ML20073F680

Responds to Listed RAI Re Test Program for GE Sbwr

ML20073F680
Person / Time
Site:	05200004
Issue date:	09/26/1994
From:	Mcintyre T GENERAL ELECTRIC CO.
To:	Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
MFN-113-94, NUDOCS 9410030200
Download: ML20073F680 (46)

Text

GENuclear Energy P. W. Mamott. Manager GeneralEbctnc Company AdvancedPhnt Technobges 175 Curtner Avenue, MC 781 San Jose CA 95125-1014 406925-6948(phone) 408 925-1993 (bcsmik)

September 26,1994 51FN No.113-94 Docket STN 524)04 Document Control Desk U. S. Nuclear Regulatory Commission Washington DC 20555 Attention: Richard W. Borchardt, Director Standardir.ation Project Directorate

Subject:

Responses to the Referenced Letters

References:

1) Letter, 51. 51 alloy (NRC P. W. hiarriott (GE), SCIIFDUIE IY)R REVIEW OF TEST AN1 '., MLYSIS PROGRAM DESCIUPTION (NEDC-32391P) FOR TIL  : NUCLEAR ENERGY (GE)

SIMP 11FIED Boll 2NG WATER luiACTOR (SBWR) AND INITIAL REQUESTS FOR ADDITIONAL INFORMATION (Q900.65-Q900.81 AND PURDUE UNIVERSITY QUESTIONS -

SETS), dated September 12,1994.

2. Letter,51. .hialloy (NRC) to P. W. Starriott (GE), REQUESTS FOR ADDII'lONAL INFORMATION REGAIU)1NG Tile TEST PROGRAM FOR Tile GE NUCLEAR ENERGY (GE) SIMPLIFIED BOILING WATER IUCACTOR (SBUR) (Q900.82-Q900.95), dated September 16,1994.

The Enclosures to this letter contain responses to Requests for Additional Information (RAls) 900.65 - 900.81, Purdue University Questions - Set 5 (Questions 1,2, and 3), and 900.83,900.87,900.91,900.93, and 900.94, which were enclosures to the Referenced letters.

Sincerely, T. R. Niclntyre, Acting hianager

. Advanced Plant Technologies

Enclosures:

1. Responses to Reference 1.

2. Responses to Reference 2.

I cc: P. A. Bochnert (ACRS)

R. W. Hasselberg (NRC) l

51. .\talloy (NRC) 4)

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(\s 9410030200 940926 CDR ADOCK 05200004 M PDR

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Enclosure I to MFN 113 94 1

i RAI Number 900.65 Quution:

Resubmit NEDC-32391P "SBWR Test and Analysis Program Description," with i clear identification of the proprietary information on each page sought to be l withheld, along with the reasons for withholding the information. Consistent -

with the staffs April 7,1994, letter from D. Crutchfield to P. Marriott regarding the quality of the SBWR application, submit a non-proprietary version of the '

report within a reasonable period of time of the proprietary version,if the ver-sions are not submitted simultaneously. Both versions of the report should be resubmitted as Revision 0 in lieu of a draft.

In the letter (s) transmitting the report, provide a page-by-page summary of any additions and corrections made to the report since the August 10,1994, version was submitted to the staff for review.

GE Response:

NEDC-32391P, "SBWR Test and Analysis Program Description," Revision A, with clear identification of the proprietary information on each page sought to be withheld was transmitted by MFN 109-94, dated September 15,1994. The ,

letter transmitting the report provided a page-by-page summary of additions and  ;

corrections made to the report. A non-proprietary version entitled "SBWR Test ,

and Analysis Program Description," NEDO-32391, was transmitted by MFN l 110-94, dated September 15,1994.

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RAI Number 900.66 Question:

Proside a point-by-point response to the staffs March 7,1994, letter from D.

Crutchfield to P. Marriott regarding concerns about the SB%R testing program.

Alternatively, provide a road map that identifies where in NEDG-32391P, "SBWR Test and Analysis Program Description," each of the concerns of the March 7,1994, letter are explicitly addressed.

GE Response:

Liint Acceptability of the Gravity Driven Cooling System (GDCS) Integral Systems Test (GIST) program data as the sole integral experimental basis for SB%R in view of the differences in configuration compared to the current SB%T design.

Response

While the physical configuration of GIST is representadve of the 1988 SB%%

design, GE considers that GIST provides GDCS performance data suitable for TRACG qualificadon. The basis for this statement is given below.

The principal difference between GIST and the current SB%1 design is that the GDCS pool is a separate entity in the drywell instead of being a part of the suppression pool. This difference notwithstanding, the test captures the interactions between multiple regions represented by the reactor vessel, drywell and wetwell. The interactions between RPV depressurization and the GDCS are properly represented. The scaling study in Appendix B of NEDC-32391P demonstrates that the major parameters governing depressurization rate and driving heads for GDCS flow are preserved even though there are differences in the configuration of GIST from the current SBWR design Testing in GIST is intended to simulate the late blowdown /early GDCS phase of  ;

the LOCA transient (Fig. 5.3.1 of NEDC-32391P), and thereby provide data for TRACG qualification of GDCS performance. The parameters of primary interest are: System pressure response which determines the timing of GDCS  ;

initiation, GDCS flow and RPV level response.

No scaling distortions have been identified in the significant phenomena which would preclude the use of the GIST data for their intended application  ;

It should also be noted that the GIST data are supplemented by data from other BWR LOCA integral test facilities (TLTA, FIST) for the early part of the blowdown.

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Absence of components / systems that could interact (IC, PCCS) .

Resbonse l GIST simulates the limiting LOCA transient without credit for the IC. No credit  !

is taken in LOCA analysis for the IC.  !

Analysis shows (NEDC-32391P, Appendix C) that the IC increases the j minimum water level for the limiting breaks (bottom break and GDCS hne break). For the steam line and feedwater line breaks, the minimum water level .

is lowered due to void collapse. However, for these breaks the minimum water level is several meters above the top of the core and the impact is not significant i to safety. Furthermore, postulated interactions between the IC and DPV i resulting in flow reversal in the IC and subsequent reduction in the depressurization rate are shown not to be possible (NEDC-32391P, Appendix C). #

Thus, the overall impact of the IC is tc. increase the margins for the limidng breaks. ,

5 Because a relatively small fraction of the dowell to wetwell flow passes ,

through the PCCS in the blowdown period, the PCCS has a minimal effect on the drywell pressure and GDCS flow (NEDC-32391P, Appendix C). The absence j of a PCCS in the GIST tests has little or no effect on the vessel transient. j Interactions between the PCCS and GDCS are important in the containment -

during the GDCS phase of the transient in that they can result in vacuum  ;

breaker opening. This leads to a return of the noncondensibles to the drywell j and subsequent recycling through the PCCS. Tests are planned in PANDA  ;

Phase 2 to address these interactions. j Issue l Insufficient characterization of GIST facility thermal hydraulic behavior Res b on s e GE a' grees that calibration data on pressure drops and heat losses would be '

desirable. However, the data are adequate for the intended applicatien. i For the TRACG calculations, the pipes, valves, elbows and orifices were treated  !

as standard hydraulic components. The pressure drops and pressure distributions were calculated based on handbook published loss coefficients (Idelchik, Crane). Based on these assumptions, TRACG calculated the GIST transient response satisfactorily. This confirms the previous good experience  !

with this approach. l t

Critical flow through the SRVs and the break is based on the minimum flow i area. The TRACG model for cridcal flow, which has been qualified over a wide  !

range of data (NEDC-32391P, Table 5.1-1) was employed for the calculation of i critical flow, i

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Heat losses were calculated based on reasonable analytical values for the natural convection heat transfer coefficient. A sensitivity study on the heat transfer coeflicient (variation of heat loss by a factor of 2) showed very little sensitivity to this parameter.

Iss ue i Lack of a quantitative scaling study l Res bon s e 1 A quantitative scaling study has been performed and is provided in NEDC-32391P, Appendix B, Attachment Bl.  ;

1ssu e l Requirement for additional data from PANDA to be included as part of the testing for design certification. Details of the test matrix and facility scaling to i be provided to the NRC.

t Resbonse GE ' agrees that PANDA data will be used as primary data for TRACG qualification. The test / qualification matrix and facility scaling are provided in i NEDC-32391P.

