ML20005C100

Forwards Amend 18 to Psar.Amend Contains Matl Responsive to Two Hydrogen Control Requirements of near-term Cp/Mfg License Regulations Which Were Adopted by NRC on 810827. Related Correspondence

ML20005C100
Person / Time
Site:	Black Fox
Issue date:	11/05/1981
From:	Gallo J ISHAM, LINCOLN & BEALE, PUBLIC SERVICE CO. OF OKLAHOMA
To:	Purdom P, Shon F, Wolfe S Atomic Safety and Licensing Board Panel
References
NUDOCS 8111180389
Download: ML20005C100 (4)

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Sheldon J. Wolfe, Esquire Mr. Frederick J. Shon l' Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing-l Board Panel Board Panel

]. U. S. Nuclear Regulatory U. S. Nuclear Regulatory l Commission Commission l i Washington, D. C. 20555 Washington, D. C. 20555 Dr. Paul W. Purdom l Administrative Judge U. S. Nuclear Regulatory Commission c/o Environmental Studies l Group

Drexel University l 32nd and Chestnut Streets

, Philadelphia, PA 19104

! Re: In the Matter of Black Fox Station, Units 1 and 2, Docket Nos. STN 50-556 and STN 50-557 l

Gentlemen

i l I am enclosing Amendment No. 18 to the Black Fox l PSAR for the information of the Licensing Board and parties.

i This amendment contains material responsive-to.the two hydrogen control requirements of the Near-Term Construction Permit / Manufacturing License regulations which were adopted by the NRC Commissioners on August 27, 1981. Those regula-tions are currently awaiting publication in-the Federal

Register.* j I I

• The remainder of Public Servico Company of Oklahoma's response to the post-TMI requirements for construction permit applicants is contained in Amendment No. 17 to the PSAR, which was transmitted to the NRC Staff, Intervenors, and the State of Oklahoma on October 5, 1981, and to the Licensing Board by letter dated October 8, 1981.

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I Sheldon J. Wolfe, Esquire Mr. Frederick J. Shon Dr. Paul W. Purdom November 5, 1981 Page Two 1 Amendment No. 18 was hand-delivered to the NRC Staff on- October 22, 1981. Copies were sent to counsel for Intervenors and the State of Oklahoma by Federal Express on October 21, 1981. At that time, counsel for the Appli-cants stated that Applicants would support any request by Intervenors and the State of Oklahoma to revise the schedule to permit thirty days from the receipt of Amendment No. 18 in which to file contentions based upon the material con-tained therein.

Sincerely,

/

/

/ osep Gallo Counsel for Public Service Company of Oklahoma Enclosure JG/pm l

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RELAIED CORRESPONDENCE l- --

BLACK FOX _SR TION SERVICE LIST m r,u c.;

USNAC Mr. Lawrence Burrell . _

Mr. Joseph Gallo Route 1,' Box 197 *81 NOV -9 Pl2:15 Isham, Lincoln & Beale Fairview, OK 73737 1120 Connecticut Avenue NW Suite 325 kl]h{,fC James H. Thessin, Esquire t,){* Washington, D.C. 20036 Office of Executive Legal gnfaics Director Mr. Maynard Human USNRC General Manager Washington, D.C. 20555 Western Farmers Electric Cooperative P.O. Box 429 Mrs. Carrie Dickerson Anadarko, OK 73005 Citizens Action for Safe Energy, Inc.

P.O. Box 924 Mr. Michael Bardrick Clarecore, OK 74017 Asst. Attorney General 112 State Capitol Building Mr. Gerald F. Diddle Oklahoma City, OK 73105 General Manager Associated Electric Cooperative, Inc. Mr. Dino Scaletti P.O. Box 754 US 3C Springfield, MO 65801 Phillips Building 7920 Norfolk Avenue Joseph R. Farris, John R. Woodard III Bethesda, MD 20014 Feldman, Hall, Franden, Reed and Woodard Mrs. Ilene Younghein ps 816 Enterprise Building 3900 Cashion Place

. Tulsa, OK 74103 Oklahoma City, OK 73112 t

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STATE OF OKLAHOMA COUNTY OF TULSA R. O. Newman, being first duly sworn, deposes and states:

That he is President, PUBLIC SERVICE COMPANY OF OKLAHOMA, the Applicant herein; that he has read the following Amendment 18 to the Black Fox Station Units One and Two Preliminary Safety Analysis Report and knows .

the contents thereof; that the same is true as he verily believes.

l DATED: This 26th day of October- , 1981 Signed s/R. O. Newman O R. O. Newman President Subscribed and sworn to before me this 26th day of October , 1981 s/ Lina HoLn Notary Public in and for the County of Tulsa, State of Oklahoma My Commission expires February 21, 1983 O

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ERRATA AND ADDENDA SHEET

AMENDMENT 18, OCTOBER 22, 1981 ,

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l Remove Page Insert Page Dated 18-102181'

. ADDENDUM II i 11, 17-100581 11 l

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18-102181

O ADDENDUM II Additional TMI-Related Requirements Addendum II to the Black Fox Station (BFS) Preliminary Safety Analysis Report (PSAR) identifies the Applicant's commitments regarding the design, construction, and operation of the BFS in response to the acci-dent at Three Mile Island, Unit 2. .

In accordance with the NRC Staff guidance contained in the July 14, 1981, generic letter (Generic Letter No. 81-26) to all pending construction permit and manuf acturing license applicants, Addendum II consists of responses to the requirerents embodied in a new paragraph (e) to 10 CFR 50.34, entitled " Additional TMI-Related Requirements. "

Commitments contained in Addendum II supersede any conflicting statements elsewhere in the PSAR where such conflicting statements were made earlier '

than the date of the current revision of Addendum II, i

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18-102181 _

10 CFR 50.34(e) Title Page

, (2)(xxvi) Primary Coolant Sources Outside the Containment 183 Area I

(2) (xxvii) In-Plant Radiation Monitoring 188 (2) (xxviii) Control Room Habitability 190 (3) (1) Procedu*es for Feedback of Operating, Design and Construction Experience 193 (3) (11) Expand QA List 203 (3) (111) Develop More Detailed QA Criteria 210 (3) (iv) Degraded Core - Dedicated Containment Pene- 242 tration

{} (3) (vi) Dedicated Containment Penetration 243 (3) (vii) Organization and Staf fing to Oversee Design and 244 18 i

Cons truction

, (2) (ix) (3) (v) Degraded Core - Hydrogen Control 271 l

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l 10 CFR 50.34e)(2)(ix)/(3)(v) RULEMAKING PROCEEDING ON DJORADED CORE l ACCIDENTS I

"(-)

v NRC POSITION:

(2) To satisfy the following require =ent, the application shall provide suf ficient info r=ation to de=onstrate that the required actions will be satisfactorily co=pleted by the operating license stage. This infor=ation is of the type custo=arily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues.

(NUREG-0713, Category 4)

(ix) Provide a syste= for hydrogen control that can safely )

acco=:odate hydrogen generated by the equivalent of a 100 l fuel-clad =etal water reaction. (NUREG-0713, II. B.8) l (3) To satisfy the following require =ents, the application shall provide sufficient infor=ation to de=onstrate that the require =ent has been  ;

=et. This infor=ation is of the type custo=arily required to satisfy '

10 CFR 50.34(a)(1) or to address the applicant's technical qualifi-cations and =anage=ent structure and competence. (NUREG-0713, Category 5)

(v) Provide preli=inary design infor=ation at a level af detail consistent with that nor= ally required at the construction l per=it stage of review sufficient to demonstrate that: (II. B . 3)

(A) Contain=ent integrity will be =aintained (i.e. , for steel contain=ents by =eeting the require =ents of the ASME i Boiler and Pressure Vessel Code,Section III, Division 1, Subarticle NE-3220, Service Level C limits, excep t that evaluation of instability is not required, considering pressure and dead load alone. For concrete contain=ents l'

by =eeting the require =ents of the ASME Soiler Pressure vessel Code,Section III, Division 2, Subarticle CC-3720, f actored load category, considering pressure and dead load l alone) during an accident that releases hydrogen generated i f ro= 100 percent fuel-clad cetal-water reaction accompanied l by either hydrogen burning or the added pressure fro =

post-accident inerting assuming carbon dioxide is the I inerting agent, depending upon which option is chosen for l control of hydrogen. As a =in1=u=, the specific code l

require =ents set forth above appropriate for each type of containment will be =et for a co=bination of dead load ano an internal pressure of 45 psig. Modest deviations fro:

l these criteria will be considered by the staff, if good cause is shown by an applicant. Syste=s necessary to ensure contain=ent integrity shall also be de=onstrated to perfor= their function under these conditions.

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3 NRC ?OSITION: 10 CFR 50.34(e)(2)(ix)/(3)(v) ggg 03 ) The contain=ent and associated syste=s vill provide reasonable assurance that unifor ly distributed hydrogen concentrations do not exceed 10 percent during and following an accident that releases an equivalent a= cunt of hydrogen as would be generated from a 100 percent fuel-clad =etal-water reaction, or that the post-accident atmosphere will not support hydrogen ec=bustion.

(C) The facility design will provide reasonable assurance '

that, based on a 100 percent fuel-clad metal-water reaction, co=bustible concentrations of hydrogen will not collect in areas where unintended cochustion or detonation could cause loss of containment integrity or loss of appropriate =1tigating features.

(D) If the eption chosen for hydrogen control is post-accident inerring: (1) Contain=ent structure loadings produced by an inadvertent full inerting (4ssu= lag . carbon dioxide), but not including seis=ic or design basis accident loadings will not produce stresses in steel contain=ents in excess of the limits set forth in the ASME Soiler and Pressure Vessel Code,Section III, Division 1, Subarticle NE-3220, Service Level A lh Limits, excep t that evaluation of instability is not required (for concrete contain=ents the loadings specified above will not produce strains in the contain=ent liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Subarticle CC-3720, Service Lead Category),

(2) A pressure test, which is required, of the contain=ents, at 1.10 and 1.15 ti=es (for steel and concrete containments, cespectively) the pressure calculated to result from carbon dioxide inerting can be safely conducted, (3) Inadvertent full inerting of the contain=ent can be safely acco==odated during plant operation.

OE) If the option chosen for hydrogen control is a distributed ignition syste=, equipment necessary for achieving and =aintaining safe shutdcun of the plant and

=aintaining contain=ent integrity shall be designed to perf or= its function during and af ter being exposed to the environmental conditions created by activation of the distributed ignition syste=.

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?SO RESPONSt: 10 C7R 50.34(e)(2)(ix)/(3)(v)

O PSO RESPONSt:

I'CRODUCTION The basis for the requirement for a hydrogen control system which is capable of dealing with rapid generation of large quantities of hydrogen is the IMI-2 accident, which resulted in the generation of hydrogen beyond the limits specified in 10 CFR 50.44 As a consequence the NRC has identified hydrogen control arising from a degraded core as deserving special attention. The Commission has imposed new hydrogen control requirements on plants about to receive operating licenses, and more recently has issued hydrogen control requirements as part of the newly issued Near-Term Construction Permit / Manufacturing License Regulations.

