ML18026A417

Forwards Revised Response to NRC Station Blackout Safety Evaluation Dtd 920114,revising Diesel Generator Target Reliability to 0.975,based on NRC Position.Resolution of Cabinet Temp Concern Will Be Submitted by 920501

ML18026A417
Person / Time
Site:	Susquehanna
Issue date:	03/13/1992
From:	Keiser H PENNSYLVANIA POWER & LIGHT CO.
To:	Chris Miller Office of Nuclear Reactor Regulation
Shared Package
ML17157B098	List: ML17157B097 ML18026A417
References
PLA-3745, TAC-M68613, TAC-M68614, NUDOCS 9203230281
Download: ML18026A417 (34)

Text

., ACCELERATED DISTRIBUTION DEMONST$A.TION SYSTEM REGULA . Y INFORMATION DISTRIBUTIO. SYSTEM (RIDS)

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'7 ACCESSION NBR:9203230281 DOC.DATE: 92/03/13 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION KEISERgH.W. Pennsylvania Power & Light Co.

RECIP.NAME REC1PIENT AFFILIATION MILLER.C.L. Project Directorate I-2 R

SUBJECT:

Forwards util revised response to Station Blackout Rule per NRC 920114 Safety Evaluation w/answers attached to all but I j

one NRC recommendation. Query on CR instrument cabinet temp to be answered no later than 920501. D, DISTRIBUTION CODE: A050D TITLE: OR COPIES RECEIVED:LTR Submittal: Station Blackout (USI A-44) 10CFR50.63, ENCL SIZE:

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05000387 05000388 A D

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-2 PD 1 1 RALEIGHiJ. 1 1 INTERNAL: ACRS 1 1 AEOD/DSP/TPAB 1 1 NRR PD2-4PM TAM 1 1 NRR/DET/ESGB 8D 2 2 NRR/DST/SELB 3 3 NRR DST/ PLB8D1 3 3 NRR/DST/SRXB8E 1 1 G FILE 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 NOTES: 2 2 D

A D

D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WAS'ONTACT THE DOCUMENT CONTROL DESK.

ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISIRIBUTION LINIS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 19 ENCL 19

Pennsylvania Power 8 Light Company Two North Ninth Street ~ Allentown, PA 18101-1179 ~ 215/774-5151 Harold W. Keiser Senior Vice President-Nuclear 215/774<194 NR l 3 1992 Director of Nuclear Reactor Regulation Attention: Mr. C.L. Miller, Project Director Project Directorate I-2 Division of Reactor Projects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO STATION BLACKOUT SAFf"TY EVALUATION PLA-3745 FILE R41-2

Reference:

RESPONSE TO THE STATION BLACKOUTRULE FOR SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 AND 2 PAC NOS. M68613 AND M68614) Dated January 14, 1992.

Dear Mr. Miller:

This letter provides the Pennsylvania Power & Light Company (PP&L) revised response to the Station Blackout (SBO) Rule as required by the referenced NRC Safety Evaluation.

This response (attached) revises diesel generator target reliability to 0.975 based on your position, and provides the requested justification to support PP&L's original position that SSES is only required to cope with a SBO event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. However, it should be noted that a thorough evaluation was undertaken to review the staff s concerns regarding the need and ability for SSES to cope for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Results of this evaluation concluded SSES has the capability to cope for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and longer if required.

With the exception of a final technical resolution to your question regarding Control Room instrument cabinet temperatures, the attachment responds in full to each of your recommendations. Our resolution to the cabinet temperature concern will be forwarded to you no later than May 1, 1992.

9203230281 920313 PDR ADOCK 05000387 P PDR

FILE R41-2 PLA-3745 Mr. C. L. Miller Questions regarding this revised response should be directed to Mr. A.K. Maron at (215) 774-7852.

Very truly yours, H. W. Keiser Attachment cc: NRC3)ocnment:Control DeaR (original)

NRC Region I Mr. G. S. Barber, NRC Sr. Resident Inspector - SSES Mr. J. J. Raleigh, NRC Project Manager - Rockville

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,9203230281 ATTACHMENTTO PLA-3745 I~NTR DUCTI N The Station Blackout Rule (10 CFR 50.63) was instituted in 1988 and required licensees to assess their ability to cope with a station blackout (SBO) of a specified duration. In 1989, PP&L submitted the results of our coping study to the NRC, concluding that Susquehanna SES (SSES) must be able to cope with a station blackout for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and maintain an Emergency Diesel Generator (EDG) reliability of 0.975 (97.5%). In February of 1991, PP&L revised its EDG target reliability value from 0.975 to 0.95 based on a spray pond bypass valve modification.

On January 14, 1992, NRC issued its Safety Evaluation of the SSES SBO submittal concluding that SSES was an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> coping plant requiring EDG reliability be maintained at 0.975. The following is an item by item response to the recommendations identified in the NRC Safety Evaluation.

t c-".STATION,::::>SL'ATCKOUT:;,::::DUR'ATION>".l NR RK P~PRL R MMENDATION: The licensee needs to change the EDG reliability target from 0.95 to 0.975 and the coping duration from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

A) Coping Duration One input to the determination of required SBO coping duration is the "return time" of extremely high winds(>125 mph). As part of our original coping assessment, PP&L contracted with Dames & Moore Consulting Engineers for the calculation of this "return time" for SSES. Dames & Moore determined this value to be -6.7E-4/yr. (about once in 1500 years) using data specific to SSES. Any return time value less than 1.OE-3/yr, coupled with our severe weather and off-site power design classification, places SSES in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping category.

The NRC evaluation did not credit use of site specific data due to this data being applicable for winds at 10 meters off the ground, rather than the required assessment height of 30 meters from the ground (average transmission tower height). It was therefore concluded, based on NUMARC Table 3.2, that the return time for SSES was more frequent than once per 1000 years and that SSES must cope with a SBO for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, Page 1

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ATTACHMENTTO PLA-3745 To address this coping duration concern, PP&L investigated the basis of Table 3.2 in NUMARC 87-00 and contracted again with Dames & Moore to determine the return time of wind speeds at 30 meters. Conversations with both NUMARC personnel and NRC staff indicated that the use of site specific data is acceptable. The NRC cautioned that the use of such data should account for wind speeds of 125 mph at 30 meters and consider National Bureau of Standards (NBS) publications 118 and 124, as well as several National Oceanic and Atmospheric Administration (NOAA) documents. Note that the use of site specific data is encouraged in NUMARC 87-00.

NBS 118 provides a method of scaling wind speeds to various heights and provides measured weather data from 129 meteorological stations across the US mainland. It is this data which PP&L and Dames & Moore believe provides the best estimates of wind speed return times at SSES. Using the method of NBS 118, the 125 mph "fastest mile" wind speed at 30 meters is scaled to a "fastest mile" wind speed of 107 mph at 10 meters (the normalized height of all reported weather data). Using the data for meteorological stations closest to SSES, NBS 118 provides the following "return times" for various fastest mile speeds:

Fastest Mile Wind Speed (mph)

Return Time ears Scranton Harrisburg 60.86 70.57 67.34 80.49 1,000 70.12 84.75 5,000 76.58 94.64 10,000 79.36 98.90 50,000 85.82 108.79 100,000 88.60 113.05 500,000 95.06 122.95 1,000,000 97.84 127.21 Page 2

ATTACHMENTTO PLA-3745 In addition, Dames & Moore have calculated the probability of exceeding various wind speeds within 1000 years, also based on the data and methods in a paper by H.C.S.Thorn:

Probability of Fastest Mile Wind Speed (mph)

Exceedance in 1000 rs Scranton Harrisburg 0.500 72 87 0.250 75 92 0.100 79 99 0.050 82 103 0.005 92 117 From the first table above, one can see that the return time of a wind speed of 107 mph at 10 meters is expected to be greater than 1 million years at Scranton and almost 50,000 years at Harrisburg. Table 2 shows'that the probability of exceeding the 107 mph wind speed within 1000 years is less than 1% at Scranton and about 3% at Harrisburg. Using the data from Harrisburg in Table 1, the expected return time of a 125 mph wind at 30 meters is -37,500 years. PP&L also-reviewed NBS, 124 for applicability. NBS 124 relies on the extrapolation of coastal weather data to infer wind speeds inland. Further, this method of extrapolation assumes intervening terrain to be open and grass covered. Since SSES is located within a valley separated from the coast by approximately 100 miles of hills and forest, the extrapolation is highly inaccurate. Thus, PP&L views NBS 124 as valid only for scoping calculations and should only be used in the absence of better techniques/data.

