ML20059C296
| ML20059C296 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 12/29/1993 |
| From: | Borchardt R Office of Nuclear Reactor Regulation |
| To: | Quirk J GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9401050084 | |
| Download: ML20059C296 (9) | |
Text
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c December 29, 1993 Docket No.52-001 Mr. Joseph F. Quirk GE Nuclear Energy 175 Curtner Avenue, Mail Code 782 San Jose, California 95125
Dear Mr. Quirk:
SUBJECT:
GE NUCLEAR ENERGY (GE) ADVANCED BOILING WATER REACTOR (ABWR)
CONTAINMENT SYSTEMS AND SEVERE ACCIDENT REVIEW ISSUES The Containment Systems and Severe Accident Branch staff has completed its safety evaluation report for the ABWR design.
Enclosed is a detailed discussion of remaining issues and the staff's final position for resolution.
Please review the information and provide a detailed plan for near-term resolution of these issues in writing not later than January 7,1994.
Should you have any questions or desire further_ discussion on the information provided in this letter, please contact Chet Poslusny at (301) 504-1132 or Son Ninh at (301) 504-1125.
Sincerely, (Original signed by Jerry N. Wilson for)
R. W. Borchardt, Director Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reacto: Regulation
~
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION w/ enclosure:
P0ocketnFile PDST R/F DCrutchfield JNWilson PDR RBorchardt MMalloy CPoslusny SNinh JKudrick, BH7 DTang RBarrett, 8H7 JMonninger, 8H7 WTravers PShea JMoore, 15B18
_MFinkelstein, 15B18 WDean, ED0 ACRS (11)- (w/o encl. )
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O NAME: PShea SNinh:
CPoslu JNWilson RBordhardt DATE: 12/40/3 12/72/93 12/28/93 12/f/93 12/ff93 (di _
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Mr. Joseph Quirk Docket No.52-001 GE Nuclear Energy cc:
Mr. Steven A. Hucik Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.
175 Curtner Avenue, Mail Code 782 Suite 300 San Jose, California 95125 Washington, D.C.
20086 Mr. L. Gifford, Program Manager Mr. Victor G. Snell, Director Regulatory Programs
. Safety and Licensing GE Nuclear Energy AECL Technologies 12300 Twinbrook Parkway 9210 Corporate Boulevard Suite 315 Suite 410 Rockville, Maryland 20852 Rockville, Maryland 20850 Director, Criteria & Standards Division Mr. Joseph R. Egan Office of Radiation Programs Shaw, Pittman, Potts, & Trowbridge' U.S. Environmental Protection Agency 2300 N Street, N.W.
401 M Street, S.W.
Washington, D.C.
20037-1138 Washington, D.C.
20460 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 l
Washington, D.C.
20036 q
l Mr. Steve Goldberg l
Budget Examiner q
725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 2
Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874
Csntainment Emeroency Procedure Guidelines (EPGs)
Heat Capacity Temperature Limit (HCTL) - The staff provided GE with two different acceptable resolution paths; either use an HCTL curve that does not i
result in elevated suppression pool temperatures or use one that does and justify that the containment can withstand the impact.
"In order for the staff to find the HCTL curve acceptable, as proposed in Amendment 32 of the SSAR, GE must demonstrate that large continuous steam plumes do not occur within the suppression pool such that the containment liner integrity could be jeopardized by the sudden unstable collapse of large steam bubbles. Large steam bubbles appeared to have been observed in a saturated pool during sub-scale experiments performed by Chun and Sonin as discussed in Dr. Sonin's paper published in Nuclear Engineering and Design (1981). The staff will find acceptable a suppression pool operated near saturation if the applicant can demonstrate that the Cross-Quencher proposed for ABWR can produce a stable steam bubble when a steam discharge could occur j
into a suppression pool operating near the saturation temperature.
Stable i
steam bubble size would then be defined as that size steam bubble or group of bubbles that may drift into a cooler region of the pool and condense such that suppression pool wall pressures do not exceed those wall pressures previously defined for unstable condensation oscillation loads from SRV actuations or a LOCA."
Low-Pressure Venting - The following items are a result of the Novem-ber 4, 1993, conference call with GE and need to be addressed to close out this item:
1.
Revise EPGs (PC/P) to show that venting is restricted to the 2-inch line in the drywell.
2.
Address suppression pool bypass mechanism through interconnection in the atmospheric control system (ACS).nd show the effect on the existing bypass analysis.
Ensure that no other bypass pathways exist that have not been accounted for.
3.
