ML13196A434

NRC Staff Answer to Motion for Summary Disposition of Contention 4B - Attachment 4B-M

ML13196A434
Person / Time
Site:	Seabrook
Issue date:	07/15/2013
From:	Atomic Safety and Licensing Board Panel
To:
SECY RAS
References
50-443-LR, ASLBP 10-906-02-LR-BD01, RAS 24821
Download: ML13196A434 (10)

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NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1)

License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-M

NUREG-1150 Vol. 1 An Assessment for Five Severe Accident Risks:

An Assessment for Five U.S. Nuclear Power Plants Final Summary Report U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research

3. Surry Plant Results following an accident is provided by the con- tile groups (iodine, cesium, and tellurium) exceed tainment spray recirculation system, whereas, approximately 10 percent (Ref. 3.11). For the by-in some PWR plants, post-accident heat re- pass accident progression bin, the median value moval can also be provided by the residual for the volatile radionuclides is approximately at heat removal system heat exchangers in the the 10 percent level whereas for the early contain-emergency core cooling system. ment failure bin not shown, the releases are lower.

The median values are somewhat smaller than 10

3. Reactor Cavity Design percent, but the ranges extend to approximately The reactor cavity area is not connected di- 30 percent.

rectly with the containment sump area. As a In contrast to the large source term for the bypass result, if the containment spray systems fail bin, Figure 3.8 provides the range of source terms to operate during an accident, the reactor predicted for an accident progression bin involv-cavity will be relatively dry. The amount of ing late failure of the containment. The fractional water in the cavity can have a significant in- release of radionuclides for this bin is several or-fluence on phenomena that can occur after ders of magnitude smaller than for the bypass bin, reactor vessel lower head failure, such as except for iodine, which can be reevolved late in magnitude of containment pressurization the accident. It should be noted that, for many of from direct containment heating and post- the elemental groups, the mean of the distribution vessel failure steam generation, the formation falls above the 95th percentile value. For distribu-of coolable debris beds, and the retention of tions that occur over a range of many orders of radioactive material released during core- magnitude, sampling from the extreme tail of the concrete interactions. distribution (at the high end) can dominate and cause this result.

4. Containment Building Design Additional discussion on source term perspectives The containment volume and high failure is provided in Chapter 10.

pressure provide considerable capacity for accommodation of severe accident pressure 3.4.2 Important Plant Characteristics loads. (Source Term)

Plant design features that affect the mode and 3.4 Source Term Analysis likelihood of containment failure also influence 3.4.1 Results of Source Term Analysis the magnitude of the source term. These features were described in the previous section. Plant fea-In the Surry plant, the absolute frequency of an tures that have a more direct influence on the early failure of the containment* due to the loads source term are described in the following para-produced in a severe accident is small. Although graphs.

the absolute frequency of containment bypass is 1. Containment Spray System also small, for internal accident initiators it is greater than the absolute early failure frequency. The Surry plant has an injection spray system Thus, bypass sequences are the more likely means that uses the refueling water storage tank as a of obtaining a large release of radioactive mate- water source and a recirculation spray system rial. Figure 3.7 illustrates the distribution of that recirculates water from the containment source terms associated with the accident progres- sump. Sprays are an effective means for re-sion bin representing containment bypass. The moving airborne radioactive aerosols. For se-range of release fractions is quite large, primarily quences in which sprays operate throughout as the result of the range of parameters provided the accident, it is most likely that the con-by the experts. The magnitude of the release for tainment will not fail and the leakage to the many of the elemental groups is also large, indica- environment will be minor. If the contain-tive of a potentially serious accident. Typically, ment does fail late in the accident following consequence analysis codes only predict the extended spray operation, analyses indicate occurrence of early fatalities in the surrounding that the release of aerosols will be extremely population when the release fractions of the vola- small. Even in a station blackout case with delayed recovery of sprays, condensation of steam from the air, and a subsequent hydro-

• In this section, the absolute frequencies of early contain- gen explosion that fails containment, Source ment failure aTe discussed (i.e., including the frequencies Term Code Package (STCP) analyses indi-of the plant damage states). This is in contrast to the pre-vious section, which discusses conditional failure prob- cate that spray operation results in substan-abilities (i.e., given that a plant damage state occurs). tially reduced source terms (Ref. 3.12).

NUREG-1 150 3-14

Release Fraction T 1 1.OE+OO Os%

- moan 1.OE-01 modi1In as 1.OE-02 1.OE- 03 1.OE- 04 M M 1.OE - 05 NG I Cs Te Sr Ru La Ba Ce z Radionuclide Group cO an C

Figure 3.7 Source term distributions for containment bypass at Surry.

