ML20093A430
| ML20093A430 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/06/1984 |
| From: | Hobbs S HYDROGEN CONTROL OWNERS GROUP, MISSISSIPPI POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| HGN-018, HGN-18, NUDOCS 8407100488 | |
| Download: ML20093A430 (67) | |
Text
___ _ _ _ - -
4
)O~'l/I' o-w7 MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi N
P. O. B OX 164 0, J AC K SO N, MIS SIS SIP PI 3 9 205 NUCLEAR LICEN5tNG & $AFETY DEPARTMENT Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Harold R. Denton, Director
Dear Mr. Denton:
SUBJECT:
Hydrogen Control Owners Group Meeting Basis Submittal HGN-018 The Hydrogen Control Owners Group (HC0G) has undertaken the 1/4 scale test program to define the thermal environment due to the combustion of hydrogen as a diffusion flame resulting from a postulated degraded core scenario. The attached information is provided to the NRC, prior to the planned meeting, for their review of the progress of the 1/4 scale test program. This submittal is to be the basis for the NRC to comment on the 1/4 scale test matrix so that HCOG and the NRC can reach concurrence on the test to be conducted in this facility. The HCOG-NRC meeting is scheduled for July 24-25, 1984 at Factory Mutual Research Corporation. An agenda for the meeting is included as Attachment 1, and the objectives for the meeting are included as Attachment 2.
The meeting will be held at the Factory Mutual Conference Center which is next to Factory Mutual Research Corporation in Norwood, Massachusetts.
(See Attachment 3 for a map of the area.)
Significant progress has been made on construction of the test facility, including pressure testing of the vessel. The NRC staff members and their consultants attending this meeting will be invited to the test site to view the facility.
(Appropriate clothing is recommended.) A bus will be provided to transport participants in the tour to and from the test site.
This submittal includes a summary of the HCOG program status (see ) which will also be discussed at the proposed meeting. The instrumentation plan (see Attachment 5) is similar to that submitted in the HCOG response to the NRC requests for additional information (RAIs) previously transmitted to the staff on April 2, 1984. However, the current instrumentation plan now shows the location of the video cameras. The HCOG has proceeded with the 1/4 scale test program as described in previous submittals on facility design and HCOG's responses to RAIs.
Additional development of the BWR Core Heatup Code has occurred since initial discussions of the code with the NRC staff. The revised code is now being used to generate the postulated degraded core accident hydrogen release time histories which will be used for testing in the 1/4 scale facility.
Conservative assumptions are being made both in scenario selection and code g
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i HGN-018 Page 2 modeling to produce hydrogen release rates which are predominantly above the threshold for establishment of steady diffusion flames. This assures that the thermal environments obtained from the 1/4 scale facility will allow assessment of effects of steady diffusion flames. Equipment required to survive the thermal environment produced by deflagrations which occur with hydrogen release rates below the threshold have been demonstrated analytically to survive. These analyses were performed assuming a conservative constant hydrogen release rate duration corresponding to the total hydrogen production equivalent to oxidation of 75% of the active fuel cladding. As a result, the focus of the BWR Core Heatup Code evaluation and the 1/4 scale test program has been and remains the characterization of the thermal environment produced by diffusion flames resulting from postulated recoverable degraded core accident scenarios.
Some results from the BWR Core Heatup Code are summarized in this submittal and will be discussed in this meeting as it relates to the proposed test matrix. An additional submittal and meeting is planned for August to provide the NRC with the opportunity to review the details of the BWR Core Heatup Code.
Information is also provided in Attachment 4 concerning the 1/4 Scale Test Program. This includes the status of construction for the facility, the objectives for the test program and the development of the test matrices for the shakedown tests, scoping tests, and production tests. HCOG believes that the completion of the planned test program will provide sufficient data to demonstrate equipment' survivability and resolve the diffusion flame issue.
Once again, it is the purpose of this submittal and meeting to discuss and resolve NRC concerns regarding the 1/4 scale test program so that the overall HCOG program may proceed toward conclusion in an orderly and expeditious manner.
' Should you have any questions, please do not hesitate to contact Mr. James R.
Haley of Mississippi Power & Light Company.
Ver truly yours,
/
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Sam H. Hobbs, Chairman Hydrogen Control Owners Group DBH/JRH:Im cc:
(See Next Page)
LAdb2
l HGN-018 Page 3
.cc:
Mr. Carl R. Stahle (w/a) (12)
-Hydrogen Control Program Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 Mr. Charles G. Tinkler (w/a) (2)
Containment Systems Branch Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. John Cummings, Project Manager (w/a)
Hydrogen Studies, Division 4441 Sandia National Laboratory Albuquerque, New Mexico. 87185 tAdb3
HGN-018 Page 4 bec: Mr. Erich Kant (IPC) (w/a)
Ms. Eileen Buzzelli (CEI) (w/a)
Mr. Erwin Zoch (GSU) (w/a)
Mr. John Hosler (EPRI) (w/a)
Mr. R. W. Evans (Enercon) (w/a)
Mr. M. Fuls (w/a)
Mr. B. Beasley (Bechtel) (w/a)
Mr. J. D. Richardson (TERA) (w/a)
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's Attachmint I HGN-018 s
HYDROGEN CONTROL OWNERS GROUP NRC MEETINC
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JULY 24-25, 1984 FACTORY MUTUAL RESEARCH CORPORATION NORWOOD, MASSACHUSETTS AGENDA July 24 8:00 - 8:15 Introduction and Meeting Objectives -
8:15'- 8:30 Overview of HCOG Program Objectives 8:30 - 9:15 Review of 1/4 Scale Test Program Status 9:15 - 10:45 Travel to Test Site 10:45 - 12:00 Tour of Test Facility 12:00 - 2:30 Lunch and Return to Factory Mutual 2:30 - 3:30 Open Discussion on Test Facility 3:30 - 5:00 1/4 Scale Test Matrix Development July 25
~8:00 - 8:30 Summary of July 24 Discussion 8:30 - 10:00 NRC Comments on HCOG Responses to RAIs 10:00 - 10:30 Application of Test Data 10:30 - 11:30 Concurrence on Test Matrix Approach 11:30 - 1:00 Lunch /NRC Caucus 1:00 - 2:30 Open Discussion 2:30 - 3:00 Summary of Meeting
.3:00 Adjourn tBdb1
Attachsent 2 HGN-018 HCOG-NRC MEETING OBJECTIVES 1.
