ML19309A458
| ML19309A458 | |
| Person / Time | |
|---|---|
| Issue date: | 08/04/1978 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Sherwood G GENERAL ELECTRIC CO. |
| Shared Package | |
| ML19309A455 | List: |
| References | |
| FOIA-80-3 NUDOCS 8003310111 | |
| Download: ML19309A458 (1) | |
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EDO:R/F Or. Glenn G. Sher.: cod, ::enager Safety and Licensir.g Operatica Ganeral Electric Co:npeny 17S Curtner Avarue San Jcse, CA 95125 02.r Dr. Sharucod:
Thank you for ycur letter of.luly 7.1970 ex.,Nsr. int: the concerns regarding t% orrposed chr;2ch to t':0 issu of Anticipchd Trcnsicats l'ithout Scran (ATUS) cs sit forth in UUMC-C'GO.
As yeu ::r~u. t':e 'f,~ ice of "; char licctor P.ac tictic.n !!as sought the recc. r'S ti:.s of the A 'visory Couitt.e on Rcactor SaTornards c.::d ti.21rr latcry Ec:tuircre.r.ts ',cvic;! Comitt te rccard!:;g the ATUS cpp. c zch i.rc" +:i in I.L EG
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He h:ill consider t!:eir recon-rcerdatic.ns in n r-ula'ing our Office recc r.:endction to the Comission.
The cc.:nts in year letter will bc given sir.ilar consideratica in thy c. :.i:,e of s*r.'. ele. ping the ::3R racc::nendation.
Si ncerely........
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!!arold R. D nton, Director Office of !:uclear r.: actor Regulatica bec:
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M xt i 3 1979 MEMORANDUM FOR: Gary L. Bennett, Chief Research Support Branch i
FROM:
M. D. Stolzenberg 1
Research Support Branch
SUBJECT:
SAFETY AND RELIEF VALVES At t% request of D. F. Ross, NRR, this mer randum and enclosures, summarize infomation available regarding relief and safety valve flow discharge (water and two-phase), operation, and testing. This memorandum completes ED0-TMI Action #18, as amended by the May 7, 1979 telephone request from D. F. Ross to you, l
RELIEF / SAFETY VALVE FLOW I
I To support NRR's evaluation of the flok through the safety valves on the B&W and GE plants under ATWS conditions RES contracted with ETEC (Energy Technology Engineering Center) to conduct a literature search to detemine if there were any addit {onal data generated in the interim between the 1975 report on ATWS and 1978.
The final report, Study of Safety Relief Valve Operation Under ATWS Conditions (NUREG/CR-068) was completed in January 1979 and publisned in Maren 1979, thereby completing the first phase of the research requented in Reference 1.
I Memorandum from E. G. Case to S. Levine,
Subject:
Request for Confirmatory Research Related to the Behavior of the Pressurizer Safety / Relief Valves During Subcooled Discharges (RR-NRR-78-10),
dated May 10, 1978.
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> Anon., " Status on Anticipated Transients Without Scram for Westinghouse Reactors," Appendix D, Division of Systems Safety, U.S. Nuclear Re ula-tory Comission, December F '
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27.o4.79 Mail Stop II30 SS 8==rt=a w Corwin / ea Washington, DC 20555 U.S.A.
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Subject:
HDR containment test bed use and KWU pressure relief valve testing l
Dear Gary,
Though rather belated, I hope you will find sufficient information enclosed in answer to your requests on HDR containment test bed use and pressure relief valve testing at KWU.
Basically HDR is quite willing to allow the U.S.
to use their containment during blow down testing as an environmental component test bed, subject only to a few scheduling and expenditure constraints. Several blowdown series will be conducted up to the end of 1981 which might be suitable for your purposes. Fluid conditions in the containment will vary with the individual blowdown types but maximum pressures of up to 6.5 bar and temperatures in excess of Testing dates and details, a containment use are in the e DUPLICATE DOCUMENT Entire document previously i
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GENER AL h ELECTRIC NUCLEAR ENERGY BUSINESS GROUP GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CAUFORNIA 95125 FORIY-SID0tm M2mILY REPORP COtERACP NO. NRC-04-76-215 BWR Blowdcun/Dnergency Core Cboling Pr@mu I
April 1979 i
SubfMRf
'Ibe TL'IA configuration proposed for BD/ECC-1B testing is still being considered 17f EPRI and NRC. 'Ihe reporting plan for BD/ECC-1A data has been documented. Preparation of the TLTA for near teIm testing is progressing on schedule and data frcm an average power test without ECC injecc. ion is reported.
TASK AA - Pr@mu Planning and Administration As a result of the March PMG Meeting, the reporting plan for the BD/Ecr-1A 4
test data has been dummmted. No formal response has been received frcm the 1
external program sponsors regarding the TLTA configuration proposed for BD/ECC-1B testing. However, following scue discussions with NBC, management mnsideration of TLTA "Non-IIEA" transient simulation capability was initiated.
TASK CC
'Ib.st Facility Design and Fabrication
'Ibe agreed upon TLTA modifications (January 1979 monthly) are being implenented. During the nonth, the nrrlified bypass inlet flow path was installed and m1ihvated in-situ. 'Ibe m1ihration indicated the need for a slight nodification'to the ficw restriction to produce the ncrninal flow split. 'Ibe nodification was made and a second calibration confiravl the sizing. New orifices were installed between the lower plenum and the guide tubes. 'Ihey were sized to simulate the corresponding flow paths in the BWR. An in-situ mlibration showed that these flow paths were correctly sized.
'1he thermal insulation in the lower plenum was visually inspected and found to be in very good condition and conforms well with the pressure vessel walls.
'Ibernoccuples will be nounted '
ness of this design in minimizing walls. 'Ibese 'IC's will be nonitorE DUPLICATE DOCUMENT aMitional differential pressure core bypass region and the ECC Entire document previously entered into system under:
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DISTRIBUTION SUBJ CIRC m 1 5 579 f1LV Kr CEJ CY Those on Attached List MLP CY LST CY WVJ CY
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Subject:
tiRC/P!.5/JAERI Annual Inforration Exchange on Codes and Cladding The next information exchange on safety related research on codes and fuel rod claddine between Projekt liuklear Sicherheit (KfK), Fuel Behavior Research Branch (HRC), and JAERI (Japan)'will be held.from June 22 to June 29,1979 at Idaho Falls. Idaho. EGLG will be our host, and the meetings will be held at the Technical Support Building, in Conference Roor A of Wing C on June 22 and 25, and the l'.ain Conference Room of Wing A on the reraining days. An agenda for the meeting is enclosed.