PP&L considers the preceding arguments and data sufficient justification for not using Table 3.2 of NUMARC 87-00 for determining our ESW category. Further, this data shows that the return time of winds in excess of 125 mph at SSES is highly likely to be greater than 1000 years.

Thus, it is concluded that the ESW category of "2" originally reported in our coping study is fullyjustified (the data actually justifies an ESW classification of "1"), and that SSES remains a "Pl" plant (per NUMARC 87-00) requiring a SBO coping time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

B) EDG Target Reliability In 1991, PP&L informed the NRC that for purposes of complying with the SBO rule our target EDG reliability was to be 0.95 (95%). In making this determination, PP&L relied on the use of "staggered operation" of RHR pumps to cool both suppression pools.

Staggered operation is required because, although in principle any two EDG s can cool both units, in actuality there are two combinations of EDG's (A and C, or B and D) which result in only one RHR pump in each unit available to alternately cool the suppression pools.

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i ATTACHMENTTO PLA-3745 e The NRC noted that the use of staggered operation did not meet the "connectability criterion" and was determined to be an unacceptable increase in operator burden. This criterion was explained in documentation provided by the NRC to NUMARC after submittal of the SSES SBO analysis. The NRC concluded that to avoid use of staggered operation, 3 of the 4 EDG's would be required.

Further, the NRC noted that if only diesels A and B start, no control structure HVAC would be available. PP&L has performed a calculation of steady state control room temperature using the method in NUMARC 87-00 and assuming that the measured, normal control room heat load exists. The result of this calculation is that the control room temperature will not rise above 111'F in the absence of normal HVAC. Because temperature remains less than 120'F, the control structure environment remains acceptable.

Based on the inability to take credit for staggered operation, PP&L concurs with the staff's position in" requiring 3 of 4 EDGs and the reliability target value of 0.975.

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;::'.SL'A'CEO'::,:,::COPINO.":..:.CAPASILITY::,::,::,'=:,',ll NR REC MMENDATI N

'he NRC made the following four recommendations based on their previous determination that PP&L had to address the need for SSES to cope with an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Station Blackout.

1) The licensee needs to conform to an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> coping duration and increase the EDG reliability target from 0.95 to 0.975.

2) The licensee should provide a procedure to refill the CST from the RWST during an SBO event.

3) The licensee should add the portable AC generator to the list of SBO equipment, provide procedures for its utilization, and apply to it an appropriate QA program.

The portable ac generator should meet the criteria in Appendix B of NUMARC 87-00. Also the licensee should replace battery 1D650 with a higher capacity battery or provide charging capability to the existing battery to extend its support for the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> SBO duration, and recovery thereafter. The licensee should include all the analyses and related information in supporting documentation that is to be maintained by the licensee for possible staff review.

4) The licensee should provide for staff review a full description, including the nature and objectives of any modification required. The analyses and related information should also be included in the supporting documentation that is to be maintained by the licensee in support of the SBO submittals.

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~R ATTACHMENTTO PLA-3745

'2 R P~PLR N R As addressed in the initial section of this response, PP&L concludes that SSES must cope with a SBO event for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This conclusion is supported by. the use of site specific weather data (at the required assessment height). As for the EDG reliability target value, PP&L has reviewed the NRC concerns and has concurred with the staff's finding that the configuration of SSES mandates an EDG reliability target value of 0.975. This reliability value has been included in the EDG Reliability Program developed in accordance with NUMARC 87-00 Appendix D.

PP&L has thoroughly evaluated the ability of SSES to cope 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with an SBO event, including all areas of concern identified in the NRC Safety Evaluation. PP&L is confident that SSES has the ability to cope for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and longer ifrequired. Since PP&L has demonstrated that SSES is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping plant this information will not be provided in support of our revised submittal, but is available for review.

';"':;EFFECTS::.".,"OF.,:,:LOSS:;OF..'",VENTILh;TION~:;i NR REC MMENDATI The licensee should: I) provide additional information and/or technical justification for the initial conditions and assumptions used in the heat-up analysis for each area of concern, 2) with regard to COTTAP computer code, provide detailed information to address the staff's concerns as identified above, and 3) re-perform the heat-up analysis for each area of concern and for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> coping duration taking into account the non-conservatism as identified in the SAIC TER.

P~PRL R N R CCPPAP2 C 1 The use of the Compartment Temperature Transient Analysis Program (COITAP) computer code has been presented to the staff as part of our submittals to resolve steam leak detection Technical Specification changes. Attachment A contains a user's manual for the COTTAP computer code and a copy of a recent paper published in Nuclear Technology which describes the methodology used in the COTI'AP program and presents some of the verification calculations which have been performed. The user's manual presents some of the calculations which were performed against problems that have exact analytical solutions. The referred paper presents the methodology along with calculations which have been benchmarked against calculations performed with the CONTAIN computer program. In addition, the program and computation package have been independently reviewed by Gilbert Associates. PP&L also maintains a Quality Assurance file/package for the COTTAP computer code.

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~t of ATTACHMPITTO PLA-3745 In the original coping assessment, two basic COTTAP2 calculations were performed: an assessment of Dominant Areas of Concern (DACs); and an evaluation of control room cabinets.

For BWRs, the DACs are the HPCI and RCIC rooms, and the main steam tunnel (NUMARC 87-00). The main steam tunnel is considered because, apparently at some plants, HPCI and RCIC are isolated on high temperature in the tunnel. At SSES, the HPCVRCIC isolations do not come from main steam tunnel temperature but from sensors located on the 683 foot elevation of the reactor building common to both HPCI and RCIC piping. During SBO, only the RCIC isolation logic is powered. Thus, for SSES, the main steam tunnel is not a true DAC. The common piping area, called the RHR piping area in the calculation, is a DAC.

PP&L recalculated the DAC temperatures using CO1TAP2 and "conservative" inputs. Inputs included use of "maximum normal" room temperatures per the FSAR. Outside air temperature was assumed to be a constant 95'F. The influence. of hot piping (including flued heads) was added to the HPCI, RCIC, RHR piping area, and the main steam tunnel. (The absence of this hot pipe loading caused the cooldown of the main steam tunnel noted in the SAIC Technical Evaluation Report). No engineering reference for a con'crete thermal 'conductivity of 0.7 could be found. However, this value was changed from 1.0 to 0.7, per the TER. The actual input deck, and the justification for all input values used, appears in the detailed calculation.

The results of the COTTAP2 calculations are presented in the tables below.

Temperature ('F)

Original Submittal: New Calculation:

ROOM 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> HPCI 113 114 114 119 RCIC 107 RHR Piping 118 117 125 130 MS Tunnel 123 117 150 171 From Table 3, the temperatures of the DACs remain less than the 180'F operability limit, even at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The inclusion of the hot pipe. loads does cause, significant increases in tunnel temperatures.