Issue Requirement for Isolation Condenser performance data from PANTHERS to be included as part of the testing for design certification, if credit is taken for the IC. j Resbonse GE agrees that PANTHERS thermal hydraulic performance data will be used for TRACG qualification. The test / qualification matrix and facility scaling are provided in NEDC-32391P. While no credit is taken for ICs in LOCA analysis, ,

interaction studies have shown no deleterious effects if ICs were to operate i (NEDC-32391P Appendix C).

Issue i Requirement for data demonstradng PCCS performance in the presence oflight  ;

noncondensible gases in an integral system test.

l Res bon se +

A combination of tests and analysis addresses the effects oflight noncondensibles (hydrogen) on the PCCS performance and containment pressure.

One of the major concerns underlying the light noncondensible gas issue is the capability of the PCCS to restart after the drywell and PCCS have been filled up with noncondensibles. A Noncondensible Blanketing Test (M7) is planned in PANDA Phase 2 to address this issue. Whether the noncondensible gas is lighter or heavier than steam does not make any difference for this demonstration, because it shows that the PCCS can purge the noncondensibles.

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l At a component level, tests will be performed at PANTHERS to determine the effect of helium buildup in the full scale PCC units. With the vent blocked, the helium will accumulate in the PCCS, and the distribution of helium and the effect on heat removal capacity can be determined.

A helium system response test program in GIRAFFE has been added as described in the response to RAI 900.67.

The effect of the hydrogen resulting from 100% metal-water reaction on the l integral system response can be bounded through calculations. ,

Issue Availability of experimental and facility data for tests run by others for GE (GIRAFFE, PANDA, PANTHERS).

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Response

GE will continue to provide the requested information or provide NRC access to the test facilities and/or test performers.

Issue Requirement for documentation of testing program in conformance with 10CFR52.47 in Section 1.5 of SSAR. l Res bon s e GE agrees to include a summary of the testing program in SSAR Section 1.5 and/or a reference to NEDO-32391.

Issue Need for additional test in properly scaled integral test facility.

Resbonse A systematic study of test and analysis needs has been performed in NEDC-32391P. A need for additional testing has been identified in PANDA for specific interactions between the GDCS and PCCS in the GDCS phase resulting in vacuum breaker openings, for interactions with ICs, and to demonstrate PCCS restart when filled with noncondensible.

( Tests M5 - M9). These tests have been added to PANDA Phase 2 testing.

Scaling of PANDA isjudged to be adequate for its intended purpose and is addressed in Appendix B of NEDC-32391P.

The GIST tests are adequate for validating vessel performance during the late blowdown /carly GDCS phase. Here there are few uncertainties and large margins to core heatup. The overall coverage of the LOCA transient by the integral tests is shown in Figure 5.3-1 of NEDC-32391P.

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4 RAI Number 900.67 Question:

GE'sJuly 1,1994, letter (MFN No. 087-94) states that GIRAFFE tests were development tests and GE intends to use GIRAFFE data to substantiate the results of PANDA and PANTHERS at another scale. Contrary to this position, NEDC-32391P, "SBWR Test and Analysis Program Description," indicates that the helium test data from GIRAFFE are to be used as part of the TRACG qualification effort. GE needs to clarify the use of helium test data from GIRAFFE vis-i-vis the position on GIRAFFE data stated in MFN No. 087-94, in particular:

(a) is the GIRAFFE helium test the only one, or are there plans for other helium tests, in GIRAFFE or in another test facility?

(b) If helium test data from GIRAFFE only is to be used, how will GE resolve the quality assurance concerns raised by the staff on other GIRAFFE tests during itsJune 21-23,1994, inspection?

(c) The recently conducted GIRAFFE helium test contained only helium.

Explain whether future tests will be more typical of post-accident conditions, include a combination of helium and nitrogen. In addition, the test duration should be based on observing at least one purge and tran-sient back to steady state operation of the Passive Containment Cooling System (PCCS).

GE Response:

(a) Since the submittal of NEDC-32319P in mid-August, GE has been pursuing negotiations with Toshiba Corporation regarding additional helium testing in GIRAFFE. These negotiations have recently been concluded. As a result of this agreement, reference to the existing GIRAFFE helium test will be removed from NEDC-32391P, and a new test program will be performed in GIRAFFE specifically to address the staff's concerns relative to lighter-than-steam non-condensable gasses in the SBWR. Facility configuration and instrumentation will be similar to the GIRAFFE Phase 2 Main Steam Line Break tests. The test objectives of the GIRAFFE Helium Test Program are:

1. Provide data that demonstrate the effective operation of the passive containment cooling system with the presence of a lighter-than-steam non-condensable gas, and

2. Provide data for qualification of containment response predictions by TRACG in the presence of lighter-than-steam non-condensable gases.

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l Four test conditions will be included. Test Condidon H1 will be a base case with nominal initial conditions the same as in PANDA tests M3 and M4, e.g., near SBWR SSAR LOCA conditions one hour into the accident scenario. The drywell will contain a mixture of steam and nitrogen at a total pressure of approximately 300 kPa. Test H2 will be a nominal repeat of test H1, but with a helium replacing the total volume of nitrogen in the drywell and PCCS. Test H3 will have the same total initial drywell pressure as tests H1 and H2, but with the initial non-condensable fraction consisting of helium / nitrogen mixture having the same propordons that would result from a 100% SBWR metal water reaction. Test H4 will start with the same initial conditions as test H1, (nitrogen and steam in the drywell), and will have constant helium injection to the drywell. The helium addition rate will be such that the helium is injected over a period of one hour, and the test will be terminated when the total mass of helium added is equal to the initial drywell helium mass in Test H3. The test will be continued to observe the venting of any residual helium from the drywell following termination of helium injection.

System response from the four tests will be compared with each other to establish the effects oflighter than steam, or a mixture oflighter-than-steam and heavier-than-steam non-condensables, on the effecdveness of heat rejection by the PCC heat exchanger. Specific test condidons are currently being finalized. No other helium testing in a facility other than GIRAFFE is planned (b) The new GIRAFFE HELIUM tests described in response to item (a), above, will be performed by Toshiba in accordance with Japanese Nadonal StandardJEAG-4101 (1990 Rev.) GE has reviewed this standard, and concluded that in all important aspects, it meets the intent of 10CFR50 Appendix B and ANSI /ASME NQA-1 (1983). GE requests that the staff review this standard for this applicadon, and concur that tests performed under it are acceptable for the application of this data to the SBWR. GE effort supporting the new GIRAFFE testing will be performed under our own, NRC accepted, QA program.

In addition to the four GIRAFFE Helium tests described in Response (a),

Toshiba will also be performing a repeat of the GIRAFFE Phase 2 Main Steam Line Test, one of the two tests described as GIRAFFE Data Group G2 in NEDC-32391P. This test will be performed using the above quality assurance requirements, and will be performed in order to reinforce (" tie-back") the validity of previous GIRAFFE testing with the NRC staff.

(c) We believe the GIRAFFE Helium test program as defined in the response to item (a) is responsive to the staffs comments as elucidated in this item.

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RAI Number 900.68 Question:

Both the staff (during a meeting with GE on August 18,1994) and the ACRS Thermal-Hydraulic Phenomena Subcommittee (during a meeting on August 24,1994) have expressed concerns regarding test instrumentation. In general, GE seems to place dependence on a limited number of pressure-temperature measurements, and then back-calculate any local conditions of interest.

Specifically, the staff is concerned with:

(a) lack of direct local heat fluxes in the PCCS heat exchangers, (b) lack of direct measurements of the pressure and/or noncondensable gas distribution along the PCCS heat exchanger tubes, (c) lack of direct measurements of local concentration of noncondensable gases.

Address the above concerns regarding adequacy of test instrumentation for PANDA and PANTHERS.

GE Response:

(a) SIET Document 00157ST92 Rev 1, transmitted to the NRC by MFN No.

086-94 datedJune 30,1994,in response to RAI 950.24, addresses the instrumentation specifically added to the PANTHERS PCC heat exchanger in order to address the ACRS concerns on local heat flux measurement.

Figure A.2.1 of the SIET document shows the location and type of instrumentation for local heat flux measurement.

