ML20128D247
| ML20128D247 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 11/25/1992 |
| From: | Wambach T Office of Nuclear Reactor Regulation |
| To: | Miraglia F, Murley T, Russell W NRC |
| References | |
| NUDOCS 9212070190 | |
| Download: ML20128D247 (10) | |
Text
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'o UNITED STATES 8
t NUCLEAR REGULATORY COMMISSION
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W ASHINGTON. D. C. 20555 November 25, 1992 Docket No.52-002 MEMORANDUM FOR:
T. Murley G. Lainas R. Zimmerman F. Miraglia J. Ace B. Boger W. Russell J. zwolinski C. Thomas J. Partlow M. Virgilio F. Congel D. Crutchfield B. Grimes E. Butcher W. Travers J. Richardson W. Bateman, EDO F. Gillespie B. D. Liaw A. Vietti-Cook S. Varga A. Thadani Operations Center J. Calvo G. Holahan C. Rossi THRU:
Robert C. Pierson, Director Wh Standardization Project Directorate J
Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation FROM:
Thomas V. Wambach, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors.
and License Renewal Office of Nuclear Reactor Regulation
SUBJECT:
DAILY HIGHLIGHT - FORTHCOMING MEETING WITH ABB-COMBUSTICN ENGINEERING (ABB-CE) FOR SYSTEM 80+ (SEVERE ACCIDENT SUBMITTAL)
DATE AND TIME:
December 2 and 3, 1992 (8:30 a.m. - 4 p.m.)
LOCATION:
Combustion Engineering, Inc.
Building 19 1000 Prospect Hill Road Windsor, Connecticut PURPOSE:
To discuss NRC staff's review of ABB-CE submittal of ALWR<FS-DCTR-33, " System 80+ Severe Accident Phenomenology and Containment Performance," dated August 1992.
The :taff questions to be addressed at the meeting are given in and the staff questions to be addressed by telecon are given in Enclosure 2.
Contact Seung Lee at (301) 504-2737.
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,2 November 25, 1992.
r PARTICIPANTS *:
ILP4 ABb-CE M. Franovich R. Schneider J. Kudrick
$1._ Jacob M. Snodderly J. Monninger A. Drozd (Original signed bd Thomas V. Wambach b aject Manager Standardization Pr: =,:t Directorate Associate Directorate ~for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
See next page
- Meetings between the NRC technical staff and applicants or licensees are open for interested members of.the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting _ Statement of-NRC Staff Policy," 43 Federal Reaister 28058,6/28/78.
Members of the.
public who wish to attend should contact me at (301) 504-1121.
DISTRIBUTION:
Docket File PDST R/F-LPlisco,--12G18 RPierson PDR JNWilson TVWambach RBorchardt-TBoyce TMiltz MXFranovich PShea EJordan, MNBB 3701 _
GGrant, EDO ACRS-(10)
OPA JKudrick, 801 MSnodderly, 801 JMonninger, 801 _
MMalloy NRR Mailroom, 12G18 P0' Dell, PTSB JMoore, 15B18 ADrozd, _. 8Dl -
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NAME: PShea g g TWYnba"chNzhdIrchar'dt-1})V/h2 Ildf/92 )
ll/Js/92 DATE:
0FFICIAL DOCUMENT COPY:
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ABB-Combustion Engineering, Inc..
Docket-No. 52-002-cc:
Mr. C. B.-Brinkman,-Acting Director Nuclear Systems Licensing Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Daniel F, Giessing U. S. Department of Energy-NE-42 Washington, D.C.
20585 4
Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Washington, D.C.
20503 Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.
Washington, D.C.
20037-1128
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QUESTIONS RELATED TO CE'S SYSTEM 80+ SEVERE ACCIDENT PHENOMENOLOGY-AND CONTAINMENT PERFORMANCE SUBMITTAL DATED AUGUST 1992 1.
Explain the development of the containment fragility curve and the accompanying calculations.
2.
3-3 ASME Service Level C loading conditions allow material strains representative of incipient yield...
3-5 For peak' strains typical of incipient yield conditions, the probability of containment failure is less than 0.05.
3-7 ASME Service Level C failure probability 0.03.
Fxplain discrepancy.
3.
3-7 Basis for assigning 5% failure probability to Ultimate Capacity ASME' pressure?
4.
3-9 Penetrations designed to withstand ASME ultimate pressure and severe accident temperature. ASME ultimate pressure (169-147 psig) assigned failure probability of.05.
Why not design to 1.0% strain-(220-204 psig) where failure probability is equal to ]? Credit is taken for no failure of penetrations between ASME ultimate and 1.0% strain.
