ML20112B745

Documents Final Selection of Hydrogen Release Histories for Use in 1/4-scale Test Program,Based on Discussions Which Have Occurred Since Aug 1984 Meeting.Final 1/4-scale Test Matrix Also Documented.Revised Final Test Matrix Encl

ML20112B745
Person / Time
Site:	Perry, Grand Gulf, River Bend, Clinton, 05000000
Issue date:	03/13/1985
From:	Hobbs S HYDROGEN CONTROL OWNERS GROUP
To:	Bernero R Office of Nuclear Reactor Regulation
References
HGN-031, HGN-31, NUDOCS 8503190301
Download: ML20112B745 (20)

Text

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l MARK 111 CONTAINMENT HYDROGEN CONTROL OWNERS GROUP som a. nodes. csoi, mon c/a Mississippi Power and Light e P.O. Dox 164] e Jackson. Mississippi 09205 601 969 2458 March 13,1985 HGN-031 U. S. Nuclear Regulatory Comnission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. Robert Bernero

Subject:

Hydrogen Release Histories and Test Matrix for 1/4 Scale Test Program

Reference:

1. Letter HGN-018 from Mr. S. H. Hobbs to Mr. H. R. Denton dated July 6,1984

Dear Mr. Bernero:

h reference letter transmitted hydrogen release histories calculated with the BWR Core Heatup Code for several assumed reactor pressure vessel reflood flow rates. 'Ihe letter identified all the principle assumptions used in calculating the hydrogen release histories. Finally the reference letter transmitted a proposed test matrix for the 1/4 scale test program.

At the time the reference letter was subnitted, the Hydrogen Control Owners Group (HCDG) had not selectal the hydrogen release histories which would be useri in the 1/4 scale test program. h proposed draft test matrix l

indicated that three hydrogen release histories would be

used in testing and these histories were designated as release histories "A", "B" and "C" for purposes of l

identification. The hydrogen release histories which l would be used for testing were selected in late July arxl early August, 1984. IKDG identified the selected hydrogen release histories during a meeting with members of your staff on August 28 and 29,1984. HCDG proposed using a hydrogen release history produced by l

l 8503190301 850313 g.2 I 4 1 fpd 8 PDR TOPRP ENVGENE, -

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Page 2 of 4 IIGN-031 reflooding the reactor pressure vessel with a flow of 300 CPM initiated at 2600 seconds after scram as release history "A". Release aistory "B" was proposed as the release history produced by reflooding the reactor pressure vessel with a high flow rate emergency core cooling system. 'Ihis release history did not include the hydrogen production during rapid quenching of the core since this amount of hydrogen did not represent a significant fraction of the total arrount of hydrogen produced. Hyde.ufsi release history "C" correspxided to a non-mechanistic constant flow rate of 0.1 lbn/sec until hydrogen production equaled 75% MRR.

Since the August 1984 meeting between IKDG and members of your staff, numerous additional discussions have occurred between your staff and HOOG. These additional discussions have related to the goals and objectives of the 1/4 scale test program and questions from your staff on the test program, and use of the BWR Core Heatup Code. The purpose of this letter is to document the final selection of hydrogen release histories for use in the 1/4 scale test program based on the discussions which have occurred since the August 1984 meeting between msnbers of your staff and IKDG. This letter also h=nts H00G's final 1/4 scale test matrix.

Attachment One to this letter is a final test matrix for the 1/4 scale test program. 'Ihis test matrix has been revised from the August 1984 proposed test matrix to include an additional scoping test which will evaluate the effect of increasing the blockage fraction present in the 1/4 scale test facility. 'Ihe hydrogen release histories which will be used for the testing are still identified in the test matrix as release histories "A", "B" and "C".

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. ,, i Page 3 of 4 H W-031~ L

. I, Attachment 'No to this letter identifies the finalized '

hyi v ,=i release histories which will be used in the 1/4 scale test facility. h asstanptions used in calculating  ;

the hyiv,=i release histories with the BWR Core Heatup 7 Code are discussed in Attachment Two. Finally, i justification for the selection of these hyivgan release histories is included in Attachment Two. Graphs showing .

the hya vyisi and steam release rates as a function of time

-.and the active core zircaloy inventory melt fraction as a- -

function of time for release histories "A" and "B" are included for your review.-

. IEDG requests that the staff review the information in ,

l Attachments One and Two to this letter and indicate that- ,

the pW hy&mi release histories and p.vrated test matrix will be acceptable to the NRC staff by March 29, i 1985. h review schedule is e ry to support the ,

IKDG's schedule for testing in the 1/4 scale test

. facility.

