ML19263D777
| ML19263D777 | |
| Person / Time | |
|---|---|
| Issue date: | 03/21/1979 |
| From: | Thadani A Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| FOIA-80-587, REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR NUDOCS 7904130248 | |
| Download: ML19263D777 (9) | |
Text
l UNITED STATES I-
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MAR 217979 MEMORANDL FOR:
T. M. Novak, Chief, Reactor Systems Branch, DSS FR0ti:
A. C. Thadani, Reactor Systems Branch, DSS
SUBJECT:
ATWS MEETING WITH B&W A meeting was held with B&W on March 7,1979 to discuss the schedules for responses to and clarification of the staff questions on ATWS transmitted to B&W on February 15, 1979.
The staff stated at the outset that all the information requested in the February 15 letter must be provided, although a significant portion need not be provided until September 1, 1979.
The staff noted one exception, i.e., B&W was asked not to combine the ATWS and the OBE loads at this time.
The staff described the minimum information required by May 1 so that the staff can proceed with its plans to prepare a policy paper for the Commission by May 31, 1979.
A summary of the meeting and an attendance list are enclosed,
,' _ ' ;, x.'.'__. /L '
j Ashck C. Thadani Reactor Systems Branch Division of Systems Safety
Enclosures:
1.
Meeting Summary 2.
Attendance List 3.
3/6/79 Memo, F.C.Cherny to A.Thadani re Feb. 15 ATWS Questions cc:
(see attached list)
Contact:
Ashok Thadani, NRR 49-27341 79041302W1
EilCLOSURE 1 B&W - ATWS ffeeting Sumary The staff met with BSW on March 7, 1979, to discuss the minimum information required by l lay 1,1979, for B&W plants so that the staff could proceed with its plans to resolve ATWS using the early verification approach described in Volume 3 of fiUREG-Qa60.
I.
Analysis Methods The staff asked B&W to perform ATWS analyses using acceptable evaluation model.
B&W expressed the following concerns with the use of their models to perform certain calculations.
1.
B&W model cannot yet stably regenerate a bubbie in the pressurizer and, therefore, B&W cannot use CADD-S for performing calculations beyond this phase of the transient.
Since the peak pressure in the B&W plants occurs approximately at 40 seconds and the bubble is re-generated at 60180 seconds, the staff asked B&W to describe the nethods used for analysis of the transient after bubble regeneration.
The staff explicitly stated that the two-phase and subcooled dis-charge rates frcm the pressurizer must be calculated using the staff recommendation in the Feb. 15 letter.
B&W is considering using a more realistic model (e.g., COBRA or LYtiX-T) for assessing the number of rods that may experience a D iB condition.
The staff asked B&W to provide justification of the calculational tool used in the analysis.
II. ATWS Events for Analysis B&W agreed to provide analyses of the most limiting (DriBR, overpressure) transients by itay 1,1979 for Alternative 3 only.
The staff asked SSW to provide the following information for Alternative 3 and 4 plants by May 1:
.' 1.
Rods in DNB 2.
Worst case scenario for releases 3.
Time for containment isolation The staff asked that B&W provide analyses for all other transients by September 1, 1979.
The staff asked B&'.I to take into consideration the potential for transients being initiated by ICS failures.
III.
Plant Conditions and Asr.umations for Evaluation of ATWS Events Th staff asked B&W to provide justification of the initial parameter values assumed in the analyses.
If conservative FSAR or Technical Specification values are assumed, no further justification is needed.
B&W agreed to provide adequate justification, although B&W cannot sub-stantiate before September 1 whether the plants would conform to the assumed value. The staff encouraged B&W to coordinate their analysis assumotions with the applicants and the licensees so that there is reason-able assurance in May that the plants would fall in the analysis envelope, although detailed confirmation need not be orovided until September.
The staff also asked B&W to take into consideration the potential impact of restrictors on the auxiliary feedwater flow since not all plants have the restrictors.
IV.
Coerative Equioment and Systcms BMI agreed to provide description and design bases for the systems relied on to mitigate the consecuences of ATWS events.
B&W would generally describe how the automatically actuated systems would conform with the criteria in Appendix C of NUREG-0460, Volume 3.
V.
Sensitivity Studies The staff agreed with B&W that sensitivity studies for parameters for which bounding values are assured need not be provided.
The staff agreed that
.' detailed senstivity studies need not be provided until September 1, although some aualititive assessment must be provided by May 1.
VI.
Classes of Plants B&W understands that specific plants have to be covered by generic analyses although B&W noted that the ultimate responsibility that the analyses represent a specific plant rests with the applicant (or licensee).
VII.
Results Documentati_on The staff agreed that the lono-term cold shutdown assessment need not be provided until September althcogh detailed short-te m analysis must be provided by May I along with qualitative assessment of the ability to bring the plant to a safe shutdewn condition.
This qualitative assessment would also describe the manual actions nccessary to achieve the safe shut-down condition.