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ATTACHMENTTO PLA-3745 0

Temperature ('F)

ROOM COTTAP2 at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> NUMARC 87-00 RHR Piping 130 MS Tunnel 171 176 Table 4 presents a comparison of the two hottest DAC temperatures as calculated by both COTTAP2 and the method of NUMARC 87-00. While it appears that the NUMARC method produces "conservative" results, it must be noted that the NUMARC calculation produces a steady state, infinite time result. The COTTAP2 results are not steady state but time dependent, and at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the temperatures in these rooms are still increasing. At longer and longer times, one would expect better agreement between the two methods. The results of the above table show that the agreement between the two methods is quite good.

The TER made reference to "oscillatory" temperature profiles. Review of the original COTI'AP2 work revealed no such profiles. The reviewers may be referring to temperature profiles which peak and drop in the short term, then continue a long term temperature rise (Figure 1). The large early peak is caused by AC motor heat loads which decay away. At later times, the room is heated by surrounding walls. This result is consistent with expected behavior.

The reviewers questioned PP&L's use of COTTAP2 for calculation of instrument cabinet temperatures and several assumptions used in these calculations. The original impetus for using CORI'AP2 to calculate cabinet temperatures was the desire to avoid opening control structure cabinet doors and not impose unnecessary operator burden.

PP&L concurs with the NRC that modifications are needed to two assumptions used in the cabinet temperature calculations. The NRC questioned our use of 120'F as the control room temperature, implying such a temperature was overly conservative. In response, the infinite time control room temperature, assuming measured normal operating heat loads, has been calculated using the method of NUMARC 87-00. The resulting control room temperature is 111'F. The TER questioned use of 180'F as the operability limit of control room instruments. Based on information received from equipment manufacturers, we currently believe the correct limit is 140'F, and are performing a reevaluation on this basis. This evaluation will be completed and submitted to the NRC no later then May 1, 1992, Page 7

<r ATTACHMENTTO PLA-3745

';:;CONTAPC41i22lT,!ISOLATION','R RE MMEND ATION'he licensee should list the valves identified in an appropriate procedure and identify the actions necessary to ensure that these valves can be fully closed, ifcontainment isolation is required during an SBO event. The valve closure should be confirmed by position indication (local, mechanical, remote, process information, etc.)

P&L R The penetrations which have been identified by the NRC as requiring to be proceduralized are the Residual Heat Removal (RHR) and Core Spray (CS) suction lines along with the Containment Spray line. Containment isolation of these lines has been addressed and approved by the NRC prior to this submittal. The following identifies that approved approach.

Susquehanna SES FSAR section 6.2.4.3.6 states in part that "Containment isolation provisions for certain lines in engineered safety feature or engineered safety feature-related systems may consist of a single isolation valve outside containment. A single isolation valve is considered acceptable ifit can be shown that the system reliability is greater with only one isolation valve in the line, the system is closed outside containment, and a single active failure can be accommodated with only one isolation valve in the line." Additionally, section 6.2.4.3.6.3 states, "Although strictly speaking the HPCI, RCIC, CS, and RHR pump suction lines do not connect directly to the primary containment, they are nevertheless evaluated to 10 CFR 50 Appendix A, General Design Criteria 56. These lines are each provided with one remote manually motor operated gate valve external to the containment and use the respective piping systems as the second isolation, barrier.. For the RHR and CS valves the hand switches are key locked".

Further investigation into this issue reveals that section 6.2.4 of the NRC Safety Evaluation Report (NUREG 0776) for Susquehanna SSES documents the NRC approval of meeting the alternative acceptance criteria specified in section 6.2.4 of the Standard Review Plan. This section summarizes these alternative acceptance criteria along with specifically identifying the lines found acceptable via this method.

Based on the above explanation we believe that containment isolation is established and containment integrity will be maintained.

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ATTACHMENTTO PLA-3745

1'R'O.CEDURFS.:;::lANDl,TRAXNING,',

RE MME ATI The staff expects the licensee to implement the appropriate training to assure an effective response to an SBO event.

PP&LR N E Appropriate plant personnel willbe trained on any new or revised procedures in accordance with the requirements of Initiative 2, NUMARC 87-00 and Reg.Guide 1.155, section 3.4.

'-;:QUALITY>'A'SSUR'A'NCE'"'.AND;:::TECHggCAL"'-::,SPECIPICATION~):'R RE MME ATI The staff expects that the plant procedures will reflect the appropriate testing and surveillance requirements to ensure the operability of the necessary SBO equipment,

'P&L' f

4 It is PP&L s intent to satisfy the Quality Assurance (QA) requirements of Reg. Guide 1.155 by upgrading an existing procedure to incorporate Station Blackout. This procedure addresses all the Reg. Guide QA requirements and will require the necessary Inspections and Tests to be performed in accorda'nce with the Operational Quality Assurance Program.

-;ED 6'!RELIA'SILIIYiPROGRAM::::..":

NR RK MMENDATI N'he licensee should complete the implementation of an EDG reliability program which meets the guidance of RG 1.155, Section 1.2 and provide a schedule for its completion. Confirmation that such a program is in place or will be implemented should be included in the documentation supporting the SBO submittals that is to be maintained by the licensee.

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ATTACHMENTTO PLA-3745 PP&L R Reg. Guide 1.155 specifies that each utility establish an EDG performance monitoring program.

NUMARC 87-00 Appendix D contains guidance for the development and implementation of such a program. PP&L has committed to implement a program of reliability monitoring and, as indicated above, PP&L must maintain an EDG reliability at or above 97.5% as part of our SBO coping strategy.

The Reg. Guide and NUMARC provide "trigger values" for determining compliance with target reliability. NRC reviewers indicated that lack of this data in our submittal hindered assessment of SSES EDG reliability. At the 97.5% reliability level, compliance is assumed if the failures to start/load are less than or equal to 3, 4, and 5 out of the last 20, 50 and 100 start attempts, respectively. As of 2/10/92 the failures to start/load in each category were 0,0, and 3, respectively. Thus, today, PP&L can accept the increased reliability target of 97.5%.

PP&L's Emergency Diesel Generator reliability monitoring program has been developed and documented in Nuclear Department Administrative Procedure-QA-0401 entitled "Emergency Diesel Generator Monitoring Program." This procedure complies with the reliability requirements delineated in Appendix D of NUMARC 87-00, Rev. 1. Reliability will be monitored against a set of "trigger values" with actions specified for various levels of trigger value exceedance.

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ROOM TEMPERATURE RESPONSE TO A STATION BLACKOUT 200 180 160 140

~ I 120 I Legend g HVAC EQUIP RM 0 EXH fAN RM 100

~ HVAC EQUIP RM 0 HVAC EQUIP RM 6 RECIRC PLENUM 80 10 20 30 40 50 60 70 80 TIME (HRS)

COTTAP: A COMPUTER CODE ~ ~ ~

FOR SIMULATION OF THERMAL TRANSIENTS IN SECONDARY CONTAINMENTS OF BOILING WATER REACTORS MARK A. CHAIKO and MICHAELJ. MURPHY "7 Pennsylvania Power & Light Company, Allentown, Pennsylvania 18101 Received December I, 1989 Accepted for Publication September 12, 1990 The Compartment Transient Temperature Analysis out the calculation on an IBM3090 computer. The Program (COTTAP) was developed by the Pennsylva- COTTAP code considers natural convection and radi-nia Power & Light Company for postaccident boiling ation heat transfer between compartment air and walls water'reactor (BWR) secondary containment thermal 'hrough a detailedflnite difference solution of the slab analysis. The code makes use ofpreviously developed conduction equations. Heat addition from hot piping implicit temporal integration methods and sparse ma- and operating equipment, and cooling effects associated trix inversion techniques to allow modeling of an en- with ventilation flows and compartment heat removal tire BWR secondary containment. Investigations were units are also included. Additional capabilities of made with a model consisting of 121 compartments COTTA P include modeling of compartment heatup re-and 767 heat-conducting slabs. The simulation pre- sulting from steamline breaks and simulation of nat-sented involves the numerical integration of20 101 or- ural circulation cooling in compartments with flow dinary differential equations over a 30-h simulation paths at differing elevations.