Address containment isolation configuration of interconnection in the ACS between the wetwell and drywell. GE should justify automatic control of the ACS over a normally closed penetration ensuring containment integrity.
4.
Address suppression pool level issue in EPGs relating to the wetwell to drywell interconnection level. The EPGs appear to be inconsistent with the design.
5.
Address suppression pool level and pressure control EPGs for injection from sources outside of containment. The EPGs appear to require conflict-ing actions in that SP/L-3.3 directs operators to stop injection from sources outside containment when the suppression pool level reaches 27.2 m.
Whereas, PC/P-6 directs operators to spray the containment when the water level reaches 27.2 m (using sources external to the containment sump shield design).
Enclosure
l EqnDinment Sump Shield Desion The staff believes that the sump shield designs proposed by GE have consider-i able merit and that some conservatism exists in the specified design criteria.
For example, the design criteria is intended to ensure that no core debris enters the sumps. However, in actuality, the sumps could withstand limited amounts of core debris.
In addition, GE did not take credit for flooding the
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lower drywell with the lower drywell flooder system or ac independent water addition system.
Based on engineering judgement, the staff believes that the sump shields would prevent a substantial accumulation of core debris and that the channels within the FDS would lead to freezing of debris within them.
l However, the analysis provided to support the proposed shield designs is not sufficient to reach this conclusion.
In particular, GE did not make use of existing experimental data and analytical tools in justifying their design.
i The staff believes that an acceptable resolution to this issue would entail the following:
1.
GE should evaluate related experimental and analytical work performed in this area to lend additional credibility to their design.
In particular, GE should address how the results of the previous work supports their design. This would include a discussion on the prototypically of the core debris, important parameters and results. The staff has performed a quick review of related work in this area and believes that it is relevant and readily available.
a.
Experiments performed at (1) KfK on ingression of molten debris into small cracks and openings, (2) Winfrith in the United Kingdom, and (3) Grenoble in France.
b.
Analytical tools such as PLUGM (NUREG/CR-3190) and BUC0 GEL (CEA).
c.
Work performed in forging and casting industries.
2.
The analysis performed by GE for sizing the FDS evaluated an oxidic melt of around 2500k and a eutectic melt of around 1700k.
However, GE used the same correlations and key parameters for both, such as thermal conductiv-ity and latent heat of fusion.
To account for uncertainty in the progres-sion of a severe accident and a range of material properties (density, melting point, thermal conduct'vity, etc.), GE should perform separate analysis for an oxidic, metallic, and eutectic melt clearly identifying the material properties and providing suitable references.
In addition, GE should identify the parameters that the shields are most sensitive to (i.e., freezing point, heat of fusion, velocity of debris in channel, atmosphere temperature, melt superheat, etc.). GE can use the results of their MAAP runs to identify the core debris composition at the time it enters the lower drywell.
In addition, GE could use the results of other code predictions (BWRSAR, MELCOR) as documented in NUREGs for similar BWRs.
l I
l
- 3.
GE should address the following; why the velocity of the debris in the FDS channels does not have to consider the initial velocity of debris from falling from the reactor pressure vessel?
4.
GE should modify the design criteria to:
a.
Specify that the EDS extends below the lower drywell floor and that both shields prevent tunneling of core debris under them.
b.
Specify sloping of the shield roof to prevent accumulation of core debris or show that the long-term debris solidification in the chan-nels is not affected by minor amounts of debris of the roof.
5.
GE should provide the thickness of the EDS corium shield necessary to withstand ablation.
DBA SuDDression Pool Bvoass l
Open Item - GE has not provided SSAR mark-ups on the use of containment sprays i
for long term containment response following a feedwater. line break and main steamline break. Section 6.2.1.1.3.2 indicates that Table 6.2-2a provides the ESF systems for post-blowdown long-term containment cooling. Table 6.2-2a l
clearly indicates the containment spray system. GE has stated that contain-ment sprays are not used in the analyses but the SSAR indicates that they were. These should be consistent.
ECCS Suction Strainers The staff believes that the actions specified by GE are appropriate; however, they do not address the potential lack of conservatism within Regulatory Guide (RG) 1.82, Revision 1 due to the deleterious effect of finely fragmented insulation.
Reducing the total amount of insulation within the containment would not resolve this problem; as the sizing criteria is based on correla-tions within the RG. Therefore, less insulation would lead to smaller strainers. The staff believes an acceptable resolution to this issue is to size the strainers in accordance with RG 1.82, Revision 1 but provide a~ factor of 3 sizing margin to account for uncertainty in the synergetic effects of i
strainer clogging from insulation, corrosion products, and other debris.