Some of the motels in Idaho Falls conveniently located are:
Westbank !!otel 475 liiver Parkway - (208) 522-3060 Driftrood liotel, 575 River Parkway - (208) 522-2242 Stardust l'.otel, 700 Lindsay Boulevard - (203) 522-2910 Also available is a toll-free call reservations center for Best Western International at 1-800-528-1234.
The meeting is in the fomat of a Zircaloy Cladding Program Review on June 25,1979 at which most of the more femal technical presentations for the meeting will be made.
Inforration workshops will be held on the other days. A final surary session will be held on Friday afternoon, june 29,1979.
We look foniard to another very informative and interesting infomation exchange.
Sincerely, W. V. Johnston, Chief Fuel Behavior Resea ranch
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Enclosure:
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.m G E N E R A L 1., c E L E CT R I C uuctexa susaav up PaOJECTS DIVISION GENERAL ELECTRIC COMPANY.175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN 145/79 MC 682, (408) 925-5040 n
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U.S. Nuclear Regulatory Commission
.g' Division of Project Management Office of Nuclear Reactor Regulation Washington, D.C.
20555 Attention:
D.F. Ross, Deputy Director 1
SUBJECT:
ACRS THREE MILE ISLAND-2 RECOMMENDA'TIONS-GE RESPONSE TO NRC RE"UEST FOR COMMENTS
Reference:
(1) Letter,1.F. Ross to Dr. G.G. Sherwood, "ACRS Recommenda-tions Relating to TMI-2 Accident," dated May 17, 1979.
Attachment:
GENERAL ELECTRIC - RESPONSE TO ACRS RECOMMENDATIONS Reference 1 requested that General Electric provide the NRC Staff with a concise discussion and position on each of the ACRS recomendations relating to the TMI-2 incident.
The information provided would be used as a basis for Staff discussions in the May 31 - June 1,1979 ACRS Sub-committee meeting on TMI-2.
The attachment is the General Electric response; it is provided for your information and use.
Based on discussions with your Staff, the General Electric response is provided in a format similar to that used in the comprehensive presentation to the full ACRS on May 11, 1979.
In addition, General Electric's conclu-sions regarding the applicability of each of the ACRS recomendations to the BWR system has been provided.
General Electric considers that the concerns and recommendations identi-fied as a result of the TMI-2 incident may not necessarily relate directly to all reactor types and designs.
To assist the NRC Staff, a summari of the inherent BWR design characteristics which should be considered during their evaluation of the ACRS recommendations is included in the attachment.
It is noted that several recommendations relate to areas outside the GE scope; consequenti d consider soliciting observations a
ties.
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. MFH 159-79 May 22, 1979 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 Attention:
Mr. V. Stello, Jr., Director Division of Operating Reactors Mr. R. J. Mattson, Director Division of Systems Safety Gentlemen:
SUBJECT:
CRD HYORAULIC CONTROL SYSTEM RETURN LINE MODIFICATION (INFORMATION TO OPERATING PLANTS)
Reference:
March 14, 1979 letter, G. G. Sherwood (GE) to V. Stello (NRC) and R. J. Mattson (NRC), subject:
Control Rod Drive (CRD) Return Line Removal The subject report is submitted to assist in the evaluation of the GE proposal to delete the CRD return line as discussed in the referenced letter.
It is forwarded per request of Mr. R. P. Snaider, of NRC Staff.
The purpose of this item is to provide utilities with basis for decisio.1 on whether or not to make a request to NRC for removal of the CRD retur'i line.
If you require additional information, please contact Alvin L. Spivak of my staff at (408) 925-3875.
Very truly yours, j
Glenn G. Sherwood, Manager Safety & Licensing Operation GGS:pab/304 Enclosure l
cc:
R. P. Snaider, NRC w/att.
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April 1979 CHANGES TO THE RETURN LINE TO THE REACTOR VESSEL ~
CR0 HYD'RAULIC CONTROL SYSTEM Intergranular stress corrosion cracking in the Control Rod Drive (CRD) hydraulic control system return line near its connection to the Reactor Pressure Vessel (RPV), -and fatigue crackin~g in the reactor vessel nozzle blend radius, have occurred at several operating BWRs.
The stress corrosion cracking has occurred in Type 304 stainless steel piping and in the RPV nozzle safe ends at weld-heat-affecte'd zones.
The' fatigue cracking in the RPV nozzles is caused by thermal cycling of the relatively cold water in the return line during different modes of CRD operation.
This memorandum is a review cf operating experience relative to this cracking, and a consideration of alternative corrective measures.
CONCERNS Crack growth'in the CRD return line and RPV nozzle can lead to the need for line replacement, vessel repair, and loss of plant availability.
Therefore, inspection and crack removal, if required, generally has been performed at the earliest outage opportunity.
To perform an adequate inspection, thermal sleeves are fully or partially removed from nozzles that are equipped with them.
The sleeve is not reinstalled if the CRD return line nozzle is capped.
If the return line nozzle is not capped, the thermal sleeve or a modified thermal sleeve is reinstalled even if the return line is valved closed with only infrequent use of the return flow capability being planned.
There is a licensing concern regarding changes in the reactor vessel water makeup capability from the CRD hydraulic control system.
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In the case of a DUPLICATE DOCUMENT in make-up capability.
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[b NUCLE AR ENERGY SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVENUE, SAN Jose, CALIFORNIA 95125 Phone (408) 2b7-3000 TWX No. 910-335-0116 MFN 108-79 April 19, 1979 U.S. fluclear Regulatory Commission Division of Systems Safety Office of Nuclear Reactor Regulation Washington, D.C. 20555 Attention:
Mr. Robert L. Tedesco Assitant Director for Plant Systems Gentlemen:
LEIBNITZ RULE AND HIGH HEAT TRANSFER IN LOCA MODELS
SUBJECT:
Loss of Coolant Accident and Emergency Core Cooling Models
References:
(1)
NED0 10329, for General Electric Boiling Water Reactors.
April 1971.
Letter, E.P. Stroupe, GE, to R.L. Tedesco, USNRC, " Application (2) of the Leibnitz Rule in the SAFE Code, " dated March 20, 1979.
Letter, E.P. Stroupe, GE, to R.L. Tedesco, USNRC, "NRC Concern (3)
About Heat Transfer Coefficient of 4 in SAFE Code", Dated Jan.29, 1979.
This letter addresses two NRC concerns about the GE Loss of Coolan evaluation model (EM) relative to the following:
Justification that the Leibnitz rule approximation as originally documented a) by GE on Pg-12, Reference 1, is appropriate.
Effect of higher heat transfer on the calculated Peak Clad Temperatuie b)
The information in this letter is supplementary to that already provided by GE in Reference 2 on the Leibnitz rule and in Reference 3 on the effect of high heat transfer and amplifies the discussion ir "-- 'alonhnne conversation of 4/17/79.