Briefly,72 thermocouples were added to the PANTHERS test instrumentation to address the ACRS concern. Four PCCS tubes, located at differing locations within the tube bundle, have been instrumented at nine elevations each. Thermocouples are located on the inside and the outside of the tubes, so that local heat fluxes may be calculated from the temperature difference across the tube wall. The algorithm to be used in data analysis is given in SIET document 00098PP91 Rev.1, transmitted to the NRC by MFN No. 098-95 dated August 16,1994.

(b) It is true that there are no direct measurements of the pressure or non-condensable gas concentration along the PCC heat exchanger tubes. We have evaluated this situation, and determined that such measurements are not necessary to determine the location within the tubes where condensation is taking place. Temperature measurements along the PCCS tubes were used successfully in GIRAFFE to determine the location of the l condensation process within the PCC heat exchanger tubes, and review of l l

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initial PANTHERS data likewise has confirmed this capability. Pressure difference measurements between the upper and lower headers of the PANTHERS PCCS have indicated very low pressure drop through the  ;

units.

(c) In PANTHERS, which is a steady state experiment, both the air and steam l flow to the heat exchangers are measured, and controlled as an independent condition of the experiment. LE has also committed to provide local non-condensable measurements in the PANDA drywell.

Our current plan is to determine the non-condensable concentration distribution by use of a combination of temperature measurements and oxygen sensors located at several locations in the PANDA drywell.

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t RAI Number 900.69 Question:

Adequacy of scaling, phenomena level versus systems interaction: During a meeting regarding the SBWR test and analysis program on August 24,1994, the ACRS Thermal-Hydraulic Pheno'mena Subcommittee expressed concerns  !

about whether preserving parameters like gravity head and local fricdon losses l is suflicient to model an integral system behavior. For example, having a " tall and skinny" test facility may affect the 3-dimensional distribution of noncondensable gases. Another example is that inappropriate modeling of global inertia terms may distort the integral system responses, like pressure ,

and water level oscillations. In the scaling analyses, did GE include these  !

" integral" or " global" effects? i GE Response: [

GE has included these effects, as noted below:

Scaling of the Global Inertia Terms in the Momentum Equation In the top-down scaling analysis presented in NEDC-32288 (Section 2.3), the transfers of mass driven by pressure differences were considered using the momentum equation integrated over a segment (piping) length. A rigorous -

analysis led to Eq. (2.31) of NEDC-32288 where a number of nondimensional groups appeared. The non-dimensional number muldplying the rate of change of the velocity is Hin (Eq. 2.32 of NEDC-32288),

=

p"L iuf / Tfp bin 3po which scales the inertial pressure drop with respect to the reference pressure drop. Considering the transit dme of the fluid in the piping, Pin can be ,

replaced by an alternadve form, Eq. (2.37) of NEDC-32288, ,

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]=Pu Ap r ,

r and the rado of the equivalent inerda to volume lengths, LI/Ly, Eq. (2.42).

Lin Li__[a n Lv Z 8ast, r

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I The inerna number and the ratic LI/LV were considered in Appendix B of )

NEDC 32391P on Scaling Applicability (Tables B1-9 to B1-11 for GIST; B1-22 to  !

B1-27 for GIRAFFE; B1-39 to B1-46 for PANDA). The Ll/Lv ratios of the prototype and of the various experimental facilities are matched reasonably  !

well. Although the experimental facility components often have difTerent Hin values than the ones of the prototype (due to differences in the flow velocities in these components), this is a very minor scaling distortion, since the relative importance of the inertial pressure drops with respect to the system response is very small. Inertial pressure drops can reach significant magnitudes only during rapid system transients when velocities change abrupty; this is not the case during SBWR transients, except during the very first moments of depressurization. (Rapid velocity changes may take place during certain specific phenomena such as chugging; the scaling of such particular effects is considered in the bottom-up analysis. Inexact scaling of local phenomena such as chugging is not expected to affect overall system behavior.)

Moreover, the scaling analysis of NEDC-32288 produced three time scales, (t , t&, and t$), which scale the rates of volume fill, ofinertial effects, and of pipe transfers, respectively (Section 2.4). Clearly, the systems considered here are made of large volumes connected by piping of much lesser volumetric capacity. The inertia and transit times, which are of the same order of magnitude, are much smaller than the volume fill times:

t >> t$ = t as shown in the NEDC-32391P tables mentioned above. It was concluded that the time scale that is controlling system behavior and therefore must be considered in scaling the system is to. The other two dme scales (controlled by the geometric characteristics L1 and Ly of the piping) are clearly of minor importance.

Three-Dimensional Effects It is evident that 3D effects cannot be simulated exactly in experiments where the aspect ratio of the system is necessarily distorted (to preserve the important heights) and the complex SBWR volume geometries are replaced by cylindrical vessels. Mixing and stratificadon phenomena in the various SBWR containment volumes are discussed in Section 3.2 of NEDC-32288, where it is shown that appropriate simulation of the discharge areas of components such as vents and vacuum breakers can preserve similarity of the phenomena.

The Grashof numbers of containment volumes controlling natural circulation are considered in Section B1-2.2.2 of NEDC-32391P. For facility components that are full-height, the Grashof numbers calculated with height as the length scale match very well. Examples are shown in Tables B1-12, B1-28, and B1-47 of

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numbers based on these cannot be matched, but study of 3D effects wm not within the scope of these tests. The horizontal dimensions of the PANDA facility approach those of the SBWR. Moreover, representation in PANDA of the Drywell and Suppression Chamber volumes by two large vessels interconnected by very large diameter pipes essentially provides two horizontal '

reference lengths: for example, the diameter of one SC vessel is close to the i width of the annular SBWR SC pool, while the distance between the opposing i ends of the two SC vessels approaches that of the SBWR SC perimeter. Thus both length scales will be present in the PANDA model.

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RAI Number 900.70 Question:

GE has identified several sources of data that may be included in the SBWR design certification database, e.g., Dodewaard startup, boron mixing tests, and l CRIEPI stability tests. For all of these sources (notjust those cited here), provide  !

detailed documentation about the tests, such as facility design, scaling, and l instrumentation; test specifications and test matrices; and test data and analyses.

Also, document specifically how GE will use these data within the test and analysis program.

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GE Response:

Section A.3.1.6 of NEDC-32391P "SBWR Test and Analysis Program Description" lists six specific sets of existing test data for which TRACG ,

analyses are being planned. Typically, this is non NQA-1 data, much ofit l several years old, but that can be used to illustrate TRACG capability to correctly  ;

predict a specific parameter; PSTF containment data for as containment main vent clearing during blowdown for example. We intend to use this data to illustrate the breadth of TRACG prediction capability and to corroborate the main body of SBWR data. The specific tests included are:

1/6 Scale Boron Mixing Test CREIPI Natural Circulation Test Dodewaard Plant Startup PSTF Mark III ,

Mark II - 4T Suppressions Pool Stratification - Mark III I These data are from tests in SBWR-like, but not necessarily SBWR unique or scaled geometries. Since, in general, these are not SBWR unique tests, specific scaling to the SBWR configuration has not been performed, and were it to be performed, it would result in the obvious; that these are not SBWR scaled tests.

We do not plan to perform any additional scaling analyses for these data sets.

Typically, the phenomena addressed are very specific, and were added to the l analysis plan for additional confidence in TRACG's predictive capability. In i cach of the six cases identified, NEDC-32391 is very specific with regard to runs to be analyzed, and the specific purpose of each of the specific comparisons to be made.  ;

References to specific test documentation and the specific data use are given in NEDC-32391P. The following are GE's comments on each of the six addition data sets:

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1/6 Scale Boron Mixing:

This data was submitted to the NRC on the ABWR docket. The report includes scaling, facility design, test matrices, and instrumentadon used.

Specific runs to be analyzed are still being defined. GE will have a detailed plan for these analyses by December 1,1994.  ;

CRIEPI Natural Circulation Test: .

We recognize that additional information will be required by the staff. GE will prepare a data transmittal on this facility and the results to be used by December 1,1994.

Dodewaard Plant Startuo:

GE will provide the NRC staff with the test reports from the Dodewaard startup, referenced in NEDC32391P by October 15,1994. Reactor description and instrumentation is included in the reports. - Scaling and test rnatrix information is not applicable in this situation.

PSTF Mark III:

The test report referenced in NEDG32391P was submitted to the NRC as part of  ;

the GESSAR docket in 1973. Scaling (to the BWR-6 design), test facility design, ,

instrumentation, and test matrices are included in the report.