These construction permit hydrogen control requirements are hereaf ter rc52rred to in this response as the " Hydrogen Control Rule."

COMMITME!C PSO commits to provide a Hydrogen Control System (HCS) which will safely accommodate, in accordance with the Hydrogen Control Rule, the hydrogen generated by the equivalent of a netal-water reaction which consumes 100 percent of the circonium metal in the active fuel cladding.

O ntSCRzeT10N Or THz RzS10NSt The following four sections of this response provide the detailed information necessary to support PSO's commitment to comply vita the requirements of the Hydrogen Control Rule. The first of these sections presents a description of PSO's long-r nge hydrogen control program, the bases for the preliminary system selection and a conceptual system description. The second of these sections presents a description of the preliminary design parameters which were used to assess the adequacy of the HCS. The third of these sections presents the preliminary sys tem description and the assessment of the system's perfor=ance. The last of these sections presents a detailed discussion of the analytical methodology used in completing the performance assess =ent.

A. HYDROGEN CONTROL PROG?lM In response to the NRC's hydrogen control require =ents, PS0 has undertaken a long-range hydrogen control program.

1. PROGRAM PLAN The hydrogen control program is proceeding in several phases, as described below.

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?SO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v) ggg o Phase 1--Preli=1 nary selection of a hydrogen control system. The results of this prelisinary assess =ent are presented in this response. PS0 vill continue its evalua-tion of the various alternative syste=s for hydrogen control, including a consideration of the various Ledus try activities in this area. The final evaluation and selection a=ong the various hydrogen control syste=s will be cocpleted and submitted to the NRC Staff two years af ter the issuance of the construction permits. An evaluation program, similar to that described in Phase 2 below, will be carried out for the final hydrogen control system selected.

o Phase 2--Freliminary evaluation of the selected system agains t the requirements of the Hydrogen Control Rule and other specific design and perfor=ance criteria. This effort has been cocpleted and the results are presented in this response.

o Phase 3--Detailed syste= evaluation which cu1=inates in a final design. The results of this effort will be subnitted with the FSAR.

PSO recognizes the existence of the many ongoing and planned research and development programs in the area of hydrogen lll control. Exa=ples of these programs are identified in Table (2) (ix)-1. As part of its long-range hydrogen control program,

?SO is coc=itted to active participt. tion in the BWR Hydrogen Control Owner's Group and to =aintaining cognizance of industry efforts in this area.

2. PRELIMINARY SYSTEM SELECTION

a. Selection Criteria A nt=ber of approaches to hydrogen control have been proposed. These approaches, as integrated into a total

3FS HCS, were evaluated against the following criteria

1) The ECS and its supporting syste=s =ust be able to safely control the hydrogen generated by the equivalent

=etal-water reaction which consuces 100 percent of the zirconium metal in the active fuel cladding, such that contain=ent integrity and safe shutdown capability will be achieved and =aintained.

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1 FSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

[~ 2) The system mus t be able to maintain the hydrogen I concentration below the detonable limits or create I an atmosphere incapable of supporting combustion.

3) The operation, including inadvertent operation, of the HCS should not endanger the health and safety of I the public. l 1

4) The inadvertent operation of the system should not result in unacceptable da= age to station safety systems or pose an undue risk to station personnel.

5) The HCS cust be able to assure that no stagnant areas exist where unintended comb'istion or detonation could result in a loss of containme.:- integrity or loss of any required =1tigating features.

6) The HCS =us t be able to function adequately over a wide variety of postulated events.

7) Th s components of the HCS, insofar as possible, should be a standard design and not require extensive development. If major components or subsystems require develop = ental work, the potential' for substantial s improvement over current performance levels should k_) exist.

8) The preliminary assessment of each alternative should be based on an appropriately conservative analysis, b.- Evaluation and Svstem Selection There is a considerable a=ount of research under way to evaluate various aspects of hydrogen control. These l activities are expected to provide valuable information l

in a ti=e fra=e that will support the detailed design and procure =ent of a final RCS for BFS. The following four potential hydrogen control systems were selected for preliminary evaluation against the above listed criteria:

o Water fogging o CO, pos t-accident inerting o Halon post-accident inerting o Distributed igniter system O

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I PSO RES?ONSE: 10 CFR 50. 34 (e) (2) (ix) / (3 ) (v)

The conclusion of the prelisinary evaluation was that a Distributed Igniter Syste= satisfies all of the above specified evaluation criteria. Base on a q2alitative evaluation of these syste=s, pSO has tentatively selected a ECS consisting of a Distributed Igniter Syste= (DIS) operated in conjunction with a spray syste= equivalent to a single loop of the containment spray operating code of the Residual Heat Re= oval CRER) syste=.

c. Conceotual Svste= Descriution The DIS would react large quantities of hydrogen with oxygen by controlled co=bustion of the hydrogen as it is released to the dryvell and contain=ent volu=es. The rate of hydrogen co=bustion within a given volume is influenced by the nu=ber and distribution of the igniting ele =ents. These ele =ents would be provided in sufficient quantities and at the proper locations to ensure that local hydrogen concentrations re=ain below the detonable range as long as the local atmosphere is capable of supporting co=bustion.

The heats of co=bustion, metal-aater reaction, and radio- g active decay would be absorbed by the suppression pool, W 1 the contain=ent sprays, and the large ther=al = ass of the -

contain=ent concrete and steel structures. This heat absorption will limit the te: cerature effects of hydrogen combustion to localized transients which will be analyzed in detail during the equip =ent survivability review. The pressure suppression effect of the contain=ent spray will act to limit the peak pressures resulting fro = the controlled deflagrations to values below the =ini e rcquired pressure used in the evaluation of containment integrity, as defined in Subsection C.2.d.

The DIS will be designed with suitable reliability such that proper functioning of the syste= is assured. The DIS will be powered from two independent sources such that each source will supply power to one-half of the igniter asse=blies.

3. PRELIMINARY DESIGN PARAMETER DEVELOPMENT

1. INTRODUCTION The three =ajor design para =eters which are used in the assessment of perfor=ance adequacy are as follows:

US -

PSO RESPONSE: 10 CFR 50. 34(e) (2) (ix) /(3) (v)

() o Hydrogen release rates o Hydrogen release points o Hydrogen combustion characteristics These design parameters were utilized in the hydrogen

=1gration and combustion analyses which are described in Section D, as part of the assessment of the adequacy of the proposed system relative to the requirements of the Hydrogen Control Rule. The selection of parameter values and analytical techniques was based on engineering judgment and experience in perfor=ing similar work in other areas. Where appropriate, parametric analyses were perfor=ed prior to selecting the base case values.

2. DESIGN CRITERIA

a. Rates of Release Conditions were postulated such that the resultant nass and energy release to the contain=ent system is accompanied by significant hydrogen generation during a ti=e fra:e which would reasonably permit recovery of core cooling prior to significant core degradation. Three systes para =eters are of major

_( significance in defining the hydrogen release.

These are:

1) Rate of Reactor Coclant Loss The ti=e to onset of hydrogen generation is determined by the rate of bicwdevn, as significant hydrogen generation does not begin until the water level in the RPV has dropped below the active core region. The rate of coolant loss also influences the rate of hydrogen generation by limiting the flow of steam and hydrogen from the core during the reactio9 period. A steam line break area of 0. L63 f t' was assumed. This is equivalent to a stuck-open safety-relief valve.

2) Rate of Makeup Depending on the rate of coolant loss and the power history of the ' core, makeup =ay be required to avoid core slu= ping prior to achieving a significant a=ount of =etal-water reaction.

yor this analysis, the reactor coolant venting 277 -

?SO RESPONSE: 10 C.F.R. 50. 34 (e) (2) (1x) / (3 ) (v) concurrently with a =akeup of 30,000 lbm/h inj ected into the lower plenus until the peak fuel centerline te=perature reached 4130 F.

This =akeup flow is equivalent to that provided by the Control Rod Drive (CRD) system.

3) Previous Core History and Core Characteristics These will determine the decay heat values and the fuel te=perature distribution. The reactor was assu=ed to be scra==ed at 100 percent power with an equilibrium power history.

In order to develop detailed, hydrogen release data, ?SO utilized the MARCH ~ co=puter code along with certain assu=p tions chosen to approxi-mate the conditions leading to a =1x1=um calculated ce=ulative release. The following assu=ptions were =ade in addition to those given above.

a) When the peak fuel centerline te=perature reached 4130 F at 34 =inutes, only the hydrogen = ass flow rate was taken fro = the g MARCH calculation and the mass and energy W release rates for stea= were based on the assumed availability of sufficient makeup to remove the total energy generated by decay heat and the metal-water reaction.

b) No fuel was allowed to slu=p into the lower plenum until all fuel reached the

=elting point. Prior to this point, the reaction beca=e steac-li=ited and the results of the MARCH calculation were modified as discussed below.

c) At 70 ninutes, the reaction rate was severely limited by the amount of flashing steam fro = the lower plenum. At this point, approxi=ately 65 percent of the active fuel cladding had reae:ed. To satisfy the Hydrogen Control Rule re-quire =ent to safely control the hydrogen generated by the equivalent of a =e tal-water reaction which consumes 100 percent of the zirconium =etal in the active fuel cladding, the reaction rate was assu=ed to be constant at 43.5 lb=/=in hydrogen until 34.2 =inutes, at which time the reaction was ce=plete.

PSO RESPONSE: 10 CFR 50.34(e)(2)(ix) /(3)(v) fg The resultant = ass, energy, and hydrogen

(_) release rates are shown in Tables (2)(ix)-2 and -3. These release rates were used for performing the preliminary evaluation of the performance adequacy of the DIS. A more detailed description of the methodology used, and the parametric analyses performed to establish this hydrogen generation rate is presented in Section D of this response.

The SWR Hydrogen Control Owner's Group has undertaken the evaluation of a BWR hydrogen source term, and of the factors which are expected to influence the design basis for evaluating the adequacy of the Hydrogen Control System performance. These factors include:

o Time to start of generation o Rate of generation o Mass and energy release to contain=ent o Location of release o Total hydrogen release The results of the BWR Hydrogen Control Owner's Group evaluation will be compared with the (S

.s) values selected for this preliminary assess =ent and any codifications which are indicated as a result of this comp 2rison will be reflected in the Phase 3 detailed design effort.

b. Release Points The 3FS contain=ent shown in Figure (2)(ix)-1 is relatively insensitive to the precise release = odes.

The two cases for evaluation are: a single asym-metric release through a safety-relief valve (SRV) dcwncomer in the suppression pool; and, a symmetric discharge through the drywell vents.