REVIEW TE ROLE OF TE 1/4TH SCALE TEST IN TE OVERALL HCOG PROGRAM.
2.
REVIEW PROGRESS ON TE 1/4TH SCALE TEST PROGRAM.
3.
REACH AGREEMENT AS TO THE ADEQUACY OF TE PRESENT TEST APPROACH TO PROVIDE THE NECESSARY INFORMATION ON DIFFUSION FLAMES.
4.
DISCUSS THE DEVELOPMENT OF THE 1/4TH SCALE TEST MATRIX.
5.
BRIEFLY REVIEW USE OF TE BWR CORE HEATUP CODE OUTPUT TO DEVELOP TIME HISTORIES TO BE INPUT TO TE 1/4TH SCALE TEST.
6.
DISCUSS NRC ' COMMENTS ON TE HCOG RAI'S SUBMITTED ON APRIL 2,1984.
7.
DISCUSS APPLICATION OF THE TEST DATA.
8-REACH AGREEMENT ON TEST MATRIX APPROACH.
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m HGN-018 MEETING BASIS SUBMITTAL The Hydrogen Control Owners Group (HCOG) has undertaken the 1/4 scale test program to define the thermal environment due to the combustion of hydrogen as a diffusion flame resulting from a postulated degraded core scenario. The information contained in this attachment is provided to the NRC for review of the progress of the 1/4 scale test program, and to allow the NRC Staff an opportunity to comment on the 1/4 scale test matrix, so that HCOG and the NRC can reach concurrence on the tests to be conducted in the 1/4 scale test facility. HCOG believes that this test program will resolve the equipment survivability issues for steady diffusion flames.
1.0 1/4 Scale Test Program The scaling basis used to design the 1/4 Scale Hydrogen Combustion Test Facility was described in Attachment 1 to Reference 1. to Reference 1 provided a detailed report which summarized the design of the facility.
1.1 1/4 Scale Test Program Objectives The goals and objectives for the 1/4 scale test program were delineated in Attachment 1 to Reference 2, which provided responses by HCOG to a number of NRC requests for additional information (RAIs).
The objectives of the test program are:
(1) to obtain suf ficient data to determine the full scale thermal environment (gas temperatures, gas velocities, and radiant heat fluxes) produced by diffusive burning of hydrogen which may be released during postulated recoverable degraded core accidents. The thermal environments will then be applied analytically to computer models of the equipment which is required to survive a burn to assess their thermal response, and (2) to assess the hydrogen mixing which occurs in the wetwell and upper containment prior to and during the hydrogen combustion transient.
Equipment survivability has already been demonstrated for the equipment required to survive a series of hydrogen deflagrations. These survivability calculations were completed for a hydrogen release corresponding to the reaction of 75% of the active fuel cladding with a hydrogen release rate of 1 lbm/sec. This evaluation is conservative since the results from the 1/20 and 1/5 scale tests, as documented in the attachment to Reference 3, showed that deflagrations will only occur for hydrogen release rates less than ev 0.5 lbm/sec.
This flow rate has been designated as the threshold for establishing continuous diffusion flames.
tBdb2
Therefore, only postulated recoverable degraded core scenarios which result in hydrogen release rates that are predominantly above the threshold for establishing continuous diffusion flames will be evaluated as part of the 1/4 scale test program.
1.2 1/4 Scale Test Facility Status The primary enclosure for the test facility described in Reference 1 is nearing completion. This includes the outer tank, the inner tank and the major flow blockages. The exterior of the enclosure has not yet been painted.
Installation of insulation on the interior surfaces of the facility is in progress. Major equipment including the data acquisition computer, the steam boiler, the air compressor, and other miscellaneous equipment has been received on site.
The current schedule for the test program calls for initiation of the shakedown tests in September, 1984. Certain scoping tests which are described in Section 2.1 of this attachment are scheduled for October, 1984. The production testing is scheduled to follow in October and be completed by February, 1985.
1.3 1/4 Scale Facility Description The capabilities of the 1/4 scale facility have been previously identified in Reference 1.
Table 1-1 reiterates these capabilities and provides additional information on the capacities for the steam injection system and for the simulated containment unit coolers. Figures 1-1 through 1-4 show the actual configuration at each elevation for each of the containment designs. Figures 1-5 through 1-8 show the 1/4 scale facility simulation of each of the containment designs.
Figure 1-9 shows the structural design of the test facility.
The test facility consists of a 31' diameter steel tank with a height of 49'.