'-"' ' ol e i s It clearly shows that the truncation in appropriate as the sensitivity on the P oppLICATE DOCUMENT calculations).
It also shows that a co the LOCA EM will result in a reduction Entire ument previous 1 d.?
A preliminary calculation in which on1 into system under:
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is conservative in the calculation of ANO].h Details for the conclusions are in the no* Of Pages:-
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GTNERAL ELECTRIC COf.iPANY,175 CURTr:ER AVE., SAN JOSE, CAUFORN'A 95125 MFN-118-79 NC 905, (403) 925-3495 April 27,1979 U.S. Nuclear Regulatory Commission Division of System Safety Office of Nuclear Rec-tor Regulation Washington, D. C.
20555 Attention:
Mr. Robert L. Tedesco Assistant Director for Plant Systems Gentlemen:
SUBJECT:
MARK III TEST PROGRAM - TEST SERIES 6002 AND 6003 General Electric is conducting 1/9 scale multivent tests in the Mark III Pressure Suppression Test Facility (PSTF) as the final part of the overall confirmation of the Mark III containment design.
Attached for your information is a description of these tests.
Test Series 6002 has been completed and a report documenting the test results is in preparation scheduled for submittal to you in July 1979.
Test Series 6003 is currently underway and will be reported later this year.
If you have any questions on this test plan, please contact me or Mr. L. D. Steinert, (408) 925-2637.
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. M. Sobon, Manager BWR Containment Licensing
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PSTF '"ZST SIT.IES 6002 AND 6003 DESCRIPTION 1.C n.TacoucTIon The Mark III Containment Confirmatory Test Program was initiated in the Fall of 1973 as a confirmation of Mark III containment design loads and analytic models used in design of the General Electric Mark III pressure Test Series 6002 and 6003, described suppression containment systa=.
These tests use a 1/9-area scale 24 herein, are part of this program.
degree seg=ent (3 cells) of a Mark III pressure suppression pool.
Test Series 6002 is designed to investigate pool swell phenomena, and Test Series 6003 is designed to investigate pool condensation and thermal stratification phenomena.
The objectives of these tests are:
Test Series 6002 1.
Confirm that single-cell pool swell results are conservative relative to multi-cell.
Evaluate the effects of cell interaction on pool surface shape, pool 2.
surface displacement, water slug thickness, velocity and wall loads.
Test Series 6003 Primarys 1.
Confir= that single-cell dynamic loads due to condensation are con-servative relative to multi-cell.
condensation Study vent interaction and its effects during stee::t 2.
phases of the pressure suppression transient.
Secondary:
3.
Observe the =ulti-vent effe DUPLICATE DOCUMENT Entire document previously entered nto system under:
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PRODUCTS DIVISION GENERAL ELECTRIC COMPANY.175 CURTNER AVE.. SAN JOSE, CALIFORNIA 95125 MC 682, (408) 925-5003 MFN 123-79 April 30,1979 U.S. Nuclear Regulatory Commission Division of System Safety Office of Nuclear Reactor Regulation Washington, D. C.
20555 Attention:
Mr. Robert L. Tedesco Assistant Director for Reactor Safety Gentlemen:
SUBJECT:
CLARIFICATION OF ODYN MODEL UNCERTAINTIES In a recent ACRS Subcomittee meeting (March 19-20, 1979, Los Angeles, California) the staff described the progress on the staff review of the ODYN model and proposed licensing basis. At this~ meeting, the staff presented their current conclusions as to the uncertainties which might exist for certain of the more significant parameters in the ODYN evalu-ation.
Although the staff differed with GE's stated uncertainties in several areas, the one most significant difference was in the neutron effective void correlation uncertainty. The staff assumed that the effect on LCPR/ICPR could be determined directly by the ratio of uncer-tainties tirres the GE calculated CPR ratio.
However, this approach is quite conservative and leads to misleading conclusions.
GE has performed a
'_ve evaluation to show that the effect of uncertainties o e A PR/ICP s closer to the values presented by GE (NEDE-24154-P ose result' g from the overly conservative staff
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approach.
information is co ained in Attachment 2 and is being o enable the staff to mo e accurately evaluate the margin present avail in the'0DYN licensing basi evaluation.
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G El:ll.l.L k' E LE CTi'IC Mr. Robert L. Tedesco Page 2 April 30,1979 The information presented in Attachment 2 contains General Electric Company proprietary information. provides an affidavit which supports the need for handling this material as proprietary in-formation.
If you have any questions regarding this transmittal. I would be pleased to review the information with you or your staff.
Very truly yours, giu. del K. W. Cook, Sr. Licensing Engineer Special Projects Licensing Safety & Licensing Operation KWC:mh/1516-1517 Attachments cc:
L. S. Gifford Fuat Odar v
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.I GENER AL @ ELECTRIC NUCLEAR ENERGY PROJECTS DIVISIO N GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN-099-79 MC 682, (408) 925-5040 ApH 1 6, 1979 U. S. Nuclear Regulatory Commission Divis'en of Systems Safety Offi:e of Nuclear Reactor Regulation Was$ington, O. C.
20555 A'.1 ention:
R. J. Mattson, Director Division of Systems Safety Centlemen:
SUBJECT:
TWO LOOP TEST APPARATUS (TLTA) TEST RESULTS
Reference:
1)
Letter, E. P. Stroupe (GE) to R. L. Tedesco (NRC),
dated January 30, 1979, subject, TLTA Information Requested by NRC 2)
Letter, R. J. Mattson (NRC) to G. G. Sherwood (GE),
dated February 9,1979 (no subject)
Reference 1 forwarded some data and information on recently completed TLTA tests.
In addition, Reference 1 addressed some concerns expressed by NRC, relating to some preliminary evaluations regarding that data.
Since that time, further communications with several members of your staff regarding this subject has taken place at frequent intervals.
Reference 2 requested further information on what appears to be enhanced vapor generation indicated by the data from TLTA as it relates to the GE Loss of Coolant Accident (LOCA) evaluation model.
Reference 2 also noted that although continued use of the present evaluation model is acceptable, resolution of what appears to be an enhanced vapor genera-tion needs to be resolved in a timely manner.
GE is reviewing the additional information requested by Reference 2 to determine a schedule for responding to these requests.
Preliminary calculations with the evaluation model have shown that the peak cladding temperature (PCT) for these tests is about 1000*F higher than that actually measured in the tests.
Further review of these preliminary calculations is necessary and the results of this additional review are l
expected to be ccepleted by April 17, 1979 and will be communicated to
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GENERAth ELECTRIC Mr. R. J. Mattson t.pril 6, 1979 Page 2 However, additional evaluation of the test data is needed for a better understanding of the results and how it may apply to members of your staff to share this increased understanding and proposes model.