Releases inside the drywell volume would channel the mass and energy of the release to the contain=ent volume through the horizontal vents at the base of the drywell into the suppress 1.,n pool. 3ecause of the symmetry of the vents, such a release would produce an axially sy= met-ic distribution of the steam and noncondensible gases to the containment volume.

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PSO RESPONSE: 10 CFR 50.34(e)(2) (ix)/(3) (v)

Potential release points outside the drywell but O

inside the contain=ent can be subcategorized into three general groups--high energy fluid syste=

lines, icw energy and s=all dia=eter piping, and large dia=eter inactive piping such as the safety-relief valve discharge lines (SRVDL) and ECCS test return lines. Of these groups, only the SRVDL creates the potential for continuous release of significant = ass and energy to the contain=ent. The SRVDL ter=1nate outside the drywell and in the suppression pool, as indicated on Figure (2)(ix)-2.

The failure of a SRV to close when required would result in a significant rate of mass and energy release to the suppression pool,

c. Hydrogen Co=bustion Para =eters The basic analytical tool used to assess the perfor=ance adequacy of the DIS is the 3&V co=puter code HY3 RID.

A functional description of ET3 RID is presented in Section D.4. of this response. EY3 RID utilizes the hydrogen release rates developed in Section 3.2.a.

and, by appropriate =odeling, inj ects this hydrogen into the contain=ent syste= at the release points llh ,

identified in Section B.2.b. To provide the pressure and te=perature ti=e histories which would result fro = controlled ce=bustion of this hydrogen, it is necessary to specify the relevant ec=bustion para =eters.

2 NUREG/CR-1561 provides a concise su==ary of the ,

current literature relative to hydrogen co=bustion.

Based on NUREG/CR-1561 and other available infor=ation,

PSO has selected base case values for the para =eters l listed in Table (2) (1x)-4 C. S?5TDf DESCRIPTION AND PERFORMANCE ASSESSMENT l 1. SYSTEM DESCRIPTION

a. Preli=inarv Lavout l The results of the hydrogen =igration analysis,

! which are described in See:1on C.2.a. of this response, fors the basis for the preli=inary layout of the DIS. The relative locations of the igniters are shown in Figures (2) (ix)-3 through (2) (ix)-3. To ensure pro =p t ignition of hydrogen exiting fro = g either a single point or distributed release, a ring W L - _ i

pSO RESPONSE: 10 CFR 50.34(e)(2)(1x)/(3)(v) of 18 igniters (9 per division) will be placed in O th vict=1=7 or th 9 t>te r= t stev tio= 576 ' 7"-

A second ring of 12 igniters (6 per division) will be placed near the HCU floor at Elevation 592' 10".

Thus, a total of 30 igniters (15 per division) will be available to provide positive ignition of hydrogen in the wet well region of the containment volume.

Above the HCU floor, the flow of hydrogen is directed by the floors and walls of the subcompartments which span the area from the drywell to the containment vessel. Hydrogen which exits the wet well region will be channeled by the steam tunnel and the suspended concrete slabs beneath the HCU modules at Elevation 592' 10" into one of the four relatively open quadrants between the sides of the concrete slabs. A total of 8 igniters (4 from each division) will be placed in these areas in the vicinity of the platform at Elevation 641' 5". A total of 8 igniters (4 from each division) also will be placed in these areas in the vicinity of the platfor=s at Elevation 618'11".

To provide for reliable cochustion of any hydrogen which reaches the containment dome, a. total of 12 igniters (6 from each division) will be placed in the volume above the polar crane.

The containment air recirculation system supplies chilled air to various general areas of the con-tainment. This system isolates on a LOCA signal but can be manually restarted by the operator. With the sole exception of the Main Steam Tunnel, subcompartments within the containment are cooled by internal fan coil units. There is very little air movement between these subcompartments and the general containment atmosphere even under onditions of full forced recirculation. To ensure controlled ignition of any hydrogen which might migrate into open subcompartments,

, 2 igniters (1 from each division) will be placed in the Main Steam Tunnel at about Elevation 525' 0" and in the Reactor Water Cleanup Demineralizer (RWCUD) pump and tank room on Elevation 641' 5". The total number of igniters to be used in the containment is 62.

To provide for controlled ignition of any hydrogen released to the dryvell, two rings of 8 igniters

each (4 from each division) vill be provided. The first ring vill be located at about Elevation 590' 0 " .' The second ring will be in the vicinity of the platform at Elevation 616' 11-1/2". As discussed 281 -c

pSO RESPONSE: 10 CFR 50.34(a)(2)(ix)/(3)(v) below, combustion in the drywell is expected to occur under hydrogen-rich rather than oxygen-rich conditions. Air is expected to reenter the drywell pri=arily through the drywell vacuu=-relief line.

To control the rate of oxygen buildup, 6 igniters (3 fro: each division) will be located in the upper part of the drywell. These igniters will also provide protection against potential pocketing in the drywe11 head region. The total nu=ber of ig-niters to be used in the drywell is 22.

b. Igniter Asse=bly Description The igniter asse=bly proposed for the preli=inary 3FS DIS is similar to that e= ployed at Sequoyah Nuclear Station and proposed for Grand Gulf Nuclear Station. The igniter, as presently envisioned, is a General Motors AC Divisien Model 7G glow plug which will be =ounted in a we'.ded steel box. The glow plug will be provided with a spray shield to protect the igniter ele =ent fro = contain=ent spray. The igniters located in the wet well region will either be pro-vided with deflectors for pool swell and froth protection, or will be shcun to have an acceptable surface temperature recovery ti=e following Ln-

=ersion. A heat shield will be provided, if necessary, to protect the igniters fro high ta=peratures.

c. Igniter Suecorts The igniter asse=blies will be adequately supported to withstand, without loss of function, the loads associated with seis=le events (SSE). and hydrody-na=le (pool swell, jet i=pinge=ent, and pressure spikes associated with pipe rupture and hydrogen ignition) and ther=al (pipe rupture and hydrogen ignition) transients.

d. Power Sucolies The igniter asse=blies will be provided pcwer frc=

two 480 7 ISF pcwer buses, one for each division.

These are Class lE power supplies which, in the event of the failure of the nor=al power supplies, will be fed fro the station's emergency diesel generators. A preliminary one-line diagra: is shown on Figure (2)(Sc)-9. All electrical components excep t the local junction bcxes will be located outside the contain=ent and are therefore accessible llh for inspection and repair, even during operation of the system.

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PSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

Ih a e. Pronosed Method of Operation

1) System Operation This DIS will be designed to prevent the ac-cumulation of detonable concentrations of hydrogen. The DIS will not be required for events which result in the generation of hy-drogen less than or equal to the amounts and release rates considered in the design of the present Containment Combustible Gas Control System (CCGC) as described in PSAR Subsection 6.2.5. It is intended that the DIS be manually actuated for all event sequences which possess the potential to generate excessive amounts of hydrogen. The design of the DIS will be such that planned or inadvertent actuation of the system will not adversely affect the operational safety of the plant nor increase the severity of a particular event.

2) Initiation Criteria

.-~ As shown in Section B.2.a. an event which will

("f) require operation of the DIS proceeds at such a rate as to allow actuation by a control room operatar in accordance with emergency operating procedures. Reactor water level is considered te be the best indication of the potential for rapid hydrogen generation in a 37R. System initiation details will be included in the FSAR.

f. Tests and Insnection

1) Preoperational Testing The DIS will be preoperationally tested to ensure correct functioning of all controls, instrumentation and wiring, transformers and igniters. The test vill consist of energizing one of the two ESF power distribution panels from the control room and verifying that all igniters powered from the associated panel are functional. The identical procedure will be followed for the re=aining igniters powered off the remaining ESF panel.

O

?SO RESPONSE: 10 CFR 50.34(e)(2)(ix) /(3)(v)

2) Surveillance Tests During plant ope ration, the igniter asse=blies, pcwer distributien panels, ins trumentation, and associated viring can be visually inspected (outside the drywell) and operationally tested at any ti=e. All igniter asse=blies will be tested periodically to verify operability. The test procedure will be st=ilar to the pre-operational test procedure discussed above.

g. Instru=entation and Controls The DIS will be =aaually initiated fro = the control roc =. Instru=entation for the DIS consists of two control roo= handswitches, one for each of the two Class lE power divisions. Each handsvitch energizes the igniters in its respective division. .

2. DISTRI3UTED IGNITION SYSTDi ?ERF0F0fANCE ASSESSMENT During the course of the preli=inary perfor=ance assess- a s

=ent described belov, other advantages of the DIS were W identified. These other advantages include:

o The hydrogen deflagration process produces pressure ti=e--histories which have frequencies which are significantly lower than the =ajor structural response frequencies.

o The syste= configuratic: is flexible. That is, should design para =eters change, the syste= can be expanded or altered (e.g. , adding or relocating igniters) with =ini=al impact on the re=ainder of the plant. Moreover, the igniters can be located in l such a =anner that the loss of one or more igniters l will not li=it the ability of the DIS to perfor= its intended function.

a. Unintended Local Co=bustion The potential for localized high concentrations of hydrogen (pocketing), which =1ght lead to unintended ce=bustion or detonation was evaluated by perfor=ing a hydrogen =igration analysis using the SOLA-DF g cc= outer code. A description of this analysis is W cuu -

PSO RESPONSI: 10 CFR 50.34(e)(2)(ix)/(3)(v)

() presented in Section D.3. The following discussion presents the results of the analyses perfor=ed.

1) Gratings The annular region of the 3FS containment volume is subdivided by several levels of grating which serve as both personnel access platforns and as supports for equipcent. The locations of the gratings (single-line shading) are shown in Figures (2)(ix)-3 through -8.

This grating has an open area of 50 percent, not counting the area occupied by equipment base pads. To determine the potential effects of gratings on hydrogen flow, two separate SOLA-DF evaluations were perfor=ed.

The first evaluation was a free rise simulation in which the hydrogen was released from the suppression pool and encountered no obstacles except the drywell and contain=ent walls.

Typical results are shown in Figures (2)(ix)-10 and 11. Each small square sy=bol represents 0.5% within a grid. Figure (2)(ix)-10 displays

() an area 18 feet wide by 30 feet high by 18 feet deep divided into a 15 by 15 by 15 grid. The bottom of the figure represents the surface of the suppression pool. The Figure (2)(ix)-10 cross-section is taken approxi=ately S.5 feet out from the drywell wall. The plume remains .

relatively compact (the half-angle of expansion is about 10 degrees). Figure (2)(ix)-ll shows i the plu=e horizontal cross section approximately l

10 feet above the surface of the suppression l pool. The plume is reasonably sy==etric and I nearly circular.

The second evaluation was a grating analysis perfor=ed by closing off alternate reus of cells in two layers, resulting in two layers each with 50 percent open area. The long axis of the top row of cells was oriented at right angles to the long axis of the lower layer.