period. Two hours of CPU time were required to carry I. INTRODUCTION ventilation system operates in a recirculation mode to promote air mixing between compartments and Under postaccident conditions, boiling water reac- to dilute locally concentrated radioactive isotopes.

tor (BWR) secondary containment ventilation systems Original design calculations for Pennsylvania Power typically isolate to prevent fission product release to & Light Company's (PP&L) Susquehanna Steam the environment. Since cooled air is no longer circu- Electric Station (SSES) assumed that air recircula-lated through the secondary containment, increased tion provided enough mixing to produce a fairly compartment temperatures result. Predictions of post- uniform temperature distribution throughout all sec-accident compartment temperatures are necessary to ondary containment compartments. For this reason, determine whether safety-related equipment is sub- a single-compartment transient. model was used in the jected to temperatures that'exceed its maximum design simulation of postaccident conditions. Recent investi-values. Safety-related equipment must be operable un- gations based on steady-state calculations have shown, der postaccident conditions in order to effect the safe however, that significant temperature variations can shutdown of the reactor. exist between compartments. These temperature After an accident, the secondary containment variations were large enough to prompt a detailed NUCLEAR TECHNOLOGY VOL. 94 APR. 199l

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS multicompartment transient analysis of the secondary and containment.

To reanalyze the postaccident transient behavior of V "

dp = Ivu Jvl g Wj(1 Y,j) + g Wg(1 YIJ) the SSES secondary containment, PP&L developed the dl jai jai Compartment Transient Temperature Analysis Pro- H~

gram (COTTAP). Development of this program began after an evaluation of available codes revealed that

+ g Wy(Y Ycj) + W~ Woold Wlo j=l none were capable of performing a sufficiently detailed (2) simulation owing to the large number of heat-conduct- where ing structures found in the SSES secondary contain-ment. For example, the CONTEMPT code,'hich is V = compartment volume (m3) probably the most widely used containment analysis t = time (s) program, can model as many as 999 compartments but is limited to 99 heat-conducting slabs. In contrast, pp= compartment air and water vapor COTTAP can model up to 1200 heat-conducting slabs densities, respectively (kg/m3) and 300 compartments. It also contains models that W mass flow rates associated with describe heat dissipation from operating electrical WJ WIJ j j'th ventilation, leakage, and cir-equipment and process piping. A COTTAP model of culation paths, respectively (kg/s) the SSES-1 and -2 secondary containment structures consists of -120 compartments and 800 heat-conduct- Y = mass fraction of air within com-ing slabs. partment The CONTAIN code2 is a more recently developed Yj, YIJ air mass fractions in donor com-containment simulation program with complex mod-eling capabilities. It is, however, designed specifically partments for ventilation path and leakage path j, respectively j

for primary containment simulation and is not well suited for secondary containment modeling because it Y~

mass fraction of air in adjoining has no provisions for energy input to compartments compartment associated with cir-.

from heat loads such as electrical panels, lighting, mo-tors, and hot piping.

culation path j Wq, = rate of steam addition due to pipe A description of the COTTAP code, including as- breaks (kg/s) sumptions, governing equations, numerical solution methods, and code limitations is given in Sec. II. Rep- Wd = steam condensation rate (kg/s) resentative results of the SSES-1 and -2 secondary con- W= rain-out rate (kg/s).

tainment analysis are presented in Sec. III, and code verification is discussed in Sec. IV. The values Wj and Wlj are positive for flow into the compartment and negative for flow out of the com-II. DESCRIPTION OF THE COTTAP CODE partment, whereas the circulation rate Wj is always a positive quantity. Ventilation paths are described by II,A. Compartment Mass and Energy Balances associated mass flow rates and identification 'heir The COTTAP code allows for air and water vapor numbers of source and receiving compartments. Ven-mass transfer between compartments by means of tilation flows can be tripped off or on at any time dur-forced ventilation, leakage, and natural, circulation ing a transient by supplying appropriate trip-logic data.

flows. A forced ventilation flow model describes heat- Leakage, circulation, and pipe break models are dis-ing/ventilating/air conditioning systems, and a leakage cussed in Sec. II.C along with other special purpose model simulates intercompartment flows that hre gen- models.

erated by pressure differentials. In addition, a natural In formulating the compartment energy balance, it circulation model simulates gravity4riven flows between is assumed that air behaves as an ideal gas. Moreover,-'or compartments connected by flow paths at differing the transients of interest, partial pressures of wa-elevations. Steam can also be added to a compart- ter vapor are typically (I atm. Therefore, it is assumed ment as a result of pipe breaks or removed through that the steam specific enthalpy depends only on tem-condensation and rain-out. Air and water vapor mass perature, i.e., the vapor enthalpy is equal to the en-conservation equations for a compartment with N thalpy of saturated steam at the temperature of the.gas ventilation paths, NI leakage paths, and N, natural cir- mixture. The partial pressure of water vapor within a culation paths are given by compartment is computed from the ideal gas equation H Nc of state, and the total compartment pressure is calcu-V g WojYoj + j~i J1 g Woj(Y<j g WgYIJ + jmi Y) lated as the sum of the air and water vapor partial pressures. With these assumptions, the compartment energy balance becomes NUCLEAR TECHNOLOGY VOL. 94 APR. 1991

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS Pb,k = total compartment pressure if pipe V PaT + PaCpa(T) contains saturated liquid (Pa)

Pb<<ak = pipe fluid pressure if pipe contains dhg(T) dT saturated steam (Pa)

+ P>r dT dr hg(Pb,k) = specific enthalpy of saturated water

= -VTC (T) dPa dC Vh (T) dP~

dj' vapor at pressure Pb,k (J/kg) hi(T) = specific enthalpy of saturated liquid VT R "+R, water at temperature T(J/kg)

T J, T> donor compartment temperatures for

+ Qligitt + Qpanel + Qmotor + Qcooler + Qpiping j ventilation path and leakage path respectively (K) j;

+ Qmisc + Qslab + Qbreak + Wbsilg(Pbreak)

Wr o ) J ( T) Wco ad hg ( T) 1 TJ temperature in adjoining compart-ment associated with circulation path

+

iVu g Wt>J[Y>>r)To)Cpa(Ttrj) + (I Yoj)hg(Tpj)]

j (K).

J>>> 1 Compartment heat loads from lighting, electrical pan-iVI els, motors, and miscellaneous equipment are main-

+ g Wlj[ Yi)TJJ Cpa (Tj)) + (

)=I 1 Yg) hg (T(J)] tained constant unless they are tripped on, off, or exponentially decayed during the transient. Hot piping Nc and room cooler loads vary with compartment temper-

+ g Wcj [ Yc)TcJCpa(TcJ)

J=l YTCpa (T) ature and can also be tripped on or off. In addition, hot piping heat loads can be exponentially decayed

+ (1 Yy)hg(T)) using the heat load decay model discussed in Sec.

