ML092870335

2009/10/13-Exhibit C - NUREG-1437, Volume 1 - Generic Environmental Impact Statement for License Renewal of Nuclear Plants

ML092870335
Person / Time
Site:	Indian Point
Issue date:	10/13/2009
From:	Office of Nuclear Regulatory Research
To:	Atomic Safety and Licensing Board Panel
SECY RAS
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ML092870100	List: ML092870115 ML092870154 ML092870206 ML092870215 ML092870265 ML092870274 ML092870289 ML092870335 ML092870361 ML092870433 ML092870444 ML092870586 ML092870592 ML092870631 ML092870640 ML092870644 ML092870672 ML092870677 ML092870678 ML092870679 ML092870685
References
50-247-LR, 50-286-LR NUREG-1437, V1
Download: ML092870335 (10)

Text

Indian Point Nuclear Generating Units 2 and 3 Docket Nos. 50-247/ 50-286-LR NRC Staff's Response in Opposition to State of New York's Motion for Partial Summary Disposition of NYS Contention 16116A Exhibit C

Generic Environmental Impact Statement for License Renewal of Nuclear Plants Main Report Final Report U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research

ENVIRONMENTAL IMPACTS OF ACCIDENTS SAMDA analyses were included in thc plant-specific basis. In effect, IPE could be final environmental impact statements for considered equivalcnt to a monitoring Limerick 1 and 2 and Comanche Peak 1 program that looks at the severe accident and 2 operating license reviews, and the performance of each licensed plant.

Watts Bar supplemental final Detailed guidance has been issued to each environmental statement for operation. licensee regarding the scope and conduct These actions are addressed below. of IPE and the reporting requircments.

NRC staff intends to review each submittal 5.4.1.1 Containment Performance and, if plant modifications not proposed by the licensee appear warranted, to pursue NRC hasexamined each of five U.S. the incorporation of such modifications via reactor containment types (BWR Mark I, NRC's backfit rule (10 CFR Part 50.109).

I1 and 111; PWR Ice Condenser; and PWR To date, 22 IPEs have been reviewed by Dry) with the purpose of examining the NRC. These IPEs have resulted in plant potential failure modes, potential fixes, and procedural and programmatic the cost benefit of such fixes. This improvements (i.e., accident management) examination has been called the and, in only a fcw cases, minor plant containment performance improvement modifications, to further reduce the risk (CPI) program and has been documented and consequences of severe accidents.

in a series of reports (NUREGICR-5225; NUREGICR-5278; NWREGICR-5528; 5.4.1.3 Accident Management NUREG/ CR-5529; NUREGICR-5565; NUREGICR-5567; NUREGICR-5575; Accident management involves the NUREGICR-5586; N WREGICR-5589; development of procedures that promote NUREGICR-5602; NWREGICR-5623; the most effective use of available plant NUREGI CR-5630). Tables 5.32 through equipment and staff in the event of an 5.34 summarize the rcsults of this program. accident. NRC has indicated its intent As can be seen from these tables, many (Generic Letter 88-20, Supplement 2) to potential changes were evaluated but only request that licensees develop an accident a few containment improvements were management framework that will include identified for site-specific review. The implementation of accident management items evaluated in the CPI program were procedures, training, and technical also included in the list of plant-specific guidance. It is expected that insights gained SAMDAs examined in the Limerick, as a result of IPE will be factored into the Comanche Peak, and Watts Bar FES accident management program. As supplements, discussed later. discussed earlier, the majority of improvements identified from the 5.4.1.2 Lnd~du'alPlant Ikaminations completed IPEs to date have been in the area of accident management or other In accordance with NRC's policy statement procedural and programmatic on severe accidents, each licensee has been improvements.

requested to perform an individual plant examination (IPE) to look for 5.4.1.4 SAMDA Analyses vulnerabilities to both internal and external initiating events (Generic Letter 88-20, Site specific SAMDA analyses were Supplements 1-4). This examination will performed for Limerick, Comanche Peak.

consider potential improvements on a and Watts Bar. A listing of the specific

ENVIRONMENTAL IMPACTS OF ACCIDENTS Table 5.32 Potential boiling-water reador containment improvements considered in the containment performance improvement program Number Potential improvement Resolution Comments 1 Enhanced ADS, low Include in IPE a pressure water supply, and backup power 2 Hardened vent Implemented for b Mark-Is, included in IPE for Mark-I1 and 111s 3 ATWS sized-hardened Drop vent 4 External filter Drop 5 Dedicated suppression Drop pool cooling 6 Alternate decay heat Drop removal 7 Core debris control Drop 8 Enhanced drywell spray ,Drop 9 Drywell head flood Drop 10 Enhanced reactor Drop building D F 11 Backup power for Included in IPE d hydrogen ignitors (Mark 111s)

Acronym: ADS = automatic depressurization system, IPE = individual plant examination, ATWS = anticipated transit without scram, DF = decontamination factor.