Except for the simulated grating, all conditions are identifical for the two cases. Typical results are shown in Figures (2)(ix)-12 and

-13. The differences between the two analyses are readily apparent, as the gratings cause the hydrogen to disperse horizontally to a nuch

('/)

~-

I l

l M

pSO RESPONSE: 10 CFR 50.34(e)(2)(ix) /(3) (v) greater degree, when cocpared to the free rise ~

si=ulation. g It has been concluded from these analyses that gratings or other large obstacles (e.g. , pipe suppo: structures, large equip =ent) can have a significant dispersive effect on hydrogen flow and should be included in detailed =igration analyses.

2) Reactor Building 360 Degree Evaluations The purpose of this series of analyses was to gain an understanding of how hydrogen which is released either at a single point or in a unifor= distribution can be expected to flov in a large, internally seg=ented structure like the 3FS contain=ent. In particular, it was considered necessary to determine the size of volume for which the assu=ption of instantaneous, ho=ogeneous =1xing would be rea-sonab le. An SRV release under the RWCU equip =ent area was selected for evaluation. A relatively coarse nodularization sche =e was used. For this relatively coarse =odel, the effects of gratings and s=all subco=part=ents were not included. This is considered to be conservative with respect to dispersal effects for evaluating unifor= distribution. The ggg effect of grating and s=all subco=part=ents is considered in the 90-degree evaluation, as discussed in section C.2 a.3.

The results of the RWCU release are shown in Figures

, (2)(1x)-14 and -15. Figure (2)(ix)-14 shows a vertical cross-section taken near the dryvell vall.

Figure (2)(1x)-14A shcws the hydrogen distribution af ter 2 minutes of release and Figure (2)(ix)-143 after 4 =inutes. As indicated, the hydrogen plume can be expected to rise fairly slow with =ini=al initial dispersal until the plu=e reaches the vicinity o' the refueling floor. After 4 minutes of release, the dc=2 hydrogen concentrations are starting to approach the lower fla==able li=it (LFL) of 4 percent, while this li=it was exceeded in the wet well tegion very shortly af ter the release started. The potential for structures and solid floors to create te=porary asy==etric flow patterns is evident.

Figure (2)(Lt)-15 shows two separate cross-sections taken af ter 12 =inutes of release. Section 1-1 0

2S6 18-102181

PSO RESPONSE: 10 CFR 50.34(e)(2)(1x)/(3)(v) indicates the start of hydrogen migration horizontally O 1=== =* vet 11 = ste - see:1on 2-2, taxen nearer the RPV center line, shows no evidence of horizontal flow into this area.

On the basis of this preliminary analysis and the grating analysis described above, it has been concluded that for a tingle point release, the homogeneous mixing assumption is reasonable for a 90 degree are cent tred about the release point.

The distributed release case was analyzed by simulating 13 equally spaced release points around the drywell wall, as shown in Figure (2)(1x)-16. The total hydrogen 'felease rate was identical to that used in the single point release evaluation. The holdup and dispersive ef facts of the Main Steam Tunnel (shown at the right center of Figure (2)(ix)-13) is indicated on Figure (2)(1x)-17. The grating analysis described above indicates the strong dispersive effect of gratings and examination of the BFS arrange =ent drawings indicates the presence of large areas of grating in all areas except the Equipment Re= oval Hatch area.

.h Based on the grating and distributed release analyses, it has been concluded that a distributed discharge

~

will result in relatively uniform concentrations in the wet well region.

3) Reactor Building 90 Degree Evaluations a) Equipment Removal Hatch. An SRV release under

the Equipment Removal Hatch area was simulated in detail. The results are shown en Figure (2) (1x)-19. This - va was selected because it is the only region m!:h offers a substantially unrestricted migration path from the suppression pool surface to the dome region, as discussed above in the distributed release analysis. The l

flow behavior observed is very similar to that seen in the free rise and grating analyses, that is, limited horizontal dispersal until grating is encountered. Based on this preliminary

analysis, it has been concluded that igniters ,

should be located as near as possible to the i suppression pool surf ace and that other igniters should be placed abo . them to ensure positive ignition of any hydtugen released or drawn into hs this area of the containment.

c w -

?SO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

O b) Reactor Water Cleanun Area. A release under the RWCU area was si=ulated because this region of contain=ent is a =1xture of grating platfor=s, solid f*.cors which extend al=ost the entire width of the annular space, and s=all subce=part=ents which =1ght pocket hydrogen. The results are shown in Figures (2)(ix)-20 through -23.

The degree of pocketing and dispersal in this area is consistent with the results obtained in the grating and Main Stea= Tunnel analyses described above. All of the R*1CU subco=part=ents are isolated fro = the Reactor Building at=osphere by solid, nor: ally closed doors, except the holding pu=p area a- Elevation 641' 5" (depicted on Figure (2)(1x)-23). This area is, in effect, an enlarged walkway. If the hydrogen release is sy==etric (as assumed in the analysis shown in Figu:es (2)(ix)-20 through -23), then hydrogen will tend to strea= past this area with little or no lateral =ove=ent. A more detailed study using asy==etric release indicates that cross-draf ts could be set up which =1ght draw hydrogen g into the helding pu=p area. This potential W 1ateral =ove=ent is slightly evident on Figure (2)(ix)-23, which shews the start of hydrogen

=igration into the walkway area between the RWCU wall and thc contain=ent shell. Therefore, igniters have been placed in these areas to ensure controlled cocbustion and to preclude pocke ting.

c) Main Stea= Tunnel Area. A release under the Main Stea: Tunnel was si=ulated beause the stea tunnel presents a large, f1:t expanse which is a nearly co=plete obstruction to upward flev. The potential for ta=porary pocketic; offered by such an obstruction sas expectec to be high in this area and is confir:ed by the analysis, as shcwn in Figures (2)(tx)-24 ahrough -23. The actual Main Stea= Tunnel construction calls for an air gap between the concrete floor and walls and the steel contain=ent.

This was si=ulated by leaving two cells open, as shown on Figures (2) (ix)-25 and -26. This gap =ay allcw hydrogen to =igrate into the tunnel, as shewn in Figure (2)(ix)-23, to be g consu=ed by the igniters placed in the upper W areas of the tunnel.

. _ _ J

I 1

1 PSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

The analysis also indicates hydrogen will readily migrate into the walkway area between the RWCU heat exchanger compartment and the steel containment. Igniters will be provided near this area to ensure controlled ignition.

4) Conclusion Based on the preliminary evaluations described above, PSO believes that a sufficient number of igniter locations (as shown on Figures (2)(ix)-3 through -8 can be provided for reasonable assurance that controlled combustion will occur in the containment and drywell well before the locali:ed concentrations of hydrogen could approach the detonable range.

b. Hydrogen Concentration The uniformly distributed hydrogen concentration should not exceed 10 percent during and following an accident or that the post-accident at=osphere should not support combustion.

A postulated accident of the type required by the Hydrogen Control Rule has three major periods:

o Initial RPV blowdown o Hydrogen generation and release with controlled combus tion o Post-accident completion of hydrogen generation i

I and release into an oxygen-depleted atmosphere.

l l The following sections describe the performance i

assessment of the proposed DIS during the hydrogen generation (period 2 above) and post-accident completion (period 3 above) periods for the two release cases considered.

1) SR7 Discharge l Figures (2)(ix)-29 through -34 depict the transient hydrogen and oxygen concentrations in each of the subvolumes used in the HY3 RID combustion analysis. Refer to Figure (2)(ix)-63 for a description of the subvolu=es. Figure (2)(tz)-35 shows the transient uniform =ixed concentrations in the contain=en*,. Inspection i

j

('~ of these figures shows that at no time during

?SO RESPONSE: 10 CFR 50.34(e)(2)(ix) /(3)(v) 2aelod 2 (hydrogen combustion) does either the 9

lots 11:ed or unifor= hydrogen concentration exceed 10 garcent while the atmosphere is capa' ole of supporting comaustion.

At the end of the hydrogen burning period, the post-accident contain=ent at=osphere is a turbulent =ixture of oxygen, nitrogen, water and water vapor, and hydrogen and other non-condensible gases. Continued operation of the spray system rapidly brings the containment at=osphere into temperature equilibrium with spray water. The hydrogen generation process continues to inj ect hydrogen into the contain-ment, raising the unifor:1y sixed hydrogen concentration to approx 1=ately 28.7 percent.

However, the uniformly mixed oxygen concen-tration has decreased to 4.4 percent. 4hich is well below the generally recognized limit for combustion.

2) Drywell Discharge

, The initial period of a dryvell discharge is different than the initial period of an SRV llh release. The blowdown period causes the drywell to pressurize and to eventually clear the horizontal vents. This allows steam and air to enter the suppression pool, where the steam condenses and the air migrates to the contain-

=ent atmosphere. At the end of the Period I b lowdown, it is assu=ed that all drywell air has been transferred to the contain=ent and the drywell atmosphere is 100 percent steam.

During Period 2, hydrogen is released to the drywell and eventually passes through the suppression pool to the contain=ent at=osphere where it is consumed by controlled combustion.

Figures (2)(ix)-36 through -41 depict the transient hydrogen and oxygen concentrations in each of the subvolu=es used in the EY3 RID combustion analysis. Figure (2) (ix)-42 shows the unifor=ly mixed concentrations in the containment. Inspection of the figures shows that at no ti=e during Period 2 does either the localized or the unifors hydrogen concentration exceed 10 percent while the local at=osphere is llk capable of supporting combustier.

PSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

Centinued release of hydrogen during Period 3 (1) raises the unifor=17 mixed hydrogen concen-scion in the containment to approximately 17 percent. The drywell hydrogen concentration is approximately 6 percent. The containment hydrogen concentration is above the 10 percent limit of the Hydrogen Control Rule. However, the uniformly mixed oxygen concentration is 2.3 percent, which is well below the generally recognized limit for combustion.

c. Equipment Qualification The burning of hydrogen in the Black Fox Station containment is expected to result in temperature spikes with high peaks but relatively short total durations. The temperature time histories for various containment subvolumes, as calculated by HY3 RID, are shown in Figures (2)(ix)-43 through -47 for the single point release case and Figures (2)(1x)-48 through -52 for the drywell release case.

These containment subvolumes are defined in Table (2)(1x)-6 and Figure (2)(1x)-63. No drywell tem-perature time histories are provided because no burns occurred in the drywell for the cases con-(]) s ide red.

While the peak calculated temperatures are signi-ficantly above the bulk or average values for the ,

containment which have been used to establish the existing envirectental qualification limits for BFS, the effects of these repeated, short temperature '

pulses and the other environmental conditions created by the burning of hydrogen need not disqualify the existing equipment for service in the containment or drywell. This determination can only be =ade on the basis of detailed evaluations. The results of the required qualification program will be described in the FSAR. The qualification program consists of seven steps:

o Establish the criteria for equip =ent selection and identify the vital equipment list for BFS.