Equipment Survivability In-Vessel Severe Accidents The applicable criterion for equipment, both mechanical and electrical, required for recovery from in-vessel severe accidents is provided in 10 CFR 50.34(f).
Section 50.34(f)(2)(ix)(C) indicates that:
Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the environmental conditions
. attendant with the release of hydrogen generated by the equivalent of a 100-percent fuel-clad metal water reaction including the environmental conditions created by activation of the hydrogen control system.
Section 50.34(f)(3)(v) indicates that:
Systems necessary to ensure containment integrity shall be demonstrated to perform their function under conditions associated with an accident that releases hydrogen generated from 100-percent fuel clad metal-water reaction.
Section 50.34(f)(2)(xvii) requires instrumentation to measure containment pressure, containment water level, containment hydrogen concentration, containment radiation intensity, and noble gas effluents at all potential accident release points.
Section 50.34(f)(2)(xix) requires instrumentation adequate for monitoring plant conditions following an accident that includes core damage.
These regulations collectively indicate the need to perform a systematic evaluation of all equipment, both electrical and mechanical, and instrumenta-tion to ensure its survivability for intervention into an in-vessel severe accident. This systematic evaluation has not been performed by GE.
The staff believes that an acceptable resolution of this issue would entail the following:
1.
GE should perform an evaluation using best-estimate means of a degraded in-vessel core damage accident that results in the reaction of a 100-percent metal-water reaction. The basis for the evaluation should be included. The evaluation should identify the most likely sequences resulting in substantial oxidation of the fuel cladding as a result of the probabilistic safety assessment. An example of an acceptable sequence would involve accident conditions where ECCS performance was degraded for a sufficient period of time to cause cladding oxidation but is later recovered to ensure a safe shutdown.
If the analysis assumes an intact primary loop, the basis for this should be supported by the results of the PSA (i.e., LOCA does not contribute significantly to core melt). The impact on the reactor system and containment system from the pressure, temperature, and radiation released should be evaluated. As an example, the safe shutdown and containment equipment identified below should be evaluated.
Plots showing pressure and temperature as a function of time should be provided.
In the event that the in-vessel severe accident environment has no effect on the equipment performance, this should be clearly indicated along with the supporting rationale.
Examples of such instances include cases where the equipment has already performed its function prior to the onset of the accident conditions or the equipment is located in an area not exposed to the environmental conditions, such as being located outside of primary contain-ment.
For equipment where environmental conditions as a result of the in-vessel severe accident are in excess of the equipment qualification range, an
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. engineering rationale must be developed as to why the equipment would survive the environment for the needed time span. This rationale could include such factors as:
limited time period in the environment; the use of similar equipment in commercial industry exposed to the same environment; the use of analytical extrapolations; or the results of tests performed in the nuclear industry or at national laboratories.
An acceptable example using this rationale is the work that GE performed for electrical penetration assemblies in Section 19F.?.2.2 of the SSAR.
In particular, GE referenced experimental tests performed at Sandia National Laboratories on actual electrical penetration assemL11es (EPAs) used in operating plants. The tests were performed at represcatative severe accident conditions with temperatures up to 700 *F and pressures up to 140 psig.
Using the results of this work, GE committed to providing EPAs which will maintain leak tightness up to containment pressure of 134 psig and a tempera-ture of 700 "F.
The end result of this is that the assumptions used for equipment performance in GE's severe accident evaluation are consistent with the as-built plant.
Safe shutdown equipment that should be addressed include:
scram equipment, HPCF motor & pump, HPCF isolation valves, HPCF controls, RCIC turbine and pump, RCIC steam valves & cables, RCIC controls, RHR, ADS, shutdown cooling, etc.
Equipment for containment integrity should include:
containment structure, CIVs - inboard, CIVs - outboard, electrical penetrations, mechanical penetra-tions, hatches, sealing mechanisms (welds, bellows, 0-ring), etc.
2.
With respect to instrumentation requirements, the staff believes that sufficient instrumentation should exist to inform operators of the status of the reactor and the containment at all times as the in-vessel severe accident is intended to be recoverable from and lead to safe shutdown with l
containment integrity maintained. The emergency operating procedures (EOPs) direct specific manual operator actions based on instrumentation readings and as such all instrumentation should exist where manual operator actions are specified within the E0Ps. As a minimum, the instrumentation identified below should be evaluated.
The instrumentation is designed to survive the environment as specified in RG 1.97.