This tank contains a smaller tank which simulates the drywell volume. Airtight modules are installed in the outer test tank to simulate enclosed volumes in the containments. Floors in the test facility have been constructed in removable sections in order to test different plant configurations. The interior walls of the test facility have been covered with insulation. The lower portions of the external tank contains the simulated suppression pool. The test facility has been designed for the pressure range of -1.2 psig to 40 psig.
The major systems required for testing in the 1/4 scale facility include the simulated containment spray system, the simulated containment unit cooler system, the hydrogen injection system, the steam injection system, and the hydrogen ignition system. Each of these systems, with the exception of the simulated containment unit cooler system, is discussed in Reference 1.
The simulated containment unit cooler system is discussed in Reference 2.
' ".i03
Hydrogen will be injected into the test facility through either simulated safety relief valve (SRV) discharge spargers or through both the simulated SRV discharge spargers and the simulated LOCA vents. The design of the simulated SRV discharge sparger was summarized in Reference 1.
This design has been modified to ensure correct scaling for the inje ion of hydrogen or the simultaneous injection of steam and hydrogen.
Reference 1 states that the GMAC Model 7G glow plug will be used in the 1/4 scale test facility to initiate hydrogen combustion. A drawing which shows the mounting arrangement for the igniter in the test facility is included in Reference 1.
The igniters are located in the test facility to correspond approximately with the igniter locations for Mississippi Power & Light Company's Grand Gulf Nuclear Station. The criteria used by Grand Gulf Nuclear Station for locating their igniters is representative of the criteria used by the other HCOG plants. Also, the 1/20th scale tests demonstrated that the exact igniter location was not a significant factor in determining diffussion flame environment. Therefore, the igniters in the facility are representative of all of the HCOG plants.
Figure 1-10 shows the test site layout. This figure shows the locations for the test facility, the steam boiler the hydrogen supply trailer, the control trailer, and access to the site.
2.0 Discussion of Test Plan The testing will be conducted in three phases. Phase one will involve facility shakedown and check out.
These tests will verify operation of facility systems and instruments.
Phase two will be scoping tests which will confirm the threshold for establishing continuous diffusion flames and will evaluate parameters to be i
considered for production tests. Phase three will entail production tests to define the thermal environment produced by postulated degraded core accident scenarios.
2.1 Phase II - Scoping Tests These tests will be used to confirm important parameters which affect definition of thermal environments from hydrogen release rate histories. Turndown tests will be performed to confirm the threshold for establishing continuous diffusion flames. Tests will also be performed to assess the effect of concurrent steam and hydrogen injection on the threshold for establishing continuous diffusica flames and the thermal environment'at the HCU floor. Tests will also be performed to assess the effect of simultaneously discharging hydrogen through the LOCA vents and through the spargers on the thermal environment at the HCU floor. A test will be conducted to verify that the slight configuration difference between the s'
Perry Nuclear Plant and the Grand Gulf Nuclear Station will have an insignificant ef fect on the thermal environment.
(See tBdb4
L f
Figures 1-5 and 1-6.)
These effects will be assessed by comparing the temperatures measured in this scoping test with the appropriate Grand Gulf Production tests.
2.2 Phase III - Production Tests After the shakedown and scoping tests have been completed, the production tests will be conducted. They are listed in a tentative test matrix see Table 2-1, and will test the effect of using different hydrogen release histories, injecting the hydrogen through the simulated ADS valves as well as an additional stuck open relief valve, the effect of no containment sprays, and the different plant configurations.
The development at the hydrogen release histories is discussed in Section 3.0.
The proposed production test matrix is based on the assumption that the following will be verified.
If the scoping tests do not verify these assumptions, the production test matrix will be revised accordingly. Assumptions:
(1) Scoping tests confirm that the effects of hydrogen released through the spargers is limiting when compared to releasing the hydrogen through both the spargers and the LOCA vents.
(2) Scoping tests will confirm that the effects of injecting steam simultaneously with hydrogen will be negligible when compared with that of the injection of hydrogen alone.
(3) Scoping tests will confirm that the Perry configuration is similar enough to that of Grand Gulf such that the data obtained is applicable for both Grand Gulf and Perry.
(4) Scoping tests will confirm that with the igniters activated, there will be no pocketing of high concentrations of hydrogen.
As shown in Table 2-1, a set of tests has been included for the plant configurations which require simulation. The test configuration for each of the configurations will be as shown in Figures 1-5, 1-7, and 1-8.
These figures show how the significant geometry differences will be accounted for.
For each of the tested configurations, two basic sets of tests with three release histories will be run. Each set simulates the appropriate number of ADS spargers for the test configuration plus a stuck open relief valve sparger. This assumption is consistent with the Emergency Procedures Guidelines which instructs the operator to open the I
appropriate number of safety relief valves and depressurize the reactor for the scenario postulated. The difference between the two sets is in the location of the simulated stuck tBdb5
l open relief valve (SORV) sparger.
In each case the location is selected to represent what is anticipated to be the chimneys with the most severe thermal environments.
Besides maximizing the thermal environment, such tests will provide results on the effect of the location of the stuck open relief valve. Tests 1 and 2 for the Grand Gulf configuration will be conducted in order to quantify the effect of sprays.
For each of the SORV sparger locations for a specific test configuration, three separate tests will be run with different hydrogen release histories. For the scenarios which produce hydrogen generation rates above the diffusion flame threshold, the hydrogen generation rate and the total amount of hydrogen produced are insensitive to the exact nature of the transient or accident which leads to conditions capable of producing hydrogen. The hydrogen production for a postulated degraded core accident scenario including rate, duration and total hydrogen production is maximized by the timing for initiation of reflood and the reflood rate. As the reflood flow rate increases, the peak hydrogen generation rate increases.
