Accident Analysis Program for LWR Power Plants -

Transmittal Document for MAAP 4.0.5, FAV03-13, January.

Femraccio, F., 2005, PVNGS MAAP and CENTS Evaluation Data, Westinghouse Letler LTR-PS-05-13 to Mark Radspinner (APS),dated February 10.

WwestinghOuSe Letter LTR-PS-05-13.

Janke, M., 2004, e-mail from Westinghouse, Boundary Conditions for Palo Verde Units 1 and 3 LOCA Analysis, to Christopher Henry (FAI), dated November 5.

PVNGS, 1992, Palo Verde Probabilistic Risk Assessment, Section 1 1 Containment Analysis, Revision 0, dated April 7.

PVNGS, 2001, Palo Verde MAAP 4.0.4 Parameter File, PVNGS Document No. 13-NS-C36, Revision 1, dated March 29.

PVNGS, 2004a, Palo Verde Nuclear Generating Station (PVNGS) Units 1 , 2, and 3 Licensee Event Report 2004-009-00 (Docket No. STN 50-528, 50-529, 50-530) (License No. NPF 41, NPF 51, NPF 74), Letter 192-01 1 5 4 - D M S M S to the U.S. Nuclear Regulatory Commission, dated September 28.

PvNGsLicensee Event Report 2004-04 PVNGS, 2004b, Simulator Scenario for 2.0 LOCA Cold Leg Break with Failed HPSI Pumps, November 9.

Simulator xenario.doc FA1/05-06, Rev. 0 7-1 Februav 2005

REDACTED VERSION Thompson, D. P.,2005, Letter from Westinghouse to F. Ferraraccio (Westinghouse), dated February 8, 2005.

pJ ..--

Weir h e i i t hi Westinghouse, 2004, Containment Sump Line Fluid Inventory, Nuclear Safety Advisory Letter NSAL-04-7, dated November 15.

FW NSAL Disb-ibution Repuest FAI/O5-06. Rev. 0 7-2 February 2005

ATTACHMENT 2-E DAR-OA-05-3, Revision 0 Report of SBLOCA Analyses with Degraded ECCS Flow After RAS Performed for Arizona Public Service in Support of Palo Verde Nuclear Generating Station Units 1, 2, and 3 REDACTED VERSION

Westinghouse Proprietaty Class 3 DAR-OA-05-3 Rev. 0 REDACTED VERSION Page I WESTINGHOUSE ELECTRIC COMPANY LLC DAR-0A-0 5-3 Revision 0 REPORT OF SBLOCA ANALYSES WITH DEGRADED ECCS FLOW AFTER RAS PERFORMED FOR ARIZONA PUBLIC SERVICE IN SUPPORT OF PAL0 VERDE NUCLEAR GENERATING STATIONS UNITS 1,2, & 3 February 2005 Originator: see bottom of paqe Date:

Mark C. Janke, Westinghouse Electric Company LLC Technical Reviewer: see bottom of paqe Date:

Tyler Upton, Westinghouse Electric Company LLC Management Approval: see bottom of page Date:

Stephen P. Rigby, Manager Operations Analysis, Westinghouse Electric Company LLC 02005 Westinghouse Electric Company LLC Official record electronically approved in EDMS 2000.

WESTMGHOUSE ELECTRIC COMPANY LLC

Table of Contents 1.0 Background / Purpose ..-....................~.....-...-.................................-.-.........-............-...-...-.4 2.0 Code Description......_._._....___.__...~........__._.._.______._......._..._............~.....~..... - ...................... ..6 3.0 Case Descrjptions & Input Parameters ..__._ ___ ~ ....__. ~ .... _............................_.........

~ -.............7 3.1 Common initial Conditions and Plant Parameters.... .._._.__._____ 7 3.1.1 Initial Plant Conditions_,__......_._.___. ............................................. . . ~ ~ . " . ~ ~ . ~ ~ . -7. ~ ' ~ ' . ' ~ ' ' . . ' ~ . ~ . ~

3.1.2 ECCS Parameters................................ .._... ................................................................. 8 3.2 Break Parameters........................-.-.........................~....~.-..............................