The BWR Hydrogen Control Owner's Group has directed the General Electric Company to under-take this effort on a generic basis. The results will be used as a foundation for identifying BFS specific vital equipment. In anticipation of the Owner's Group report, PSO has performed

() a preliminary review of BFS and established a preliminary list of safety-related systems and 2fu maasbvut l

PSO RESPONSE: 10 CFR 50.34(e)(2)(tr) /(3)(v) cocponents which are located inside containment and are necessary for achieving and =aintaining the safe shutdown of the plant and/or =aintaining contain=ent integrity. All systems which are located, totally or partially, inside the contain=ent vessel were considered. Those

, preliminarily identified in Table (2)(ix)-5 were selected on the following basis:

Function A--Syste= or component =us t function to recover the reactor core.

Function 3--Systec or cocponent cust function to

=aintain contain=ent pressure boundary.

Function C--Syste= or component mus t function to mitigate the consequences of the post-accident events.

Function D--Systecs cocponents whose failure could negatively af fect syste=s or cocponents identified as necessary in accordance with Function A, 3 or C.

Function E--Syste=s or cocponents whose function

=1ght be desirable, e.g., to monitor the course of g the event.

These syste=s and components, which will be reviewed for potential exposure to post-accident environ = ental conditions, are listed by function in Table (2)(ix)-5:

o Calculate the environ = ental para =eters.

This step will establish the transient te=perature and pressure profiles.

o Determine the equipment para =eters. This step vill establish the external equipment parameters (geo=etry, co= position, e=issivity, etc.), the internal para =eters (geo=etry and cocposition), the present qualification limits, the critical components and the expected failure =echanis=s, and the equipment environment (location, exis ting ther=al shielding, etc. ) .

o Evaluate the response of vital equiptent to repeated hydrogen burns and docu=ent the survivability status of the equiptent.

PSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

{sg/ o Compara the results of the analytical models with the performance of equipment exposed to hydrogen flames. PCO notes that tests recently completed at Fenwal Labora-tories on behalf of TVA indicate that typical samples of equipment and materials normally used inside the containment show resistance to the effects of repeated hydrogen burns, o Take corrective action as necessary.

Those vital equipment items for which survivability cannot be demonstrated will be upgraded. Examples of possible actions are:

I= prove the equipment's resistance to surface heat transfer and pressure.

Provide separate ther=al shielding.

Upgrade critical components.

Relocate equipment.

( -

Replacement of equipment with units of demonstrated survivability, o Qualification of the DIS igniters.

d. Containment Integrity A preliminary evaluation of the current BFS contain-ment vessel, including vessel anchorage in the Reactor Building foundation mat, indicates, based on the Hydrogen Control Rule, that containment integrity will be maintained during a condition of an internal pressure of 45.0 psig. This pressure envelopes the peak calculated pressure generated by an accident that releases hydrogen generated by the equivalent of a metal-water reaction which consumes 100 percent of the zirconium metal in the active fuel cladding accompanied by controlled hydrogen combustion. The basis for that conclusion is set forth in the following text.

O 293 18-102181

?SO RESPONSEe 10 CFR 50.34(e)(2)(ix)/(3) (v)

1) Description of the Contain=ent The contain=ent vessel is a free-standing fixed O and vertical cylindrical steel pressure vessel with an ellipsoidal head and a flat bottom steel liner plate. The cylindrical shell is anchored into the concrete foundation =at. The cylindrical steel shell is backed by reinforced concrete in the suppression region to ci:igate structural response due to the hydrodynamic effects of the suppression pool. The physical d1=ensions of the contain=ent vessel are as follows:

o Inside dia=eter of 120'-0".

o Shell height to tangent of 153'-7".

o Ellipsoidal head with a ratio of 2:1 with an inside height of 30'-0".

The contain=ent vessel, including all penetra-tion sleeves, velded attach =ents, and the reinforced concrete backing in the suppression pool area are designed to act as an independent structural co=ponent within the Shield Building. l (l)

Anchorage of the contain=ent vessel is acco=-

plished by extending the vessel shell into the concrete foundation =at for an approxi= ate distance of 6 feet.

• 'ithin a the suppression pool area of the bottos liner plate is a leaktight =e=brane which is designed te resist the hydrodyna=ic effects of the suppression pool. For all other areas, the bottom liner plate serves as a leaktight =e=brane.

The liner plates are continuously supported by the foundation cat. The bottc= liner plate, excep t in the suppression pool area, is covered by concrete which for=s the internal structures to the Reactor Building and which protects the liner plate frca the Reactor Building enviren=ent.

A torodial knuckle plate for=s the transition piece from the containment cylinder to the flat botta= plate in the suppression pool.

Eajor attach =ents and appurtenances to the contain=ent vessel cylinder and head include two personnel air locks, an equip =ent hatch, polar crane girder, fluid and electric syste:

lll penetration sleeves, supports for internal f ra=ing and platfor=s, equip =ent and co=ponent supports, and inspection platforts and ladders.

pSO RESPONSE: 10 C7R 50.34(e)(2)(tr)/(3)(v)

() The base material for the vessel shell, stif-feners, and the bottom liner plates conforms to SA 516 Grade 70. For this evaluation, the vessel shell was assu=ed to have a uniform shell thickness of 1-3/4 inch, which is the maximum shell thickness permitted by the ASME Code without post-weld heat treatment, and no external shell stiffeners. The actual thickness of the vessel shell and extent of the use of stiffening of the vessel will be determined during the final design process. PSO antici-pates that the final vessel configuration which accommodates the design conditions outlined in

?SAR Subsectica 3.8.2 and these supplemental requirements will utilize thinner shell thick-nesses and vessel stiffening to qptimize the vessel design.

2) Applicable Codes, Standards, and Specifications a) Codes. Standards, and Specifications. In

" order to conform with the Hydrogen Control Rule, the following codes, standards, and specifications are used in this evaluation:

O o ASME Boiler and Pressure Vessel Code,Section III, Division 1,1980 Edition l and Addenda through summer 1980, l Subarticle NE-3220, Service Level C limits, except that evaluation of l instability is not required, consi-dering pressure and deal load alone.

o ASME Boiler.and Pressure Vessel Code,Section III, Division 2,1980 Edition with Addenda through summer 1980, Subarticle CC-3720 Liner (Tactored Load Category only).

3) Structural Acceptance Criteria For this evaluation, the structural acceptance criteria for the steel containment vessel are based on the limits for primary stresses defined in Subarticle NE-3220, Service Level C. The i following allowable limits are considered:

O l 295 13-102181 l _ _ . _ _ _ _ _ _

?SL RES?ONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

Pri=arv Stress h

General Membrane Larger of 1.2 S or 1.0 s c Y Local Me=brane Larger of 1.3 S or 1.5 S c Y Sending plus Local Me=brane Larger of 1. 3 S or 1.5 S C 7 Where:

S is the allowable stress intensity for the s? eel caterial S

y is the minimum yield strength for the steel material The structural acceptance criteria for the bottom liner plate are based on the allowables as defined in Subarticle CC-3720, Liners, considering the allowables for factored load conditions .

For concrete and reinforcing steel in the Reactor Building foundation cat and the reinforced concrete backing in the suppression pool, the lll acceptance criteria are based on the allowables as defined in Subarticle CC-3420, Allowable Stresses for Factored Loads. In particular, the allowable stresses for compression, shear, and bearing in the concrete are as specified in paragraph CC-3421. The allowable stresses for tension and cocpression in reinforcing steel are as specified in paragraph CC-3422.

4) Loads and Load Combinations For the evaluation perfor=ed in response to the Hydrogen Control Rule, the following loads and load cocbinations are considered:

a) Containment Internal Pressure.

The contain=ent internal pressure ? ' is the pressure which results from eit$er of l the two following conditions, whichever l

produces the larger load effect for the cocponent being considered.

O l

l l

l

pSO Resp 0NSE: 10 CFR 50.34(e)(2)(Lc)/(3)(v) 3 o Recuired Pressure. The required sj containment static pressure is 45 psig.

o Calculated Internal Pressure. The calculated internal pressure for the 31ack Fox Station containment is the pressure which occurs during an accident that releases hydrogen generated by the equivalent of a metal-water reaction which consumes 100 per cent of the zirconium metal in the active fuel cladding accompanied by hydrogen combustion. The pressure is a time-dependent function. The pressure time-history is computed using the methods discussed in Section 3 and 5 of this response. The contain-ment model consists of six volumes:

two in the suppression pool and wet well region, two in the subcompartment area directly above the respective I suppression pool region, one for the volume above the refueling' floor, and e one for the drywell. Pressure time-j(.)g histories, based on the combustion analyses, were computed for each volume for both the single release t point and the distributed (axially I symmetric) release point cases.

Figures (2)(ix)-53 through (2)(ix)-53 l show the preliminary pressure time l histories for the six volumes inside

! the containment for the SRV release case. The pressure wave for=s are characteristically overpressure impulses of approximately 5 to 10 l seconds in duration. The maxi =um l observed peak pressure is approximately 16.8 psig (31.5 psia) and occurs as the result of a burn in the suppression pool area. The pressure time histories for each cocpartment during the period when this peak pressure occurs have been superi= posed and are presented on Figure (2)(ix)-59. The total period of the impulse is approximately l

10 seconds. The fundamental period

! of the steel containment vessel is l

l 297 13-102131

PSO RIS?ONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v) 0.06 seconds. Therefore, the rela-tionship of the forcing function to lll the dynamic characteristics of the vessel indicate that the effect of the pressure i= pulse is quasi-static and can be co= pared directly to the design pressure stipulated in "Re-quired Pressure" subsection described above.

In addition, inapection of Figure (2)(tc)-5f indicates that significant pressure differentials, i.e., 1.0 psid, do not exist between the various containment volu=es. Theref ore, asy==etric pressure distributions due to burning of the hydrogen are negligible.

The calculated peak pressure of 16.3 psig (31.5 paia) is below the required pressure of 45 psig specified by the Hydrogen Control Rule.

Figures (2)(1x)-60 through (2)(ix)-65 show the preli=inary pressure ti=e histories for the six volu=es inside the contain=ent vessel for the drywell lll release case. The pressure wave for=s are characteristically over-pressure i= pulses of approx 1=ately 5 to 10 seconds duration. The =axi=u=

observed peak pressure is 27.8 psig (42.5 psia) in the contain=ent and 29.3 psig (44 psia) in the drywell.