However, RG 1.97 only ensures that the instrumentation will survive in the worst environment resulting from a design bases event and not a severe accident. Therefore, engineering rationale must be developed as to why the instrumentation would survive the environment. This rationale could include such factors as:
limited time period in the environment; the use of similar i
equipment in commercial industry exposed to the same environment; the use of analytical extrapolations; or the results of tests performed in the nuclear industry or at national laboratories.
Instrumentation should include: neutron flux, RPV water level, RPV pressure, sup pool temperature, sup pool level, DW/WW H2 conc, DW/WW 02 conc, DW temperature, DW pressure, WW pressure, WW yemperature, DW water level, etc.
j
. Ex-Vessel Severe Accidents The applicable criteria for equipment, both electrical and mechanical, required to mitigate the consequences of ex-vessel severe accidents is provided in the Equipment Survivability section of SECY-90-016. This section indicates that features provided only (not required for design basis acci-dents) for severe-accident protection (prevention and mitigation) need not be subject to the 10 CFR 50.49 environmental qualification requirements,10 CFR Part 50, Appendix B quality assurance requirements, and 10 CFR Part 50, Appendix A redundancy / diversity requirements. The reason for this judgement is that the staff does not believe that severe core damage accidents should be design basis accidents in the traditional sense that DBAs have been treated in the past.
However, mitigation features must be designed to provide reasonable assurance that they will operate in :he severe-accident environment for which they are intended and over the time span for which they are needed.
In cases where safety-related equipment (equipment provided for DBAs) is relied upon to cope with severe accident situations, there should be reasonable assurance that this equipment will survive accident conditions for the period that is needed to perform its intended function.
According to SECY-90-016, GE was to review the various severe accident scenarios analyzed and identify the equipment needed to perform its function during a severe accident and the environmental conditions under which the equipment must function.
Equipment survivability expectations under severe accident conditions should include consideration of the circumstances of applicable initiating events (e.g., station blackout, earthquakes) and the environment (e.g., pressure, temperature, radiation) in which the equipment is relied upon to function. The staff concludes that GE has not performed the i
evaluation as outlined by SECY-90-016.
The staff believes that an acceptable resolution of this issue would entail the following:
1.
GE should provide an evaluation of the dominant accident sequences identified in Section 19E.2.2 of the SSAR.
For each accident sequence, GE should identify the mitigation features. Mitigation features should include ADS, ACIWA and RCIC, as appropriate.
In addition the specific environment profile (pressure, temperature, radia-tion fields) should be specified.
For each mitigation feature an assessment of survivability should be done using ground rules similar to those specified above for in-vessel accident.
At least the following mitigation features should be evaluated:
SRVs, containment structure, vacuum breakers, CIVs -
inboard, CIVs - outboard, electrical penetrations, mechanical penetrations, hatches, sealing mechanisms (welds, bellows, 0-rings), passive flooders, COPS, COPS CIVs, etc.
2.
With respect to instrumentation requirements, the staff believes that sufficient instrumentation should exist to inform operators of the status of the containment at all times, and of the reactor during the early
. stages of the accident to ensure reactor failure at low pressure or allow for low-pressure injection from the ac independent water addition system.
As a minimum, the list of instrumentation identified below should be evalu-ated. Where extended ranges of operation of the instrumentation is needed, it should be identified along with the environment to which the instrumenta-tion will be exposed.
The instrumentation is designed to survive the environment as specified in RG 1.97.
However, RG 1.97 only ensures that the instrumentation will survive in the worst environment resulting from a design bases event and not a severe accident. Therefore, engineering rationale must be developed as to why the instrumentation would survive the environment.
This rationale could include such factors as:
limited time period in the environment; the use of similar equipment in commercial industry exposed to the same environment; the use of analytical extrapolations; or the results of tests performed in the nuclear i
industry or at national laboratories.
At least the following instrumentation should be evaluated:
RPV water level, RPV pressure, sup pool temperature, sup pool level, DW/WW H2 conc, DW/WW 02 conc, DW temperature, WW pressure, WW temperature, etc.
Eauipment Tunnel Bvoass Mode
]
Guidance provided within SECY-90-016 and SECY-93-087 indicated that contain-ment structures should be protected from direct contact with core debris.
The j
equipment tunnels are located on the periphery of the lower drywell at a mid-level elevation. Core debris exiting the reactor vessel has the potential to reach the tunnels. An accumulation of core debris within the tunnels could lead to melt-through and development of a suppression pool bypass mechanism.
The staff believes that an acceptable resolution to this issue would be for GE to provide reasonable assurance that an appreciable amount of core debris would not enter the tunnels. This can be done by showing that the existing equipment within the lower drywell provides a tortuous pathway to the lower drywell periphery or providing an additional shield structure over the tunnels.