However, both the duration of the hydrogen production and the total amount of hydrogen produced by the scenario decrease as the reflood flow rate is increased.
After evaluating various reflood cases, the HCOG will select three hydrogen release histories to be included in the 1/4 scale test matrix. The 1/4 scale test matrix identifies these three histories as Cases A, B and C.
These three histories will be selected to be representative of the range of reflood rate cases with hydrogen generation rates above the diffusion flame threshold. The range of hydrogen generation rates is characterized by long duration with low peak hydrogen generation rates to short duration with high hydrogen generation rates. This will provide a range of release histories for analyzing the thermal response of small and large equipment.
The hydrogen release histories as generated by the BWR Heatup Code shows the hydrogen generation rates continuously varying with time. For the production tests these rates will be simulated in.032 kg/sec increments which have been
=
mathematically calculated from the BWR Heatup Core Output. An example of this method is shown in Figure 3-1.
3.0 Development of the Hydrogen Release Time Histories
3.1 Background
The hydrogen release time histories are being developed for use in the 1/4 scale test matrix. The development of these time histories have considered postulated recoverable degraded core accident scenarios which result in hydrogen release rates predominantly above the continuous diffusion flame threshold.
Due to the exothermic zirconium-steam reactions which produce tBdb6
these hydrogen release rates there is significant heating of the core. These scenarios cannot result in a total hydrogen production equivalent to oxidizing 75% of the active fuel cladding and still maintain a recoverable core.
Scenarios which could potentially result in oxidation equivalent to 75% of the active fuel cladding with a recoverable core would result in hydrogen release rates below the threshold for continuous diffusion flames. Therefore, these scenarios are outside the scope of the 1/4 scale test program.
The types of postulated scenarios which lead to hydrogen release rates above the threshold for continuous flames have a common sequence of events. The postulated scenarios begin with any transient or accident which results in relatively fast vessel depressurization, loss of significant vessel inventory, and inability to inject makeup to the vessel. Very little hydrogen would be produced during the initial depressurization of the vessel and not until the steam cooling becomes ineffective. The fuel is assumed to be 3/4 uncovered and would begin heatup before any appreciable oxide layer would be formed on the cladding surface. This type of scenario is conservative because formation of an oxide layer would reduce the amount of oxidation which could occur and therefore produces less hydrogen. This scenario is consistent with the emergency procedure guideline because if no injection flow is available to provide makeup to the vessel, the emergency procedure guideline requires that the vessel be depressurized. This will be accomplished by opening at least seven safety relief valves and would occur before any hydrogen is generated. This also assures that multiple hydrogen release points will be available.
No reflood is assumed to occur as core heatup progresses until hydrogen production in the vessel becomes limited by the availability of additional steam to feed the oxidation. This point in an unmitigated scenario represents the peak hydrogen generation rate. Reflood is then assumed to be initiated which produces a maximized hydrogen generation rate.
For such scenarios, the hydrogen generation rate and the total amount of hydrogen produced are insensitive to the exact nature of the transient or accident which leads to conditions capable of producing hydrogen. The hydrogen production for a postulated degraded core accident scenario including rate, duration and total hydrogen production is maximized by the timing for initiation of reflood and the reflood rate. As the reflood flow rate increases, the peak hydrogen generation rate increases.
However, both the duration of the hydrogen production and the total amount of hydrogen produced by the scenario decrease as the reflood flow rate is increased.
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3.2 BWR Core Heatup Code Modeling Although the BWR Core Heatup Code has been utilized in calculating hydrogen release histories for degraded core accidents, the details of this code including modeling capability, assumptions and core nodalization will be presented to the Nuclear Regulatory Commission in a future submittal from HCOG. However, to provide a preliminary basis for assessing the hydrogen release histories predicted by the BWR Core Heatup Code, some details on code modeling are included in this submittal.
The BWR Core Heatup Code provides detailed modeling of the core components and geometry as well as the thermal hydraulic conditions in the reactor pressure vessel. The core geometry is assumed to remain intact during core heatup which maximizes the surface area available for oxidation.
Zircaloy oxidation is modeled with the oxidation terminated at high temperature, in the oxidizing nodes. The oxidation is assumed to be terminated because in reality the cladding has melted and would relocate forming a geometry which would provide much less surface for oxidation. The code has been modified to account for oxidation of the exterior fuel channel walls and control blade oxidation which occurs due to steaming in the bypass region. A conservative assumption presently being used in the code is that the nodes above and below a melted node are still available for oxidation.
In-fact, the flow of steam to these nodes would be restricted if not totally shutoff and
,3 would stop all hydrogen production in that fuel channel.
Termination of Zircaloy oxidation at high temperatures has been observed in numerous experimental investigations.
Experiments conducted by Hagen at KFK included simulation of fuel rod oxidation and meltdown. These experiments showed that oxidation was terminated in the temperature range of 1'
2200-2300 K.
The Power Burst Facility test included a test (SFD-1) which showed that the oxidation was terminated in the temperature range of 2300-2400 K.