23, 1979 in Bethesda, Md.
a meeting for that purpose the week of Apr.il It would also appear appropriate to have personnel from the Divisio Operating Reactors present at this meeting to assure full commu with all interested personnel in NRC.
to discuss the schedule for any additional delineation of the te
~
LOCA results if it is needed.the NRC staff with an update on the status of the.G model modifications and proposed changes.
If you have any further questiens, please contact R. N. Woldstad of staff, (408) 925-2539.
Very truly yours, O
Glenn G. Sherwood, Manager Safety & Licensing Operation GGS:mh/1531-1532 L. S. Gifford, Bethesda cc:
M. W. Hodoes, NRC
~V. 5teilo, NKC
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ZIRCALOY CLADDING DERITTLDENT, RECO}DENDED CRITERIA
- by T. F. Xasener, H. M. Chung, A. M. Garde, and S. Mdjundar
}bterials Science Division ARGONNE NATIONAL LABORATORY Argonne, Illinois 60439 FOR PRESENTATION AT THE SIXTH WATER REACTOR SAFETY RESEARCH INFORMATION MEETING National Bureau of Standards Gaithersburg, MD November 6-9, 1978 l
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i Water Reactor Safety Research Heat Transfer Highlights November 1978
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NOTICE This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Nuclear Regulatory Commission, nor any of their employees, nor any of their contractors, subcontractors, their employees, makes any warranty, expressed or implied, or I
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assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, j
product or process disclosed, or represents that its use would not infringe privately owned rights.
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3)) E R A1. h ELE CT31C suCtema susacy PROJECTS DIVISION f-l GENERAL ELECTRIC COMPANY.175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 f1FN-121-79 MC 905, (408) 925-3495 i<
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U.S. Nuclear Regulatory Commission D
Office of Nuclear Reactor Regulations Washington, D. C. 20555 Attention:
Mr. V. Stello, Director Division of Operating Reactors Gentlemen:
SUBJECT:
MARK I CONTAINMENT PROGRAM GENERAL ELECTRIC REPORT NEDE-21983-P, " SUBMERGED STRUCTURES MODEL - MAIN VENT AIR DISCHARGES EVALUATION REPORT" Twenty copies of the report NEDE-21983-P, " Submerged Structures Model -
Main Vent Air Discharges Evaluation Report" are being provided by the i
General Electric Company on behalf of the Mark I Owners Group as part of the Mark I Containment Program, Task 5.14.1.
This report compares predictions of an analytical model for estimating drag loads on submerged structures due to main vent air discharges following a postulated loss-of-coolant-accident to the results of a test program.
O The report, NEDE-21983-P, contains information which General Electric o
Company customarily maintains in confidence and withholds from public The information has been handled and classified proprietary
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disclosure.
Accordingly, by General Electric as indicated in the attached affidavit.
we hereby request that NEDE-21983-P be withheld from public disclosure in accordance with the provisions of 10CFR2.790.
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Very truly yours, J
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i?t L. J. Sobon, Manager i
BWR Containment Licensing LJS:p /23 Q
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Gifford (GE) - Bethesda C. I. Grimes (NRC) j WL 996 5lyd 29 2.)
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s, G F.N ER AL h E LECTRIC uuctean susnay P R OJ E CTS DIVISI O N GEMERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE. CAUFORNIA 95125 MFN-124-79 May 2, 1979 U. S. Nuclear Regulatory Comission Division of Systems Safety Office of Nuclear Reactor Regulation Washington, D. C. '.0555 Attention:
Mr. Robert L. Tedesco Assistant Director for Plant Systems j
Gentlemen:
SUBJECT:
LEIBNITZ RULE AND HIGH HEAT TRANSFER IN LOCA MODEL
References:
(1) Loss of Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors.
April 1971.
(2)
Letter, E. P. Stroupe, GE, to R. L. Tedesco, USNRC, " Application of the Leibnitz Rule in the SAFE Code", dated March 20, 1979.
(3) Letter, E. P. Stroupe, GE, to R. L. Tedesco, U$NRC, "NRC Concern About Heat Trar.sfer Coefficient of 4 in SAFE Code", dated Jan. 29, 1979.
(4) Letter, E. P. Stroupe, GE, to R. L. Tedesco, USNRC, "Leibnitz Rule and High Heat Transfer in LOCA rnodels", dated April 19, 1979.
Reference 4 supplied preliminary results pertaining to the Leibnitz Rule application and high heat transfer in LOCA models.
This letter provides the verified results which like the preliminary results, show the current LOCA evaluation model is conservative in the calculation of peak clad temperature and is in conformance with 10CFR50.46/ Appendix K.
For ease of rwading, the original changes which have resulted as a ults delineated by a vertical line in s
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%.%.'....f May 8, 1979 General Electric Company ATTil: Mr. H. T. Watanabe M/C 682 175 Curtner Avenue San Jose, California 95125 Gentlemen:
RE:
INFORMATION REQUIRED TO COMPLETE NRC REVIEW'0F NEDE-21821 AND NEDE-21821-01, BWR FEEDWATER N0ZZLE/SPARGER. FINAL REPORT AND SUPPLEMENT 1 Mr. R. P. Snaider of the NRC staff had, on April 25, 1979, notified Mr. H. Watanabe of the GE staff that there were still two major out-standing items which must be resolved prior to our considering the review of the subject documents to be complete.
These items are:
1.
Vibration testing - In attempting to show freedom from vibration by flow sweeping over the range of conditions expected in service, I
inherent narrow band peaks in the sparger response may have been hidden.
These peaks, if allowed to develop, coul d cause structural instabilities.
We are concerned that apparently no attempt was made to detect such peaking by analytical techniques or to verify that the fundamental frequencies encountered were indeed the only ones to which the sparger will respond.
In addition, much of the vibration teminology and all of the sensor response plots and tables are unclear and contain unde-fined parameters. This makes evaluation of the report conclusions extremely difficult.
2.
System changes - on page 4-295 it is stated that:
" Examining the breakdown of crack growth due to the various subcycles in Table 4-33 indicates that about 80 to 90 percent of the total is due to the on/off flow cycling during hot standby conditions. Thus system improvements aimed at mitigating this aspect of the thermal cycling were pursued since they have i
the highest payoff in terms of potential reduction in crack growth.
In particular, the most promising system change was identified as the jdition of an improved feec.:ater low ficw j
controll er. "
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General Electric Company May 8, 1979 Furthermore, Figure 4-148 shows that if operation to a 1-inch crack depth were permitted, the improved sparger design would extend the allowable number of cycles from approximately 10 to 100 whereas adding the flow controller would extend the permissible operations by another 400 cycles.
Since the greatest benefit is apparently obtained by the addition of a low flow controller, we wish to evaluate whether the low flow controller and/or other system modifications.would provide adequate means to avoid cracking in the feedwater nozzles.