3.3 Core Decay Heat _. .. ._.........__..............._.............. ~ .-.......

3.4 ContainmentSpray Pumps .......................................................................... ~

3.5 Operator Actions 12 3.6 Sensitivity Cases................................................................... ..._......... ~ - ....................

3.6.1 Sensitivity Case 1: [ ] ...... 13 3.6.2 SensitivityCase 2: [ ] ....... 13 J ....... 13 3.6.3 SensitivityCase 3: [

3.6.4 SensitivityCase 4: [ ] ....... 13 3.6.5 Sensitivity Case 5: [ ] 14 3.6.6 SensitivityCase 6: [ ] _...._. 14 3.6.7 Sensitivity Case 7: ] ....._. 14 4.0 I ......-........-.................-...--.-..-..-.............-....-.................-....-................15 5.0 Case Results .................................................................................................................... 17 5.1 ] ____.......,._._.___.._.____._. 17 5.2 Discussionof IndividualCase Results................................................................................... 20 5-21 Series 1 & 2 Cases: Cold Discharge Leg (CDL) Breaks.............................................. 20 5.2.2 Series 3 & 4 Cases: Suction Leg (SL) Breaks............................................................... 23 5.2.3 Series 5 Cases: Sensitivity Cases..................................................... ~ .......................... 25 5.3 Case Summary .............................................................................................. ___..__...__.__ 27 6.0 Conclusions ............._._..................--..--........-......-..---.........-.........-................-................... 29 7.0 Figures .............................................................................................................................. 30 7.1 Series 1: CDL Breaks. Failed HPSl after RAS ____._...._..___..__... _.. ................................ 30

^^

7.1.1 CDL-1.__...... ..._.___ ...._..___. ...... .................................................................................................. ju 7.1.2 CDL-2.......................... ..................................................................................................... 33 7.1.3 ... ....-.. ......- .... ........ ........ -..--... ......- ................ ...............

7.1.4 CDL-4.._._.... ~ ................................................................................... ~ ........... ......................... .. 43

~

7.1.5 __..__.___._..___._....... ......_.__ ................................................... -...-................-

7.1.6 ..................................................................... 5 1 7.1.7 ...-............... ...............- .............................

~ 55 7.1.8 7.1.9 7.1.10 _............................................................... 62

~

7.2 Series 2: CDL Breaks, DegradedHPSl after RAS ............................ ............................. 64 CDL-1 DH (Case n& Required) .............................................................................. . .. 64 7.2.1 7.2.2 CDL-2 DH (Case not Required)......_..... ................................... 64 7.2.3 CDL-3 DH.. 65 7.2.4 CDL-4 DH.. ..--.......................... ................ ............................................

7.2.5 CDL-5 OH.. ........-....-......--.-. ......................................... 71 7.2.6 COL-6 DH.. ..................................................................... 74 7.2.7 CDL-7 DH ._..___..___...._._.._....... ...................................................................... 77 7.2.8 CDLS DH.. ................................................................ ..-....-......-....--..-...-... .. 80 7.2.9 CDL-9 DH........................................................... ~ ...................................................

7.2.10 CDL-10 DH ....................................................................................................................... 84 7.3 Series 3: SL Breaks Failed HPSl after RAS .._....._......._................................................ 86 7.3.1 SL-1....................................................................................................... .~"~~.'~.'..~......'~...' 86 WESTMGHOUSE ELECTRlC COMPANY LLC