Finally results from experiments performed by H. Chung at Argonne National Labs have been extrapolated and indicate that oxidation terminates at a temperature of approximately 2200 K.
i The apparent cause of oxidation termination is that the Zircaloy, which melts at a temperature of 2173 K, interacts with the uranium oxide fuel. The liquid zirconium-uranuim oxide mixture slumps into a geometry which essentially eliminates the surface area available for oxidation. The physical state of the core is assumed to be altered when the oxidation cutoff temperature is reached. In the BWR Core Heatup Code, the oxidation termination point is assumed to be 2400 *K.
When this temperature is reached in a given node, the oxidation from this node is terminated irreversably based upon the assumption that the geometry for the node is no longer intact and available for further oxidation.
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Based upon the previous discussion, the scenarios evaluated with the BWR Core Heatup Code have several common elements.
Therefore, the BWR Core Heatup Code has been modeled to reflect the following assumptions:
(1) The reactor pressure vessel is depressurized to two atmospheres.
(2) The vessel water level is asaumed to be lowered until at least 3/4 of the core is uncovered.
(3) The fuel is free of any oxide buildup until after the core is 3/4 uncovered.
(4)- The maximum hydrogen generation rate is reached before any additional water is injected.
(5) The injection occurs at constant reflood rates.
(6)- The core geometry analytically remains intact which maximizes the surface area available for oxidation.
(7) The Zircaloy oxidation terminates at 2400
'K.
3.3 BWR Core Heatup Code Results The BWR Core Heatup Code is being used to evaluate different reflood rates. Plots (See Figures 2-1 thru 2-20) are provided for reflood rate cases (0, 80, 150, 300, and 500 gpm) to show the core water inlet and steam outlet flows, the hydrogen generation rate, the Zircaloy oxidation fraction (NWR H,
fraction) and the active core Zircaloy melt fraction. Reflood flow rates above 5000 gpm are more probable since more systems are available for recovery with flow rates in this region including all of the ECCS systems. Reflood transients with injection flow equal to 5000 gpm or above produce very little hydrogen because the core is rapidly quenched. Total hydrogen production is less than.200 pounds.
The 0 gpm reflood case produces peak hydrogen release rates at or just below the threshold for diffusion flames and therefore cannot be used for evaluating the effects of diffusion flames in the 1/4 scale test facility. However, it is used for determing the peak hydrogen generation rate for timing the start of'reflood as described previously and for comparison against the reflood cases. The 80, 150, and 300 gpm reflood cases have hydrogen production rates above the diffusion flame threshold; however, these injection rates are inadequate to prevent excessive clad melt (greater than 50% melt) before the core is cooled.
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4.0 Application of 1/4 Scale Test Data As shown in the revised instrumentation plan included in Attachment 5, a significant amount of instrumentation has been included in the 1/4 scale test facility in order.to measure gas velocities, temperatures, and radiant heat fluxes. From these measurements the thermal environment will be defined for the areas with equipment of concern. The' test instrumentation has been strategically located based upon the experience gained'in the 1/20th scale tests.
The convective heat transfer coefficient for a specific piece of equipment will be determined using standard methods with consideration of the flow Reynolds Number (based on the measured and appropriately scaled temperatures and flow velocities) and the geometric characteristics of the equipment. The radiative flux to equipment will be determined using Hottel charts with consideration of the view factor from the radiation source (flames and/or hot metal surfaces) to the equipment and a gas / metal source temperature based on measured data. The equipment thermal response will be determined by applying the thermal environment to the equipment computer models.
The approaches described above for determining convective and radiative heat loads will be used to predict the full-scale thermal environment from the 1/4 scale test data. This same nethod and thermal environment will be analytically applied to models of the sphere calorimeters included in the test instrumentation. The predicted thermal response of these calorimeters will be then compared with the observed response of the calorimeter.
It is expected that this will demonstrate that the standard heat transfer methods are adequate for definition of a thermal enviroment based on measured data.
REtERENCES 1.
-Letter from HCOG to NRC, HGN-012, dated August 12, 1983.
2.
Letter from HCOG to NRC, HGN-016, dated April 2, 1984.
3.
Letter from HCOG to NRC, HGN-014, date February 9,1984.
tBdb10
Table 1 - 1 TEST FACILITY CAPABILITIES Simulation of plant geometry variations.
Maximum / minimum pressure: 40/-1.2 psig.
Controllable H release rate:
0-4 lb/s (full-scale).
2 Total H release / test: 2500 lbs (full-scale).
2 H release through spargers and/or LOCA vents.
2 Steam and air purging.
Containment sprays at 3800 gpm (full-scale).
Pool temperature range 70-185*F.
Steam injection flow rate: 0-159 lbs/see @ 100 psi (full-scale).-
Containment unit cooler heat removal rate:
23 M Btu /hr (full-scale).
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iy a d-TABLE 2-1 .E TENTATIVE PRODUCTION TEST MATRIX 1 A!. ( ^I Th, t' 'tp v:. TEST H / STEAM NO./ LOCATION SPRAYS OR ' CONFIGURATION -NO-RELEASE HSITORY[g) OF ACTIVE SPARGERS(2) COOLERS SIMULATED 9 'T" Grand Gulf / Perry (4) 1' "A- 9 (8 ADS-+ 312*) 0FF 7 "A" ON i /J ~ "B"
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"C" 1V \\ 5 "A" 9 (8 ADS + 32*) 6 ;I '"B" se 't 7' "C" l E II 1 8 '"A" 8.(7 AGS + 36*)(3) Clinton 4M. 9 "B" ^ "C" 1F k 10 8 (7 ADS + 323*)(3) 11 "A" ~c 12 "B" 1F II 13 "C" 14 "A" 8 (7 ADS + 28*)(3) River Bend 15- "B" 16 "C" if y* ;;t 17 "A" 8 (7 ADS + 323 )(3) ~ 18' "B" p 3r 1 r u4 _ 19 "C" y
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[. ; O 4 NOTES FOR TABLE 2-I (I) TE HYDROGEN / STEAM RELEASE HISTORIES ("A," "B" AND "C") SHOWN IN TE ATTACHMENT ARE FOR GRAND GULF. TE CORRESPONDING RELEASE HISTORIES FOR TE OTHER PLANTS WILL BE DERIVED BY MULTIPLYING THOSE HISTORIES BY TE FOLLOWING RATIO: NO. OF FUEL BUNDLES IN THE OTER PLANT CORE NO. OF FUEL BUNDLES IN GRAND GULF'S CORE (2) TE SPARGER LOCATIONS IN TE TEST FACILITY ARE EVENLY SPACED AT-IS' INTERVALS. TE TEST SPARGER CLOSEST TO TE ACTUAL LOCATION OF TE ACTIVE PLANT SPARGER WILL BE UTILIZED. (3) SINCE THIS PLANT IS BEING SIMULATED AS A " MIRROR IMAGE" 0F TE TEST FACILITY, TE TESTED LOCATIONS WILL BE IN " MIRROR IMAGE" POSITIONS TO THOSE SHOWN. (4) A TEST WILL BE CONDUCTED AS PART OF TE SCOPING TESTS TO VERIFY THAT TE SLIGHT DIFFERENCE IN CONFIGURATION BETWEEN GRAND GULF AND -PERRY WILL HAVE MINIMAL EFFECT ON THE THERMAL ENVIRONMENT AND TEREFORE, TESE PRODUCTION TESTS WILL BE APPLICABLE TO BOTH GRAND GULF AND PERRY. O l-l i I l tBdbl4 i t.-
BNR-6 DEGRADED CORE ACCIDENT HYDROGEN PRODUCTION 4.5000 HCOG01 3/4IJNCOVER ED i i CORE D GPii REFLO DD R=STE RM 00TLI.T FLOW 4.0000 B= CORE INLET FLON 3.5000 3.0000 Eo 2.5000 A Ee $ 2.0000 e v. 3.5000 1.0000 pt { I 0.5000 N A / 0.0000 l t 2000. 3000. 4000. 5000. 6000. I TIME,' 5E COND S l NATER INLET-STERM OUTLET FLONS l o FIGURE 2-1 i l -__-_--._-_.--,-,_.----.--,-.---.,,.----,-,-,,.n-n.,.,,-,,,--n-.---r- - -, ~.. - - - -,,,. -,.
i l BWR-6 DEGRADED CORE RCC] DENT HTORCGEN PRODUCTION 0.2000 r HCOGP 1 3/4IJNCOVER ED l CORE O GPli REFLO DD j 0.1800 0.1600 1 l l C.1400 i l l E 0.1200 o ( u E s I E 0.2000 E c f; a { l i sp 0.0800 f / D.0600 I \\ 0.0400 1 0.0200 t 0.0000 f
- 5000, 6000.
2000. 3000. 4000. ~ ~ TIME, SECONDS HYDROGEN GENERRTION RATE i FIGURE 2-2 i ,-----n,,,,_y -y --,,.,_.--,_e m,,.w--, +-
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i I i BWR-6 DEGRRDED CORE RCCIDENT HYDROGEN PRODUCTION i 0.2200 HC0Gll1 3/4 UNCOVER ED l CORE O GPii REFLO DD D.2000 7 i l' 0.1800 l l l 0.1600 i \\ 0.1400 l 3 l g 0.1200 { l e [ 0.1000 i i 0.0800 I l 0.0600 l t l l 0.0400 D.0200 l r 0.0000 ) i 2000. 3000. 4000. 5000. 6000. TIME, SECONDS ZIRCRLOT OXIDATION FRRCTISM t l FIGURE 2-3 C I I i f
.......a l BNR-6 DEGRADED CORE ACCIDENT HYDROGEN PRODUCTION 0.9000 I HCOGUI 3/4 UNCOVER ED CORE D GP.i REFLO 3D 0.8000 f 0.7000 + i 0.6000 / f I 0.5000 l z E Ee t u. D.4000 i I D.3000 i l i / 0.2000 i 4 l 0.1000 l i D.0000 1 2000. 3000. 4000. 5000. 6000. TIME, SECONDS RCTIVE CORE ZIRCALOY MELT FRRCTION FIGURE 2-4 { e l F i
I a I l SWP-S DEGP.RDED C08.E ACCIDEhi HYOP.00F.h PP.DDUJi!Ch f S.0003 <~ = HCOGP2 3/4 IFC0 VERE 6 I CORE SD GPit REFLODD 'l a" ' 4.5000 1 9=COR E INLET F.ON i ( plL 4.0003 / / 1( i 3.5000 s / 1 i I z 3.0003 u O m i t {
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r L t l l l l SdR-9 Dr.0Rn0E0 C3rf. ROCIDENT MirROGEN PP.0000T!ch l l 0.3200 f HCOGP2 3/4 UNC0VEh!i j LORE BD GF11 HLF 0 JI I l 0.2900 5 / 0.2400 i j I ( l \\ i 0.2000 i l n 3 z l o r n l U J: l s Yl [ E 0.1600 f a 1 i r N L L*s (n I Y N O.1200 l b ( 4 I f \\ 0.0800 l \\ i 4 l i \\ 0.0400 f s s 0.0000 l 2000. 3000. 4000. 6000. 6030. TIME.5ECONDS HYDROGEN GENERRTION RRTE FIGURE 2-6 l i t i s: - - ..9 ..e.....
