ML20062J672
| ML20062J672 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/31/1980 |
| From: | Jensen W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062J670 | List: |
| References | |
| NUDOCS 8011040473 | |
| Download: ML20062J672 (5) | |
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0'llTED STATES OF A" ERICA f;UCLEAR REGULATORY C0r*lSS10N BLf 0RE THE AT0!ilC SAFETY AT;D LICENSING EDAP.D In the fatter of
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E f*ETROPOLITAli EDISON C0:4PANY, ET AL.
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Docket flo. 50-289 (Three liile Island fluclear
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Station,Unittio.1)
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NI C STAFF TESTiliONY OF ll ALTON L. JENSEN, JR.
RELATIVE TO BOARD QUESTION REGARDING UCS CONTENTION 8
' Peard question regarding UCS Contention 8:
The board directs the staff and the licensee to present experts and the fundamental documents involved in the small break LOCA analysis, and to have very complete testimony on this subject.
The reco rnendations of fiUREG-0565 and HUREG-0523 should be addressed.
It appears from the small break LOCA analysis that there is a large amount of reliance upon operator action and on non-safety-grade equipment.
The board wants that issue explored by testimony, including why such reliance is proper.
Tr. 2374-85.
Response
Testimony on the subject of small break LOCA analyses and supporting fundar. ental
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documents have been filed by the NRC in response to UCS Contention 8.
This testimony did not specificaljy address the recommendations of NUREG-0565,
' Ceneric Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock
& Wilcox Designed 177-FA Operating plants," January 1980 or those of NU":EG-0623,
" Generic Assessment of Delayed pump Trip During Small Break Loss-of-Coolant Accident in pressurized Water Reactors," l'ovember 1979. With one exception, inplementation of the recomendations relating to small break LOCA analysis 4
8011048 V U T
2-ucre not required by the order for THI-1 restart. The exception is the rcquirement of !;UREG-0623 for tripping the reactor coolant pumps which has been implempnted at TMI-1 as discussed below. All !!UREG-0565 and !!UREG-0623 N
reco:nnendations will be impicc.icnted as part of the "!!RC Action Plan developed as a result of the TMI-2 eccident " HUREG-0660, Task II.K Item 3.
A letter fron D. Eisenhut, NRC, to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits dated Septc:..ber 5, 1930, provided preliminary clarification of TMI Action Plan requirements.
The implementation of these requirements is addressed in the NRC staff testimony of Robert W. Reid concerning Board Question 5.
Seven recomnendations of NUREG-0565 relate s;'ecifically to small break LOCA analysis and are listed in Table 2-1 of that document.
(Items 2.2.2.a.2.2.2.c, 2.6.2.a and 2.6.2.9 concern the com,' uter models and Items 2.2.2.b, 2.6.2.c and 2.6.2.d concern additional analyses).
f UREG-0565 Items 2. 2. 2.a. 2.2.2.c, 2.6.2.a 2.6.2.g The concerns by the B&O Task /orce regarding the small break LOCA models,
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involve the need to confirm specific model features against applicable experi-mental test data. The recent tests against which present small-break LOCA medels can both qualitatively and quantitatively assessed include the entire semiscale small-break test series and LOFT test L3-1 and L3-2.
Other separate effects tests. (e.g., ORiil core uncovery tests) and future tests, as appropriate, should also be factored into this assessment. As discussed in the NRC response to UCS conteittion 8, a considerable margin exists between the calculated core
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3 conditions for postulated small breaks at THI-l and the core damage limits of 10 CFR 50.46. The HRC staff believes that further refinement of the small break LOCA r;odels is desirable in understanding the sequence of events during Ib the accident but that the current model and calculations are in conformance with 10 CFR 50.46.
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[;UREG-0565 Itens 2.2.2.b, 2.6.2.c, 2.6.2.d If model deficiencies are discovered as a result of the above data ccmparisons, the GRC will require that the ECCS models be revised and the small-break spectrum
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analyses for THI-l be repeated (Item 2.2.2.b).
The additional analyses reconnended in Items 2.6.2.c and 2.6.2.d involve multiple system. failures t.hich would fall within the range of the break spectrum already analyzed.