DAR-OA-05-3 Rev . 0 7.3.2 Westinghouse Proprietary Class 3

. REDACTED VERSION SL-2 .................................................................................................................................. 90 Page 3 7.3.3 SL-3 ................................................................................................................................ 94 7.3.4 SL-4 ................................................................ ................................................. 99 7.3.5 SL-5 ............................................................................................................................... 103 7.3.6 SL-6 ................................. ....................................................................................... 106 7.3.7 SL-7................................. .............................................................. 109 7.3.8 SL-8 ........................................ .................................................... 112 7.3.9 SL-9 ........................................................ .......................... ..................... 114 7.3.10 SL-10.......................................................................................................................... 115 7.4 Series 4: SL Breaks, Degraded HPSl after RAS ............................................................. 116 7.4.1 SL-1 DH (Case not Required)........................................................................................ 116 7.4.2 SL-2 OH (Case not Required)........................................................................................ 116 7.4.3 SL-3 DH............................................................................................................................. 116 7.4.4 SL-4 DH............................................................................... ............................ 119 7.4.5 SL-5 DH....................................................... ................ ........................... 121 7.4.6 SL-6 DH............................................................................... ........................... 124 7.4.7 SL-7 DH............................................................................................................................ 127 7.4.8 SL-8 DH ............................................................................................................................. 129 7.4.9 SL-9 DH .............. ........................................... ....................................................... 131 7.4.10 SL-1ODH........ ......................................................................................... 133 7.5 Series 5: Sensitivity C .

ed HPSl after RAS 2" SL Break................................. 135 7.5.1 SL-2 75F CD..................................................................................................................... 135 7.5.2 SL-2 1 HPSL .................................................... ....................... 137 7.5.3 St-2 SIT Gamma = 1 ...........................................

7.5.4 [ ] ............................................ ........................... 143 7.5.5 [ J .................................................... 145 7.5.6 CDL [ ......................................... 149 7.5.7 CDL [ J ........................................................... 151 WESTINGHOUSE ELECIXC COMPANY LLC

Westing house Proprietary Class 3

+MwEWw DAR-OA-05-3 Rev. 0 REDACTED VERSION Page 1.O Background / Purpose This report was prepared by Westinghouse Electric Co. for Arizona Public Service (APS) in support of Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 & 3 . This analysis is part of a project to determine the past operability of the PVNGS units with air in the Emergency Core Cooling System (ECCS) suction lines to the containment sump.

If a Loss of Coolant Accident (LOCA) were to occur with air in the ECCS pump suction line to the sump, it is postulated that the High Pressure Safety Injection (HPSI) pump operability could be compromised due to air binding in the pump volute. This is postulated to occur at the time of the Recirculation Actuation Signal (RAS), when the HPSl and containment spray pump(s) suction shifts from the Refueling Water Tank (RWT) to the containment sump.

Two different scenarios of HPSl pump degradation have been analyzed. In the first scenario, LOCAs of various break sizes are analyzed with complete failure of the HPSl pumps afler RAS initiation. Since the Low Pressure Safety Injection (LPSI) pumps de-energize at RAS, the plant operator is assumed to restart one LPSl pump to maintain Reactor Coolant System (RCS) makeup flow, in accordance with plant emergency operating procedures. In the second scenario, the same LOCA transients are analyzed with degraded HPSl pump flow, for a duration of four minutes, after which the air in the pumps has been discharged into the system and pump performance is considered to return to normal. For this second scenario, there is no operator action to restart a LPSl pump. The degraded HPSl flow condition is based upon pump performance tests performed for this project at Wylie Corporation which is documented in an APS letter to the NRC, # 102-05195GRO/DGM/RAS, dated 12/27/2004.

Since this analysis is intended to look at past operation, best estimate conditions are assumed. This analysis is in no way considered to be part of the PVNGS licensing basis nor has it been performed to satisfy any requirements of 10CFR50.46.

The purpose of this report is to describe any detrimental effects (core uncovery) that occur or are exacerbated by the HPSl pump degradation (total loss of operability and / or degraded operation) during various small and medium break size LOCA events.