5.
3-9 Discuss the Equipment Survivability section of SECY-90-016 in relation to the penetrations, seals, any equipment, and instrumentation relied upon in severe accident prevention or mitigation.
Provide a list of all equipment, instrumentation, seals etc., that would be relied upon for severe accidents or be exposed to severe accident conditions, and the accompanying parameters for which they are designed to tolerate.
Does the containment spray system, cavity flood _ing valves etc., have _ any special requirements for severe accident operation?
6.
3-10 What credit for DBAs and severe accidents is taken'for holdup'and filtration in the containment annulus of the shield building?
Is the annulus filtration system included in other CE plants (Palo Verde)?
What was the impedus for its installation? - AVS designed for design basis-fission product loadings, what about severe accidents?.Can the AVS be powered from the combustion turbine generator?
7.
3-13 What is the inter-relationship between the CFS and containmentL spray system for providing an inexhaustible continuous supply of water?-
8, 3-13 Provide preliminary indications of the. accident management guidance' for-when to manually actuate the cavity-flooding system.
Credit can not be taken for a pre-flooded cavity unless there is assurance of the indications used for flooding.
9.
3-13 What is the purpose of the holdup volume tank? Why not flood directly from IRWST to reactor cavity?
10.
3-15 Provide timing for filling the HVT, reactor cavity, and final level in each.
Do any accident sequences result in a core mass in dry cavity or only partially filled cavity?
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3-15 Any equipment 'qualific? tion requirements for the CFS valves?
12.
3-15 What features are discussed in "Section 4.3.2"?.ls this the correct reference?
13.
3-15 How will prolonged and irreversible core uncevery be determined?
Will the cavity ever be flooded when core damage is not known to have occurred?
14.
3-15 What amount of water is required in the cavity prior to introduction of the core melt, for the assumptions made regarding coolability and core concrete interaction to be'true?
15.
3-15 What is the impact on containment performance and core debris coolability if the CFS works but the containment spray system does not?
10.
3-16 Provide the 100% metal-water. reaction pressurization calculation.
Does it assume one burn,-intermittent burns, continuous burn, etc.
What pre-existing containment pressures are assumed and why?
17.
3-19 Reactor coolant system pressure, where direct containment heating is no longer a concern (anticipated corium dispersal-threshold value).
18.
3-20 RD capability to depressurize the RCS from 2500 to 250 psia prior to reactor vessel melt-througn. How long is this? -How long-to core melt and vessel melt-through in representative sequences?
19.
3-20 Battery sized to power loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. What about 580 coping period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />? Sequences and timing for when depressurization is needed.
1 20.
3-20 Provide results of the MAAP runs referenced.
21.
3-23 1s the instrument shaft referenced, the same as the 101 chase in Fig. 3.6-27 22.
3-24 DCH steam exists through louvered vents under the refueling pool.
Are there any paths where the pressures associated with a HPME would-force any material (insulation). out of the way creating a different4 flow path?
23.
3-24 Provide reactor cavity floor area and representative core debris.
depths for various sequences or amounts of core debris.
24.
.3-27 Provide a copy of Reference 3.12 and basis for conclusion that supports coolability in the long term.
25, 3-27. How will cooling of the upper layers of corium retard any concrete -
attack?
26.
3-29 What type of concrete-is used in the basemat?
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3-29 Provide analysis for 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> for corium to contact the containment liner.
28.
3-29 Provide discussion on how the reactor vessel and RCS are supported and what structures are required for this.
29.
3-29 What in-cavity structures might be damaged by an ex-vessel steam-explosion?
30.
3-29 Discuss the area above the refueling pool not prone to missiles.
What amount of concrete is below the core debris chamber and HVT sump?
Are any sumps located within the reactor cavity?
31.
3-32 What is the external source of water for the containment spray-system? (fire water, tee provided?)
32.
3-33 Are there dead-ended regions of the cavity where water from the containment spray system would not be recirculated to the HVT and IRWST?
33, 3-33 Can the containment spray system be powered from the combustion turbin_e generator? What is the impact on containment performance with no containment spray systems available? How 3 the containment spray system used in the MAAP analyses?
34, 4-1 What is the significance of defining early containment failure as I hour after the core debris penetrates the reactor vessel?
35.
4-1 Are there any areas where the System 80+ design departs from the EPRI Requirements for severe accidents?
36.
4-3 Discuss the SECY-90-016 criteria for high pressure melt ejectirn, such that a depressurization system should provide a rate of-RCS depressurization to preclude molten-core ejection-and to reduce RCS pressure sufficiently to preclude creep rupture of steam generator tubes.