I Mark II-4T:

The test report referenced in NEDG32391P was provided to the NRC under the Mark II Containment Program in 1976. Scaling (to several Mark II containment configurations), test facility design, instrumentation, and test matrices are included in the report.

Suppression Pool Stratification - PSTF:  ;

The two reports referenced in NEDG32391P were provided to the staffin 1977 and 1978 under the GESSAR docket. These reports are specifically data analysis reports from PSTF Mark III testing.

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i RAI Number 900.71 Question:

Explain the rationale for excluding shutdown events and beyond-design-basis events from the SBWR design certification test program. Shutdown events must be evaluated for the SBWR, and presumably will be analyzed using the same computer code (s) used for design-basis analyses. As far as beyond-

. design-basis accidents are concerned, the staff must determine the robustness of ,

the passive safety systems to deal with events nominally beyond the design  ;

basis (e.g., multiple failures) and the possibility of reliance on active, non-safety systems to deal with the consequences of these events. Note: "Beyond-design- i basis" in this context is nol equivalent to severe accidents.

GE Response: l Beyond-design basis and shutdown events were not explicitly considered for ,

the study that led to the definition of the Test and Analysis Program. In ,

response to this RAI 900.71, these scenarios have been considered and GE  ;

concludes that they are covered by the dermed programs.

a) Beyond-design basis events:

GE takes this set of events to mean those event and equipment failure combinations which are defined by the PRA success criteria (Attachment 19AA to the SSAR). In these events, core uncovery occurs but cladding temperature remains below 2200 F. The dominant phenomena introduced in these events (beyond the design basis events) relates to core uncovery for a period of time followed by recovery as cooling systems are restored. These phenomena are already included in the PIRT tables (e.g., Cll, Cl3, C14, C15, i C24, C25). Tests which cover these phenomena include the TLTA boiloff test, and small and large break tests in TLTA and FIST. All these tests were performed with a simulated full scale BWR fuel bundle and cladding heatup occurred over a range of temperatures and system pressures. TRACG has been qualified against these tests with excellent results (NEDE-32177P). No additional tests or analyses are needed to cover these events. .

b) Shutdown events Plant shutdown to the hot standby condition is accomplished by bypassing steam to the main condenser and through the use of the RWCU/SDC system for decay heat removal. The ICs can also be used for decay heat removal ,

during this phase of the transient. No new phenomena are introduced in this transient, beyond those already considered. Cold shutdown is achieved .

. through decay heat removal by the RWCU/SDC system. If these systems are  !

not available, other core injection systems (e.g. FAPCS), can be used for decay ,

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heat removal. One train of the RWCU/SDC system is suflicient to remove the decay heat, but two trains are engaged for the first 8 days to keep the cold leg ,

temperature of the RCCW at 95 F. Again, no new phenomena are introduced. -

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RAI Number 900.72 Question: ,

Explain how GE can rely solely upon analysis to resolve the issues of systems interactions during the early phases of transients and accidents, when there are essentially no integral systems test data either exisdng or planned that cover '

such conditions during that time. Note that PANDA is not scaled to represent the early phase of SBWR accidents, and is incapable of representing the " worst-case" sequences for the SBWR, that is, bottom drain line and Gravity-Driven Cooling System line breaks.

GE Response:

GE is not relying solely on analysis to resolve systems interaction issues in the early phases of the LOCA transient. Figure 5.3-1 of NEDC 32391P illustrates the i coverage of various portions of the transient by different integral systems tests.

- Section 5.3 of NEDC-32391P discusses this figure. The systems interaction analysis was performed to identify the needs for tests where systems interactions might be important, where possible adverse interactions rnight occur and where there could be uncertainties in the analysis. This led to the >

definition of the PANDA Phase 2 tests. It is true that PANDA does not have the power supply capability to simulate the decay power at 10 minutes into the transient, and that the GDCS tanks do not have sufficient capacity to simulate the full capacity in the SBWR. However, test procedures will be developed to -

minimize the impact of these parameters on the system transient response.  :

This is addressed further in response to RAI 900.73. It should be recognized that the purpose of these tests is to provide representative data for code validation of the key phenomena and interactions. Thus, in the early GDCS period of the transient, the emphasis is on the interactions between the heat remosal by the PCCS combined with the effects of steam condensation within the reactor and drywell. The key phenomena related to drywell depressurization, sacuum breaker opening, recycling of noncondensibles, PCCS purging and restoration of PCCS performance will all be maintained even if there are scaling distortions '

in some of the parameters.

Incidentally, the bottom drain line break and GDCS line break are limiting for the minimum water level in the reactor vessel. In the PANDA tests, the focus is on the containment performance and the large steam line break is t: - limiting l break.

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4 RAI Number 900.73 Question:

Specify as precisely as possible at what time in the accident sequence the  ;

PANDA tests that are to represent the "carly" phases of main steam line breaks will begin.

GE Response:

Although the detailed procedures for the PANDA Integral Systems Tests with an early start have not been completed, it appears that these tests (M7 and M8) ,

can simulate the SBWR containment response to a steam line break as early as 'I 10 minutes into the transient. ,

At approximately 10 minutes into a main steam line break accident, the RPV .:

pressure is calculated to have dropped to approximately 300 kPa and is nearly equal to the drywell and wetwell pressures. The PANDA vessels and '

connecting piping have the capability to model this transient directly from this time on except for the decay heat and the GDCS inventory addition to the RPV. -

The PANDA power supply is capable of providing 1.5 MW to the electrical  :

heaters in the RPV. The SBWR scaled decay heat at one hour after scram is '

approximately 1 MW. The remaining 0.5 MW is available to simulate the RPV l structural stored energy for those tests beginning at one hour into the simulated SBWR accident. 1.5 MW matches the scaled SBWR decay heat at approximately 20 minutes following scram. j The PANDA GDCS was designed to provide good simulation of the PCCS condensate drain discharge geometry and discharge conditions after draining  ;

of the initial GDCS inventory to the RPV has stopped. Representation of the full  !

GDCS capacity was not an objective for the PANDA design. As a result, the capacity of the GDCS is approximately 40% of the scaled SBWR GDCS volume.

The approach in PANDA for modeling the SBWR transients prior to one hour after scram will take advantage of the fact that a significant fraction of the SBWR decay heat during this period is used to heat the subcooled GDCS water  ;

which has drained into the RPV. By running the PANDA tests with a constant  !

power of 1.5 MW for the period simulated prior to one hour and adjusting the initial conditions in the RPV and the GDCS, it is expected that the test start time can correspond closely to 10 minutes into the SBWR main steam line break. ,

As stated above, the detailed test procedures for M7 and M9, the PANDA l Integral Systems Tests with an early start, have not been completed. For test l M7, however, the approach described above will provide data to demonstrate the PCC capability to start-up when it is initially filled with air and RPV conditions  :

are representative of SBWR conditions immediately following blowdown. For l test M9, the RPV and GDCS conditions will be adjusted to cause vacuum  !

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breaker opening and reintroduction of air to the dr>well and PCC. Test M9, therefore, will demonstrate the PCC startup capability if air is reintroduced to the drywell via the vacuum breakers early in the transient.

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RAI Number 900.74  !

Question: {

PANDA tests will be initiated "on the run", therefore, a transient condition will  !

be established which is intended to simulate a particular reactor transient. How l will this be accomplished without significantly affecting the transient under  ;

study?

l GE Response:

t The initial condidons for the PANDA tests will be based on calculated  :

conditions in the SBWR at the time in the transient corresponding to the test l start time. For the transients to be simulated, the SBWR pressures, temperatures, j liquid levels, and non-condensible gas concentrations which will be the basis i for the PANDA initial conditions are not varying rapidly with time. Therefore, l establishing inidal conditions based on the calculated values for these slowly  !

varying parameters will not affect the test transient. l l

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RAI Number 900.75 i

Question:

The staff has previously requested detailed test matrices for the PANTHERS  !

Isolation Condenser (IC) tests. These have never been provided and the i information in Appendix A of NEDC-32391P, "SBWR Test and Analysis  !

Program Description," is not sufliciently detailed (e.g., noncondensable gas j concentrations, test duration, test cycles, etc.). Provide this information for ,

review. In addition, address the concerns raised about instrumentation for i PANTHERS PCCS testing (Q900.68 above) for the IC tests.