Break sizes of 1 to 10 inches in diameter are analyzed in both the cold discharge leg (CDL) and the Reactor Coolant Pump (RCP) Suction Leg (SL). Breaks smaller than one inch are not analyzed because they do not cause a Containment Spray Actuation Signal. Thus. sprays pumps are not needed and the time to RAS is sufficiently long to allow a plant cooldown and shifl to shutdown cooling. For these small breaks, pressurizer level is regained without RCS water levets dropping below the level of the hot legs. Breaks greater than 10 inches in diameter are not analyzed because RCS pressure is well below the LPSl pump shutoff head at the time of RAS. Therefore, Row from the LPSl pump, restarted by the operator after RAS, is greater than normal HPSl pump flow from two pumps. Thus, break sizes greater than 10 inches in diameter are not considered limiting. Only the two cold leg break locations are analyzed because any breaks in the hot leg would allow venting of steam produced in the reactor core directly to the containment, without need for loop seal clearing or WESTlNGHOUSE ELECTRIC COMPANY LLC

REDACTED VERSION DAR-OA-05-3 Rev. 0 Westinghouse Proprietary Class 3 -=fmwWw Page 5 draining. RCS depressurization occurs without depressing water level below the top of the core. Thus, cold side breaks are limiting regarding core uncovery. A sensitivity case with a break in the pressurizer was performed to verify the limiting nature of cold leg breaks.

This report was prepared according to Westinghouse Procedure WP 4.25,Rev.

2, 11/30/04, and is supported by Westinghouse Calculation Note CN-OA-05-1, Rev. 0, dated 02/11/05.

WESTINGHOUSE ELECTRIC COMPANY LLC

REDACTED VERSION Westinghouse Proprietary Class 3 DAR-OA-05-3 Rev. 0 Page 6 2.0 Code Description The Westinghouse CENTS computer code has been utilized for this analysis.

CENTS is an interactive, best-estimate simulation computer code that calculates the transient behavior of a PWR for plant maneuvers, accidents and operator actions, in a wide range of variations in plant state, from steady state to severe upsets, as well as lower mode operation at mid-loop.

A modular node-flowpath network models the primary system thermal-hydraulics.

Within each node, the model supports full thermal non-equilibrium, local pressures and thermodynamic properties, phase separation, bubble generation, flow regime dependent steam condensation, and transport dynamics of non-condensable gases, boron and radio-nuclides. A point kinetics model receives reactivity feedbacks from moderator, fuel, boron and rods, and an input axial shape. The core heat transfer model employs boiling curves over the full range of conditions, and calculates axialhadial temperature distributions in the fuel rod.

The primary sides of the steam generators (SGs) have detailed representation of the thermal profiles, accounting for forward and reverse heat transfer from relevant correlations. The coolant levels and their effect on the fluid state, heat transfer area and heat flux are modeled on both the primary and secondary sides. The secondary system representation provides sufficient detail for accurate modeling of the recirculation phenomena and the downcomer and evaporator water levels.

A modularwontml system.provides for generic-- _ _ definition of control logic for scram channels, rod control, emergency safety signais, primary and secondary system relief and makeup, and ancillary systems. Detailed control systems are designed via input from modules that perform standard arithmetic, integral-differential and logical transforms.

The origin of CENTS is the SBLOCA code, CEFIASH-4AS. CENTS is the RCS model set for several full scope simulators, including the NRC simulator for the CE 2700 Mwt design plant. It has since been licensed by the USNRC for Chapter 15 (non-LOCA) safety analyses of PWRs designed by CE and Westinghouse. There is an SER limitation placed on the code when used for referencing in licensing actions with respect to the calculation of transient behavior. It states that due to a lack of benchmarking provided in the topical report, that CENTS should not be used for LOCA licensing analysis for demonstrating compliance to 10CFR50.46 criteria. Nor shall it be used for severe accident analysis. However, it is acceptable for use in modeling small breaks for non-regulatory acceptance criteria. (Note that small breaks are usually defined to be approximately 1.Of? or less)

The analysis performed with CENTS in this report meets the above limitation, in that this is considered a best estimate analysis that is not used to assure WESTMGHOUSE ELECTRIC COMPANY LLC

DAR-OA-05-3 Rev. 0 Westinghouse Proprietary Class 3 -

REDACTED VERSION compliance with 10 CFR 50.46 criteria. Nor is the code being used to define core temperature conditions to ascertain if severe accident conditions exist.