"Analpie showed thet potential improvement may be c a t beneficial.

• cost beneflclal for Mark-Is.

'Not cost effective-potential improvement will be loo expensive with too little benetit.

dMay be cost beneficiel.

ENVIRONMENTAL IMPACTS OF ACCIDENTS Table 5.33 Potential pressurized-water reactor ice condenser improvements considered in the containment perforname improvement progrnm Potential improvement Resolution Comments Reactor cavity flooding Drop Not cost beneficial. Might cause ex-vessel steam explosion.

Backup water to the Drop Not cost beneficial containment spray system Backup power to the air Drop Not cost beneficial. May increase return fan system containment pressurization Reactor depressurization Include in Currently being pursued as a viable aceident accident management strategy management Improved hydrogen ignitor Include in Most cost beneficial of all alternatives sptcm (backup power) individual plant considered (although it still does not examination meet the backfit test). To be looked at (IpE) within the IPE program Containment inerting Drop Not cost beneficial, may reduce accessibility for maintenance Filtered vent Drop Not cost beneficial Ex-vessel core debris curb Drop Large uncertainty as to effectiveness Steam generator tube Further research Being examined in separate Nuclear rupture improvements-- needed Regulatory Commieasion program by the increased testing Materials Engineering Branch, RES Containment bypass Included in Being examined as part of a separate improvements generic issue6 interfacing system loss of coolant program accident generic isaue (GSI 105)

Table 534 Potential pressurized-water reactor (PWR) large, dry containment improvements considered in the containment perEormance improvement Program Potential improvement Resolution Comments Operator depressurization Drop No conclusive findings on its benefit to risk using power-operated relief reduction valve Addition of a cavity flooding Drop Not cost beneficial. The effect of a flooded system cavity on the direct containment heating threats may be beneficial or detrimental, depending on each plant Addition of hydrogen Assess in Recommend all dry PWR containments control system individual assess the likelihood of local hydrogen plant detonation in the IPE examination (IPE)

SAMDAs reviewed for applicability to Peak were justified for the purpose of Limerick is provided in Table 5.35. The Further mitigating environmental concerns staff examined each SAMDA (individually as they relate to severe accidents. Recently, and, in some cases, in combination) to the staff evaluated SAMDAs for the Watts determine its individual risk reduction Bar Nuclear Plant. As in the Limerick and potential. This risk reduction was then Comanche Peak analyses, no plant compared with the cost of implementing modifications were justified for the purpose the SAMDA to provide cost-benefit of further mitigating severe accident risk evidence of its value. Considering that the and consequences.

estimates of risk at Limerick used by the staff in these evaluations were considered Several important items from these to be high and that the uncertainties analyses should be noted.

associated with the costs, effectiveness, and/or operational disadvantages of some 0 First, the SAMDAs considered at SAMDAs were large, the staff concluded Limerick, Comanche Peak, and Watts' that there was no clear evidence that Bar covered a broad range of modifications to Limerick were justified for accident prcvcn tion and mitigation the purpose of further mitigating severe features. These features included the accident risks. items that were evaluated for all containment types as part of the CPI The staff made a similar assessment of Program.