5. Sequoyah Plant Results effects except for station blackout sequences. In most accident sequences for Sequoyah, there is However, when power is recovered following substantial water in the cavity that can either pre-a station blackout, if the igniters are turned vent core-concrete attack, if a coolable debris bed on before the air-retum fans have diluted the is formed, or mitigate the release of radionuclides hydrogen concentration at or above the ice during core-concrete attack by scrubbing in the beds, the ignition could trigger a detonation overlaying water pool. As a result, a large release or deflagration that could fail containment. to the environment of the less volatile radionu-These blackout sequences, however, repre- clides that are released from fuel during core-sent a small fraction of the overall frequency concrete attack is unlikely for the Sequoyah plant.

of core damage.

In the station blackout plant damage state, con-tainment failure can occur late in the accident as

3. Lower Compartment Design the result of hydrogen combustion following power.

The design and construction of the seal table recovery. Figure 5.7 illustrates the source terms is such that if the reactor coolant system is at for a late containment failure accident progression an elevated pressure upon vessel breach, the bin in which it is unlikely that water would be core debris is likely to get into the seal table available to scrub the core-concrete releases. In room, which is directly in contact with the this case, decontamination by the ice bed is im-containment, and melt through the wall caus- portant in mitigating the environmental release.

ing a break of containment. The design of As discussed previously, for very wide ranges of the reactor cavity, however, does have the uncertainty covering many orders of magnitude, potential to cool the molten core debris and one or more high results can dominate the mean also mitigate the effects of potential direct such that it falls above the 95th percentile.

containment heating events for those se- 5.4.2 Important Plant Characteristics cuences where water is in the reactor cavity. (Source Term) 5.4 Source Term Analysis 1. Ice Condenser In addition to condensing steam, the ice beds 5.4.1 Results of Source Term Analysis can trap radioactive aerosols and vapors in a The absolute frequencies of early containment severe accident. The extent of decontamina-failure from severe accident loads and of tion is very sensitive to the volume fraction of containment bypass are predicted to be similar for steam in the flowing gas, which in turn de-the Sequoyah plant (Ref.. 5.2). Figure 5.6 illus- pends on whether the air-return fans are op-trates the release fractions for an early contain- erational. For a single pass through the ice ment failure accident progression bin. The mean condenser with high steam fraction, the values for the release of the volatile radionuclide range of decontamination factor used in this groups are approximately 10 percent, indicative of study was from 1.3 to 35 with a median of 7 an accident with the potential for causing early fa- for the in-vessel release and less than half as talities. The in-vessel releases in these accidents effective for the core-concrete release. For can be subject to decontamination by the ice bed the low steam fraction scenarios with a single or by containment sprays following release to the pass through the ice beds, the lower bound containment. The sprays require ac power and was approximately 1.1, the upper bound 8, are, therefore, not available prior to power recov- and the median 2. The values used for multi-ery in station blackout plant damage states. The ple passes through the ice bed when the con-decontamination factor of the ice bed is also af- tainment is intact and the air-return fans are fected by the unavailability of the recirculation running are only slightly larger, with a me-fans during station blackout. dian value of 3. Thus, the credit for ice bed retention is substantially less than the values The location and mode of containment failure are used for the decontamination effectiveness of particularly important for early containment fail- suppression pools in the BWRs.

ure accident progression bins. A substantial frac- 2. Cavity Configuration tion of the early failures result in subsequent bypass of the ice bed. In particular, if the contain- The Sequoyah reactor cavity will be flooded ment ruptures as the result of a sudden, high- if there is sufficient water on the containment pressure load, such as from hydrogen deflagra- floor to overflow into the cavity. If the con-tion, the damage to the containment wall could be tents of the refueling water storage tank are extensive and is likely to result in bypass.

NUREG-1 150 5-12

Release Fraction 1.OE+OO 95%

mean 1.OE-O1 median 6%

1.OE-02 up w-1 1.OE-03 1.OE-04

<A (n

D PA M 0 1.OE-05 --- -- -- ...

NG I Cs Te Sr Ru La Ba Ce ZI

(-I d Radionuclide Group M

c3e OI 0 Figure 5.6 Source term distributions for early containment failure at Sequoyah.