A modification was made to the CENTS code to provide a loop seal model for the RCP suction lines. This was considered appropriate for this analysis because timing of the loop seal clearing for the cold discharge leg breaks is important in Page 7 determining RCS pressure response and core two-phase level at the time that loop seal clearing occurs. In support of the loop seal model added to the code, the PVNGS base deck configuration was re-nodalized in the suction leg regions to employ two nodes for each suction leg from the steam generator to the RCP.

As a check on the CENTS loop seal model design, benchmark cases were run against an analysis performed with the CEFLASH-LSAS REM code, a Westinghouse best estimate SBLOCA code. These benchmark cases were performed for the Waterford -3 plant, which is a CE design PWR of similar size, power ,level and loop seal design to that of the PVNGS Units. Three inch CDL and SL breaks were analyzed for this benchmark. The transient attributes of interest in the benchmark were the behavior of the loop seal (Le. the timing of the loop seal clearing and its affect on break flow and enthalpy) and the overall RCS pressure and core level during the transient. The benchmark showed that the behavior of the loop seal model was in good agreement between the two codes.

This supports the acceptability of the loop seal modeling modifications made to the CENTS code.

3.0 Case Descriptions & Input Parameters 3.1 COMMON INITIAL CONDlTfONS AND PLANT PARAMETERS 3.1.1 Initial Plant Conditions The initial plant conditions are identical for all cases and represent nominal full power parameters.

0 Core Power Level: 100% (3876 MM)

-_ Core Inlet Temperature: 553'F

-- - - --. 0 Pressurizer Pressure: 2250 psia

. Core Flow: 45500 Ibmkec Pressurizer Level: 21 -2ft Steam Generator Level: 37.3 ft 0 Feedwater Enthalpy: 408.4 BTU/lbm WESTINGHOUSE ELECTRIC COMPANY LLC

REDACTED VERSION Westinghouse Proprietary Class 3 7 DAR-OA-OS3 Rev. 0 Page 8 3.1.2 ECCS Parameters The initial ECCS conditions and assumptions are the same for all cases except the sensitivity cases. These parameters are as foliows.

WESTINGHOUSE ELECTRJC COMPANY LLC

DAR-OA-05-3 Rev. 0 Westinghouse Proprietary Class 3 REDACTED VERSION Page 9 3.2 BREAK PARAMETERS 3.3 CORE DECAY HEAT WESTMGHOUSE ELECTRlC COMPANY LLC

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3.5 OPERATOR ACTIONS Operator actions are in accordance with the APS emergency operating procedures. In particular, the Loss of Coolant Accident Procedure [ I , was used to determine the simulated operator responses to the transient. The actions taken are similar for all t h e cases analyzed, though the timing of some actions is different for each case. The actions are summarized as follows:

Secure two RCPs 5 minutes after reactor trip. The "Trip 2 - Leave 2" strategy is based upon step 7 of the procedure.

Secure all RCPs if subcooling is ~ 2 4 ° F .This action is also based upon step 7 of the procedure. The initial waiting period of 5 minutes for securing RCPs is based upon the time for operator diagnosis of the situation.

Cooldown the plant to Shutdown Cooling Entry conditions. This is assumed to start at 1500 seconds, based upon a reasonable delay for the plant operator to assess the situation and take immediate post trip actions, etc. This is based upon step 22 of the procedure. An aggressive cooldown rate of 9O"Flhr is assumed. It is assumed that the operator uses the core exit temperature at 1500 seconds as the starting point. Thereafter, the operator will not perform any action to exceed the 90°F/hr rate. Note that for many of the cases analyzed, particularly the larger breaks, the cooldown rate may greatly exceed the procedural limit of 1OO"Flhr due to energy loss out the break and not due to operator action. In this case the operator is only cooling the steam generator secondary by relieving steam through the automatic dump valves, but this does not affect the RCS cooldown, as long as steam generator secondary temperature is greater than RCS temperature. [Note that the actual cooldown rate setpoint used in the CENTS code controllers was set at 85"Flhr since the controllers simulate action by the operator every 300 seconds. Since there is this set frequency of action, 85"Flhr is conservatively used to help assure that 90°F/hr is not exceeded for certain time intervals.]

