ML19220B932
| ML19220B932 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/02/1974 |
| From: | Tedesco R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7904270672 | |
| Download: ML19220B932 (9) | |
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Docket No. 50-320 Voss A. Moore, Assistant Director for Light Water Reactors, Groun 2, L REQUEST FOR ADL fl0NAL I';TORMATICN FOR THREE MILE ISLAND NUCLEAR STATION, UNIT 2 Plant Nane: Three Mile Island Nuclear Station, Unit 2 Decket No.:
50-320 Licensing Stage: CL MSSS Supplier: Babcock & Wilcox Architect Engineer: Burns and Poe coneninmant Type: Dry Responsible Branch & Project Manager: LWR 2-2; B. Washburn Requested Completion Date: August 2, 1974 Applicant's Response Datei October 11, 1974 Review Status: Awaiting Information The enclosed request for additional information (Q-1) for the Three Mile Island, Unit 2 operating license raview has been prepared by the Containnent Systens Branch af ter having reviewed the applicable portions of the FSAR.
In the coursa of our review we have identified the following as significant review itens:
1.
The applicant has not addressed potential additional energy release from the steam generators af ter reflooding of the core in his enalysis of a postulated reactor coolant system cold leg pipe break. We vill require the applicant to include post-reflood, steam generator energy release in his centain--
ment response nnnlyses; and 2.
The applicant has not completely described his methods of analyses to deternire the subecupartment pressure response.
A draft of these questions was provided the Project Manager on.!uly 5, 1974. A meeting was held with the applicant on July 25, 1974 to discuss the significant unresolved areas.
Origint signed by:
Robert L. TMa m 7 0 0 4 2 7 0 6~17._
Ro3ert t. Tedesco, x,ststant 31 rector for Containm,nt Safety Directorate of Licensing
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ep tated cc:
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Voss A. Moore.
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F. Schroeder S. Hanauer J. Glynn R. Klecker D. Eisa9.ut S. Varga J. Carter B. Washburn G. Lainas D. Shtra L - Reading CS - Reading CSB - Reading Docket File No. 50-320 9
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03-1 03.0 CONTAI' BENT SYSTDIS 31GCH
- 03.1 The response to Question 9-5 in the preliminary review (6.2.1.1) report concerning the contain=ent subco=part=ent pressure response is ince=plete.
Provide the following information:
(1) As a minimum, the diff erential pressures resulting f rom the following pipe brez.ks and break locations should be analyzed:
(a) Hot and cold (pump suction and discharge) leg RCS ruptures in the steam generator cc=partment, reactor cavity and prt:ary shield wall pipe penetrations.
(b) Pressurizer spray and surge lines in the pressuri-zer compartment.
(c) Steam, feedwater, and/or any other lines carrying high energy fluid with the potential to over-pressurize co=part=ent walls, barriers and floors the failure of which might effect the perfor=-
ance of equip =ent necessary for the safe opera-tien and shutdown of the plant or contain=ent integrity.
(2) For each compartment analyzed:
(a) Describe the nodalization sensitivity studies per-formed to determine the mini =ua at=Scr of vole =e nodes required to conservatively predict the maxime= pressure for each subcompart=ent.
The nadalization sensitivity studies should include consideration of spatial pressure variation; i.e.,
pressure variations circe:ferentially, anially and radially within the subcompart=ent, par-ticularly in the reactor cavity.
(b) Provide schematic drawings showing the nodali-sation of each subce=partment or compart=ent sub-division indicating nodal net f ree volu=es and interconnecting flow path areas.
(c) Provide sufficiently detailed plan and section drawings for several views showing the general arrange =ent of subcompartment-structures, components, piping, and other major obstructions fres which suber=partment volumes and flow areas can be detarrin22.
84~076
03-2
- 03.1 (d) Provide and justify the break type and area (6.2.1.1) used in the analysis.
(e) Provide and justify values of vent loss co-efficients and/or friction factors used to cal-culate flow between nadal volu=es. When a loss coefficient consists of more than one component identify each co=ponent, its value and the flow area at which the loss coefficient applies.
(f) Discuss the =anner in which sovable obstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated.
Include analytical justification if credit is taken for the re= oval of such ite=s to obtain vent area. Provide assurance that vent areas will not be partially or completely plugged by displaced objects.