This submittal was campiled by HODG frczn the best information available for submittal to the Nuclear ReJ0Jatory Consnission. 'Ihe submittal is believed to be 4 S$lete and accurate, but it is not subnitted on any M specific plant docket. h information contained in this letter and its attachments should not be used for j ' evaluation of any specific plant unless the information

has been endorsed by the appropriate member utility.

HOOG members may individually reference this letter in r whole or in part as being applicable to their specific i plant. ,

HOOG appreciates the NRC staff's support to date in '

, maintaining ambitious schedules for review. 'Ihe H00G wishes to express our appreciation in advance for supporting the short review schedule requested for this letter and attached information.  :

1 Sincerely, b N h~

s. H. Hoees Chairman, HOOG 4

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b Page 4 of 4 HGN-031 SHH/jlw Attachment ,

oc: Mr. Carl R. Stahle H A vvai Control Program Mamger U. S. Nuclear Regulatory Otmunission Office of Nuclear Reactor Regulation Washington, D. C., 20555 Mr. Charles G. Tinkler Containment Systems Branch U. S. Nuclear Regulatory Otmunission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Mr. John Claunings, Project Manager H A vvai Studies, Division 4441 Sandia National Tahnratory Alhrp=rque, PM 87185 d

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- - i is ATTAONENP ONE 3

'IO HGi-031 [

1/4 Scale Test Matrix Table 1 Js the Hydrogen Control Owners C40up's final scoping test matrix. HAvf=1 c release history "A" is identified in '

Attachment Two to this letter. 'Ihe purpose of each scoping +

test is discussed below. r A series of three identical tests (S-1 through S-3) using ,

hydrogen release history "A" calculated by the BWR Core  ;

Heatup Code will be performed to assess the digrue of  !

repeatability of the themal environment prMari during the i tests. 'Ihese tests will provide a basis for evaluating the '

statistical significan of variations nhaarved during other  !

scoping tests. h tests will also provide a basis for l determining the uncertainty in the themal environment data  !

obtained during production testing. i Test S-4 will be perfomed to confirm the threshold flowrate "

at which steady diffusion flames will be established. h test will be performed by establishing a continuous flame through 9 spargers with an initial h A vy=i flowrate of 1 l lbn/sec full scale which is well above the threshold. 'Ihe hy & vy=i flowrate will then be decraaaai in steps until continuous diffusion flames can no longer be maintained. '

h effect of concurrent steam and hydrocyn release will be determined in test S-5 bysimultaneously injecting the t

hWuf=> release history used in the repeatability test with .

the steam release history predicted by the BWR Core Heatup Code. If concurrent steam injection affects the themal environment to a greater extent than would be expected based on data scatter nhaarved during repeatability tests, steam injection will be included in the production tests.

Test S-6 will be completed to establish the effect of releasing hylvgai through the simulated IDCA vents in combination with the simulated SRV spargers. 'Ihe 1/20 scale .

tests indicate that release of h A vy=i through the spargers is . limiting relative to release through both the spargers and the IDCA vents. h hyl vy=i release history used in the repeatability tests will be injected into the facility with the hy& + i relansed simultaneously through the spargers and the IDCA vents. Seventy percent of the

hy& vysi will be dischat W through the SRV spargers and l thirty percent will be discharged through the simulated IDCA vents.

. Test S-7 will be completed with a section of grating near the suppression pool surface above an active SRV i (

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Attachment 1 Page 2 of 3 IlGN-031 sparger. h grating section will be located in the sector between 318' and 332*. h purpose of this test is to evaluate the effect of grating near the suppression pool surface on the combustion transient. A separate scoping test (S-14) to evaluate the effect of grating near the suppression pool surface will be empleted in the River Bend Station geometry. For this test, the grating section will be located in the sector between 28' and 44*.

Test S-8 will be empleted with a blockage fraction of 50%

at the 17.2 foot elevation in the facility which is the the floor above the simulated IICU floor. h blockage included in the 225' chimney will be only 35% which is representative of the blockage fraction in this chimney in the Mark III plants. 'Ihis test will allow for an assessment of the effects of additional blockages such as equipnent, piping and cable trays which are not modeled in the 1/4 scale test facility. h blockage will be distributed uniformly in the facility using radial metal strips.

All of the repeatability tests will be performed with the Perry Nuclear Power Plant geometry. The first set of production tests will be performd with the Grand Gulf Nuclear Station configuration. If the thermal environment measured for the Grand Gulf production tests is within the data scatter observed during the repeatability tests, the production tests performed with the Grand Gulf configuration will be applicable to Perry.

A separate set of scoping tests (S-9 through S-14) will be completed for the River Bend Station plant geometry. h se tests will be empleted with simulated unit coolers in operation instead of the containmnt sprayo which will be used during all of the scoping tests performed with the Perry plant gemetry.