The advantages of such modifications, compar'ed to sparger changes and clad removal, are that the radiation exposure of workers would be greatly reduced and any degradation of the system modifications could be monitored and more easily repaired.
Our concern is that the radiation exposure resulting from these modifications be kept as low as reasonably achievable and that there be some assurance of the
' continued performance of these modifications to avoid feedwater nozzle cracking.
Therefore, we request that you include in your report an assessment of the relative merits of only the system modifications cocpared to the improved sparger and clad removal so that we would be able to single out the merits of improved feedwater controls; re-routing and maximizing cleanup water flow; and the turbine roll at reduced temperature.
Because resolution of the BWR feedwater nozzle cracking is considered a critical matter, we request that your personnel obtain the answers to'the above questions as soon as possible.
We are ready to meet with you at the earliest possible opportunity. _ Please inform Mr. Snaider (301/492-8414) when such a meeting can be arranged.
Sincerely, t
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Darrell G.s E enh t, epLty Director Divisicat t: 1 G.E. ' as cor.itted to ; reviding the balar:ce of
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the o rly verific:tien re,p ntes for tim't h bro Olitional suttittals, g
cne in early July snd the other in the Fall cf IN9.
5 intend to complete
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4 U;11TED STATES I4UCLEAR REGULATORY COMMISSIO:!
0FFICE OF INSPECTIO:: A!!D Ef!FORCEriEilT I
WASHlf!GTON, D.C.
20555 May 29, 1979 IE Information Ifotice Ho. 79-13 INDICATION OF LOW WATER LEVEL IN THE OYSTER C Summarr A loss of feedwater transient at the Oyster Creek facility on Ma h
t resulted in a significant reduction in water inventory within t e reac or (triple low core shroud area as measured by one set of wa i
(Figure 1).
area indicated water levels above any protective feature setpo nt The water level within the core shroud area was reduced belo level" setpoint of 4-feet, 8-inches above the top cf the fuel.
i Subsequent analysis by the licensee has determined that the l was 1 to water level (solid, without steam voids) over the top of the fue 1-1/2 feet.
Coolant sample anal,ses and offgas release rates indicate that damage occurred.
General _
The plant Oyster Creek is a non-jet pump BWR with licensed power of 1 was first i.ude critical May 3,1969.
Status Before Transient _
l Operating at near full power with tbc main parameters at leve follows 1895 MWt power level 79" Yarway (13'4" over top of fuel) 1020pgigreactorpressure
- /hr feed flow 7.1x10 gym recirculation flow rate (4 pumps) 4 14.8x10 DUPLICATE DOCUMENT e
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 May 29, 1979 IE Information Notice No., 79-13 INDICATION OF LOW WATER LEVEL IN THE OYSTER CREEK REACTOR Sumary A loss of feedwater transient at the Oyster Creek facility on May 2,1979, resulted in a significant reduction in water inventory within the reactor core shroud area as measured by one set of water level instruments (triple low level), while the remaining level instruments, sensing from the reactor annulus area indicated water levels above any protective feature setpoint (Figure 1).
The ' water level within the core shroud area was reduced below' the " triple low level" setpoint of 4-feet, 8-inches above the top of che fuel.
Subsequent analysis by the licensee has determined that the minimum collapsed water level (solid, without steam voids) over the top of the fuel was 1 to 1-1/2 feet.
Coolant sample analyses and offgas release rates indicate that no fuel damage
=3:
occurred.
q==
General Oyster Creek is'a non-jet pump BWR with licensed power of '930 MWt.
The plant was first made critical May 3,.1969.
Status Before Transient Operating at near full power with the main parameters at levels as follows:
1895~ MWt power level 79" Yarway (13'4" over top of fuel) reactor water level 1020 psig reactor pressure 7.1x106 #/hr feedflow 14.8x104 spm recirculation flow rate (4 pumps) 12 psid core op i
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PROJECTS DIVISIO N GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CAUFORNIA 95125 MC 682, (408) 925-5040 June 29,1979 MFN-177-79
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U. S. Nuclear Regulatory Commission '"*, * ' -
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"r, Office of Nuclear Reactor Regulation Washington, D. C.
20555 Attention:
Mr. Frank Schroeder, Acting Director
~
Division of Systems Safety Gentlemen:
SUBJECT:
TWO LOOP TEST APPARATUS (TLTA) RESULTS
Reference:
1)
G. G. Sherwood (GE) letter to F. Schroeder, dated 6/15/79, "Two Loop Test Apparatus (TLTA) Results" 2)
R. J. Mattson (NRC) letter to G. G. Sherwood (GE),
dated 2/9/79 (no subject)
INTRODUCTION In December 1978, General Electric received a verbal request from the NRC to perform a comparison of TLTA results and General Electric licensing evaluation model results.
Although this required extensive resources to apply the licensing evaluation model to the TLTA facility, General Electric committed to perform a comparison of the measured TLTA peak cladding temperatures with the General Electric licensing evaluation rodel and provide it to the NRC by June 29, 1979. This letter transmits those^ final comparisons, thereby completing the General Electric commitment.
The preliminary results were presented at the May 24 meeting and were documented in Reference 1.
The final results support the preliminary conclusion that the evaluation model conservatively predicts the average power test (with and without ECC) by approximately 1,000*F.
TLTA/EM PEAK CLAD TEMPERATURE COMPARISON Figures 1 and 2 show the final comparisons for the average power test with and without ECC.
The extremely large margin for both tests is primarily due to the relatively long dryout delay (approximately 40 seconds) observed in the test, combined with significant steani cooling from lower plenum flashing and better heat transfer during the ECC phase of the test than that conservatively assumed in the evaluation model. g
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SUMMARY
General Electric believes that (1) based on the positive results from TLTA which show the approved BWR LOCA evaluation model significantly overpredicts the test results (on the order of 1,000*F), and (2) the completion of the GE commitments made in Reference 1 that the NRC issues identified in Reference 2 will be considered closed.
If further clarification is required, please contact R. N. Woldstad of my staff at (408) 925-2539.
'Very truly your,
Glenn G. Sherwood, Manager Safety and Licensing Operation GGS:gmm/406-407 cc:
L S. Gifford (Bethesda) i 1
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. l ; l 1 l. L L' NUCLEAR ENERGY PROJECTS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CAUFORNIA 95125 MC 682 (408) 925-3141 MFN 185-79 July 20,1979 DiW 00l"'aLI 1 U.:.m t v;"f )
a U. S. Nuclear Regulatory Commission x.
Division of Systems Safety Office of Nuclear Reactor Regulation Washington, DC 20555 Attention:
Frank Schroeder, Acting Director Division of Systems Safety Gentlemen:
SGJECT:
TWO LOOP TEST APPARATUS (TLTA) INFORMATION
Reference:
1.