(g) Provide a table of blowdown mass flow rate and energy from that postulated that would result in tne highest differential pressure for each compart=ent.
(h) Provide a curve of differential pressure as a function of ti=e indicating spatial response for the analysis of the reactor cavity.
(i) Provide the design differential pressure, peak calculated differential pressure, and time of peak pressure for each co=part=ent.
(3) With regard to rethod of analysis for subcompart=ents:
(a) Provide the na=e and a detailed description of the blowdown and pressure transient code (s) used in the analysis.
The description should include all mathe=atical correlation to deter =ine the subsonic and sonic vent flow.
(b) Justify the blowdown =odel used shewing that it maximices the mass and energy release rate.
(c) Provide and justify the critical flow =odel used in the blowdown analysis and the break discharge coefficient applied.
(d) Provide and justify, preferably by comparison with crperi= ental data, the equation or correlstion used ta
- ' "'~
etween m._ro.__ents.84-077
03-3
- 03.0 Include a discussion of the critical flow (6.2.1.1) odel and discharge coefficient applied to critical flow.
(e) Discuss the method of treating the air-stea=-
water =1xture in subco=part=ent ther=odyna=ics.
- 03.2 If sand plugs are utilized as radiation shieldings inside (6.2.2.2) the contain=ent, provide the following infor=ation:
(1) Describe the sand plugs identifying the materials of construction. Provide drawings to indicate design details of the structures.
(2) Describe and discuss the design provisions =ade to ensure that the sand plugs do not becc=e da= aging
=issiles following a LOCA.
(3) Discuss the design provisions of the contain=ent su=ps, spray no::les, pu=ps and other safety equip-
=ent to accc==odate the release of sand to the con-tair=ent at=osphere.
03.3 Table 9.2-2 shows that a significant amount of energy (None) stored in the Steza generators has not been released to the containment during post-reflood period. Justify that this energy re=ains in the stea= generators.
'Je find that nucleate boiling will exist on the stea= generators primary side for reverse heat flow during the blowdown, reflood and post-reflood periods.
- 03.4 Provide the following information for a double-ended pu=p (s.2.1.3) suction break:
(1) Analyses of = ass and energy rates released to the containment to include the following:
(a) =ultinoding in primary =etal slabs; (b) no quenching frc= LPI pe=p flow; (c) no quenching fro = EPI pu=p ficw or flood tank flow af ter blowdown; (d) nucleate boiling on primary side of stea=
generators and condensation on secondary side for reverse heat flow; (e) all available seca: gaaeratoc cau;gy usin; the assumption that nucleate boilin; exists in the
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stea: generntor cu es for revarse acc: flo..
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O 03-4
- 03.4 (2) The results of the containment pressure transient analyses to include the above energy release.
- 03.5 Thc TSAR indicates that the cocputer programs, FLASH and (6.2.1.2)
PRIT were used for hot leg break analyses and CRAFT was used for cold leg break analyses to predict the mass and energy releases to the contai=nent during blowdown and post-blowdown.
Because CRAFT has been found by us to be acceptable for these analyses, justify the use of the FLASH and PRIT codes for the hot leg break mass and energy release rates to be adequately conservative for containment pressure analyses.
- 03.6 Describe and justify the =esh spacing used to determine the (6.2.1.3) heat transfer to the containment concrete heat sinks. We currently believe that about 0.1 inch mesh spacing should be used for the initial 3 to 4 inches of concrete.
- 03.7 Describe and justify the assumptions used to =odel the heat (6.2.1.3) transfer at the contain=ent liner to concrete interface; i.e., consideration for contact resistaace or effects of air gaps.
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03.8 The max 1=us heat transfer coefficient of 620 Stu/hr-f t
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(6.2.1.3) obtained frca Kolflat's work was utilized in your contain-ment pressure response analysis during the blowdown phase of a LCCA. This initial high heat transfer coefficient for a large contai=nent was not substantiated by the work reported by Tagami.
Provide an analysis of the contain=ent peak pressure utilizing heat transfer coefficients during the blowdown phase of the accident which are consistent with the data reported by Tagami.
In addition, provide the heat transfer coefficient which was used in your pressure response analysis as a function of tine after LOCA.