Secure 1 of 3 Charging Pumps when RWT level approaches 50%, secure a second Charging Pump at 40% RVVT level and secure the third pump at 30% RWT level. This is based upon step 48 of the procedure.

0 When RAS occurs, if both HPSl pumps completely fail, it is assumed that the operator will restart a LPSl Pump. This is based upon functional recovery guidelines to maintain a source of reactor makeup water. In the cases of this analysis, restarting a LPSl pump is assumed to occur as soon as the HPSI pumps are lost. For those cases where the RCS pressure is below the LPSl shutoff head, this means flow is never lost. If RCS pressure is above the LPSl shutoff head as it is for the smaller break sizes, then ECCS pump flow ceases till pressure drops.

WESTMGHOUSE ELECTRIC COMPANY LLC

REDACTED VERSION Westinghouse Proprietary Class 3 DAR-OA-05-3 Rev. 0 Page 13 3.6 SENSITIVITY CASES Seven sensitivity cases have been analyzed to support this analysis. The sensitivrty cases are intended to show the effects on overall case acceptability for those parameters which play an important role in the transient and could vary in some significant way from the values chosen for the various series of cases.

Details are discussed below.

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DAR-OA-05-3 Rev. 0 5.0 Case Results WESTMGHOUSE ELECTRIC COMPANY LLC

..,I.-.

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DAR-OA-05-3 Rev. 0 Westinghouse Proprietary Class 3 REDACTED VERSION 2o 5.2 DISCUSSION OF INDIVIDUAL CASE RESULTS In the discussion below, the failed HPSI and degraded HPSl cases are discussed together. Prior to RAS these cases are identical. After RAS, it is useful to compare how the relative ECCS flows affect the remainder of the events.

A review of the figures showing ECCS flow provides some perspective on the overall effect of degraded HPSl flow for four minutes, after RAS. As an example, for the [ I CDL break with degraded HPSI,Figure 7.2.3.3shows the ECCS flow.

RAS occurs shortly after [ 1 seconds. A visual review of the degraded HPSl flow indicates that the depleted flow is a very small portion of the integrated flow over the course of the event. It would be expected to have very little effect on event results. This fact is supported by the Sensitivity case [ 1 which show that nominal vs. degraded HPSl flow does not significantly change case results.

5.2.1 Series 1 8 2 Cases: Cold Discharge Leg (CDL) Breaks CDL -1 WESTINGHOUSE ELECTRIC COMPANY LLC

WESTMGHOUSE ELECTRlC COMPANY LLC WESTMGHOUSE ELECTRlC COMPANY LLC DAR-OA-05-3 Rev. 0 5.2.2 Westinghouse Proprietary Class 3 Series 3 8 4 Cases: Suction Leg (SL) Breaks REDACTED VERSION Page 23 3

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DAR-OA-05-3 Rev. 0 5.3 CASE

SUMMARY

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WESTINGHOUSE ELECTRIC COMPANY LLC Westinghouse Propietary Class 3 REDACTED VERSION DAR-OA-05-3 Rev. 0 Page 29 6.0 Conclusions The series of cases described above show that degraded HPSl flow caused by the air in the ECCS sump suction line will not lead to situations where core uncovery would occur. Two cases with total HPSl pump failure at RAS led to some partial core uncovery for an extended period of time, due to a depletion of RCS inventory. Those were the 3" and 4" CDL breaks. There were no cases with degraded HPSl pump flow which had any partial core uncovery associated with the degraded ECCS Row.

There were some additional cases, both failed and degraded HPSl flow cases, that showed short periods of partial uncovery due to loop seals filling and clearing; however, this phenomenon is expected for both CDL and SL breaks and is not due to the degraded flow in the ECCS system. This was verified by Sensitivity Case 7.

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Westinghouse Proprietary Class 3 REDACTED VERSION DAR-OA-05-3 Rev. 0 Page 30 7.0 Figures 7.1 SERIES I: CDL BREAKS, FAILED HPSl AFTER RAS 7.1.1 CDL-3 WESTJNGHOUSE ELECTRIC COMPANY LLC

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