Roger J. Mattson (NRC) letter to G. G. Sherwood (GE) dated February 9,1979 (no subject) 2.
G. G. Sherwood (GE) letter to Frank Schrceder (NRC) dated June 15,1979, "Two Loop Test Apparatus (TLTA)
Results" 3.
R. H. Buchholz (GE) letter to Frank Schroeder (NRC) dated July 13,1979, "Leibnitz Rule in LOCA Models" (MFN-183-79)
F.eference 1 requested General Electric to provide information pertaining to pre'vious TLTA tests.
Reference 2 provided r, ort of the information requested in Reference 1 and defined additional information which would be provided by July 31. The purpose of this letter is to further clarify t ;at additional information.
Ey July 31, Gener-1 Electric will provide the following information:
1.
A writeup to support the May 24 slides - General Electric will provide a descriptive writeup to accompany the slides used at the May 24 presentation. A description of what was learned from the TLTA tests with and without ECC will be provided.
l Differences and similarities between the tests with and with-l out ECC will also be shown.
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Steam separator AP anc break flow discussion - General Electric will describe the mass balance for the tests with and without ECC and will discuss the mass energy balance (analyses) which leads to the conclusion that more liquid is going out the break for the case with ECC than for the case without ECC.
This shows that break flow is the major contributer to the variation in the difference of the depressurization rate seen in both tests. General Electric will also discuss the analysis which demonstrates that total flow out the steam separator was substantially lower for the case with ECC than for the case without ECC. Separate sensitivity studies which show that the flow through the steam separator was basically all vapor will also be presented.
3.
TLTA scaling discussion - General Electric will provide a discussion of scaling in general in the TLTA and describe where compromises were needed and where the simulation is representative of the BWR.
4.
TLTA/1974 vaporization data base comparison - General Electric will provide a further explanation of the facility design and the method used for the 1974 test which provided the data for the presently approved vaporization correlation.
i 5.
Side entry orifice CCFL - General Electric will provide a writeup to show that neglecting this effect is conservative.
6.
Grio spacer water accumulation discussion - General Electric will provide a writeup which will discuss whether it is possi-ble to have water accumulation at the grid spacer during different phases of the transient.
7.
H=4 to H=12 additional discussion - General Electric will provide additional clarification concerning the application of the heat transfer coefficient during core spray initiation.
8.
Discussion of plant choices for Leibnitz Rule study - This was provided by reference 3.
In addition, General Electric will provide:
l.
The measured pressures from the average power TLTA tests (with and without ECC) and the pressures from the calculation of the TLTA test with GE licensing evaluation models and 2.
The indirectly measured break flow from the TLTA tests and the break flow predicted by the GE licensing evaluation models.
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GEf:ER AL G ELECTRIC U. S. Nuclear Regulatory Commission Page 3 General Electric will continue to work with the NRC staff to resolve the TLTA issue as defined in Reference 1.
This will be accomplished as described in Reference 2 and as amplified by this letter.
Other TLTA information for which General Electric is contractually obligated per contract NRC-04-76-215 will be provided on the presently negotiated NRC/EPRI/GE schedule. It is General Electric's intent to supply suf-ficient information to close out the eight (8) clarification items listed above. We believe the commitments made i this letter should allow the NRC to close out the TLTA issue in the near futJre.
If fur-ther clarification is required, please contact R. N. Woldstad of my staff at (408) 925-2539.
Very truly yours, C g E. P. Stroupe, Manager
'7g7lg,u BWR Project Licensing Safety and Licensing Operation MC 682, Ext. 53141 RW:ggo/97-8 L. S. Gifford (Bethesda) cc:
L. E. Phillips (NRC) l f
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- 3. 6-L:9 Mr. Marshall E. Miller, Esq.
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Board Panel Nuclear Regulatory Commission qq,ig @
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Washington, D.C.
20555
Dear Mr. Miller:
The Department of Energy (DOE) has been following the proceedings before the Nuclear Regulatory Commission (NRC) involving the' operating license applications of Texas Utilities Generating Company for its Comanche Peak nuclear plant (( Docket Nos. 50-445A and 50-446A) and of Houston Light and Power Company for its South Texas Project (Docket Nos. 50-498A and 50-499A).
is the position of DOE that any intrastate-only operati It ng provisions associated with participating in the Comanche
?eak and South Texas Project nuclear facilities would
~
3.dversely affect future coordinated planning and operation of power supply in the Southwest.
Such an impediment is not in the public interest.
From a system economy and reliability perspective, such provisions also make it more difficult to develop and implement contingency plans for interregional power movements in the context of energy emergencies and fuel supply disruptions.
Further, it is the position of DOE that intrastate-only provisions associated with participating it. the Comanche Peak and South Texas Project would adversely limit the range of competitive possibilities in putting together hulk pcwer supply options.
By limiting the number of choices and the associated c,mpetitive processes in raking these choices, such an impediment prevents bulk power suppliers from combining options in the most effi-cient manner in terms of both planning and operations, and hence, is likely to lead to an increase in the overall censumers' charge for electricity.
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Please advise me if DOE can be of further assistance to NRC in this regard.
i Sincerely,
) l7 lPfeffer/-//l'~
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i 1: ting Assistant Administrator Economic Regulatory Administration 1
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P!blic Service Company of Oklahoma February 2, 1979 File:
6212.125.3500.21L E_ack :ox Station SF.V 3thble Oscillation Loads Oceket SIN 50-556 and SIN 50-557 Cf fice of Nuclent Reactor Regulation rivisien of Project Managenent Light '<!ater T.cactors Branch No. 4 1~. S. Nuclear Regulatory Commission Cashington, D. C.
20555 I
Attn: Steven A. Varga, Chief Gen:1eren:
Nring eur reeting of January 23, 1979 with Dr. Roger Mattson, Director, Civision of Systens Safety, Applicants agreed te provide a commitment related te the nethcdology to be used for conbining the leads that occur when nultiple safety relief valves (SRV's) actuate, specifically leads from escillating bubbles in the suppression pool.
On the basis of that discussion and agreenent Applicants commit to the following:
- 1) Centainment structures will be designed to accc.codate the leads associated
.cith the sinultaneous actuations'of all 19 SRV's with all the bubbles assured to escillate in phase in the suppression 2001.
- 2) Design of the affected equipment and components will be done utilizing those techniques described in the G.E. Report 22A'.365 "Interin Containment Loads 7.eport - Mark III Centainment" Revision 2 (1CLR Rev. 2) Appendix M and revisen, as a result of the regulatory staff's generic review, currently undeneay anc
- e be completed the first quarter of 1980. The ICLR Rev. 2 is contained on the Elack Fox Statica docket as Appendix 3C to the PSAR, Amendnent 14 da:ed February 2,19 o.