37.
4-6 is this a generic graph or specific to the System 80+ design?
38.
What are the specific details of the cavity / containment design that will influence the consequences of an EVSE event and how does this relate to the statement on page 4-21, " Proper location of support structures and cavity wall design can effectively eliminate the containment threatening potential of steam explosions,"? (4.1.2.2' 2, page 4-20) 39.
What is the basis for statement, "the actual mass of corium expectec: to be-involved in any one explosion is small (under 20 kg)"? (4.1.2.2.2.3,.
page 4-21) 40.
Why was the applicability of the BETA V6.1 experiment not addressed?
41.
Provide the analysis that establishes the cavity design strength to be approximately 225 psid. (4.1.2.2.4, page 4-22)
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4 42.
The cavity design strength mentioned above is for a static load yet dynamic loads are more likely in the case of a steam explosion.
What is the cavity design strength for a dynamic load associated with a pressure impulse lasting 5 ms?
43.
Easis for stating the most likely RV failure mechanism will be via instrument tube failure? (4.1.2.2.4, page 4-22) 44.
The submittal states, "The energetics of this type of an event (FCl) were estimated in (D0E/ID-10271 "Frevention of Early Containment failure due to High Pressure Melt Ejection and Direct Containment Heating for Advanced Light Water Reactors," March 1990.) to~ produce localized cavity loads in the vicinity of 10 bar.
D0E/ID-10271 refers to a part of the-Zion Probabilistic Safety Study that determined this cavity load.
How is this analysis directly applicable to CE80+7 (4.1.2.2.4, page 4-22) 45.
Whatexperimentaldataoncoriumquenchingindicates}hatthequenching process exhibits maximum heat fluxes of up to 30 Mw/m for short time periods? (4.1.2.3.2.3, page 4-23) 46.
We would like to review the 21 psi pressure spike calculation and why it was assumed that the initial containment loading was at design limits (49 psig) at the time of vessel breach. (4.1.2.3.3, page 4-24) 47.
Are there any other means for compensating for loss of steam inerting besides the igniters?
It is not clear how the 80+ design will.
i compensate for a loss of steam inerting once the containment sprays are l
activated. (4.1.3.1.2.1.1, page 4-27) l l
48.
Furthrr discussion of Figure 4.1-2: ALWR Combustion Potential.
l 49.
We would like to review the analysis that establishes for 100%. Metal Water Reaction complete AlCC hydrogen burns result in peak containment pressures of about 140 psia. (4.1.3.1.4, page 4-32) 50.
We would like to review the analysis that the venttd IRWST hydrogen L.
concentrations are only 2 v/o greater than the overall containment l-concentrations. (4.1.3.1.4, page 4-32) 51.
What is the basis for the statement, " Igniter burns should produce pressure spikes less than that associated with a 50% core wide oxidation"?
52.
What is the basis for assuming the hydrogen burn will be initiated from-a 30 psia base pressure? (4.1.3.1.5, page-4-34) 53.
Further discussion of Table 4.1-4: Summary of PRA' Assumptions for System 80+ Hydrogen Deflagration Induced Loading and 4.1-5: Summary of System 80+ Containment Failure Probability Due to Hydrogen Deflagration.
54.
Section 4.1.4.6.1 describes the containment bypass phenomena.
What about fai'aure of containment penetrations such as the personnel and I
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equipment hatches? (4.1.4,6.1, page 4-54) 55.
Containment failure before core melt represents 62% of the containment failure frequency and 55% of the total risk for the CE System 80+.
Provide an analysis for inclusion of a filtered venting system and direct venting system for the CE System 80+,
A direct venting system-could be considered, if scrubbing though the IRWST is expected.
Provide insights to scenarios were this scrubbing may be effective and the expected decontamination factors. A filtered venting system should be l
considered for sequences that release directly to the containment or if the IRWST is determined to be ineffective in scrubbing.
56.
Containment bypass represents 28% of the containment failure frequency and 40%-of the total risk. Containment bypass consists of interft.cing system LOCA and steam generator tube rupture with unisolable path to the atmosphere.
Provide an indication of the benefits of inclusion of the-SECY-90-016 criteria for addressing interfacing system LOCA and areas where the criteria has not been met.
Provide an analysis for directing the steam from secondary side relief valves back to the containment and through the IRWST.
57.
Provide the analysis of basemat melt-through including assumptions of heat fluxes, amount of core, temperature, ablation rates etc.
58.