Table 2 is the !!ydrogen control Owners Group's final production test matrix. !!ydrogen release histories "A",

"B" and "C" listed in the production test matrix are identified in Attachment Two to this letter.

'Ihe production test matrix is predicated upon several assm ptions concerning the outcome of scoping tests. It is assumed that the scoping tests confirm that sparger release is limiting relative to sparger and IOCA vent release. h initial production test is expected to

Attachment 1 Page 3 of 3 1 2 -031 confirm that the Perry configuration is sufficiently similar to the Grand Gulf configuration such that data obtained in the Grand Gulf configuration is applicable to Perry. The hydrogen, oxygen and e ter vapor concentration measurements obtained with the igniters activated are assumed to be sufficient to resolve the mixing issue for

{u Mark III containments. All tests will be completed with Y

all igniters energized.

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TABLE 1

• SCDPING TEST MA3RIX TEST H,/S'IEAM NO./IDCATICE SPRAYS / CDNFIGURATION NO REudiSE HISitRY T ACTIVE SPARGERS CIXIJRS SIMJIAH D REMARKS .

S-1 "A" 9 (8 ADS + 329*) (N PERRY TESTS S-1 'IHROUGH S-3

'IO ARRPR9 REPEATABILITY S-2 "A" S-3 "A" S-4 INITIALLY AT CINFIRM URESHOLD H 1 IDt/SEC STEP FIDW RR DIFMJSION DOWN 'ID URESKID FLAMES V

S-5 "A" 00NQJRRENT STEAM /H2 RELEASE S-6 "A" 8 ADS + IDCA VENTS FIDW SPLIT BEIWEEN ADS SPARGERS AND IDCA VENTS (70% SPARGERS, 30% LOCA VENTS)

S-7 "A" 9 (8 ADS + 329*) GRATING INSTALLED ABOVE POOL FROM 318' 'IO 332*

S-8 "A" 9 (8 ADS + 329*) y INGEASED ARF 1. BIDGAGE INSTALLED AT ELEVATION 17.2' S-9 "A" 8 (7 ADS + 28*) RIVER BEND REPEATABILITY WITH COOLERS S-10 "A" S-11 "A" S-12 "A" y CDNOJRRENT STEAM /H 2

S-13 "A" 7 ADS + IDCA VENTS FIDW SPLIT BEIWEEN ADS SPARCERS AND LOCA VENIS (70% SPARGERS, 30% IDCA VENTS S-14 "A" 8 (7 ADS + 28*) Y V GRATING INSTALI2D ABOVE Ne_ Sam

• TABIE 2 PRODUCTION TEST MATRIX

' TEST H / STEAM NO./IOCATION CDNFIGURATION NO HISTORY gg ),OF ACTIVE SPARGERS(2) SPRAYS ORSIMHATED COOLERS 1 "A" 9 (8 ADS + 312*) OFF GAND GULF / PERRY (

2 "A" CN i3 "B" 4 "C" y 5 "A" 9 (8 ADS + 32*)

6 "A" V 7 "A" 9 (8 ADS + 152*) (3) CLINION 8 (7 ADS t 36*)

8 "B" V 9 "A" 8 (7 ADS + 323*) I3I 10 "A" 3) y

-11 "A" 88 (7(7ADS ADS + 186*)

+ 28*) ( (3) RIVER BEND 12 "B"

13 "C" V 14 "A" 3) 15 "A" 8 (7ADS 8 (7 ADS + 323*)

+ 230*) ( (3)y y

• See tM following page for Explanation Notes.

NorES EUR TABLE 2 (1) '1he hydrogen / steam release history ("A") shown in Attachment Two is for Grand Gulf. h corresponding release history for the other plants will be derived by multiplying this history by the following ratio:

No. of fuel bundles in the other plant core No. of fuel bundles in Grand Gulf's core (2) 'Ihe sparger location in the test facility are evenly spaced at 15' intervals. h test sparger closest to the actual location of the active plant sparger will be utilized.

(3) Since this plant is being simulate 1 as a " mirror image" of the test facility, the tested locations will be in

" mirror imaga" positions to those shown.

?

(4) This is the Grand Gulf configuration. Data frm these tests will.be considered applicable to Perry pending verification that the minor configuration differences between Perry and Grand Gulf do not affect the thermal environment. This verification will be made by comparing results for the first scoping test and test number two in the production test matrix.

i

, ATTACliMENT 'IWO

'IO HGN-031 HYDROGEN RELEASE HIS'IORIES 'IO BE USED IN THE 1/4 SCALE TEST PROGRAM The !!ydrogen Control Owners Group (HOOG) has utilized the NR Core lleatup Code to calculate hydrogen release histories which could be produced by degraded core accidents. We asstunptions used in calculating hydrogen release histories were identified in reference one. A meeting was held on Oct*r 3 aM 4 between the IKDG and the Nuclear Regulatory Comnission (NRC) staff to discuss use of the WR Core Heatup Code. As a result of this meeting, IKDG has implemented some minor revisions to the i use of the NR Core Heatup Cbde.