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- 3) Affected equipment and cc=ponents will not be per anently installed until 2 is available the generic resolutien of the staff review of ICLR Rev.
(during the first quarter 1980) for use in design.
In the event that the ultitate staff resolution is not forthcoming by April.1,1980, Applicants
- their, ill proceed with. installation of af f ected equipnent and components at i
o-m risk taking into consideration interis staff reports of ::ethedology acceptability, i
- ie believe. that these ce==itzents fairly reflect the sens. of c,ur neeting.
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Y' T. N. Ewing Manager, 3FS Nuclear Project i'
T :E:VLC:id Attachtant 4
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._E_!?[rFlX'STATICNSEF.VICELIST XC:
L. D:4 Cavis, Esquire Joseph R. Farris, Esquire William D. Paton, Esquire John R. Woodard, III, Esquire
' Colleen Woodhear, Esquire Green, Feldman, Hall & Woodard Counsel for NRC Staff 816 Enterprise Building U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74103 Neshington, D. C.
20555 l'r. Cecil Thomas 1437 South Main Street, Suite 302 U. S. Nuclear Regulatory Commission Tulsa, Oklahoma 74119 Phillips Building 7920 Ncrfolk Avenue Mrs. Ilene H. Younghein Bethesda, Maryland 20014 3900 Cashion Place Oklahoma City, Oklahoma 73112 Mr. Can A. Norris Enviror.nental Projects Branch 3 Mr. Lawrence Burrell U.S.. Nt: lear Regulatory Ccr.. mission Route 1, Box 197 Faillip: Building Fairview, Oklahoma 73737 7920 N -folk Avenue Eeth2sda, Maryland 20014 Mrs. Carrie Dickerson Citirans Acticn for Safe Enercy, Inc.
"r. Villiam G. Hubacek P. C. Box 924 U.S. NL: lear Regulatory Commission Claremore, Oklahoma 74017 Cffice of Inspection and Enforcement Ee;i:n IV 611 Fyin Piaz2 Drive, Suite 1000 Arlir.3t:n, Tsxas 76:12 l'r. Gerald F. Diddle Ger.erai "ar.acer
- sso:iited Eie:tric Ccoperative, Inc.
F. O. E:x 75*
E;rir.gfield, Missouri 65E31 Mr. Ma;. nard Human Ge eral P.anager Wes ter: Fer=ers Electric Ccoperative P. O. I x 429 Ar.ada rio, Oklahoma 73305 Mi:hael I. Miller, Esq.
- s.am, Lincoln & Esale r.s '.s: National Plaza SLite '203
^Picag:, Illinois 60533 "r. J::e;h Gallo
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NvCLEAR REGULATORY COMMISSION LVGossick
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December 28, 1978 t
ccm.u N BHayden SHHanauer VPanciera VStello The Honorable James C. Corman HShapar Ur.ited States House of Representatives Millie Groff
ashington, D. C.
20515
Dear Congressman Connan:
I at pleased to respond to your letter of November 7,1978, which requests-the ste'us of the fire protection programs in nuclear power plants and state::, 'he view that fire prevention measures, such as installing flame retardant coatings to electrical cables, are preferable Ehd acre cost effective than fire suppression methods.
I.eculd like first to address the status of the fi're protection programs.
Since the occurrence of the fire at the Browns Ferry Facility in March 1975, the{RC has undertaken a substantial upgrading of fire protection
- rovisions at nuclear pcwer plants.
This included a thorough study of the Ercv.ns Ferry fire to determine the lessons to be learned, and to
- r svide recon
- aendations needed to develop licensing guidance for both r.e
- nd operating plants.
This initial effort resulted in the issuance cf IREG-0050, Recomiendations Related to Browns Ferry Fire, a copy of
.chich is attached for.your information.
This report, which forms the
': asis for our present regulatory guidance concerning fire protection, sets forth a defense-in-depth concept which emphasizes echelons of deTense against fires.
These echelons of defense consist of:
1.
Preventing fires from getting started.
2.
Detecting and extinguishing quickly such fires as do get started and limiting their damage.
3.
Designing the plants to minimize the effect of fires on essential functions.
The report emphasizes that no one of these echelons can be perfect or ccmplete and that it is their multiplicity, and the depth thus afforded, that provide a high degree of safety in spite of lack of perfection in any civen system.
The concept of defense-in-depth has been applied by the staff in the folicw-up measures taken on all piants following the Ercar.s Ferry. Fire, and in developing fire protection guidelines for both new and operating plants.
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, Follow-up measures taken by the staff included (1) issuance of Inspection and Enforcement Bulletins directed toward control of ignition sources and improved housekeeping procedures; (2) special inspections of fire stops in operating plants; (3) improved inspection procedures for cables and penetration seals for plants under construction; and (4) alerting licensee's upper management to this concern.
These measures provided a prompt increase in fire protection of all plants while the' more complex aspects of fire protection were under consideration.
This interim protection has been strengthened by fire protection Technical Specification requirements for operating plants which cover existing fire protection systems, improved administrative controls for fire brigade strength and training, and controls for the use of combustibles and ig,nition sources within the plant.
Du. ring 1975 and 1976 the staff developed an extensive set of guidelines for fire protection considerations in new plants.
These guidelines were issued in the Standard Review plan as a Branch Technical Position (BTP)
ASB 9.5-1.
They were also issued for public. comment as Regulatory Guide 1.120 in June 1976.
These guidelines place emphasis on all three echelons of defense-in-depth, the first of which is fire prevention.
The implementation of these guidelines for.new plants assures that fire protection is taken into consideration as a basic part of the design rather than as an add-on after the design is otherwise completed.
Combined with the NRC and the nuclear industry's conservative approach to consideration of other potential accidents, we believe that this program will result in an acceptable level of fire protection for ncw nuclear power plants.
A second aspect of the staff's effort involves the evaluation of operating reactor facilities to assure that a level of fire protection commensurate to that required for new facilit'es is provided. We have concluded that there is no basis to restrict t.e operation of these facilities for cublic safety reasons.
This conclusion is based on our judgment that the likelihood of disruptive fires of the magnitude of the Browns Ferry event is small.
This judgment does not derive solely from statistical reescaing but rather from the fire protection conditions existing in the plants in 1975 and the subsequent positive and effective actions taken to i?. prove fire protection programs at these facilities since that time.
These actions have dec eased. the likelihood of severe fires through ~
effective fire protect.cn +nd improved fire fighting ability. Appendix A to BTp 9.5-1 which was issued in September 1976 sets forth the guidelines for this evaluation.
All licensees have been requested to compare their facility's fire protecticn program to the new guidelines and to prepare an extensive ~ " fire hazard analysis." These analyses were submitted to
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. the NRC between late 1976 and early 1978.