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GE Response:

The PANTHERS Test Requirements and Test Specification were sent to NRC in MFN 119-92, dated May 27,1992. Rev. 2 of this specification was transmitted by MFN 101-94, dated August 31,1994.

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. . i RAI Number 900.76 Question:

The scaling analysis submitted with NEDC-32391P, "SBWR Test and Analysis  ;

Program Description," is an improvement over previous documentation provided by GE; however, additional work is required to demonstrate that for t each of the important phenomena identified in the phenomena idendfication '

and ranking table (PIRT), the range of thermal-hydraulic conditions expected in the SBWR is covered by one or more tests in the test program.

GE Response:

Data has or will be obtained for all of the phenomena marked 'High' in the phenomena identification and ranking table (PIRT). The data comes from a  :

combination of testing programs and plant data. The type of data used for each  ;

phenomena is indicated by the test coverage matrix shown in Figures 5.5-1 and  ;

5.5-2 of the TAPD. Table 1 of this RAI includes the information in those tables  !

along with scaling information on the phenomena. More detailed information  ;

showing specifically which test or tests are used to obtain data for each l phenomena is also contained in the tables in chapter 5.  ;

Data for some of the parameters have been obtained from operating BWR's. .l Therefore the data will be over the same ranges as expected in the SBWR. )

Additionally, data from BWR simulator facilities such as SSTF, TLTA, FIST, PSTF and the Boron Mixing Facility have been used. These facilities were design to simulate operating BWR behavior for accidents and transients which ,

are very similar to those for the SBWR. A description of each of these facilities  ;

is included The TRACG Qualification document, NEDE-32177P, Rev 1. Data  ;

obtained from this category is marked in the "BWR facility" column of Table 1.

In addition, for those parameters that were considered to be particularly important, a detailed review of the test ' data and ranges used for coverage was performed. This information is contained in the Qualification Data Base (QDB) that supports the TAPD. Table 1 indicates which PIRT phenomena are reviewed in the QDB. Phenomena that is covered by data from GIST, GIRAFFE, PANDA or PANTHERS has already been scaled in Appendix B of the TAPD. These phenomena are indicated by checks in the " Scaled in App. B" column of Table 1.

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Table 1. PIRT phenomena data coverage Test Coverage Scaling i F

\ $ h i e a a F I e i i t E

]

! d a 3 [ s 5 I

PIRT # Phenomena issue t

E t

E E c

I 3

)m A1 jlP flashing / redistribution X!XlX X X A2 LP heat slab stored energy _ XlX X X X X A3 Inlet orifice uncovery IX X X A4 lLP void fraction X X X. X X A5 lLP void collapse / inlet subc. X X X X X A9 lLP stratification X X X X X B1 Byp_as_s flashing X X X X X B2 Bypass level X X X X B4 jCCFL at bottom of bypass X X X BS CCFL at top of bypass X X X B6 Channel to bypass leakage X X X X X B7 l Bypass refill X X X X C1AX Void coefficient X C1BX , Doppler _ coefficient X C1CX l Scram reactivity X C2AX Interfacial shear and h.t. X X X X X X C2BX Subcooled boiling X X X X X C3AX Fuel pellet power dist. X X C3CX Fuel gap conductance X X C4 Core flashing. X X X X X C5 Inlet orifice uncovery X X X C6 Inlet orifice CCFL X X X C7 Upper tieplate CCFL X X X X C8 Multibundle flow dist. X X X X C8X . Core void _ collapse iX X X C10 !Core void distribution X X X X X C11 Charinel to bypass leakage X X X X X C12 Natural circulation flow X X X X X C13 ;Dryout/ boiling transition X. X _

X C14 IFilm boiling (Iow flow) X! IX l X f

e Table 1. PIRT phenomena data coverage (cont'd)

I Test Coverage Scaling i' e

$ ji 5 l e. p S 2

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s

• l I&r Igi l 5%I i

- I PlRT # Phenomena lssue ,,

C15 ! Film boiling _(disp. drop.) Xl X X C23 l Core pressure drop X X _

X X X X C24 ! Decay heat iX X X X C25 l Fuel stored energy _ X X X X C26 ! Critical po_wer for 9 ft core X l X D1 lGT flashing X X X! X X D2 lCCFL at top of GT X X X D4 l Refill oTGT X X X E1 D C break uncovery X X X X E2 D C void profile X X X X E3 LGDCS_ interaction X X E5 lD C heat slabs X X X X X E6 D C flashing X X X E7 IC interaction E8 D C break flow X X X X F1 Chimney void distribution X X X F2 . Chimney flow distribution X X X F4 Mixing.at top of chimney X X F5 Geysering during startup _i X X 11 3

Separator CU/CO IX' X 12 jSeparator inertia _

X 13 !Sep_arator pressure drop Xl X L1X Steamline pressure drop lX X L2X Steamline acoustic effects X X L1 SRV/DPV critical flow X' X X X X L2 jDroplet entrainment X X X X L3 iTransition to unchoked flow X X X X L5 lMultiple choked locations !X X Q1 IIC pressure drop Xl X X b

i Table 1. PIRT phenomena data coverage (cont'd)

Test Coverage Scaling F i

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I g 2 s-hi! gd, "w I - "

{

PIRT # Phenomena lssue $!h .,, 55 O2 jlC capacity ___ X l XlXl 03] Stratification in IC_ drums _ _

X _j_ X'X 04 ilC pool stratification . !Xl X X 05 l Secondary side heat transfe.r. iXl X X  ;

ST1 ! Hydrodynamic stability X( l X

')T? lCorewide stability _ _

l X X

'g@_jR_egional stability i X XL ATWi[ Boron mixing in bypass _

{X _

X ATW2fBoron stratification to LP X X ATW3! Boron delivery to core _

X' X ,

jlnteraction between multiple XL1 ilC modt.les and units X. _X

~

' System inieraction -

GDCS/ System  !

XL3 ,depressurization X,,_ X CONTAINMENT ~~

r EH IBTeak mas's flow Critical flow X X X. X ,

X d_ _ _ . _ .

Friction X X Xj_ Xi X j, _ _

Entrainment X X X X X ,

MV1 l Main vent flow X X. X j MV3 ! Vent clearing time Xl Xl X XT

~

SO1 SRV flow X X X DW1 Flashing /evapo_ ration in DW X X DW2 , Condensation on DW walls X X X X

_!Degredation of conduction X '

X X X Wall / Structure conduction X! X X X Phase i

}

DW3 3-D effects distribtalon X X'X i Noncondensable

! s distribution X X X l Buoyancy /

! natural

! circulation X X X i

Table 1. PIRT phenomena data coverage (cont'd)

Test Coverage Scaling i

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{ k i

1 i g g h  !! i i r 5 m i g [

PIRT #, Phenomena lssue  !$$! .,, (5 5 f Condensation on reactor Interfacial Heat  !

DW4 outflows Transfer _

X X X

! Degredation by j j N/C X X X  !

jCondensation/ evaporation Interfacial Heat l WW1 ;of main vent discharge _

Transfer X; X X X l Degredation by i N/C _

X X i Condensation / evaporation Interfacial Heat j WW2 of SRV discharge Transfer X X  !

Condensation / evaporation Interfacial Heat  ;

WW3 of PCC vent discharge Transfer X X X X  !

Free surface condensation / Interfacial Heat  :

WW4 evaporation

~

Transfer X X  !

Degredation by N/C X X X  !

Condensation on l

WW5 Heat sources / sinks WW walls X X X X Conduction through WW .

walls X X X X .

Degredation by l N/C Xl X X- X i Bouyancy/

natural WW6 Pool mixing and stratification circulation X X X X X  !