(1 Y) hg (T)], (3) II.C.7.

where II.B. Slab Model T = compartment gas temperature (K)

In the secondary containment of a BWR, compart-Cp,(T) = specific heat of air at temperature T ment walls, ceilings, and floors are generally concrete (J/kg K) slabs that range in thickness from -0.3 to -2 m. To hg(T) = specific enthalpy of saturated water determine the heat transfer rate between a compart-vapor at temperature T (J/kg) ment atmosphere and the bounding concrete slabs, the one-dimensional heat conduction equation R, = ideal gas constant for air (288.7 J/

kg K)

(4)

R = ideal gas constant for water (461.4 J/kg K) is solved for each slab. Here, T, (K) is the slab temper-Qligbt, Qpanel>> Qmotor> Qcooler>> Qpiping > Qrnisc ature, and x (m) is the spatial coordinate. Since the

= compartment heat loads due to light-thermal diffusivity ns (m /s) is supplied as input for each slab, materials other than concrete can be mod-ing, electrical panels, motors, air eled provided that slabs are of uniform material com-coolers, hot piping, and miscellane-ous equipment (J/s) position. This one-dimensional description assumes that slab edge effects do.not significantly affect the Q,i,b = rate of heat transfer to compartment overall rate of heat transfer.

air/water vapor mixture from sur- Boundary conditions on slab temperature are given rounding slabs (J/s) by Qb,k = heat transfer rate to air/water vapor aT, h, mixture from liquid exiting break as [Tl (r) Ts(0 r)]

it cools to compartment temperature Bx o ks (J/s) and Wb, mass flow rate of steam exiting break (kg/s) 46 NUCLEAR TECHNOLOGY VOL. 94 APR. 1991

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS where free convection from a vertical plate. For horizontal slabs, free-convection coefficients depend on whether T> (t), T2(t) = temperatures of compartments ad-the surface is being heated or cooled by the surround-jacent to the slab ing gas mixture. As recommended by Holman,4 the k, = slab conductivity (J/m s K) correlation of Fujii and Imuras is used with the mod-ified characteristic length proposed by Goldstein et al.~

L, = slab thickness (m) to compute the coefficient for an arbitrarily shaped h~, h2 = heat transfer coefficients (J/ slab with heated surface facing upward or cooled sur-mz s K). face facing downward. In cases where the upper sur-The solution of Eq. (4) subject to Eqs. (5) and (6) gives face is cooled or the lower surface is heated, the the rates of energy transfer from the slab surfaces to correlations of Lloyd and Moran7 are used.

the adjacent gas mixtures. Diatomic gases such as nitrogen and oxygen are es-The coefficients hi and hz account for natural sentially transparent to thermal radiation; however, the convection, radiation, and condensation heat transfer. emissivity of water vapor with respect to thermal radi-In the absence of condensation, the coefficient hl can ation is significant. In COTTAP, radiant energy ex-be expressed as change between a slab surface and water vapor contained within the surrounding gas mixture is modeled through hi ht + h/p, (7) the use of an effective radiation heat transfer coeffi-where h>and h~, are the natural convection and ra- cient [see Eq. (7)]. For the applications of interest, tem-diation components, respectively. perature differences between a slab surface and the Natural convection coefficients are expressed in surrounding gas mixture are relatively small (typically terms of the Nusselt number, which in turn is a func- (5 K). Therefore, the following approximate relation tion of the Rayleigh and Prandtl numbers. For the co- proposed by Hottel and Sarofim for small tempera-efficient hl, the appropriate relation is ture differences is used to compute the radiation coef-where Nu =

h)Ct. =

k f(Ra,Pr),

~ITIC (g) ficient:

h/I: '4

(~, +1)

+ a + b c)elm,auaTau ~ (10)

Ct. slab characteristic length where k = gas thermal conductivity o = Stefan-Boltzmann constant (5.669 x 10 s J/

mz s K4) and the Rayleigh and Prandtl numbers for the gas mix-ture are, respectively, defined by e, = slab emissivity gpCI.l Ts(0, t) TI (t)l vier d p k

T,= average Tau = [(T" + Tsurf)/2) ',

temperature, which is defined by (l1)

(9) where where T = gas temperature (K) g = acceleration due to gravity (9.8 m/sz) T~ = slab surface temperature (K) p = coefficient of thermal expansion (K ') e,= emissivity of water vapor evaluated at T,u.

v = kinematic viscosity (mz/s)

The Cess-Lian'quations, which give an analytical n = thermal diffusivity (m /s) approximation to the emissivity charts of Hottel and Egbert," are used to compute the water vapor emis-p = dynamic viscosity (kg/m s) sivity. In Eq. (10), c has the value 0.45, and a and b are Cv = specific heat of the airhvater vapor mixture obtained through differentiation of the Cess-Lian emis-

- ~

(J/kg K). sivity equations Gas mixture properties used in the calculation of free 81n[e(T,PP,PL )]

convection coefficients are evaluated at the thermal a (12) boundary layer temperature, which is taken as the av-Bin(PL )

erage of the slab surface temperature and the bulk gas and temperature.

For vertical slabs, coefficients are calculated from 8 ln [e(T, PP, PL,)]

the correlation proposed by Churchill and Chu3 for 8 ln(T)

NUCLEAR TECHNOLOGY VOL. 94 APR. 1991 47

Chaiko and Murphy FOSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS where isolation of a pipe break (due to valve c losure', for in-stance) a compartment begins to cool a nd condensa-P, = a'r pa t'al prcssu e (Pa) tion continues to occur on surroundin g walls. For a-P= water vapor partial pressure (Pa) sufficiently fast cooldown rate, conde nsation alone does not prevent compartment air from becoming sat-urated, and thus moisture droplets (rain-out) form Condensation on a slab surface occurs when the within the g a s mixture. To maintain co mpartment rel-surface temperature drops below the dew point (the ative humidity less than or equal to unity, the rainout saturation temperature of water evaluated at the par- rate W (kg/s) is calculated from the following empir-tial pressure of water vapor in the compartment) of the ical model:

air/water vapor mixture. Heat transfer coefficients for = 200 (RH 0.99)max(WC,i) condensation conditions are calculated using the exper- W, imentally determined Uchida" correlation, which in- if RH) 0.99 cludes the diffusional resistance effect of noncondensible and gases on steam condensation rates.

In COTTAP, initial compartment temperatures, W, = 0.0 if RH s 0.99, (16) pressures, and relative humidities are specified as in-where put data. An initial slab temperature profile is deter-mined by computing the steady solution to Eqs. (4), RH = relative humidity (5), and (6) corresponding to the initial compartment Ws = total steam flow rate into the compartment conditions. This implies that compartments have been maintained at their initial conditions long enough for (kg/s) slabs to attain steady-state temperature profiles. C,i = constant that is supplied as part of the input data (kg/s).

II.C. Special Purpose Models The COTTAP code includes specialized models to II.C.2. Hot Piping Model simulate the effects of pipe breaks, hot piping, and In many secondary containment compartments, compartment air coolers. Leakage and natural circu- the major heat source consists of piping that contains lation models are also included to describe intercom- reactor steam or coolant. The heat addition rate to a partment mass transfer. In addition, the code includes compartment airhvater vapor mixture from a hot pipe a simplified slab model, a heat load decay model, and is calculated from a compartment model in which temperature, pressure, and relative humidity are specified as a function of alp(<g Up 7rLpDp [Tj'(t )] > (I 7) time. where

.~

Up = overall heat transfer coefficient (J/m2 s K)

II.C.l. Pipe Break Model Within the scope of the present model, pipes may L~ = pipe length (m) contain steam or saturated liquid water. Input data de- D~ = outside diameter of the pipe (or insulation if fine the total mass flow through the break Wb, (kg/s) the pipe is insulated) (m) along with the time at which the break develops and the length of time over which fluid loss occurs. For Tj pipe fluid temperature (K) pipes containing saturated liquid, the steam flow rate T = compartment temperature.