A revised model of the fuel rod has been hnplemented which incorporates omments identified by NRC consultants at Oak Ridge National Iaboratory (CRNL). his model corrects a minor deficiency in the code when a single node is used to represent the fuel. A revision has also been made to the fuel cladding gap nodeling. his revision also incorporates commnts identified by an NRC consultant from ORNL. These modifications were discussed with the NRC staff during the October 3 and 4, 1984 meeting. Details of the modifications will be subnitted to the NRC staff in a future HCDG submittal.

he irreversible oxidation cutoff was applied in earlier analysis using a cosine function over a 50*K tsuperature band around the irreversible oxidation cutoff temperature.

This approach was used to assure . stability (i.e. avoid numerical ringing) in code results' as the irreversible oxidation cutoff temperature was approached by a node. HOOG

'has now determined that this representation of the oxidation cutoff is not required for code stability. Rather than eliminate the gradual termination of oxidation, HOOG has decided to conservatively use a linear function over the last 10*K before the irreversible oxidation cutoff is reached.

All of the assumptions used in previous calculations of hydrogen production and identified in reference one have been used in final calculation of hydrogen release histories to be used in the 1/4 scale test program with one exception. Previous analyses have utilized four radial unit cells while the final analysis utilizes eight radial unit cells.

Hydrogen release history "A" which will be used in the 1/4 scale test program is produced by injecting 300 GPM into the. reactor pressure vessel at 2530 seconds after scram.

%e timing for reflood was established to assure that the

Attachment 2 Page 2 of 3 HGN-031 accident remains recoverable with no more than 30% of the active core zircaloy inventory exrw=H ng the zircaloy melting temperature at any one time. h reflood flow rate was established by reviewing the systerts which could be available during rcooverable degraded core accidents. h basis for selecting the 300 GPM flow rate was reviewed during a meeting between HCDG arri the NRC staff on August 28 and 29, 1984. 'Ihis hy&weu release history is believed to represent a limiting diffusion flame thermal environment for recoverable degraded core accidents. Figures 1, 2 and 3 show the hyivg1 release, steam release and zircaloy melt fractions as functions of time for this event.

Hydrogen release history "B" which will be used in the 1/4 scale test program is prea W by injecting 5000 GPM into the reactor pressure vessel at 3400 seconds after scram.

'Ihe timing for reflood was established to assure that the accident remains recoverable as described for release history "A". h reflood flow rate is representative of the capacity of any large EOCS flow rate. 'Ihis release history represects the most probable hydrogen generation event since it is most probable that a large capacity EOC system would be recovered. Figures 4, 5 and 6 show the hyivpn release, steam release and zircaloy melt  :

fractions as functions of time for this event.

h peak hyi v pi production rate shown in Figure 4 exceeds the capability of the 1/4 scale test facility to simulate hyivpfa production at the hydrogen release points. h HOO3 intends to simulate hylvpi release history "B" until the predicted hy& wen release rate reaches 1.0 kg/sec (full scale) which is the peak hyl v gi injection rate in the 1/4 scale facility. The hyivpi injection rate will be held at this value until the total auxint of hyivpa injected equals the amount in the hyi vgen production " spike". h hyivpi injected to the facility will then correspond to the release history shown in Figure 4 through the conclusion of the transient.

Hydrogen release history "C" which will be used in the 1/4

- t ain corresp mds to injecting .1 lbm/sec. in the scale piv3 facility for a pericd of 45 minutes, full scale. The data obtained from this test will be used exclusively for verifying the analysis of deflagrations using the CIASIX-3

Attachment 2 Page 3 of 3 HGN-031 code. 'Ihe time duration for this test was established to assure that sufficient hyl vyui will be injected into the facility to produce either global hyivful burns in the

, wetwell ctrnpartment or possibly a global burn in the entire facility.

'Ihe IKDG has selected a constant hydrogen injection rate of

'O.1 lbn/see to verify the deflagration analysis. 'Ihis injection history was selected to assure that the combustion phenomena would be well separated from combustion phenomena which occur near the threshold for steady diffusion flames.

Based upon previous testing, the threshold hydrogen flow rate for establishing steady diffusicn flames is ;

approximately .4 lbm/sec. A hydrogen injection flow rate of 0.1 lbn/sec should increase the probability of achieving voltane burns which are comparable to the. combustion phenomena treated by the CLASIX-3 code.

/

References

1. Ietter HGN-018 from Mr. S. H. Hobbs to Mr. H. R. Denton dated July 6, 1984.

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