7. ZION PLANT RESULTS 7.1 Summary Design Information corporating some methods and issues (such as common-cause failure treatment, electric power The Zion Nuclear Plant is a two-unit site. Each recovery, and reactor coolant pump seal LOCA unit is a four-loop Westinghouse nuclear steam modeling) used in the other four plant studies.

supply system rated at 1100 MWe and is housed in a large, prestressed concrete, steel-lined dry The objective of this study was to perform an containment. The balance of plant systems were analysis that updated the previous Zion analyses engineered by Sargent & Lundy. Located on the and cast the model in a manner more consistent shore of Lake Michigan, about 40 miles north of with the other accident frequency analyses. The Chicago, Illinois, Zion 1 started commercial op- models were not completely reconstructed in the eration in December 1973. Some important de- small-event-tree, large-fault-tree modeling method sign features of the Zion plant are described in used in the study of the other NUREG-1150 Table 7.1. A general plant schematic is provided plants. Instead, the small-fault-tree, large-event-in Figure 7.1. tree models from the original ZPSS were used as the basis for the update. These models were then This chapter provides a summary of the results revised according to the comments from Refer-provided in the risk analyses underlying this report ence 7.3 and were enhanced to address risk issues (Refs. 7.1 and 7.2). A discussion of perspectives using methods employed by the other plant stud-with respect to these results is provided in Chap- ies.

ters 8 through 12.

This study incorporated specific issues into the 7.2 Core Damage Frequency Estimates systems and accident sequence models of the ZPSS. These issues reflect both changes in the 7.2.1 Summary of Core Damage Frequency Zion plant and general PRA assumptions that Estimates* have arisen since the ZPSS was performed. New The core damage frequency and risk analyses per- dominant accident sequences were determined by' formed for this study considered accidents initi- modifying and requantifying the event tree models ated only by internal events (Ref. 7.1); no exter- developed for ZPSS. The major changes reflect nal-event analyses were performed. The core the need for component cooling water and service damage frequency results obtained are provided water for emergency core cooling equipment and in tabular form in Table 7.2. This study calculated reactor coolant pump seal integrity. The original a total median core damage frequency from inter- set of plant-specific data used in the ZPSS and nal events of 2.4E-4 per year. Zion Review was verified as still valid and was used for this study. Additional discussion of the 7.2.1.1 Zion Analysis Approach Zion methods is provided in Appendix A.

The Zion plant was previously analyzed in the 7.2.1.2 Internally Initiated Accident Zion Probabilistic Safety Study (ZPSS), per- Sequences formed by the Commonwealth Edison Company, and in the review and evaluation of the ZPSS A detailed description of accident sequences im-(Ref. 7.3), commonly called the Zion Review pre- portant at the Zion plant is provided in Reference pared by Sandia National Laboratories. 7.1. For this summary report, the accident se-quences described in that reference have been Since previous analyses of Zion already existed, it grouped into six summary plant damage states.

was decided to perform an update of the previous These are:

analyses rather than perform a complete reanalysis. Therefore, this analysis of Zion repre-

• Station blackout, sents a limited rebaseline and extension of the dominant accident sequences from the ZPSS in

• Loss-of-coolant accident (LOCA),

light of the Zion Review comments, although in-

• Component cooling water and service water

'In general, the results and perspectives provided here do induced reactor coolant pump seal LOCAs, not reflect recent modifications to the Zion plant. The benefit of the changes is noted, however, in specific places in the text (and discussed in more detail in Section

• Anticipated transients without scram 15 of Appendix C). (ATWS),

7-1 NUREG-1 150

7. Zion Plant Results state frequency-weighted average,

• the mean con- 7.4 Source Term Analysis ditional probabilities from internal events of (1) early containment failure from a combination of 7.4.1 Results of Source Term Analysis in-vessel steam explosions, overpressurization, The containment performance results for the Zion and containment isolation failures is 0.014, (2) (large, dry containment) plant and the Surry (sub-late containment failure, mainly from basemat atmospheric containment) plant are quite similar.

meltthrough is 0.24, (3) containment bypass from The source terms for analogous accident progres-interfacing-system LOCA and induced steam gen- sion bins are also quite similar. Figure 7.5 illus-erator tube rupture (SGTR) is 0.006, and (4) trates the source term for early containment fail-probability of no containment failure is 0.73. Fig- ure. As at Surry, the source terms for early failure ure 7.4 further displays the conditional probability are somewhat less than those for containment by-distributions of early containment failure for the pass. Within the range of the uncertainty band, plant damage states, thereby providing the esti- however, the source terms from early containment mated range of uncertainties in these containment failure are potentially large enough to result in failure predictions. The principal conclusion to be some early fatalities.

drawn from the information in Figures 7.3 and 7.4 is that the probability of early containment The most likely outcome of a severe accident at failure for Zion is low, i.e., 1 to 2 percent. the Zion plant is that the containment would not fail. Figure 7.6 illustrates the range of source Additional discussion on containment perform- terms for the no containment failure accident pro-ance is provided in Chapter 9. gression bin. Other than for the noble gas and io-dine radionuclide groups, the entire range of source terms is below a release fraction of 10E-5.