Due to the dynamics of vent clearing, these drywell and containment pressure transients are separated slightly in ti=e, producing a =ax1=u= drywell-to-containment pressure diffeTential of 5.5 psid which is significantly belev the 30.0 psid design pressure for the drywell. These peaks occur as the result of a burn in the contain=ent region. The pressure time histories for each co=part=ent during the period when this peak occurs have been superi= posed and are presented in Figure (2)(ix)-66 for the contain=ent burn. The shape of this curve is very similar to that resulting frc=

the SRV release, i.e., the pressure lll 298 13-102181

?SO RESP 0NSE: 10 CFR 50.34(eT (2)(ix) /(3)(v)

(~T time-histories are quasistatic and

\/ pressure differentials are cegligible.

The calculated peak pressure of 27.3 psig is below the required pressure of 45 psig specified by the Hydrogen Control Rule.

b) Dead Loads. The deadicads (D) consist of the following:

o Weight of the steel of the contain=ent vessel and its appurtenances.

o Crane weight.

o E=pty weights of attached piping.

o Weight of electrical connections,

=echanis=s, ladders, and platfor=s contributory to the contain=ent vessel shell.

In addition, an equivalent hydrostatic pressure of 25 feet 10 inches in the

(]) suppression pool area, corresponding to the suppression pool inventory following the upper pool du=p which occurs with the Loss of Coolant Accident, is considered.

c) Lead Combinations. The folleving supple = ental load combination applies to this evaluation.

(1) D + ?,'

This load combination is considered in addition to the require =ents of the ASME l Code,Section III, Subarticles CC-3000 and l NE-3000, and Regulatory Guide 1.57.

5) Design and Analysis Procedures

(

a) Steel Contain=ent 7essel. The analysis of

( the containment vessel was carried out by using the contain=ent- vessel =odel developed by Chicago Bridge and Iron Cc=pany (C31). This j :odel is based on a proprietary finite ele =ent computer code, C3I Program E1374, es for the solution of proble=s involving

(_) shells of revolution. This progras cal-culates the deflections, forces,

PSG RESPONSE: 10 CFR 50.34(e)(2)(1x) /(3) (v)

o=ents, and stresses for each output point in the model. lll b) Feinforced Concrete in the Suopression Pool Area. The evaluation of the reinforced concrete in the suppression pool area was perfor:ed using finite-element co=pute r code, Black & 7eatch Progra= S73. In this evaluation, shell elemencs are used to represent the steel contain=ent vessel and axisy==etric quadrilateral ele =ents are used for reinforced concrete backing in the suppression pool region. The progra

calculates the. ti=e histories and the

=ax1=um values for displace =ents, forces.

=o=ents, and stress for each output point.

6) Results A preliminary evaluation of the 31ack Fox containment wassel indicates that the contain=ent integrity can be =aintained within the acceptance criteria outlined in Subsection C.2.d.3 (pages 295 and 296) when the contain=ent vessel is subjected to the required pressure of 45 psig. As indicated above, the required pressure envelops the effects of the peak pressure lll resulting fro = an event that releases hydrogen generated by the equivalent of a =etal-water reaction which consuces 100 percent of the circonium metal in the active fuel cladding accompanied by controlled co=b us tion. Therefore, this preli=inary evaluation -

satisfies the require =ents of the Hydrogen Control Rule Subpart (A) to Hydrogen Control Rule (3 ) (v) .

D. DETAILED ANALYTICAL METHODOLOGY

1. INTRODUCTION The analytical approach used to assess the perfor=ance l

adequacy of the proposed DIS consists of three =ajor parts. These are discussed in detail in the following sections.

In su==ary, the MARCH cc=puter code, supple =ented by hand calculations, was used to derive the = ass, energy, and hydrogen release rates to be used. Using the hydrogen relesse rates as input, the modified SOLA-DF co=puter code was used to evaluate the potential for pocketing under idealized conditions and without considering co= bus tion- llg l

300 18-102181

PSO RESPONSE: 10 CFR 50.34(e)(2)(1x)/(3)(v) induced turbulence. The HY3 RID coc:puter code was used to determine the pressure and temperature response of the containment environment to controlled combustion.

2. FTDROGEN GENERATION RATES PSO has performed a preliminary parametric analysis and has used the results of this analysis as a basis for selecting the release rate shown in Tables (2)(ix)-2 and (2)(ix)-3,

a. Cccouter Code The only publicly available computer code to analyze the combined phenomena reactor heat-up, boil-off, and gydrogen production under degraded core conditions is MARCH .

MARCH was de. nloped by Battelle-Colu= bus for the Probabi-listic Analysis Branch of the NRC Staff. .The develcceent of the MARCH code is an ectension of the meltdown analysis work perfor:3d by Battelle-Colu= bus for the Reactor Safety Study in which the original BOIL code, a srbroutine of MARCH, was developed. Most of the models used in the BOIL subroutine of MARCH art the same as those reported in the Reactor Safety Study.

() 1) Description of MARCH Model The 50IL subroutine calculates core heat-up in an accident where the fission-product decay heat boils the water out of the pressure vessel and uncovers the core. The reactor water volu=e is divided into two regions: sters and liquid. The core is divided into s=all volu=es or nodes. Calculations are perfor:ed to determine the heat produced in each node by perfor=ing heat balances between the fuel ,

and coolant. A steam bo11off rate and the water-steam =ixture level in the reactor core is also calculated.

The reactor core in the Black lox calculations was

=odeled in the BOIL subroutine using 10 radial and 24 axial power zones. The appropriate axial and radial pcuer distributions were used to simulate the axial and radial region pcwer peaking factors. Core nodes in the =ixture region are assumed to be well cooled. Nodes in the stea space are convection cooled by the steam boiling out of the =ixture regions.

kg) 301 18-102131

PSO RES?ONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

The 3 OIL subroutine =odels radiation heat transfer from the top fuel nodes in the core to structures above the core and from core nodes just above the

=1xture to the water region. Four heat structures were modeled above the core for the Black Fox calculation.

These heat structures represented the nenactive top of core; the core shroud doce and steam separators; the steam dryer; the reactor pressure vessel steel; and miscellaneous piping. Three heat structures were modeled directly below the core. These heat structures represented the nonactive bottom of core; the guide tubes and shroud support legs; and reactor pressure vessel steel.

2) Discussion of the MARCH Model The MARCH co=puter code uses conservative assu=ptions and approxi=ations to =odel the behavior of a reactor core and containment system undergoing a degraded core accident. The result is that calculations using MARCH are expected to be conservative with regard to the rate and a=ount of hydrogen generated prior to significant core degradation.

The BOIL subroutine uses the Dittus-Foelter correla- 9N tion to model forced convection steam cooling heat trans fe r. The Electric Pcwer Research Institute (E?RI) and the NRC have undertaken an extensive experimental progras# to determine the actual heat transfer mechanis=s and to establish both best-esti= ate and licensing bases for modeling core heat -

removal. The results obtained to date# indicate that for the conditions expected to exist in a core undergoing significant hydrogen generation, the Dittus-Boetler correlation is conservative. That is, the steam is expected to be more effective at l cooling the rods than predicted by the MARCH code and, therefore, the rate of hydrogen generation should be lower than predicted.

The circonium =etal-water reaction rate is modeled in the BOIL subroutine using the Baker-Just gaseous diffusion for=ulation.6 In each of the fuel nodes, the metal-water reaction is generally a two-step process which is initially controlled by the gaseous diffusion of water vapor toward the hot fuel rod and l

by the gaseous diffusion of the hydrogen away from the fuel rod. At a later ti=e, as deter =ined by the diaceter of the fuel rod, the thickness of the g oxidized layer, and the te=perature of the steam and fuel rod, the reaction rate becomes controlled by the solid-state diffusion of oxygen into the cladding.

i l

l

\

d

?SO RES?ONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

() at which the thickness of the oxidized layer in-creases when solid-state diffusion controls is calculated by the 3aker-Just solid-state diffusion formulation. The use of the Baker-Just diffusion correlations has been found to predict twice the rate of hydrogen nental results. ' production as obtained in experi-i MARCE calculations generally predict that 30 to 60 percent of the active fuel cladding is oxidized during fuel heat-up, prior to the time the core collapses. The range of cladding oxidation results f rom uncertainties in modeling assumptions and the type of fuel heat-up being analyzed. Additional cladding oxidation may occur when the core collapses into the lower plenum water of the reactor vessel.

This additional oxidation would occur very quickly and generally does not produce a large amount of additional hydrogen since the water in the lower plenum would quickly cool the cladding, thereby quenching the reaction.

b. Parametric Analysis

() To evaluate the effects of break size on the hydrogen get.eration rate, three different breaks were postulated.

1 o A 0.163 ft2steam line break, equivalent to a full-open SRV.

o A 2-inch steam line break.

4 l o A 1-inch line break equivalent in size and location to an R?V instrument line, so that the j blowdown would be subcooled liquid until the water level had dropped below the level of the break.

Cumulative hydrogen generation curves as a function of time for each break are shown in Figure (2)(ix)-

67, with the times normalized to the s; art of hydrogen generation. The curve for the 0.163 f t' break envelopes the other two curves over sost of the range.

3. HYDROGEN MIGRATION ANALYSIS The possibility of localized high concentrations of hydrogen (pocketing) in subcompartmented contain=ent structures has

pSO RESFONSC: 10 CFR 50.34(e)(2)(ix)/(3)(v) been identified as an ite= of concern in the Hydrogen Control -

Rule. To address this concern, pSO has perfor:ed a preli=inary hydrogen =igration analysis for the 3FS contain=ent. The g

results of the analysis ere presented in this response.

a. Obf ectives of the Analvsis o To deter =ine the rate of hydrogen buildup in various contain=ent subco=part=ents.

o To ev&3 uate the potential f or hydrogen =aldistribu-tion and pocketing.

o To provide a rationale for selecting the nu=ber and location of igniters.

o To assist in developing an igniter control philosophy.

b. Cc= cuter Code The evaluation of =ulti-co=ponent gas flows featuring both asy==etric (SR7) and axially sy==e tric (dryvell vent) dischargss into a subco=part=ented closed sgructure was perfor:ed using a =odified version of SOLA-DF , a public do=ain solution algorith= for nonequilibriu= two-phase flew. In short, SOLA-DF is a finite difference code which uses the i=plicit continuous fluid Eulerian g'

cethod to solve the = ass, =ccentu=, and energy equations which describe the syste= under evaluation. To provide a

= ore cocplete and flexible analysis of the hydrogen

=igration crebles, the original SOLA-DF code has been

=odified a include three-di=ensional capability, rec-tangular as well as cylindrical coordinates, and the necessary constitutive relationships to describe the behavior of hydrogen-air =ixtures.

For each of the cases described below, the contain=ent was the area of interest, since the drywell is effectively sealed to ordinary gaseous inflows. Hydrogen gas (all stea= vas assu=ed to be condensed) was released at a te=perature of 100 F into an initially quiescent at=osphere also at a te=perature of 100 F. No heat transfer between the air and the various structures or the suppression pool was per=itted. The release rate for hydrogen was the sa=e as that used for the cocbustion analysis. The area of release for the single point release was twice the ci;-

cu= scribed area of the SR7 quenchers, or about 160 ft'.