SAMDAs for the Comanche Peak Steam Electric Station. A list of the SAMDAs Second, the Limerick analyses were reviewed in this evaluation is provided for a plant at a high population site.

in Table 5.36. As with the Limerick Since risk to the public is generally evaluation, the staff had no basis for proportional to the population concluding that modifications to Comanche surrounding thc plant, one would

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ENVIRONMENTAL IMPACTS OF ACCIDENTS Table 535 Severe accident mitigation design alternatives (SAMDAs) considered for the Limerick Generating Station

1. Installation of alternative means to maintain suppression pool subcooling to improve plant's capability to remove decay heat and prevent containment overpressure challenge

2. Provision of an alternative means of decay heat removal 3a. Installation of containment vent of sufficient size to prevent containment overpressure due to an anticipated transient without scram event 3b. Installation of containment vent and filter of sufficient size to prevent containment sverpresaurc due to an inability to remove decay heat 3c. Installation of containment vent (no filter) of sufficient size to prevent containment overprcssurc due to an inability to remove decay heata

4. Installation of core debris control devices to prevent corelconcrete interaclion and remove decay heat from the core debris 5a. Provide enhanced drywell spray capability to increase the reliability for removal of heat Emm the drywell atmosphere .end the care debris, thereby minimizing the! thperrt of containment failure due to overpressure 5b. Provide modification for tlooding of the drywell head to help mitigate accidents that result in leakage through the drywell head seal

6. Provide the capability lor diesel-driven, low-pressure makeup to the reactor to help in mitigation of core damage resulting from accident sequences In which the reactor vessel is depressurized and all other means of injecting water to the vessel have been lost

7. Improve the reliability of the automatic depressurization system to reduce the probability of vessel failure at high pressure during a severe accident

8. Establish an improved decontamination factor for secondary containment through enhancement to the fire protection system and/or the s t a ~ d b ygas treatment system hardware and procedures to improve fission product removal "l'hls SAMDA has been Implemented lor plants having Mark I conlainmcnts.

ENVIRONMENTAL IMPACTS OF ACCIDENTS Table 5.36 Listing of severe accident mitigation design alternatives considered for the Comanche Peak Steam Electric Station

1. Additional Instrumentation for Bvpass Sequences: Install pressure-monitoring or leak-monitoring instruments (permanent pressure sensors) between the first two pressure isolation valves on low-pressure injection lines, residual heat removal (RHR) suction lines, and high-pressure injection lines. The additional instrumentation would improve the ability to detect valve leakage or open valves, and would decrease the frequency of interfacing system loss-of-coolant accidents (LOCAs).

2. Deliberate Ignition Svstem: Provide a system to promote ignition of combustible gases (hydrogen and carbon monoxide) at low concentrations. The ignition system would prevent large-scale deflagrations or detonations in events involving gradual releases of combustibles (such as from cladding oxidation or core-concrete interactions) but may be ineEfective for rapid releases of hydrogen that could occur coincident with reactor vessel failure at high pressure.

3. Reactor Coolant Svstem Deuressurization: Provide a capability to rapidly depressurize the reactor coolant system. Reactor depressurization would allow injeclion using low-pressure systems and would reduce the threat of direct containment heating and induced failures of steam generator tubes and primary coolant piping in the event low-pressure injection systems are not available. Depressurization could be achieved by a system specially designed to manually depressurize the reactor vessel or by actuation of existing pressurizer power-operated relief valves, reactor vessel heal vent valves, and secondary system valves.

4. Indeuendent Containment Sprav Svstem: Provide an independent containment spray system, using the existing spray headers if appropriate. The spray system would cool the containment and the core debris, thereby reducing the challenge to containment from overtemperature and long-term overpressure by steam. However, unless the sprays ,

terminate core-concrete interactions, the noncondensable gases released from the concrete are expected to cause the containment to eventually fail by overpressure.

5. Reactor Cavitv flood in^ Svstem: Provide a capability to flood the reactor cavity before and after reactor vessel breach. Cavity flooding would promote debris coolability, reduce core-concrete interactions and noncondensable gas production, and provide fission product scrubbing.

6. Filtered Containment Venting: Provide a capability to vent the containment through a vent path routed to an external filter. The filtered vent would mitigate challenges to containment from long-term overpressure and hydrogen burn (by reducing the baseline containment pressure) but may not be effective for mitigating energetic events such as hydrogen burns coincident with reactor vessel failure.

ENVIRONMENTAL IMPACTS OF ACCIDENTS Table 5.36 (continued)

7. Additional Diesel Generator: Provide an additional diesel generator with cross-ties to both Class 1E buses. This modification would increase the availability of the AC power system and reduce the frequency of station blackout sequences.

8. Additional D C Batterv Capability: Provide additional DC batte~ycapability to ensure eight hours of instrumentation and control power, as opposed to four in the event of a station blackout. This would extend the time available for recovery and reduce the frequency of long-term station blackout sequences.