7.3.2 Important Plant Characteristics (Containment Performance) Additional discussion on source term perspectives Characteristics of the Zion design and operation is provided in Chapter 10.

that are important to containment performance include: 7.4.2 Important Plant Characteristics (Source Term)

1. Containment Volume and Pressure Capa-bility 1. Containment Spray System The combined magnitude of Zion's contain- The containment spray system at the Zion ment volume and estimated failure pressure plant is not required to operate to provide provide considerable capability to withstand long-term cooling to the containment, in con-severe accident threats. trast to the Surry plant. Operation of the spray system is very effective, however, in re-

2. Reactor Cavity Geometry ducing the airborne concentration of aero-sols. Other than the release of noble gases The Zion containment design arrangement and some iodine evolution, the release of ra-has a large cavity directly beneath the reactor dioactive material to the atmosphere resulting pressure vessel that communicates to the from late containment leakage or basemat lower containment by means of an instru- meltthrough in which sprays have operated ment tunnel. Provided the contents of the re- for an extended time would be very small.

fueling water storage tank have been injected The source terms for the late containment prior to vessel breach, this arrangement failure accident progression bin are slightly should provide a mechanism for quenching higher than, but similar to, those of the no the molten core for some severe accidents containment failure bin illustrated in Figure (although there remains some uncertainties 7.6.

with respect to the coolability of molten core debris in such circumstances). 2. Cavity Configuration The Zion cavity is referred to as a wet cavity,

'Each value in the column in Figure 7.3 labeled "All" is a in that the accumulation of a relatively small frequency-weighted average obtained by calculating the amount of water on the containment floor products of individual accident progression bin condi- will lead to overflow into the cavity. As a re-tional probabilities for each plant damage state and the ratio of the frequency of that plant damage state to the sult, there is a substantial likelihood of elimi-total core damage frequency. nating by forming a coolable debris bed or 7-9 79NUREG-1 150

z 0

co Release Fraction 0d OR 1.OE+OO ED 95%

mean 1.OE-O1 median 5%

1.OE-02

-i3 0

1.OE-03 1.OE-04 1.OE-05 NG I Cs Te Sr Ru La Ba Ce Radionuclide Group Figure 7.5 Source term distributions for early containment failure at Zion.

7. Zion Plant Results REFERENCES FOR CHAPTER 7 7.1 M. B. Sattison and K. W. Hall, "Analysis of 7.5 R. A. Chrzanowski, CECo, "March 13, 1989 Core Damage Frequency: Zion Unit 1," Letter from Cordell Reed to T. E. Murley,"

Idaho National Engineering Laboratory, NRC, NRC Docket Nos. 50-295 and NUREG/CR-4550, Vol. 7, Revision 1, 50-304, August 24, 1990.

EGG-2554, May 1990.

7.6 H. J. C. Kouts et al., "Special Committee 7.2 C. K. Park et al., "Evaluation of Severe Ac- Review of the Nuclear Regulatory Commis-cident Risks: Zion Unit 1," Brookhaven Na- sion's Severe Accident Risks Report tional Laboratory, NUREGICR-4551, Vol. (NUREG-1150)," NUREG-1420, August 7, Draft Revision 1, BNL-NUREG-52029, 1990.

to be published.*

7.7 Stone & Webster Engineering Corporation, 7.3 D. L. Berry et al., "Review and Evaluation of "Preliminary Evacuation Time Study of the the Zion Probabilistic Safety Study: Plant 10-Mile Emergency Planning Zone at the Analysis," Sandia National Laboratories, Zion Station," prepared for Commonwealth NUREG/CR-3300, Vol. 1, SAND83-1118, Edison Company, January 1980.

May 1984.

7.8 Federal Emergency Management Agency, 7.4 Cordell Reed, Commonwealth Edison Co. "Dynamic Evacuation Analyses: Independ-(CECo), "Zion Station Units 1 and 2. Com- ent Assessments of Evacuation Times from mitment to Provide a Backup Water Source the Plume Exposure Pathway Emergency to the Charging Oil Coolers," NRC Docket Planning Zones of Twelve Nuclear Power Nos. 50-295 and 50-304, March 13, 1989. Stations," December 1980.

7.9 USNRC, "Safety Goals for the Operation of Nuclear Power Plants; Policy Statement,"

• Available in the NRC Public Document Room, 2120 L Street FederalRegister, Vol. 51, p. 30028, August NW., Washington, DC. 21, 1986.

7-19 NUREG-1150