Operator action in dealing with these events will be included in Task I.c.1 of the TMI-l Action Plan.
NUREG-0623' Ti.e staff review of analyses by the reactor vendors regarding tripping of the reactor coolant pumps is described in NUREG-0623. These analyses predict that if the coolant pumps are tripped during certain periods during a small break LOCA a greater degree o*f core uncovery could cccur than if the pumps were tripped immediately. The staff therefore issued IE Bulletin 79-05C to the owners of B&W plants requiring an immediate manual pump trip following an indication of HPI actuation anii requiring an additional operator to be in the control room to perform the action. The manual trip requirenant is reflected in the TMI-l tmall break LOCA procedures and the NRC concluded that TMI-l was in conformance with the short-term requircments of IE Bulletin 79-05C.
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. Dulletin 79-05C also contained a long-term requirement for automatic tripping.
of the reactor system pumps. The automatic pump trip was reco:r.nended in I;UREG-0623 s,o that the pumps trould be tripped for LOCA events but tould not
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be tripped for non-LOCA events. 1:UREG-0623 reco;nized an uncertainty in the thermal-hydraulic phenomenological modeling of small breaks with the pv.ips ' running. To cvaluate this uncertainty Action Plan Item II.K.3.5 of NUREG-0060 t:as revised in i'ay 1930 to require a continued study of the criteria for early reactor coolant pump trip. Holders of approved ECC models have been required to analy7e the forthcoming LOFT test (L3-6). The capability
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of the industry models to corectly predict the experimental behavior of this test will have a strong input on the staff's determination on t: hen and how the reactor coolant pumps should be tripped.
The imple:nentation of the Action Plan is discussed in the NRC testimony of R. Reid to Board Question 5.
The assumption that the operator manually trips the reactor coolant pumps it..ediately following a small break LOCA is the only reliance on non safety grade equipment and the only operator action assumed in the analyses of small break LOCAs described in the NRC responst to UCS Contention 8.,As discussed.
above an additional operator will be available in the TMI-1 control room to trip the reactor coolant pumps. The operators will be trained to perform this action as discussed on page Cl-16 of. NUREG-0680.
Four operational transients in PURs (North Anna Unit 1 Prairie Island, Arkansas Nuclear One Unit 2, and Crystal River 3) which occurred in 1979 and 1980 h ve indicated that operators have acted promptly in tripping the reactor coolant pumps when safety injection signals were received at those facilities. The NRC staff believes that the manual trip requirement at TMI-1 is adequate in the interim period while the need for an ' automatic trip is evaluated under the Action Plan.
WALTON L. JENSEN, JR.
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PROFESSIONAL QUALIFICATIONS I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear Regulatory Coraission.
In this position I am responsible for the technical analysis and evaluation of the public hea'lth and safety aspects of reactor-f systems.
Frca June 1979 to December 1979, I was assigned to the Bulletins and Orders Task force of the Nuclear Regulatory Cocaission.
I participated in the preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant Accident 9ehsvior in Babcock & Wilcox Designed 177-/A Cperating Plants."
From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the i
NRC.
In these positions I was responsible for the development,and evsluation of computer programs and techniques to calculate the reactor system and centainment system response to postulated loss-of-coolant accidents.
Frem 1967 to 1972, I was empicyed by the Sabcock and Wilcox Company at Lynchburg, Virginia. There I was lead engineer for the development of loss-of-coolant corputer prograEs and the qualification of these programs by comparison with ex;erimental data.
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Frem 1963 to 1967, I was employed by the Atomic Energy Commission in the Division of Reactor Licensing.
I assisted in the safety reviews of large power reactorf, and I led the reviews of several small research reactors.
I received an M.S. degree in Nuclear Engineering at the Catholic University of America in 1968 and a 8.5. degree in Nuclear. Engineering at Iississippi State University in 1963.
I am a graduate of the Oak Ridge School for Reactor Techr. ology, 1963-1964 I am a me:ter of the American Nuclear Society.
I am the author of three scientific papers dealing with the response of B&W reactors to Loss-of-Coolant Accidents and have authored one scientific paper dealing with containment analysis.
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