Temperature WW7 3-D effects in gas _ space distribution X X X noncondensable distribution X X X

! Interfacial shear i X X ,

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Table 1. PIRT phenomena data coverage (cont'd)

Test Coverage Scaling f

$ E I l, !s 4 E

l 9 ;I h I

0I; t{ 2 5-3' m l

PlRT# Phenomena lssue !l$

h h ,,

l

! Mixing, entrainment intojets X X t Bouyancy/natur f

_g _

al circulation X X l Phase ,

! separation _X_{X

{ Containment spray Interfacial Heat WW8 condensation Transfer X X X i Degredation by  ;

N/C X X i fContainment hydrodynamic WW9 ,lo_ ads Pool Swell X X X Condensation oscillation X X Chugging ,

X  ;

l_ SRV Discharge _j i__X X X l GD2 'GDCS flow __

l X X X j PC1 !PCC flow / pressure drop X X X X  ;

Condensation on primary l PC2 side Interfacial H.T. X X X X X Degradation by n/c X X X X X  ;

Shear Enhancement X ,

PC3 Secondary side heat transfer Pool temp. dist. X X X X  ;

Poo! void dist. X X X X Natural  ! ,

circulation !X X X X I

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Table 1. PIRT phenomena data coverage (cont'd) l Test Coverage Scaling F

{

l i E C i a 4 i g e E E h l

i h

l 5

g a j

PlRT# Phenomena issue I$!$ '

.,, 5 .,, j Secondary side l entrainment X X- X X

]

PC4 Parallel PCC tube effects Friction X X X X I

l Void fraction X X X X  !

l PCS ' Parallel P_CC unit effects Friction X X X X l

Void fraction X X X X iPCC lan component PC6  ! separation X Purging by l PC8 PCCS startup with n/c presssure diff. X X X Degredation by

~

N/C X X X Leakage Aetween drywell  ;

DWB1 and wetwell X X X X l Vacuum breaker flow VB1 characteristics X X EQ1 Equalization line flow X X ]

EQ2 Equalization line sloshing X X i Heat transfer to safety  ;

OC1 envelope_  ;

DPV1 Mass flow in DPVs Critical flow X  ;

Friction X Entrainment X

Containment liner gap CW1 ! conductance ~

l Concrete properties at high CW2 ' temperature X X

Table 1. PIRT phenomena data coverage (cont'd)

Test Coverage Scaling

! P

{,lj fE' E i' il gF 3

g E i

I-> _ ij s![ r I

i PIRT # Phenomena Issue  !-lr r!e E E c Q

, $l5

Operation in i i l Passive Autolitic hydrogen rich  !

l PAR 1 <Recombiners environment XI X Added heat load j l from recombination PAR 2, -_ _

reaction i

XC1_ System interaction IC/DPV/PCCS . _

X i

l IC/DPV/

(lX l

XC2 jystem interaction GDCSPCCS lX X i i XC4 l System interaction FAPCS/ PCCS X X l multiple PCC l

l modules and XC5 Jystem interaction units lX i X l light i noncondensable XC6 ! System inter _ action _ DW/PCCs/WW X X containment i system

response XC7 ' System interaction (DW/WW/MV) X X I

p O RAI Number 900.77 Question:

Responses to the staffs previous requests for additional information (April 11, ,

1994) are also needed to determine the adequacy of NEDC-32391P, "SBWR Test and Analysis Program Description." Of particular interest are responses to Q901.23 through Q901.27. j i

GE Response: l Responses to the referenced RAIs have been sent in to the NRC by letter MFN 096-94, SUBMITTAL OF ADDITIONAL INFORMATION ON LICENSING ,

TOPICAL REPORT (NEDE-32177P and NEDE-32178P), dated September 20,1994.  !

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l RAI Number 900.78 1

Question: 1 1

The SBWR is unique from the standpoint of suppression pool thermal capacity.

It is designed only for the first hour of decay heat energy, unlike previous designs which could accommodate all of the decay heat energy. Therefore, discuss the interactions expected between the PCCS and the suppression pool during transient periods such as PCCS purging, return to steady state operation, and vacuum breaker opening. Specifically, discuss the potential of opening the main vents for short periods, thereby sending mass and energy to the pool and possible instabilities as seen in the single tube condensation tests at the University of California (Berkeley). This discussion should rely on test data as much as possible.

GE Response:

The statement in this RAI that the SBWR design is unique from the standpoint of suppression pool thermal capacity is incorrect. All BWR pressure suppression pools are sized to accommodate the primary system blowdown energy. None of the suppression pools in existing GE pressure suppression containment types are designed to accommodate all of the decay energy without resort to some other energy removal system. In the absence of such a system, the pool will continue to heat with time as decay energy is added.

In earlier containment designs, the suppression pool temperature is limited by operation of the pool cooling mode of the Residual Heat Removal (RHR)

System. The suppression pool absorbs the blowdown energy prior to operator initiation of RHR. Energy addition to the suppression pool continues by flow of drywell steam (generated by decay heat) through the main vents, and energy is removed from the pool by the RHR system to the ultimate heat sink. The peak pool temperature is established by the relative rates of energy addition and extraction. Typically, a maximum pool temperature near 190 degrees F occun about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the accident scenario, when the RHR heat exchanger delta T is sufIicient to remove energy at the rate of energy addition to the pool from decay heat. The suppression pool temperature then slowly decreases as the decay energy addition decreases.

In the SBWR the situation is similar. During the blowdown period, the suppression pool absorbs the majority of the primary system energy, although there is some energy extraction by the PCCS. Depending on the break scenario, the blowdown period lasts from about 10 to 30 minutes. Following blowdowm, GDCS reflood of the vessel causes subcooling ofits fluid contents, and little steaming occurs until about I hour into the accident scenario. At this time, the PCCS is capable of rejecting all of the decay heat. In this way, the PCCS is analogous to the RHR system.

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1 A critical element of SBWR design is the PCCS heat exchanger vent j configuration. The PCCS vent exits into the suppression pool at a shallower submergence than the top main vent. This geometry is important, because the  !

SBWR pressure suppression containment, like all earlier containments, is a I forced flow, pressure driven system, not a temperature driven natural convection system. In all pressure suppression containment systems, mass and

+ wrgy are added to the dgwell from the break in the primary system, and the anwell pressure increases. The pressure will continue to increase, lowering the water level in the vent system, until a flow path is established between the drywell and the wetwell. The wetwell pressure is set by the thermodynamic conditions in the wetwell, including partial pressure of the original wetwell air (or nitrogen in the case of the SBWR), the partial pressure of the air purged over from the dowell to the wetwell air apace, and the vapor pressure of steam 1 corresponding to the suppression pool surface temperature. Once the vents have cleared, the drywell pressure is equal to the wetwell pressure, plus the l submergence head of the vents, plus any flow head losses in the vent system. l There would be flow from the dnwell to the wetwell even if there are only non-condensable gases in the dqwell. (In fact, some of the containment '

testing performed in the 1970's and '80's was performed with only non-condensables.) Once sufficient mass and energy are added to the drywell so I that the vent submergence head is overcome, flow will occur. This holds true ,

whether the flow is through the main vents, or through the PCCS heat exchangers l l

Early in the LOCA scenario, mass and energy addition rates into the dowell l from the primary system are larger than the heat removal capacity of the PCCS.

During blowdown, the dgwell pressure is such that both the PCCS vent and the main vents are cleared, and flow goes to the suppression pool via both paths. As primary system steaming decreases, the drywell pressure will decrease, 1 eventually allowing the top main vents to re-flood and flow to the suppression j pool will stop. Flow will still occur, however, through the PCC heat exchanger l and PCC vent. It is the difTetence in submergence between the main vents and PCCS vent that preferentially directs flow through the heat exchanger, and shifts the primary LOCA heat sink from the suppression pool to the PCCS pool.

Table 1 illustrates both the similarities and differences in suppression pool design as containment configurations have evolved. This table gives the ratio of pool volume to core rated thermal power. Both blowdown energy and decay heat are a direct function of core rated power. Thus the ratio of pool volume to ,

core thermal power is a direct indication of the suppression pool's ability to  !

absorb the total primary system accident energy. The value given for the SBWR is the highest of all the containment types listed. The design is ven' robust. The relatively high value of this parameter for the SBWR is the result of l two factors, (1) the potential for thermal stratification in the SBWR suppression pool, which has no safety grade system capable of mixing the pool, and (2) the requirement that the pool absorb both the blowdown energy and that small fraction of the excess decay energy that is released, until the PCCS system is capable of assuming the full load at about one hour.