Wb, exiting the pipe (kg/s) is calculated from the en-The overall heat transfer coefficient is calculated by the ergy balance code based on initial compartment conditions; the co-Wbihy(P>) = Wbslig(P) + (Wbi Wbs)h/(P), (14) efficient is then maintained constant throughout the which describes the isenthalpic expansion of fluid from transient.

pipe pressure P~ to compartment pressure P. The liq- II.C.3. Air Cooler Model uid fraction, which does not flash as it leaves the pipe, is assumed to cool to compartment temperature, and Cooling units are used in a number of secondary the dissipated sensible heat is transferred directly to the containment compartments to remove heat generated compartment air/water vapor mixture. For the case by equipment such as emergency core cooling systems where a pipe contains steam, all of the mass and energy (ECCS) injection pumps and high-voltage buses and exiting the break is deposited directly into the compart- transformers. Heat removal rates of cooling units are ment gas mixture. calculated from Rain-out phenomena can be important in compart-ments containing pipe breaks. For example, following Qcool(t ) Ccool (T(t) Tcool(t ) j ~ (I g) 48 NUCLEAR TECHNOLOGY VOL. 94 APR. 199t

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS where This model also describes intercompartment, gravity-driven circulation flows that can develop at open door-T,/(t) = average of the inlet and outlet cooling ways (see the analysis of Brown and Solvason'.

water temperatures C,/ = constant that is computed from spec- II.C.6. Thin Slab Model ified initial values of the cooling load The detailed slab model discussed in Sec. II.B is Q,/, the inlet cooling water tempera- not required to describe heat transfer through thin ture, the cooling water flow rate, and slabs that have little thermal capacitance. Slabs of this the compartment temperature T.

type, e.g., refueling floor walls, have nearly linear tem-An energy balance on the cooling water yields the out- perature profiles, and thus the heat flow through a thin let cooling water temperature. slab can be calculated by the use of an overall heat transfer coefficient U. The rate of heat transfer II.C.4. Leakage Models through a thin slab is obtained from The COTTAP leakage model simulates pressure- q/s(r) = UisA [T1 (>) T2(/)], (21) induced intercompartmental mass transfer through openings such as doorways and ventilation ducts. In- where tercompartment leakage is calculated by balancing the A= thin slab heat transfer area (m )

pressure differential between the compartments with an irreversible pressure loss. Thus, the leakage rate sat- Tj Tz = temperatures of the compartments sepa-isfies rated by the slab (K).

Values of U(J/m s K) are supplied as part of the P (t) P, (t) Kinesia(t)[ Win(t)] code input data (one value for each vertical slab and (19) 2p//(l)Aw two values for each horizontal slab). For horizontal where slabs, two values of Uare required because free-convection film coefficients depend on the direction, P1, Pz = pressures of the compartments associated upward or downward, of heat flow through the slab.

with the leakage path (Pa)

WN = leakage rate (kg/s) II.C.7. Heal-Load Decay Model K~ = irreversible pressure loss coefficient Cooling of a component such as a pipe filled with A//, leakage area (m ) hot stagnant fluid or a pump that has ceased operat-ing is simulated through the use of a lumped-param-p//, = gas density within the compartment sup- eter heat transfer model. Most compartments in the plying the leakage flow (kg/m ). secondary containment have a large thermal capacity because of the bounding concrete slabs. It is therefore It is assumed that inertial effects do not significantly assumed that the component temperature changes on affect leakage rates.

a faster time scale than the compartment air temper-ature; i.e., the air temperature is assumed to remain II.C.5. Natural Circulation Model fairly constant during the cooldown of the component.

With this assumption, the component heat dissipation A natural circulation model simulates gravity- rate Qc(t) is governed by driven mixing in compartments connected by flow paths at differing elevations. The circulation rate W, (kg/s) is obtained from 7

'Q'" =-Q(/) (22) d/

W g['() '()](" ) where K//[Alp2(t)] + KN/[ANpi (/)] J Qc(/o) = Qco (23) where and 7, (s '), the thermal time constant of the compo-p1, pz densities of the air/water vapor mixtures nent, is given by within the two adjacent compartments McCw (kg/m ) (here it is assumed that p2 is the Vc (24) gas density for the cooler compartment)

UcAc E,E/elevations of the upper and lower flow where paths (m) M, = mass of the component (kg)

A,A/upper and lower flow path areas (m ). Ci~ = specific heat of the component (J/kg K)

NUCLEAR TECHNOLOGY VOL. 94 APR. 1991 49

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS U, = overall heat transfer coefficient (J/m2.s K) tions, fourth-order central difference formulas are used to compute T t at interior grid points:

A, = component heat transfer area (m2).

In Eq. (23), to (s) is the time at which the cooldown

( Tsi-2 1

process begins, and Q<<, which is supplied as input Tsxxi = + 16Tsi-i 30Tsi + 16Tsl+i 128, 2 12 data, is the heat dissipation rate prior to cooldown. So-lution of Eqs. (22) and (23) gives the exponential-decay Ts'+2) + O(h ), (26) approximation used in COTTAP to model heat dissi-where pation of cooling components. The component time constant y, is specified as input data except in the case i =3,4,...,N-2 of hot piping, where it is calculated by the code from 6 = spacing the piping description data. between grid points.

II.C.8. Time-Dependent Compartment Model A six-point sloping difference formula is used to ap-proximate T at i = 2 and i = N 1:

p With the time-dependent compartment (TDC) model, environmental conditions within a compart-ment are specified as a function of time; i.e., temper-ature, pressure, and relative humidity versus time are Tsxx2 = I 2 (10Tsi 15Ts2 4Ts3 + 14Ts4 supplied as tabular input data. This model is particu- '6T,s+ T,6) + O(~ ) (27) larly useful for representing outside air conditions, in- and cluding solar and thermal radiation effects. The influence of solar and long-wave atmospheric radiation 1 on exterior buildup surfaces can be described by spec- TsxxN l 2 (10TSN 15TsN-l 4TsN-2 ifying the effective Sol-Air temperature'n the TDC 6TsN-4 +

instead of the actual outside air temperature. In sec- + 14TsN-3 TsN-5) ondary containment analysis, the TDC model is also +O(~4) . (28) useful for describing transient conditions within the primary reactor containment, which are generally For the end points, where the normal derivatives known from the results of detailed licensing basis cal- are specified through convective boundary conditions, culations. the following finite difference approximations, recom-mended by Pirkle and Schiesser,'3 are used to com-II.D. Numerical Solution Nlethods pute T An energy balance and two mass balances are solved for each compartment to determine gas temperature, air mass, and water vapor mass. In addition, the one-Tsxxt = I 1262 415 Ts i + 96T2 6

s 36T3+ S 32 3

T4s dimensional heat conduction equation is solved for 3 each slab. Before computing the numerical solution of Tss 50t3,Tsxl + O(h4 ) (29) the governing equations, partial differential equations describing heat flow through slabs are approximated and by sets of ordinary differential equations (ODEs). This is accomplished through application of the method of lines (MOL). In the MOL, a finite.difference approx-TssxxN= I 415 TN+ 96TN i 36TN 2 imation is applied only to the spatial derivative in Eq. (4), giving +

32 3 TsN-3 TsN-4 + 506TsxN 2

d Tsl

=GT s sxx p (25) +O(a) (30) where In Eqs. (29) and (30), the normal derivatives Tsxi and i = 1,2,3,...,N, the number of equally T~ are evaluated in accordance with Eqs. (5) and (6),

spaced the convective boundary conditions; i.e.,

grid points Tp slab temperature at grid point i Tsxl hi (Tl Ts I )

s T p

= finite difference approximation to the second-order spatial derivative at grid pointi. and Following the approach used by Pirkle and Schiesser'3 in the MOL solution of parabolic equa- Tsx2 h2(TsN T2) (31) s 50 NUCLEAR TECHNOLOGY VOL. 94 APR. 199l

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS dL'32)

Allgovernirig equations are now expressed in terms of 2. Cooler modeling does not describe moisture re-ODEs of the form moval under conditions where the cooling coil temper- .

ature is below the dew point of the inlet gas mixture.

dy =

F(y,t) with y(0) =ye 3. Pipe break modeling is valid only for lines con-taining steam or saturated liquid; breaks involving the release of subcooled liquid cannot be described.