l ,1,0.30 i / SWP.-6 Dr.0P.OEG r.DPr RCCIDENT HTDP.0 GEN PR0)Ud!Dh l / HCOGh2 3/4 IJNC3VER ED / c.2003 b u nf. t y er. r.i.f a.uLu ~ i { } i n.24co t D.2000 I f l / l I .u. C.16C0 I ~ I I l 2 l E h 0.1200 f i I C. Ceca I l r I 0.0400 / 0.0033 l t 2030. 3000. 4000. 5200. 50:3. TIRE. SECONDS IIR;R'_0T CX!DATICN FRRCTICh i l FIGURE 2-7 -1 ~ m _. - -.. = 0... o oc,
) 9WR-S DEGP.RCED CDP.E RCCIDENT HTDP.0 GEN FP.00Uti!.1N \\ ,.,o3e NCOG'.'2 L 4 UNCOVECCD CORE 83 GPit REFLCDD s'~ r ~-W D.1000 i i D.6000 j I D. BOO 3 'z .u ( y 0.4000 l z m I l D.30SD I i l C.2G03 / \\ \\ l 2.3000 1 8 I D.0000 l 2933. 2002. 40D3. 5003. 5000. t i!Mi.SECSND5 ( i RCTIVE CSP.E ZIRCRLOT MELT FP.R0i!SN l l [, FIGURE 2-8 i L i
F 1 BdP.-3 DE'JP.RDF.D 20P.E R;;!5r.hi !:YDROGEN Pp.0Ddci!Ch l ^ l 13.0000 l [ H"00.'3 3/4 UNc3VEP. r. 0 CCRE 150 P1 REFLC 20 9.C000 a-3 ?E" at'-- '! y au! - l 8=COR E INLET F.ON i t.0000 i 7.C000 \\ [ ts x G.CCO3 r P; l .a n ] i f Q i.0003 I et T-i e,3 p =' \\ / [: 4.0033 3.0000 t n r \\ i I 2.2000 ( I i 1.0000 l P f '\\[ C.C003 2000. 3C00. 40CO. SC00. 800C. l TIME.SiCONDS NRTEP. IN!.ET-STERM OUTLET FLOWS i FIGURE 2-9 l l f .,---...------..._..,,,-._-..._-,..-.-..,,-.,.-...,.--,,,,,,__,,_.,-.-,-,._---.,.n,
i l i BWP-d DE3P.R3E') CO*.E n:CIDFKT HYDP.CGEN PP.00U.:ilch C.5000 1l' HC3GP3 3/4ljN T.OV C#. '.3 l COBE 150 GP1 P.EF t.0 00 l 0.4500 I l C.4000 l l C.3503 E 0.3000 ) Id 0 i { t \\ Q 0.2s00 c: i E t i u s e.ma e D.1500 i D.1000 t 0.0600 c.0000 2000. 3CO3. 4000. 50C0. $000. [ t TIME.5E00Kn3 HYJP.0 GEN GENERRTION fLRTE FIGURE 2-10 i L f L t i (
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i l WCOGN 3/4 IJNCOVERED CORE 300 C-Plt RF.F10 DD 10.8300 "-5?' "" !"'L! Ty t 9-CORE INLET F.0W t 16.0000 i f I4.(1800 oa 12.0000 i "s f E 10.0000 l l a 5 o y t.0000 f I I i i 6.0000 i i I I f 4.0000 f i f \\ 2.0000 l 8.0000 I t i 2000. 3000. 4000. 5000. 6000. TINE. SECONDS l NRTER INLET-STERM OUTLET FLOWS l a FIGURE 2-13 f 1 ? t
~ t i i l t BNP.-6 DEGRADED CCRE RCCIDEhT HYDROGEN PRODUCT!ch f C.900C HCOC44 3/4 UNCOVER ED CORE 300 GPM REFLoDD 3 . t l O.7000 [ i l C.seco f o I z I o l 8 0.5000 i en i ti E l E i C.4C00 I d v r f C.3000 i i l 0.2000 \\ D.3000 i \\ l i ^ 0.o000 l anoo. 3000. 4eco. soon. sooo. 1!ME.5ECONDS l HYDP.0 GEN GENEP.RT!CN RRTE [ t FIGURE 2-14 l i i b I .--..-.4 .__.-._m.. . _ _ _, -, _, _.. -. -.. - _ _,., _ _ _..,. ~. _, _ _ _. _ _ -.. _, _.... _
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[ / \\ 0.2400 l 0.2200 l \\ f 0.2000 C.1800 0.1600 l E l 8 { y 0.1400 l e I e t C.1200 [ l i i f 0.1000 I 0.0800 l i D.0500 I 0.0400 / ? A C.0200 C.0000 i 2000. 3000. 4000. 6000. 8000. t 11ME. SECONDS i ZIRCRLOY OXIDRTION FRRCTION i t FIGURE 2-15 l l i i t . _ _ _.. _., _. _ _... -,,... _.. -. _. _ _ _. - - _ ~ _ _ _. -. _ -. _ _. _ _ _ _ _ _ _ _ _ _.. _. _ _ - _ _ _ _. _. _ _ - _ _ _ _ _ _ __ _ _ _ _
l i l BdP.-6 DEGRADED CCP.E ACCIDENT HYDRCCEN PRCDUC1!Ch l 0.8000 NC00:'4 3/4 UNCOVERED 0.5500 '-. CORE 300 Gali REFl.0]D i \\ 0.5000 g r 0.4500 C.4000 { \\ 0.3500 0 0.3000 ~ 2 e ) e k D.2500 0.2000 l l 0.A500 0.1C00 l h 0.0500 C.0000 i 2000. 3000. 4000. 5000. GC00. TIME.SECCNDS ET!bE CCRE ZIP.CRLC7 MELT FRRCTICN l FIGURE 2-16 r t ? -..