9. Alternative Means of Core Iniection: Provide a capability for makeup water to the reactor using a low-pressure, diesel-driven pump of sufficient capacity and associated piping hardware and procedures. The diesel-driven pump would serve as a backup to the front-line, low-pressure injection systems and could also be used to maintain core cooling in the event of a LOCA

10. Improved Availability of Recirculation Mode: Provide a system to automatically switch the suction of the safety injection and centrifugal charging pumps to the RHR pump discharge when the refueling water storage tank is depleted. Automatic switchover would reduce the potential for operator error and improve the availability of core cooling in the recirculation mode.

1 1 Additional Service Water Pump: Add a third 100 percent service water pump to improve the availability of the station service water system. This would reduce the frequency of sequences involving failure of vital plant equipment due to loss of cooling.

generally expect SAMDAs for plants mitigation improvements, the ongoing at high population sites to have the regul~toryprograms related to severe most favorable cost-benefit ratio. accident mitigation (i.e., individual plant Since SAMDAs were Eound not to be examination/individual plant examination of justified at Limerick, it is unlikely that external events and Accident Management) they would be justified for plants at have not been completed for all plants.

other sites. Since these programs have identified plant programmatic and procedural Third, plant procedural and improvements (and in a few cases, minor programmatic improvements (rather plant modification) as cost effective in than plant modifications) were the reducing severe accident consequence and only cost-beneficial improvements risk, it would be premature to generically identified from tf~eseanalyses. conclude that a consideration of severe accident mitigation is not required for 5.4.1.5 Conclusion license renewal.

Although NRC has gained considerable However, based on the experiences experience regarding severe accident discussed above, the NRC expects that a

ENVIRONMENTAL IMPACTS OF ACCIDENTS site-specific consideration of severe 5.5.1 Impacts from Design-Basis Accidents accident mitigation for license renewal will only identify procedural and programmatic The environmental impacts of postulated improvements (and perhaps minor accidents were ev:vtilunled for the license hardware changes) as being cost-beneficial renewal period in GEIS Chapter 5. All in reducing severe accidcnt risk or plants have had a previous evaluation of consequence. Therefore, a site-specific the environmental impacts of design-basis consideration of alternatives to mitigate accidents. In addition, the licensee will be severe accidents shall be performed for required to maintain acceptable design and license renewal unless such a consideration performance criteria throughout the ha6 already bcen included in a previous renewal period. Therefore, the calculstcd EIS or related supplement. Staff releases from design-basis accidents would evaluations of alternatives to mitigate not be expected to change. Since the severe accidents have already been consequences or these events are evaluated completed and included in an EIS or for the hypothetical maximally exposed supplement for Limerick, Comanche Peak, individual at the time of licensing, changes and Watts Bar; therefore, severe accidcnt in the plant environment will not affect mitigation need not be reassessed for thesc these evaluations. Therefore, the staff plants for license renewal. concludes that the environmental impacts of design-basis accidents are of small significance for all plants. Because the 5.5 SlJMhURY AND CONCLUSIONS environmental impacls of design basis aceidenta are of small significance and The foregoing discussions have dealt with because additional measures to reduce such the environmental impacts of accidents impacts would be costly, the stag concludes during operation after license renewal. The that no mitigation measures beyond those primary assumption for this evaluation is implemented during the current term that the frequency (or likelihood of license would be warranted, This is a occurrence) of an accident at a given plant Category 1 issue.

would not increase during the plant lifetime (inclusive of the license renewal 5.5.2 Impacts from Severe Accidents period) betcause regulatory controls ensure the plant's licensing basis is maintained and 5.5.21 Atmospheric Releasea Improved, where warranted. However, it was recognized that the changing The evaluation of health and dose effects environment around the plant is not caused by atmospheric releases used a subject to regulatory controls and prediction process to identify those plant introduces the potential for changing risk. sites that are bounded by cxisting analyses.

Estimation of future severe accident Existing analyses represent only a subset of consequences and risk was based upon operating plants. A particular portion of existing risk and consequence analyses this subset, specifically those plants having found in FES for recently licensed plants severe accident analyses in their respective because thesc include severe accident FESs, was uscd in this evaluation. EI analyses and constitute a representative set (which is a function of population and of plants and sites for the United States. wind direction), in conjunction with the FES severe accident analyses, was then used to develop a means to predict