~he dispersive effects of buoyancy, =ocentu=, convection, diffusion, te=p e rature, pressure, and gravitation were included in the analysis.

g Turbulence induced by the sprays and the controlled com-bustion of hydrogen were not included in this analysis.

PSO RESP 0NSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

Neglecting these s!fects in the migration analysis is a

() conservative assu=ption as these effects are expected to increase hydrogen dispersal and thereby further reduce the potential for local pocketing.

750 has performed a high te=perature release analysis, in which a ' hot" hydrogen release was si=ulated by discharging the hydrogen at 1642 F into an atomosphere at 10 0 F.

The upward velocity of the hydrogen increased by a factor of about 3 over the low temperature case, and horizontal dispersal was reduced.

c. Description of Completed Cases To meet the obj ectives of the preli=inary analysis, the follcwing cases have been evaluated:

o Three-dimensional evaluation of pocketing due to grating in the annular volume between the suppres-sion pool and the refueling floor resulting from single-point discharge of hydrogen gas in thermal equilibrium with the suppression pool.

o Three-dimensional 360-degree evaluations of the containment without fans operating. A single-point

'(]) release under the RWCU equipment area was simulated.

For the drywell release case, uniform discharge through the vents was simulated.

o Three-dimensional, 90-degree evaluations of the annular region for three single-point releases.

These were the RWCU equipment area, the Main Steam Tunnel area, and the Equipment Removal Hatch area.

These finely nodalized studies provide a detailed evaluation of the pocketing potential in both cpen and congested areas.

4 HYDROGEN COM3USTION ANALYSIS The BFS contain=ent pressure / temperature analysis was done using the Computer Code HY3 RID. The HY3 RID code models the suppression pool and vent flow between the drywell and sup-pression pool. Other features of the code include a variable heat transfer spray model, entrained water fallout model, heat transfer to specified heat sinks, and the simulation of various engineering safeguard equipment such as fans and heat exchange rs.

HY3 RID can simulate the multicompenent (H , 02 ' N , and CO gas and two-phase fluid transfer between compartments 7 2 due !o)

() the burning of hydrogen and/or due to the mass and energy release from a pipe break. The computer code can simulate multicc=partment (up to 100 volumes) transis .c pressure and temperature responses and track the distribution of the non-condensible gases.

pSO RESPONSE: 10 CFR 50.34(a)(2)(ix)/(3)(v)

The =odel used to deter =ine the pressure /te=perature responses due to controlled burning by the DIS was a =ulti-node =odel lll which divides the 3FS Reactor Building into discrete volu=es based on flow area and natural divisions to flow. The nodal diagra= is presented on Figures (2)(ix)-63, and the associated volu=e descriptions are given in Table (2)(ix)-6. The HY3 RID

odel contains six co=partments, a ce=pression pool at the botto= of Volu=es 5 and 6, vents connecting the drywell to volu=es 5 and 6, containment spray with spray carry-over, and a vacuu: breaker st=ulation.

The flow paths connecting the ce=part=ents are represented as shown on Figure (2)(ix)-63 by arrows pointing in the direction of allowed flow. The junction (flow path) para =eters are presented on Table (2)(ix)-7. The junction flow areas repre-sent the =ini=u= flow area of the connecting co=part=ents.

The junction loss coefficient calculations were done by eva ing the obstruction losses using the handbook by Idel'chik. { gat-Included in the loss coefficient calculations are the losses through grating and other losses due to =iscellaneous obstruc-tions. Ef fects of flow inertia were also included in the HY3 RID calculations.

The vents and suppressica pool and related para =eters are shown in Table (2)(ix)-8. Included are the total voluce of vater at the ner=al vater level, pool surface area in the wet lll well and drywell, nor=al pool height above base =at floor, and the drywell weir height above the nor=al water level. Other related para =eters include specifications for the vents, drywell holdup vciu=e and upper pool du=p para =eters.

The drywell vacuum breaker is repiesented as a one-way flow cath in the diagra= presented on Figure (2)(ix)-68. The drywell vacuu=-relief line opening is a function of the vacuu:

breaker valve, butterfly valve, and associated controls. At 2.0 psid (contain=ent to drywell), or at 0. 2 psid (contain=ent to drywell) if the drywell pressure is above 2.0 psig, a signal is generated to open the butterfly valve. Af ter a 3-second delay, the butterify valve opens within 5 seconds and re=ains open until closed by operator action. Once a butterfly valve is open, the vacuu: breaker valves are activated. At 0.2 psid (contain=ent to drywell), the =agnetic latch on the vacuu: breaker valve releases and the disc i:=ediately swings wide open. The disc re=ains fully open (area is 0.5475 f t')

until the differential pressure falls to about 0.1 psid.

Selow 0.1 psid, the disc is partially open until it resents at approxi=ately 0.02 to 0.03 psid. The equivalent loss coeffi-cient for the vacuu=-relief lines is 5.51.

O

pSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

• he containment spray system parameters used in the 3FS HY3 RID

> calculations are presented in Table (2)(ix)-9. The spray is released from tao spray rings located in the contain=ent dome.

The inner ring is located approximately 59 feet above the operating floor, and the outer ring is located approximately 48 feet above the operating floor. The spray falls from the spray rings through the containment dome until the spray pattern is disturbed by various obst' ructions. These obstructions consist of storage pools, the reactor well, drywell head storage area, reactor head storage area, R*JCU heat exchanger removal hatches, and gratings. Part of the spray will collect in the upper pool and is assumed to drain directly into the suppression pool. A large part of the spray will strike the obstructions and agglomerate, forming large masses or sheets of water which either flow directly down into the lower compartments or run down the walls of the compartments. It is assumed for these calculations that a small fraction of the spray remains as the initially specified spray droplets, and the effective carry-over spray flow fraction was conservatively esti=ated at 10 percent.

The passive heat sinks used for these calculations were the containment steel shell adjacent to the containment and asso-ciated compartments and the concrete and steel in the drywell.

/' - Only steam-condensing heat transfer was taken into account, using the Uchida correlation. Radiant heat transfer was not considered in these analyses. Noglecting radiant heat transfer produces higher compartment temperatures to minimize the heat transfer to the structures, as was done in these analysis.

Two analyses were performed to determine the effectiveness of

, the DIS to reduce the hydrogen concentration, namely the

+ ' stuck-open SR7 which releases the hydrogen directly into the l suppression pool, and the same release directed to the drywell-l (see Figure (2)(ix)-63 for release points with respect to the l RYBRID burn model). The mass and energy release rates for water and hydrogen used for both analyses are presented in Tables (2)(ix)-2 and (2)iv)-3, respectively.

The results of the HY3 RID burn analyses are presented in Section C. of this response.

307 18-102131 l . _ - . . . - _ _ - - -. .- . - . . . _ . _ - . - - _- - .-

- PSO RESPONSE: 10 CFR 50.34(e)(2)(ix)/(3)(v)

REFERENCES

1. Roger O. Wooton and Halil I. Avci, " MARCH (Meltdown Accident Response Characteristics) Code Description and User's Manual," BattelIe Columbus Laboratories, NUREG/CR-1771, BMI-2064, October, 1980,

2. M. P. Sherman, et_ al. , '*Ihe Behavior of Hydrogen During Accidents in Light Water Reactors," NUREG/CR-1561, SAND 80-1495, R3 (August, 1980).

3. NUREG-75/Oll, WASE-1400, " Reactor Safety Study, An Assessment of Accident Risk in U.S. Commercial Nuclear Power Plants," October, 1975.

4 Wooton, R. O. , " Boil 1, A Computer Program to Calculate Core Heatup and Meltdown in a Coolant Boilof f Accident," Battelle Columbus Laboratories (March,1975).

5. Separate-Ef feets Tests and System Ef feets Tests," EPRI NP-1460, h " Analysis of the FLECHT SEASET Unblocked Bundle Steam--Cooling and

5. "FLECHT SEASET (Full-length E=ergency Core Cooling Heat Transfer-Separate-Ef fects Tests and System Effects Tests," EPRI NP-1460,

" Analysis of the FLECHT SEASET Unblocked Bundle Steam--Cooling and Bo11of f Tests," NUREG/CR-1533, May,1981.

6. L. Baker and L. C. Just, " Studies of Metal-Water Reactions at High Temperatures III. Experimental and Theoretical Studies of the Zirconium-Water Reaction. " Argonne National Laboratory, ANL-6548 (May , 1962) .

7. Duncan, J. D. and Leonard, J. E. , '*Ihermal Response and Cladding Perfor=ance of Zircaloy Clad Simulated Fuel Bundles Under High Temperature Loss-of-Coolant Conditions," GEAP-13174, (May, 1971) .

8. R. O. Wooten, et al. , " Analysis of the Three Mile Island Accident and Alternative Sequences," Battelle Columbus Laboratories, NUREG/CR-1219 (January,1980).

9. C. W. Hirt, et al, "SOLA-DF: A Solution Algorith= for Nonequilibrium Two-phase F17,"JUREG/CR-0690, LA-7725-MS (June, 1979).

10. I. E. Idel'chik, Handbook of H'rdraulic Resistance, U. S. Department of Commerce, AEC-TR-6630 (196o).

308 13-102131

n TA3LI (2)(ix)-l

. c =. c v. m

.~s_=,<

v) g _ _~w .. , =. .. =_D , C .'. w

' _v. ;G , .s' .',2 .o '._s' .\"..".D T. _* a=. .~a Condue:c: Soonsor Status Goal T'.'A !7A Completed Enamine perfor=ance of GM glow and plug, and conduc- 1'#a **-e Cngoing tes:s on various ignitars.

Fenwal AI?/DUKI/T7A Co=pleted H., combustica tests designed to look at condi-ions si=ulating in contain=en:

environ =ent (=ostly quiescent cha=ber tests)

LLNL NRC Cc=pleted H., co=bustion tests to look at LFL under high stea=

Icadings in a well-nixed quiescent chacher.

AICL EPRI/AEP/ Cnzoing

- Iffect of :urbulence on E 2

DUKZ/TVA cochustion; igniter effectiveness studies ( ostly well-cixed chamber tests)

HEDL EPRI/AI?/ Ongoing Mixing, stratification, and DUKI/TVA distribution of H.,

following LOCA acciden: (verv large cha=ber bu t no:

. cocbustion)

ACURIX IFRI/AIP/ Ongoing Igniter location effect during ACUREX/DUKZ/ dynamic injection of H I 2

TVA suppression characteristics of microfogs during dynamic injection of H.3; equipnen:

survivability. "

SANDIA NRC Cngoing Easic and applied research into H co=bustion; effects 2

of microfoss in vell-=ixed quiescent cha:bers.