Considerable delay in the submittals from licensees was caused by the complexity of the required analyses and the need to obtain qualified fire protection engineers.
The NRC staff review of licensee's fire protection analyses began in early 1977.
The staff reviewers have been assisted by fire protection consultants having expertise in unique aspects of fire protection engineering.
The reviews are extensive and address such things as control of combustibles, detection systems, suppression systems, fire retardant coatings and administrative controls. The fire p.rotection reviews for all operating reactor facilities are targeted to be completed by June 1979; the target date to complete the implementation of fire protection modifications at As of December 1, m st operating nuclear power plants is October 1980.
1978,. reviews have been completed for 31 of the 70 operating reactor f~a cili ti es.
'P th respect to your view concerning fire prevention measures, the f:regoing description of the staff's fire protection program illustrates the heavy emphasis that has been placed on fire prevention through administrative controls.
For example, where possible combustibles have baen eliminated and in certain instances where it has not been possible eliminate the combustible inventory, coatings have been applied to-1:(1) reduce ignitability of cable groups, and (2) reduce fire propagation inrough cable groups. Appendix A permits the use of. flame retardant coatings and these coatings have been applied or are proposed to be applied to certain electrical cables of a large number of operating plants.
As discussed above, fire prevention is one aspect of the NRC's defense-in-depth approach.
However, an effective fire protection program must include measures to ' detect and extinguish promptly fires that do occur.
- n some areas of the plants we require that automatic fire suppression systems be installed, but the capability to manually fight a fire has o: been neglected.
Finally, steps are being taken to mitigate the effects of fir,es on systems and equipment that provide essential functions for safe ' plant operation and shutdown.
In instances where a fire could
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- pardize the safe shutdown of the plant the staff requires the provision
'or.either an alternate plant shutdown capability independent of the fire area or fire protection systems that can be shown to protect redundant electric cable divisions from electrical and exposure fires.
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_4 If I can be of further assistance, please 1st r.e know.
Sincerely,r t -
seph M. Hendrie
Enclosures:
Recor.endations Related to Browns Ferry Fire Re gulatory Guide 1.120 e
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3.,. 1 g 3 f.'UCLEAR REGULATORY COMt.USSION c.
- v. ASHIN G TON, D. C. 20355 DEC 2 71978 General Electric Company ATTN:
Mr. L. J. Sobon, Manager BWR Containment Licensing MC905 17S Curtner Avenue San Jose, CA 95125 Gentlemen:
SUBJECT:
REVIEW 0F GENERAL ELECTRIC TOPICAL REPORT NEDO-21052,
" MAXIMUM DISCHARGE RATE OF LIQUID-VAPOR MIXTURES FROM VESSELS" We have completed our review of the General Electric Topical Report NEDO-21052, " Maximum Discharge Rate of Liquid-Vapor Mixtures from
- l:.ssels," as it is to be applied to determine the mass and energy release rate resulting from a design basis accident for Mark I containment respense analyses.
Based on our review, we conclude that the model described in NED0-21052, in conjunction with its method of application for Mark I containment response analyses, are acceptable for reference as specified in the enclosure.
During the course of our review, we determined that additional justi-fication would be necessary to support application of this model to break sizes and types other than the double-ended rupture of a recircu-lation line in a Mark I plant.
We understand that you wish to pursue the application of this model for other sizes and types of breaks.
Therefore, when you provide the additional information required, as discussed in our letter dated January 30, 1978, the staff will i
continue its review of the subject topical report.
Such information should be submitted to the Division of Project Management.
The staff does not intend to repeat its review of this topical.
report when it appears as a reference for a Mark I containment response analysis, except to assure that the model is applicable to the specific plants involved.
Should the regulatory criteria or regulations change such~ that our conclusions concerning this topical report become invalid, you will be notified and will be civen the opportunity to revise and resubmit your topical report for review, should you so desire.
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_2 In accordance with established procedure, we request that General Electric issue a revised version of the topical report to include any supplementary information provided for our review, this acceptance letter, and the fiRC staff evaluation.
S ncerely, D. Eisenhut cting Assista t Director for Systems and
- Projects, Division of Operating Reactors
Enclosure:
Topical Report Evaluation Mr. L. Gifford cc:
General Electric Company 4720 Montgomery Lane Bethesda, MD 20014 e
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! CLOSURE I
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TOPICAL REPORT EVALUATION REPORT N0.:
NEDO - 21052 REPORT TITLE:
Maximum Discharge Rate of Licuid-Vapor Mixtures fron Vessel s REPORT DATE:
September, 1975 ORIGINATING ORGANIZATION:
General Electric Company i
REVIEls'ED BY:
Analysis Branch, DSS i
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b DUPLICATE DOCUMEN" b
Entire docurant previously entered int ystein u a
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N o.
of pages:
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l FEB 121979 i
Docket flo. STN 50-447
$6E.
' [Rof /3'rs. -
f Dr. Glenn G. Shcrwood, !!anager Safety & Licensing Operation i
Concral Electric Company 175 Curtner Avenue San Jose, California 95125
Dear Dr. Sherwood:
l
SUBJECT:
REQUEST FOR WITHliCLDINC I!iF0EMATION FROi FUBLIC DISCLOSURE Cy your applic'ations cnd affidavits dated April 21,1978 and !;ovc=ber 15, 1978 you sube.itted revisions 1 and 2 to the report' 22A4365AB, " Interim Centaincent Loads Report (ICLR) !* ark III Contain:r. cat," and requested that they be withheld from public disclosure.
Your reasons for requesting our withholding of this information were that the material contained in revisiens 1 and 2 to the report was obtained at l
considerable expense to the General Electric Cc:npany and that the release of such inforcation would seriously affect your competitive position.
Uc have reviewed your applications and material based on the require':ents I
cnd criteria of 10 CFR 2.790 cnd have determined that the above-centioned docucents sought to be withheld contain trade secrets or confidential or proprietary comercial infornation.
We have also found at this time that the right of the public to be fully apprised as to the bases for and effects of the proposed action does not outweigh the demonstrated concern for protection of your compet-itive position. Accordingly, uc have determined that the inforaation should '
be withheld fro: public discicsure.
He therefore approve your request for withholding pursuant to 10 CFR 2.790 and are withholding report 22A4355AB from public inspection as
~
i proprietary.
Uithholding fro: pubife inspection si$all not affect the right, if any, of persons properly and directly concerned to inspect the documents.
If the need arises, we cay send copics of this information to our consultants working in this tres. Ue will, of course, assure that the consultants have signed t c appropriate agreements for handling proprietary
[
ge data.
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NRC 70RM 318 (9 76) NRCM 0240 Tr u. s. covram a.aar pmentias crricca tore - sae ea4 s.
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