. ~. f r ) ENR-S DEGRADE 0 CO8tE RCCIDENT HYDROGEN PRODUCT!ON l 320. i 11C0007 3/4 In#VER ED ..l LORE 50000fM RjfLU DD 4=57E RM cuiLI.T FI OW 220. S-CS* : P". E ! '.S? i: 240. I i 200. o z o u i na t en N t I E 160. e i c r o U I t 320. I I I l 80. l + k; 40. \\ l s l C. I I f l 2000. 3000. 4000.
- s000.
6000. j 11ME. SECONDS NRTER INLET-SiERM OUTLET FLONS I FIGURE 2-17 i f t f l t I l l.~
EWR-6 CEGRRDED C0p.E ACCIDENT HYDROGEN PRODUCT!Ch HCOGP7 3/4 1*C0VER F.D i 3.4003 g, g g g, 3gggggn g 7,7, y ;g i t 1.2000 t 1 1.0000 0 l 2= i u W j 0.2000 m i r aeo u 1
- g 0.0000 r
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- 6000, f
l I TIME. SECONDS M7DROGEN GENERRTION RATE l FIGURE 2-18 t j
t b Bd.9-6 DEGRRDED CORE RCCIDENT HYDP.0C-EN PP.00UCTICN 0.csso t HC00' ? 3/4 l 'NC VEP. ED i j 0.0500 '"'E CC m ov rri e m f. I i D.0550
- .l t
f . D.0500 l C.0450 I 0.0400 E 0.0350 ( i L' t u E 0.0300 f i k i e I 0.02S0 I l i t 0.0200 i / 0.0150 l 0.0100 l l i 0.0050 i i I C.0000 r[ 2000. 3000. 4003. 3000. 6000. TIME. SECONDS IIRCRI.0T DXIDRTION FRRCTIDN ,.j i i FIGURE 2-19 i t t I y sy-wwii---v-- e gr-----ne-y- -w t mgest e ---e-+---ww-w---w--vv- ,-+--3 a--e n------w.ww--ww--mmew-e-ww-*--w- + + - -w-e-re---e----e-ew--r--------
_ u. ._ 2 _ I i i l l l i BNR-6 DEGRADED CORE ROC! DENT HYDROJEN P8.00UCTION l l D.0900 i j HC00l '73/4 UN00VER F.D CORE 50000-Pit REFLO 23 .l i l i O.0803 i l 0.0700 l i i f 0.0000 I 0.0500 t z .o. r Wu em c. 0.0400 l 0.0300 l l t l 0.0200 i f O.0100 I r I 0.0000 2000. 3000. 4C03. 8a000. 6000. TIRE.$ECONDS f t RCT!VE 00RE IIRORLOY MELT FRRCTION l 1 i FIGURE 2-20 f i-t l l t
t ? l l i BWR-6 DEGRADED CORE RCCIDENT HYDROGEN PRODUCT!cN C.secc l t HCOGL4 3/4 UNCOVERED CORE 300 GP! t REFLODD 0.8000 C.7000' l p r 0.6000' o t o d O.5000 m t E E e i C.4CCO j d l' l i v. t i l C.3000 i 0.2000' l A ? t F y j 0.3000' i \\ T
- A I
A f 3 O.0000 .2000. 3000. 4000. 5000. 6000. f 11ME. SECONDS HYDROGEN GENERATION RATE i FIGURE 3-1 NOTE: Example of how the hydrogen generation rate will be simu-lated in the test facility. r
HGN-018 b2 R-v= /h b Dat ee : / i PROPOSED INSTRUMENTATION LAYOUT FOR, EPRI 1/4-SCALE FACILITY o============================================================================= t
SUMMARY
@YM2DL EXTENDED DESCRIPTION
- OF LOCATIONS
=
= _ _
M VIDE o c.AME itAS E o Gas Thermocouples. Type K, mostly 3-5 mil Wire. 118 I ( Surface Thermocouples (on~ Insulation or Grating). 12 Type K, mostly 32 mil Wire. l T Total-Energy, Wide-Angle Radiometers. 5 V Gardon-Type Total Heat Flux Gages. 4 I O Sphere Calorimeters. lo () Bidirectional Velocity Probes, Mounted Horizontally. lo (8 multiplexed, 2 continuous). O Bidirectional Velocity Probes, Mounted Vertically. 16 (all multiplexed) O Continuous Gas Analysis (Hydrogen and Oxygen). 3 Multiplexed Gas Analysis Lines 18 l (Hydrogen, Oxygen and Water Vapor). b Totg1 Pressure. 2 Other Temperatures (pool water E23, hydrogen E13, steam E13, 7 water spray E13, igniter [13, instrumentation area [13) Flow Rates (steam E23, water spray E13). 3 e_-- l TOTALS 213 NOTES:
- 1) Location of active spargers indicated in plan view (X) and in elevation (T ).
- 2) Event lines used for control purposes not shown in the list.
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