T.'.RC EPRI/AZF Cngoing LFL of H, in :he presence of DUKE /T7A nicrofogs.

7ARICUS TIC (IOCCR) Planned Develop adequate technological p

s basis for decision :sking for Degraded Core Rulenaking.

3K@@ R#sil#4URL

TA3LI (2) (ix)-2 3 LACK FOX RIACTOR CCC1.tr .*>.SS A::D II.T.RGY RELIASE R.CIS

(-)

v Ti=e Mass Release Rate Energy Release Rate **

(Min)

(lbm/=in) (3tu/ Min) 0.0 1.3 6 x 10' 1.62 x 10 7

.106 2.48 x 10' 2.95 x 10 5.94 2.43 x 10' 2.95 x 10' 6.0 7 1.51 x 10' 1.83 x 10 4

10.0 1.29 x 10 1.54 x 10 7 14.0 9.31 x 10 1.16 x 10 7 3 6 13.0 5.8 x 10 7.80 x 10 22.0 3.77 x 10 5.40 x 10 6 26.0 2.61 x 10 3.92 x 10 0 30.0 1.38 x 10 2.96 x 10 0 34.0= 3.58 x 10 3 4.27 x 10 6 38.0 4.32 x 10 3 5.75 x 10 6

/~ 42.0 6.15 x 10 3 7.34 x 10 6

b 46.0 7.19 x 10 3 3.59 x 10 6

30.0 6 7.98 x 10 9.53 x 10 54.0 7.21 x 10 3 8.61 x 10 6

6 53.0 7.54 x 10 9.00 x 10 62.0 0 6.62 x 10 7.90 x 10 70.0 0 4.96 x 10 5.92 x 10 34.2 4.91 x 10 3 5.86 x 10 6

34.21*x= 2.09 x 10 3 2.50 x 10' 150.0 1.65 x 10 3 1.97 x 10 6

• March cu:put =odified at 34 minutes to include enough ICC flev to recove energy produced by decay heat and retal-vater reaction.

"~ne fissica product decay heat was not identified separatel/ since the total decay heat was included in :he energy release rates.

• • • Eydr: gen generatica ends.

O 310 13-102131

(

.s

..%14. 0,4)(,1%,*J

.) ,

Os -

• T %'**fs -e.V. a JV *-*' *. *.*.v e r*ua.VI *%A_

. **.0

4 .s* b r. .A 1 .t .m w *m

. O.* .3 .'*L**

. s*

., *. *W .%; O Ti e Hyd: ren Ra' ease 22:e Te : era:ure (sin) (15 /:in) (?)

0.0 0.0 534

~

12.wa

. 1 . .' ', x . 10 7. *eS. ,

r

. 0 449 1,001 1'

. .O 1. .? 4 34.0 14.3 1,642.0*

2 3 . n. .s.*

i si - - '.n 42.0 63.3 1,64 I-6.0 32.,7 1,642.0 I

50.0 97.9 1,642.0 i

54.0 35.3 1,642.0 l 58.0 92.0 1,642.0 i i 62.0 76.4 1,642.0 1

70.0** 48.5 1,642.0 i.

l l 24.7 4c^.E 1 4.4

{

l 2 .,s . .e .' o.o 1,44,.o

..u i 150.0 0.0 1,642.0 i

I "Hyd::ge: :e:pera:ure was assu=ed ccustan: when core begins to nelt, i

I

"*A: 70.0 minutes the =etal-vater reac:1:n bece:es severely " ' > b*r I the a:cun of flashing stea=. ?::= :his pein: onward, the reaction rate

! vas assured :: be constant until all of -he ac-ive :ir::nium clad had reac:ed.

311 .-. .

13-102131

1 TA3LE (2)(ix)-4 i

3I.ACX FOX STATION 3ASE CASE COM3USTION PARAMETERS Hydreren Lean Volu=e percan: 3 for initiation 8.0 percen:

3

Vole =e percan

O g for initiation 5.0 percen:

7elu=a percen: H, fer propagation '

greater than 0.0 percen:

(upward) , greater than 9.0 percent (horicantal and downward Fla=e speed 6 feet per second 3urnup (of available hydrogen)* 85 percen:

i i

3urnup (of available oxygen)* 100 percen:

S:ea= effects Jilutent only=*

i

(~

, Orvren Lean (Dev Well Oniv)

! Volu=e percent H f: initia:1 n Less than 90 percen:

2 Volu=e percen: 0 f: initiati n 5.0 percent 2

! Volu=e percent stea= for ini:ia:1on Less than 60 percen:

i

! Fla=e speed 6 feet per second I Surnup (of available oxygen) = 100 percent I

• For individual ignitions, s:cichio=etry is =aintained. For exa=ple, if :here is sufficient oxygen to initiate a burn but insufficient to consu=e all available hydrogen, the total burnup is li=ited by the available oxygen.

• =Staa and water vapor are treated as dilutants for co=bustion purposes. The effects of va:ar as a hea:-absorbing =aterial are included in ca.'.culating the pressure and te=perature response.

i i

1 1

312 13-102131

TA3LI (2)(ix)-5 (VD SYSTIMS, COMPO:E . S, OR STRUCMIS RIQUIRED FOR w,. . -: . c ac D c+a-s.,, A.sO u_

. h- ,.u.Lr.s 4.s.-.u L. 0.%.,. ..

. sm. .s. . . .%- u n.u Svste= Tunction*

(A) (3) (C) (D) (I)

Autc=atic Depressuricatics Syste= I I RER Syste: ,

Centain=en: Spray Mode I Suppression ?cci Cccling Mode I Shu:devn Ccoling Mede I Standby Service Wa:e Syste= I Centain=er: and Reac:cr Isolatic: Syste=s I I Centainnent Vacuu: Kelief I Ory Well vacuu: Relief I

()

Suppression ?cci Makeup I MSIV ieakage- Centrol I Distributed Igni:ica Syste= I Eydrogen Recc=biners I Fes: Accident Moni:cring Systa=** I Centain=en: A:nospheric Moni:cring Syste I Eigh Pressure Core Spray I Low Pressure Core Spray I Ilectric Power Distribution (Cable) I 5:andby Gas Treat =ent Syste= I "The func icns A, 3, C, D, and I are defined in the text of See:ica C.2.c.

""Exa=ples cf desirable post-accident noni:cring instrumentation loca:ed

'g- inside contain=ent include: centain=en: pressure, reac:cr water level, N suppression pocl vater level, and hydrogen moni:cring instrunenta:icn.

TABLE (2)(ix)-6  :

O stiCx 10x Ccxrian:Ex: ossCarrT10ss Vole =e aelative Initial Nu=her* Volume Teenerature Humiditv Pressure Descriotion

(., . 3 ) (F) (psia) 1 551,623 90 .70 14.7 Containment Doce Above l El 666'-5" l 2 274,310 13 5 .70 14.7 Drywell l

3 142,911 90 .70 14.7 Containment between El 610'-4" and El 666'-5" frem Az.

46 degrees to Az.

- 314 degrees 4 49,587 90 .70 14.7 Containment between El 666'-5" and El 610 '-4" f rom Az .

314 degrees to Az.

46 degrees

();

5 126,115 90 .70 14.7 Containment between El 610'-4" and top cf water from Az. 46 degrees to Az. 314 degrees 6 43,759 90 .70 14.7 Containment between l

El 610'-4" and top of l water from Az. 314

! degrees to Az. 46 degrees i

l *See Figure (2)(tz)-63 for nedal diagram.

1 1

314 18-102381

1 Y

i

() TA3LE (2)(ix)-7 DESCRIPTION OF SLACK FOX JUNCTION PARAMETERS Junction Volu=e to volace Minimus K ** K **

Nu=ber*

i d Area id di L/A. ft

-1 (f t )

i 1 1 3 1,251 3.05 3.25 .0288 2 1 4 402 3.05 3.25 .0879 1

3 3 5 1,053 1.34 1.72 .008 4 4 6 366 1.84 1.72 .023 5 3 4 251 434 .370 .233 6 5 6 388 .042 .040 .161

• See Figure (2)(ix)-63.

• • Loss coefficients are based on =inimum area. ,

4 1

1

()

I 315 18-102131

l

[

l TA3LI (2) (ix)-8 3 LACK FOX SU??RI5520N PCOL A?O RILATED ?A?.AldETERS Fool Water l

l 2ensi y (ib=/f ) 62.2 2

i Volume (f:') 133,672 Te=perature (F) 100 Heat capacity (3:u/lb-7) 1.0 Fool surface area 1 ve: well, f:2 5,399 9

? col surface area in dryvell, f ~ 432 l

i Nor=al pool height above base =at floor, ft 20.2 i

Weir height above water level, f: 5.92 Vents Row 1 Row 2 Row 3 l Nc=ber of vents 40 40 40

! 7 Flow area per vent, f:~ 4.12 4.12 4.12

( Vent length, f 5.0 5.0 5.0 Water depth at ven bo::c=, f: 8.4 12.9 16.3

, Iquivalent ven: length added 'or l inertia effects, f: 2.86 2.86 2.S6 i

l Turning less coefficien: 1.2 1.2 1.2 Dry Well Holdup voluce*, f: 44,855 Holdup surface area, f:2 2,336 Cecer Pcol**

Volu=e du= ped to suppressien pool, f: 34,150 Cu=p ti=e, =inutes 6 Water ta=perature, F 100

• Net free volume belev and inside : p cf weir vall in drfvell.

)

"*The upper peci vill be du= ped aut::a:1cally 30 minutes af:er a LOCA ce after a LCCA when the suppressi:n pool level d. ps c 1:v-1:v va:er level.

TA3LE (2)(ix)-9 3 LACK FOX S? RAY SYSTO. PA?r.ETIRS b

v F10v rate, sp= per spray loop 5,250 (See Note 1)

Te=perature, ? 132 Crop dia eter, =icrons 350 (See 'bte 2)

Fall ti=e, seconds See Note 3 Hea: ::ansfer coefficient, 3:u/h-ft'-? See !b t e 4 Contain=en: dc=e to lower cc= par =en:s carry-over fraction .1 Initiation See Note 5 Ti=e to at:ain full flew, =inu:es 3 Ter=ination Operator Actica

bres:

1. Only one loop was assu=ed operational for these calculations.

2. The arith =e:ic =ean drop dia=eter for a Sprace 1713A nos:le.

(])

3. Fall ti=a is dependent on local conditions in the co=part=en: and height of co=part=ent.

4 This value is based on the local conditions of the co=part=ent.

5. Spray ini

iation is when 10 =inutes have elapsed af:er the drywell i reaches 2 psig, or the spray will be auto =atically initiated if the i containment pressure is grea:er than or equal to 9 psig, or =anually l initiated by the operator.

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                                                                                                                                                                                                                                                    'iGilRE (2)(ix)-12

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