- 03.9 With regard to the reactor building spray syste=, provid e (6.2.2.2) the following infor=ation:
(1) A discussion of Le NPSH require =ents for the R3 spray pu=ps with supporting infornation (i.e., static elevation head, the friction head loss in the suction piping, the vapor pressure of the fluid and the reactor building pressure) to show the cargin between the required and available NPSH to demonstrate conform-ance to the guidelines of Regulatory Guide No. 1.1.
(2) A discussion cf the design erevisiens -hich rermit the spray water that may enter the refueling cavity 84-079
03-5
- 03.9 and the reactor cavity following a loss-of-coolant (6.2.2.2) accident to be drained to the contain=ent su=p.
- 03.10 With regard to the housing and ductwork for the reactor (6.2.2.2) building fan cooler system, provide the following informa-tion:
(1) An analysis of the pr2ssure differential for the housing and ductwork.
(2) The design pressure and the calculated differential pressure as a function of ttne for the housing and ductwork.
(3) A detailed description of the analytical =odel, assenptions and appropriate bases used in calcu-lating the pressure differentials.
03.11 Analyze the x:ount of contain=ent atmosphere that will be (9. 4.15. 2) released to the environment if the containment purge vent valves are open at the time of a loss-of-coolant accident.
03.12 Describe the essential specifications used in the design (9.4.15.2) of tne contain=ent vent and purge valves (i.e., design temperature, pressure, diff erential pressure, radiation, and dynamic forces). Describe analyses or tests that are conducted to demonstrate that the valves will operate as specified.
- 03.13 Describe the provision for =enitoring hydrogen in the (6.2.5.2) reactor building following c LOCA.
The discussion should include the following de.=ign considerations:
(1) the number and location of sanpling points within the reactor building; (2) seismic design classificati2n of the system.
03.14 Identify those syste=s or porticas of syste=s which will (6. 2. 4) be open or vented to atmosphere and/or drained of fluids to assure that isolation valves vill be exposed to contain-ment air pressure as required by Appendix J.
Those systems not vented or drained and which for= a part of the contain-sent boundary shculd be identifiea and the reason with justification for not venting and draining should be stated.
- 03.15 Airlocks should be press"ri ed to ? folicving the sane (6.2.1.4) schedule as preoperationsi and per10di7 Type A tests.84-080
03-6
- 03.15 Every six =1nths during periods when airlock doors (6. 2.1. 4) are not opened the airlock door and penetration seals should be tested at the door manuf acturers reco== ended test pressure. During periods of door use when the reactor contains fuel, the door seals should be leak tested every three days.
- 03.16 Specify the acceptance criteria for the =easured contain-(6. 2.1. 4)
=ent leak rate CL ) at peak pressure and the acceptance criteria for the $5nta1==ent integrated leak rate verifi-cation test.
- 03.17 You state in Section 6.2.1.4.2.3 that the measured leakage (6.2.1.4) rate (L(E ).
shall not exceed the =axi=um allowable leakage
)
We will require in accordance with Appendix
- rate, J that (t ) shall not exceed 0.75 L.
- 03.18 Tor the Type A test you indicate that (16) te=perature (6.2.1.4) detectors and (2) dewcalls will be utilized. To provide more accurate ta=perature and hesid}ty values, one temperaturedetectorper100,0g0ft of contai= ent volu=e and one devcell per 500,000 ft of volu=e should be provided.
- 03.19 You indicate that for the contain=ent integrated leak test (6.2.1.4)
(Type A) the reduced pressure test will be performed sub-sequent to the peak pressure test.
In order to minimize stabilization time and reduce off-gassing due to entrain-
=ent the following test sequence should be provided:
s (1) reduced pressure test; (2) structural integrity test; (3) peak pressure test.
- 03.20 With reference to Technical Specificatien Section 4.4.1.2.3, (16.4.4.1) you state that local leak testing shall he perforned at each refueling.
The frequency of local leak testing should not exceed two years as stated in Appendix J.
- 03.21 The Technical Specifications shculd state co=pliance with (16.4.4.1)
Appendix J of 10 CFR 50.
- 03.21 With reference to Section 4.4.1.1.5.C, you state that if (16.4.4.1) two consecutive Type A tests f ail to meet acceptance criteria, a Type A test shall be performed at each unit shutdotin for refueling. The followup test 12 required by Appendix J te be performed at each plant shutdown for refueling or approxi=ately every is months whicnever occurs first.84-081
03-7
- Information requested of si=11ar plants.
Reco=2 ended positions e
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84 ~ 08'3