Provide a best-estimate analysis of the impact on containment performance of continued core-concrete interaction for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
59.
What are the DCH assumptions in Figure 4.1-1 for the best estimate dry cavity with RCS water case?
60.
How much radial and axial ablation can the reactor cavity withstand without failure?
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RAls on CE System 80+ for Containment Performance 1.
The statements under the item " Design Basis Pressure Capacity" section in page 3-3 of Reference 1 should be revised in accordance with the responses of RAls 220.45 through 220.48 and 270.44, 2.
From the Response to RAI 270.42 for general membrane stress for 3-D finite element model for the steel containment vessel with openings, the maximum stress intensity calculated for the testing load condition (load combination is D + L + P T ) at 53 psig is 24,614 psi and the allowable stress intensily+ fo,r Level C Service Limit is 52,480 psi at design basis accident temperature of 290 *F.
Since the stresses are at or below yield, a linear calculatica for the allowable' Level C Service
.imit pressure can be determined as a check from the testing load condition as follows; 24,614 : 53 - 52,480 : X, X - 113 psig for 3-0 model.
The temparature is assumed to be 290*F.
From Figure 3.1-2 of Reference 1 for 2-D axisymmetric thin shell model, the level C Service Limit pressure is 145 psig.
Since the internal pressure is dominant in the resulting stresses, the difference in the results (145 psig vs. 113 psig) needs to be explained.
3.
In page 3-5 of Reference 1, it states, "The material properties were represented by a bilinear stress-strain curve which was assumed to be essentially elastic-perfectly plastic in nature." which means the stress is maintained at yield, while the strain is increased.
However, use is made of a 5% strain hardening modulus in SA-537 Class 2 Stress-Strain Curve (Ref. 2).
Provide-the reasoas why a bilinear stress-strain curve with a 5% strain hardening is chc an.
4.
In page 3-5 of Reference 1, it states, "the strain at the maximum pressure of 193 psig is approximately 0.003 in/in."
Explain how this strain and pressure can be cbtained beyond the yield point using von Mises theory when it is valid only up to the yield point.
5.
In page 3-5 of Reference 1, it states, "The exact value varies depending upon element location and whether the midsurface or inner / outer surface is examined."
Explain how the membrane strain can be varied with location and across the plate thickness?
6.
Assuming a bilinear stress-strain curve, the stress calculations at 0.003 in/in strain with 5% strain hardening can be performed as a $e -
o 3 oy+ (0.003 - o /E) x 0.05 x E, and P,co3 - o,co3(2t)/r which give t o
o values in folle* wing table, o
E P
P (psi)
(psi)
(ps M (ps*k) 110*F 59,500 29.00E6 177.5 193 290*F 52,480 28.35E6 157.8 169 350*F 51,100 27.80E6 153.7 161 450*F 48,800- 27.30E6 147.1 147 Provide the reasons why System 80+ analyses produce-higher pressures.
7.
In page 3-6 of Reference 1, it states, "the 0.02 in/in actual tensile failure point of SA537 Class 2 material used in the containment shell construction."
Is the 0.02 in/in strain true tensile failure point for i
SA537 Class 2, or should it be 0.2 in/in7 l
o.'
8.
Explain how the extrapolation method in page-3-6 of Reference 1 can be used to get the ultimate pressures of 0.003 in/in strain at higher temperatures using a bilinear stress-strain curve?
9.
Provide the bases for 3% and 5% of probabilities of failure at Level C and ultimate presstres, respectively, in page_3-7 of Reference 1.
10.
Provide the following information:
a.
Material strength uncertainty, modelling uncertainty, and pressure distribution for containment fragility curve, b.
Material characteristics for penetration seals _ to ensure minimal containment leakage a" higher pressures and temperatures.
It should be noted the containment should fulfill not only structural integrity function but also the leaktightness function.
Structural integrity is necessary but not sufficient, because a 3/8"p hole in the containment may not fulfill its function of restricting the release of radioactive materials in case of a reactor severe accident even though it is structurally sound. Therefore, it is essential to establish the leakage through the seals of various penetrations.
11.
Typographical errors:
a, In Page 3-1 of Reference 1, the material is SA537, not SA357.
b.
In the column title of membrane strain in page 3-7 of Reference 1, the unit is in/in, not % in/in.
References:
1.
ALWR-FS-DCTR-33, Rev. O, " System 80+ Severe Accident Phenomenology and Containment Performance," Combustion Engineering, Inc., August 1992.
2.
Meeting handout, " System 80+" Steel Containment Vessel Code Design Activities," Combustion Engineering, Inc., April 22 and 23, 1992.
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