Solutions of Eq. (32) exhibit rapid initial adjust-ments in compartment air temperature caused by the 4. Compartment flooding events cannot be simu-relatively small thermal capacitance of the air contained lated because all liquid is assumed to exit through com-within the compartment. Moreover, slab temperatures partment floor drains.

undergo rapid initial changes in narrow regions near the boundaries, resulting in the formation of spatial thermal boundary layers. In the numerical integration III. RESULTS OF SSES SECONDARY CONTAINMENT of Eq. (32), small time steps are required to simulate ANALYSIS FOR POSTACCIOENT CONDITIONS these initial transients. As the initial transient response decays, however, it is desirable to increase step sizes in This section gives representative results for a COT-order to reduce the computation time required to fol- TAP simulation of the combined SSES-1 and -2 sec-low the slowly varying part of the solution. Equations, ondary containments under postaccident conditions.

such as Eq. (32), which exhibit initial temporal bound- The thermal responses of the Units 1 and 2 secondary ary layer structures are termed stiff differential systems containments are coupled by heat transfer through (see the discussion in Ref. 16), and because of stabil- common walls that separate the two structures. The ity limitations, they cannot be solved efficiently with SSES model consists. of 105 compartments, 16 time-explicit integration schemes. For this reason, an im- dependent compartments, 767 slabs, 38 thin slabs, and plicit scheme was selected for COTTAP. 505 heat loads. The simulation was carried out for 30 h Numerical integration of the governing Eq. (32) is and required 124 min of CPU time on an IBM 3090 carried out with the LSODES code,'hich uses the computer. Note that most of the CPU time is required implicit backward differentiation methods proposed by to simulate the rapidly varying part of the transient Gear for the solution of stiff systems. The LSODES that occurs within the first few hours of the event.

code also employs sparse'matrix inversion techniques Thus, substantially longer simulation times do not sig-in solving the implicit finite difference equations. With niflcantly increase CPU time requirements.

these numerical integration features, it is feasible to For this analysis, it is assumed that a loss-of-carry 'out the integration of the large differential sys- coolant accident (LOCA) occurs in SSES-1 and a false tems that arise in the simulation of secondary contain- LOCA signal (a spurious signal that indicates loss of ment transients. As an illustration of the problem reactor coolant and leads to ventilation system isola-dimension, simulation of the SSES-1 and -2 secondary tion and operation of ECCS injection pumps) is gen-containments under postaccident conditions required erated on SSES-2. Under postaccident conditions, the solution of 20101 coupled ODEs. ECCS injection pumps comprise the key equipment For these large-scale problems, reevaluation of within the secondary containment structure. The ECCS code-calculated slab heat transfer coefficients at every consists of the residual heat removal (RHR), core time step leads to unacceptably long computation spray, and high-pressure coolant injection (HPCI) sys-times. To alleviate this difficulty-, the frequency of re- tems. These systems receive electrical power from high-evaluation (number of steps between reevaluation of voltage buses contained within emergency switch gear coefficients) is a parameter supplied as input to the and load center rooms. Figure I shows the calculated code. Sensitivity calculations on small-scale problems temperature response within a SSES-1 RHR pump representative of postaccident secondary containment room (each unit contains two RHR pump rooms and transients indicate that coefficients can be reevaluated two core spray pump rooms). Initially, the air temper-as infrequently as once per ten steps without introducing ature increases rapidly because of the small thermal ca-significant errors in the results. The CPU time require- pacitance of the air within the compartment. As air ments were reduced by a factor of 4 when coefficients temperature increases, a balance between compartment were reevaluated at every tenth time step. heat sources and losses to compartment air coolers and slabs begins to develop. At this time, air.temperature II.E. Code Limitations. in Modeling Accident Scenarios starts to increase on the slow time scale governed by the slab thermal capacity and transport properties. An The following modeling limitations have been iden- initial rapid temperature rise followed by a much tified in the current version of the COTTAP code: slower temperature increase is characteristic of all com-partment heatup transients. After 1 h of operation, this

1. Fission product transport among compartments particular RHR pump switches from the injection is not modeled. mode of operation to the suppression pool cooling NUCLEAR TECHNOLOGY VOL. 94 APR. 1991 51-

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS 317 hC

~ 316.

320

~~ 315 E 316 I- I- 314 E E 3 316 o 313 E 312 o 314 Q 311 X

312 310 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Time (h) Time (h)

Fig. 1. Simulation of postaccident temperature response Fig. 3. Simulation of postaccident temperature response within SSES-I RHR pump room for LOCA on within SSES-I HPCI pump room for LOCA in SSES-I and false LOCA on SSES-2. SSES-I and false LOCA in SSES-2.

mode. As a result of increased compartment heat loads HPCI system, which also begins operation at the start associated with the change in operating mode, the tem- of the accident. In this case, however, compartment perature again increases rapidly until a new balance temperature continues to increase when the system between the heat-generation and heat-loss rates is at- ceases operation at I h into the transient. This occurs tained. because piping heat loads within this compartment are The temperature response within a SSES-I core substantial. When HPCI pump operation stops, an as-spray pump room is shown in Fig. 2. Core spray op- sociated room cooling unit also ceases operation. Upon eration begins at the start of the event and ceases I h shutdown of the cooling unit, slowly decaying piping later. Temperature decreases rapidly at this point be- heat loads rapidly increase compartment temperature cause, once pump operation is terminated, no signif- until a balance between heat generation and heat losses icant heat loads remain in the compartment. Figure 3 to compartment slabs is approached. Figure 4 gives the illustrates the temperature response of the SSES-I temperature within a SSES-I load center room that

~ 317 309 P

~

j e

316 E

I- E 308 E I-3 315 E CC 3

cc 307

~ 8C~

314 O

V) 313 306 O 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Time (h) Time (h)

Fig. 2. Simulation of postaccident temperature response Fig. 4. Simulation of postaccident temperature response within SSES-I core spray pump room for LOCA in within SSES-I load center room for LOCA in SSES-I SSES-I and false LOCA in SSES-2. and false LOCA in SSES-2.

52 NUCLEAR TECHNOLOGY VOL. 94 APR. 1991

I Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS supplies electr ical power to emergency equipment. In 315 this compartm ent, heat loads remain essentially con-stant throughout the transient.

From the results of this analysis, it is determined that under postaccident conditions, some of the equip-ment within the secondary containment would be ex- K 310 posed to temperatures that exceed their qualification P values. Consequently, components were reassessed for operation at higher temperatures, and in some in-stances equipment was relocated to compartments with less severe environmental conditions. Furthermore, a ~~ 305 procedure was developed to instruct plant operators to COTTAP shed nonessential electrical loads within 24 h after an accident in order to moderate the temperature re-CONTAIN sponses within secondary containment compartments.

300 0 2 4 6 8 10 IV. EVALUATION OF CODE ACCURACY Time (h)

Fig. 5. Comparison of COTTAP and CONTAIN compart-As part of the verification process for the COT- ment temperature simulations for test problem I, TAP code, calculational results were compared with those obtained with the CONTAIN (Ref. 2) program, which has been verified through comparison with ex-perimental data.' Although the CONTAIN code 0.20 does not accommodate a direct heat input (such as from operating mechanical or electrical equipment) to a compartment, useful problems can nevertheless be 0.18 formulated in order to investigate the modeling and computational accuracy of COTTAP. Two such prob-lems were formulated for code verification. The first 0.16 problem tests the CO%I'AP compartment mass and en-ergy balance calculations and the slab heat transfer simulation. This problem consists of a single compart-ment that has a 1000-m3 volume and contains air at 300 K and 101 325-Pa initial temperature and pressure.