The staff agreed that maintainina the containment integrity can be demonstrated by the assessment of mass and energy releases to the contain-ment being significantly below those expected to occur in the event of a LOCA.
VIII. Additional Guidance on Requirements 1.
Engineering Considerations The staff gave a copy of a note from F. Cherny to A. Thadani (see ) which describes the mini.m information required by
'!ay 1.
B&W agreed to provide this in.ormation by May 1 subject to the utilities providing B'W the necessary data on components.
B&W would also provide, subject to the licensees and applicants authori-zation, a detailed analysis plan by May 1 and submit a series of reports
_4 during the summer of 1979 although B&W did not commit to providing all the analyses by September 1.
The staff gave encouragement to the following approach suggested by BSW:
a.
Leakage from isolation valves need not be assumed if their integrity can be maintained.
b.
Valves sinilar to the one for which inteority and operability has been demonstrated by detailed analyses can also be assumed to maintain their integrity and coerability if the similarity is based on consideration of geometry, manufacturer, and material.
c.
Hydro tests should be assessed for their applicability in denon-stration of integrity.
2.
Fuel Damace Considerations B&W would provide an estinate of fuel failure for the worst case ATWS event by flay 1 and the remainder analyses by September 1.
B&W would provide justification of the code used for estirating the number of rods, if any, experiencing a DNB condition.
3.
Radiological Considerations See Section II of these minutes.
4 Reactivity Feedback Considerations The staff asked B&W to provide justification for the chosen values of the moderator temperature coefficient (MTC) for Alternative 3 and a plants.
The staff also asked B3W to respond to the following by May 1:
. a.
A commitnent that the assumptions used to develop initial MTC values would be applicable and adhered to by specific.31 ants.
b.
Changing from " rodded" to " feed and bleed" reactor should make the MTC more positive.
c.
Ninety-five percent MTC may be 1 pcm more positive than a BOC, 100% power, equilibrium xenon base MTC.
IX.
Information Recuired in Electrical Areas, and X.
Diversity Considerations B&W agreed to provide criteria, preliminary descriptions, and a discussion of diversity considerations by May 1.
The staff agreed that detailed wiring diagrams are not needed until implementation stage but the staff requires the remainder of information by September 1.
ENCLOSURE 2 B&W ATUS Meeting Attendance List NRC B&W A. Thadani H. Ferris D. Thatcher D. LaBelle T. flovak F. Burke F. Cherny 0.E. Lee F. McPhatter A. McBride
EriCLOSURE 3 yn,,
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!379 i;0TE TO:
A. Thadani, Reactor Systems Branch, JSS - c FRCM:
F. C. Cherny, Section Leader,itechanical Engineering Eranch, DSS
SUBJECT:
FEB. 15, 1979 AT'lS QUESTIONS
Reference:
A. Thadani Note of March 5, 1979 I have reviewed the questions posed in your March 5 note and have discussed them with 3. D. Lian.
Recognizing that considerable analyses may have to be perforned in order to adequately respond to all of the concerns raised in section VIII.B of the Feb. 15 questions and taking into account the 3 " bins" discussed by R. flattson at the March 1 meeting, I feel the only possible schedule compromise that can be made in the structural / operability area would be as follows (using designations in your 3-5 note.):
a.
All the information in section VIII.B of the February 15 questions must be supplied and approved before the final ATUS rule goes out for public comment.
b.1 Minimum information to be supplied in the first bin (by May 1,1979) should include:
1.
For Alternative 3 plants, a list of all components that will exceed level C allowable pressures and the pressure the component would be exposed to under the worst case postulated ATWS event using Vol. 3 t;UREG-0460 assumptions.
In addition for such components a comprehensive qualitative description of the analysis to be perforced to demonstrate that the component's structural integrity will be maintained upon exposure to the ATWS environment. These discussions shall specifically indicate how the analysis to be performed will take into account the general and component specific concerns noted in section VIII.B.l.b of the February 15 questions.
2.
Valve Operability Analyses and Tests - Section VIII.B.l.c of tne February 15 questions - Provide by itay 1 a comprehensive description of analyses and tests to be performed to sub-stantiate isolation valve operability for both Alternate 3 and Alternate 4 plants. The specific concerns noted in this section of the February 15 questions should be addressed.
e A. Thadani -
I379
'" # 0 3.
Also provide by iiay 1 a list of all Alternate 3 P'.iR plants where stean generator tube plugging criteria could be affected as a result of taking the maximum calculated AT..S pressure into account for deteraining minimum allowable tube wall thickness.
b.2 All analyses and test results required to respond to section VIII.3 of the February 15 questions shall be furnished in the second bin i.e., by September 1, 1979.
b.3 :!ot applicable for section VIII.B questicas.
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F. C. Cherny, Section/ Leader itechanical Engineering 3 ranch Division of Systens Safety cc:
R. !!attson J. Knight R. Bosnak
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