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Table 1 Ratio of Suppression Pool Volume to Core Thermal Power Containment Type Ratio (Cubic Feet per AfWth)

SBWR 57.58 ABWR 32.20 Sik III (GESSAR) 36.21 Afk III (Grand Gulf) 35.48 hik II (Nine Afile Point 2) 46.58 Aik I (Browns Ferry) 37.35 In the SBWR, reopening of the main vents is not expected to occur following GDCS reflood of the RPV. If, due to the addition of non-condensables, for example, the PCCS heat rejection capability temporarily drops below the decay energy level, the drywell pressure will increase. However, before the drywell pressure reaches the point where the main vents will reopen, a pressure difference will exist that will clear the PCCS vent, effectively purging the non-condensables and reestablishing PCCS performance.

Even ifit is postulated that flow through the main vents is somehow reestablished, the amount of energy added to the pool before an effective PCCS purge of non-condensables causes only a small increase in suppression pool temperature. A bounding calculation of an event of this type was transmitted to the NRC staff by hiFN No. 214-93. This calculation was based on the bounding assumption that all the decay heat energy is absorbed by the suppression pool during the time period required to purge all the hydrogen produced by a 100%

metal-water reaction from the drywell to the wetwell via the PCCS. The resulting additional pool heatup for this scenario is 3 degrees K.

At one hour into the accident scenario, the steam generated by decay heat in the SBWR is about 12 kg/sec.. The top vent area of the SBWR is 3 square meters, yielding a mass flux of about 4 kg/m 2-sec. The condensation regime has been observed to change from steady to intermittent (chugging) at mass 2

fluxes lower than 10 lbm/sec ft2 (48.9 kg/sec m ). Therefore, even if main vent flow were to reoccur, it would be within the chugging regime. Cyclic flooding and re-clearing of the main vents that occurs during chugging results in improved suppression pool mixing, and a reduction of pool thermal stratification. The SBWR design uses conservative assumptions for suppression pool stratification, based on limited mixing. While the effect of chugging in reducing thermal stratification is difficult to quantify, it is certainly present, and the effect of the energy addition to the pool would likely be less than the 3 degrees K estimated above.

Instabilities were seen in the first single tube condensation experiments performed at UC Berkeley. This experiment, reported in NEDC-32310 " Single Tube Condensadon Test Program", was performed in a natural circulation loop.

Subsequent single tube experiments utilizing forced circulation loops, including two experiments at UC Berkeley and two at the Massachusetts Institute of Technology, have not shown any evidence of flow oscillations or instabilities.

Thermocouple instrumentation of the heat exchanger tubes in the PANTHERS experiment (see response to RAI 900.68) make it possible to monitor for instabilities in this prototype heat exchanger test. No evidence of instabilities has been identified in data reviewed to date, which include conditions that span the PCCS flow regime. Given that the SBWR is a forced flow design, and that no instabilities have been seen in forced flow experiments, they are not expected to be a factor in SBWR performance. Also the condensation instabilities seen in the UC Berkeley natural convection experiments were local in nature, and did not greatly effect the overall heat rejection within the tube.

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1 RAI Number 900.79 Question:

Heat loss has proven to be a significant problem in evaluating the GIRAFFE data. Therefore, provide the heat loss evaluation of both the PANTHERS and j PANDA facilities and discuss how these losses will be considered in the i evaluation of the test data.

1 GE Response: i i

Reference:

PANTHERS-PCC TEST PLAN AND PROCEDURES, SIET  ;

Document No. 00098PP91, Revision 1, July 12,1994, sent to the NRC in MFN I No. 098.94, dated August 16,1994. l Section 8.1.2.5 of the PANTHERS Test Plan & Procedures (see reference) gives  ;

the equation for the global energy balance of the PCC at PANTHERS. The equation includes the heat losses of the inlet and outlet lines. However, these i heat losses were measured during the shakedown of the test facility and found l to be negligible (i.e., less than 50 kW) compared to the total thermal power (around 1 to 14 MW). Therefore, the condensation thermal power formula will be simplified to that shown at the end of the referenced section.  ;

i Quantification of heat losses for PANDA is a planned item in the test facility -

startup program. The measurements have not yet been performed. PANDA is very heavily insulated, and heat losses are not expected to have a major effect i on the results. The design goal is to limit heat losses to 10% of the decay heat at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the LOCA scenario. Calculations indicate losses will be j substantially less than the target values.

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RAI Number 900.80 I

Question:

i Interaction between the ICs and the PCCS may have a profound impact on the j performance of the system. Discuss the possibilities of tests considering both  ;

units operational. In particular, the early in dme test to obtain GIST-type data ,

should be one of the tests considered.

GE Response: i i

The systems interacdon studies performed as part of the SBWR test reassessment and reported in NEDC-32391P indicated that the minimum RPV water level was slightly effected by the presence of the IC and PCC for some i postulated break scenarios. However, there was essentially no effect on system performance. The SBWR is a very robust design from the standpoint of core '

cooling. Minimum accident water levels are calculated to be approximately 1 to 4 meters above the top of the fuel, and peak clad temperatures are essendally unchanged from steady state performance values. Overall system performance ,

would only be effected if the water level dropped below the top of the fuel, and -

even then there would be very significant margins to 10CFR50.46 and Appendix K temperature limits.  ;

Appendix A of NEDC-32391 defines the tests GE has concluded are technically '

adequate for SBWR certificadon. PANDA test M6 was added to the matrix specifically to address IC efTects. As a result of staff comments from the meedng on August 18, we are considering adding IC operation to PANDA tests ,

M8 and M9 as well. As noted in the response to RAI 900.73, PANDA tests M7 ,

and M9 will be started approximately 10 minutes into the accident scenario.

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RAI Number 900.81 Question:

Transient behavior is of particular concern to the staff, therefore, each PANDA l

test duration should include at least one purge cycle of the PCCS. Confirm if i that is the case.

GE Response:

It is unclear what is meant by a " purge cycle" in this RAI statement. Every PANDA test begins with some air fraction within the drywek Over time, this air fraction will decline, but there will always be some small residual air content in the dqwell. Tests M1 through M4, are of this type. Tests M5 through M9 have test conditions defined to address specific TRACG qualification needs as defined in NEDC-32391P. Some, but not all, of these tests will result in the vacuum breaker opening, and re-enty of non<ondensables into the drywell. In these cases, the purge of these non-condensables into the wetwell will be investigated. Again, there can be no assurance that all the air will be purged from the drywell in any given test.

Superimposed on these system purges may be short cycle variations in the non- )

condensable content with the PCC heat exchanger. These will be investigated, should they occur.

If the staff will be more specific in what they mean by "PCCS purge cycle" we can respond more fully.

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4 Purdue University Questions - Set 5  !

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1. Provide the SBWR drywell spray flow rate and water temperature. j

2. Provide the SBWR wetwell spray flow rate and water temperature.  ;

3. What were the droplet size, flow rate, and water temperature of drywell  !

and wetwell sprays that were assumed in TRACG analyses? -l l

i GE Response: j I

1. The maximum allowable differential pressure across the containment l liner determines the drywell depressurization rate and consequently the maximum allowed drywell spray flowrate. These parameters have not i been finalized yet.

The Fuel & Auxiliary Pools Cooling System (FAPCS) pumps are variable j 3

speed pumps and can provide a flow rate between 257 and 422 m /h in the l drywell spray mode. If these flow rates are too high, it can be reduced to - i 3

150 m /h for long term operation without causing problems with the l pump. If an even lower flow rate is required, the flow can be partly i bypassed by opening the valve in the discharge line to the suppression j Pool-  !

The spray temperature has been calculated to be 55 C with a spray 3

flowrate of 346 m /h with the supression pool water (source of drywell  :

spray water being cooled by the FAPCS heat exchanger) at 79 C. f

2. The maximum allowed wetwell spray flowrate has not been finalized yet. li i

The Fuel & Auxiliary Pools Cooling System (FAPCS) pumps are variable  !

3 speed pumps and can provide a flow rate between 307 and 445 m /h in the -l wetwell spray mode. If these flow rates are too high, it can be reduced to i 3

150 m /h for long term operation without causing problems with the  !

pump. If an even lower flow rate is required, the flow can be partly i bypassed by opening the valve in the discharge line to the suppression .

pool. l i

The spray temperature has b en calculated to be 55 C with a spray l 8

flowrate of $46 m /h with the supression pool water (source of wetwell  ;

spray water being cooled by the FAPCS heat exchanger) at 79 C.