Concrete slabs, which range in thickness from 0.1 to 1 m, form the walls of the compartment. Allslabs have Q-0.14 0.12 CONTAIN COTTAP a uniform, initial temperature of 300 K. To add heat 0.10 to the compartment, the air in contact with the outer 0 2 4 6 8 10 surface of one slab (the slab that is 0.1 m thick) is sud- Time (h) denly increased to 400 K at t = 0. In addition, at 50 s into the transient, air with a temperature of 500 K is in- Fig. 6. Comparison of COTTAP and CONTAIN.compart-jected into the compartment at a 0.26 kg/s flow rate. ment pressure simulations for test problem l.

Outer surface temperature rise and air injection con-ditions were selected to effect significant, but not ex-cessive, temperature and pressure response.

Figures 5 and 6 present a comparison of the COT- ification involves modeling of compartment tempera-TAP and CONTAIN calculation results for the first ture and pressure behavior under conditions where test problem. The temperature and pressure simula- high-energy steam is injected into the compartment. In tions both show excellent agreement; note that the this problem, condensation effects strongly influence pressure response curves given in Fig. 6 completely the rate of temperature and pressure increase. Com-overlap. In -Fig. 5, the initial temperature increase, partment physical description data are the same as that which is due to injection of hot air into the compart- for test problem 1. In this case, however, the only heat ment, begins to level off at -0,5 h. Heat addition by source is the steam entering the compartment at a means of conduction through the externally heated slab 0.20 kg/s flow rate and a 2.7756 x 106 J/kg enthalpy.

then begins to occur, causing a further but less rapid This flow rate and enthalpy are characteristic of a increase in temperature. small steam leak within a secondary containment com-The second test problem considered for code ver- partment. Figures 7 and 8 show a comparison of the NUCLEAR TECHNOLOGY VOL. 94 APR. I99t 53

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS 450 ACKNOWLEOGMENTS The authors thank Jack G. Refling, James E. Agnew, Mark R. Mjaatvedt, and Leonard J. West for their many helpful suggestions during the course of this work. We also

~ thank Lisa Walsh for typing the manuscript.

4oo'50 CONTAIN COTTAP REFERENCES

1. C. C. LIN, C. ECONOMOS, J. R. LEHNER, G.

MAISE, and K. K. NG, "CONTEMPT4/MOD4: A Multi-compartment Containment System Analysis Program,"

BNL-NUREG-51754, Brookhaven National Laboratory (1984).

300 0 5 10 15 20 2. K. K. MURATAet al., "User's. Manual for CONTAIN Time lh) 1.1: A Computer Code for Severe Nuclear Reactor Accident Containment Analysis," NUREG/CR-5026, Sandia Na-Fig. 7. Comparison of COTTAP and CONTAIN compart- tional Laboratories (1989).

ment temperature simulations for test problem 2.

3. S. W. CHURCHILL and H. H. S. CHU, "Correlating Equations for Laminar and Turbulent Free Convection from a Vertical Plate," Int. J. Heat Mass Transfer, 18, 1323 (1975).

0.6

4. J. P. HOLMAN, Heat Transfer, 4th cd., p. 250, McGraw-Hill Book Company, New York (1976).

os 5. T. FUJII and H. IMURA, "Natural Convection Heat Transfer from a Plate with Arbitrary Inclination," Int. J.

Heat Mass Transfer, 15, 755 (1972).

g 0.4 I 6. R. J. GOLDSTEIN, E. M. SPARROW, and D. C.

JONES, "Natural Convection Mass Transfer Adjacent to Pn 0.3- Horizontal Plates," Int. J. Heat Mass Transfer, 16, 1025 Q. (1973).

COTTAP 0.2 CONTAIN

7. J. R. LLOYD and W. R. MORAN, "Natural Convec-tion Adjacent to Horizontal Surface of Various Planforms,"

ASME 74-WA/HT-66, American Society of Mechanical 0.1 Engineers (1974).

0 5 10 15 20

8. D. Q. KERN, Process Heat Transfer, p. 690, McGraw-Time (h) Hill Book Corupany, New York (1950).

Fig. 8. Comparison of COTTAP and CONTAIN compart-ment pressure simulations for test problem 2. 9. H. C. HOTTEL and A. F. ballot'tM, Radiative Transfer, McGraw-Hill Book Company, New Yoit (1967)

'1~

10. R. D. CESS and M. S. LIAN, "A Simple Parameteriza-tion for the Water Vapor Emissivity," Int. J. Heat Transfer, 98, 676 (1976).

COTTAP and CONTAIN simulation results. The re-sults show good agreement even though the codes em- 11. H. C. HOTTEL and R. B. EGBERT, "Radiant Heat Transmission from Water Vapor," Am. Inst. Chem. Eng.,

ploy considerably different approaches in the 38, 531 (1942).

calculation of condensation rates on slab surfaces. The COTTAP code uses the experimentally determined 12. H. UCHIDA, A. OYAMA, and Y. TOGO, "Evalua-Uchida'ondensation coefficient, while CONTAIN tion of Post-Incident Cooling Systems of Light-Water carries out a detailed computation of the thermal re- Power Reactors," Proc. 3rd Int. Conf. Peaceful Uses of sistances associated with the gas boundary layer and Atomic Energy, Geneva, Switzerland, 1964, Vol. 13, p. 93, the condensate film. United Nations (1965).

54 NUCLEAR TECHNOLOGY VOL. 94 APR. 1991

Chaiko and Murphy POSTACCIDENT BWR SECONDARY CONTAINMENTTHERMAL ANALYSIS

13. W. G. BROWN and K. R. SOLVASON, "Natural Con- 17. A. C. HINDMARSH, "ODEPACK, A Systematized vection Through Rectangular Openings in Partitions-I Ver- Collection of ODE Solvers," Scientific Computing, Vol. I, tical Partitions," Int. J. Heat Mass Transfer, 5, 859 (1962). p. 55, R. S. STEPLEMAN et al., Eds., IMACS Transac-tions on Scientific Computation, North-Holland Publishing

14. ASHRAE Handbook 1985 Fundamentals, American Company, Amsterdam (1983).

Society of Heating, Refrigerating and Air-Conditioning En-gineers, Atlanta, Georgia. 18. K. K. MURATAand K. D. BERGERON, "Experimen-tal Validation of the CONTAIN Code," Proc. 11th LWR

15. J. C. PIRKLE, Jr. and W. E. SCHIESSER, "DSS/2: A Safely Information Mtg., Gaithersburg, Maryland, October Transportable FORTRAN 77 Code for Systems of Ordinary 24-28, 1983, SAND-83-1911C, Sandia National Laborato-and One, Two and Three-Dimensional Partial Differential ries (1983).

Equations," presented at 1987 Summer Computer Simula-tion Conference, Montreal, Canada, 1987. 19. K. K. MURATAet al., "CONTAIN: Recent Highlights in Code Testing and Validation," Proc. Int. Mtg. Light Water

16. C. W. GEAR, Numerical Initial Value Problemsin Or- Reactor Severe Accident Evaluation, Cambridge, Massa-dinary Differential Equations, Chap. 11, Prentice-Hall, En- chusetts, August 28-September I, 1983, American Nuclear glewood Cliffs, New Jersey (1971). Society (1983).

Mark A. Chaiko [BS, 1980, and MS, 1983, chemical engineering, Penn-sylvania State University (PSU); PhD, applied mathematics, Lehigh Univer-sity, 1989] is a project engineer-nuclear systems at the Pennsylvania Power &

Light Company. His current technical interests include boiling water reactor stability analysis and thermal-hydraulic modeling of reactor systems.

Michael J. Murphy (BS, mechanical engineering, 1982, and MS, nuclear engineering, 1986, PSU) is a project engineer-nuclear systems with the Penn-sylvania Power & Light Company. He is currently involved in simulation of anticipated transient without scram and severe accident analysis.

NUCLEAR TECHNOLOGY VOL. 94 APR. 1991 55