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3. The drywell and wetwell sprays are simulated with the use of a TRACG PUMP component and a component representing the system heat .

exchanger. The flow rates used for the two spray modes were 321 m3/hr and 307 m3/hr for the drywell and wetwell, respectively. The temperature of the spray is not prescribed. The water is circulated through the simulated heat exchanger, characterized by a heat transfer area of 386 m2, an overall heat transfer coeflicient of 1510 Wm2-K, and a sink temperature of 313K. It is expected that the outlet temperature will be only slightly above the sink temperature. Spray droplet size is not prescribed. It is determined by TRACG as the value implied by a critical Weber number of 6.5, based on relative velocity, or 0.2 mm, whichever is larger.

As an example, for containment conditions of 300 kPa and 100% steam, the relative velocity is about 5 m/sec, yielding a droplet size in the range of 7 to 8 mm.

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0- Enclosure 2 to MFN 113-94 i

RAI Number: 900.83 Question:

Discuss how TRACG models mixtures of steam and non-condensable gases, including mixtures with more than one species of non-condensable gas (e.g.,

steam, nitrogen, hydrogen).

GE Response:

In addition to steam, TRACG solves the mass conservation equation for a second gas species. In a given computational cell, the two vapor species are perfectly mixed so that they have the same temperature and velocity. Thus the noncondensible gas is transported with the steam to the next cell with the same velocity as the steam. The concentration and partial pressure of the noncondensibles are tracked in every cell. Conventional donor cell techniques are used to calculate the flow of the noncondensibles from one cell to the next.

The assumption of perfect mixing within a cell can make the results sensitive to the cell size, and the cell size must be chosen appropriately for the problem being solved. In a three (or two) dimensional grid, buoyancy effects due to larger concentrations of a light gas in certain regions can be properly accounted for within these assumptions.

Currently, TRACG allows for only one gas field other than steam. A mixture of two species would be treated as a gas with averaged properties. For this specific analysis of the containment design basis (100% metal-water reaction) event, a mass conservation model for a second gas field is being implemented in TRACG.

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Enclosure 2 ta MFN 113-94 RAI Number: 900.87

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Question:

Provide details of the CSAU study related to SBWR containment analysis.

GE Response: i GE intends to follow the CSAU methodology developed by Boyack et al (Quantifying Reactor Safety Margins, NUREG/CR-5249,12/89). The 14 step 1 methodology developed by this team is outlined in the attached figure from the i above reference. Currently, GE is at Step 8 in the process. The test and analysis - l needs have been defined and the Separate Effects Data analysis is completed.  :

The remaining steps involve the determination of bias and uncertainty in the i TRACG calculations (Step 9), establishing whether there is a scale effect (Step .;

10), and accounting for the effects of uncertainties in the plant operating ,

parameters (Step 11). Under Step 9, all the parameters identified as High in the .!

PIRT tables ( e.g.,4.1-2(a)) will be addressed. It is expected that a much smaller subset of this list will show significant effect on the containment pressure and temperature response in the preliminary sensitivity studies. For this reduced set 4 of sensitive parameters, reasonable ranges will be defined for the subsequent  !

statistical analysis in Steps 12 and 13. The model and plant parameters will be  !

perturbed from their nominal values in a set of TRACG calculations. These r calculations will serve to define the upper 95th percentile pressure and  !

temperatures, which will be compared with the allowable design limits.  !

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I Enclosure 2 to MFN 113-94  ;

l RAI Number: 900.91 1 1

Question:

The staff is concerned that assumptions termed as " licensing basis" which are used for calculations of accidents and transients in the SBWR do not represent the actual operation of the plant which would be expected in such cases. These analyses routinely exclude operation of safety systems that would be expected to operate, such as the isolation condenser. It is also possible that selected non-safety systems could operate and change the integral plant behasior. GE should include in its test programs a range of test conditions to ensure that the data will represent a sufIiciently broad basis for code assessment assuming both

" licensing basis" conditions and realistic plant conditions during accidents.

GE Response:

The " licensing basis" calculations do not take credit for equipment not classified as Engineering Safeguards. Also, single failure assumptions are required in the analysis. However, GE has performed analysis to show that scenarios where such equipment is available improve the accident response, and that the licensing assumptions do in fact provide bounding results.

In NEDC-32391P, calculations have been performed with the ICs available.

Cases have also been run with active systems operating (CRD and FAPCS).

Based on these calculations (Appendix C), testing needs have been defined. The PANDA tests will include tests where the ICs are operational. The effects of the FAPCS in the drywell spray mode will be simulated by adding cold water to the drywell. The GIST tests included one (A05) in which the CRD system was simulated. The ICs have a beneficial effect on the limiting LOCA transients and were not simulated in GIST.

Enclosure 2 ta MFN 113-94 '

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RAI Number: 900.93 Question:

Provide a discussion of vacuum breaker actions for analyzed transients and accidents, including a Gravity-Driven Cooling System line break and include discussion of assumptions made for both expected and " licensing-basis" scenarios. In addition, detail why failure to close (after actuation) of a drywell-to-wetwell vacuum breaker is not, in GE's view, a credible failure.

GE Response:

Vacuum breaker cycling has been predicted for nearly all LOCAs, following GDCS initiation. The injection of subcooled GDCS water into the vessel reduces the pressure in the vessel and drywell to below the setpoint of the vacuum breakers and they cycle open, returning noncondensible gases to the drywell.

Predictions indicate the vacuum breakers remain open for only brief periods, and can cycle several times during the GDCS injection period of the transient.

The LOCA transient which is predicted to provide the most vacuum breaker cycles is the GDCS line break. This accident dumps the inventory of one GDCS pool directly into the drywell, which produces vacuum breaker openings.

Later, as the GDCS flow from the unbroken lines fills the vessel to the level of the break and spills over into the drywell, additional vacuum breaker openings are predicted. Predictions of this transient indicate that as soon as the decay heat boiloff resumes, the drywell is re-pressurized, flow through the PCCS resumes and the noncondensibles are slowly purged through the PCCS, back to the wetwell.

Differences in the ' licensing basis' and expected LOCA calculations such as those presented to the NRC relate to availability of additional safety systems. As was shown, the use of intermittent drywell spray, while reducing the drywell pressure, also produces additional vacuum breaker cycling. For all cases analyzed to date however, the PCCS was able to return the recycled noncondensibles to the drywell and retain part of the pressure reduction benefit resulting from the use of the spray.

The assumption of the reliability of vacuum breaker operation is based on the design requirement of the vacuum breaker. The vacuum breaker valve design reliability objecdve is to fail to open or close less than once in every ten thousand demands. To achieve this objective, simplicity of design was,used.

The design configuration selected is a vertical poppet valve opening with high wetwell pressure and closing by gravity plus drywell pressure. The valve has double sealing surfaces une hard and one soft. The sealing surfaces are designed so that a design basis seal obstruction could be accommodated on one seal without the failure of the second seal. To demonstrate reliability, the prototype valve has undergone extensive testing. Before the valve reliability test was begun, the valve was aged and degraded to simulate sixty years of service.

Aging consisted of soft seal irradiation, whole valve thermal aging, whole valve

Enclosure 2 to MFN 113-94 1

dynamic aging, design basis accident steam aging and ingestion of grit to coat seal and moving surfaces. . The valve was then cycled three thousand times without failure. Using a 13ayesian statistical approach, three thousand cycles without failure was shown to demonstrate a high probability of meeting the reliability objective of one failure in ten thousand.

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o' Enclosure 2 ta MFN 113-94 RAI Numbcr: 900.94

. Question:  ;

f Provide a listing of the TRACG code version used for each TRACG run i analyzed and presented during the " scaling" part of the August 18,1994,  ;

meeting, including a discussion of any differences in the results obtained with j the " preliminary" and the " Level 2 " versions of the TRACG code.

GE Response:

The results discussed at the meeting are contained in Figures B.3-1 to B.3-4 for GIST and Figures B.3-5 to B.% for GIRAFFE. For GIST, TRACG calculations are i shown for the test, for the current SBWR design and the 1988 SBWR design. Of ,

these, the test predictions and the calculations for the current SBWR design 4 were made with the Level 2 version of the code, while the calculations for the ,

1988 SBWR design were old calculations. Calculations made with the  !

preliminary code version and the Level 2 version have shown very little  !

differences for other similar calculations. The GIRAFFE test predictions as well as the corresponding calculations for the SBWR in Figures B. 3-5 and B.% l were all made with the Ixvel 2 version of the code.  ;

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