ML20209D535

Informs That Phrase Estimate for Cost Benefit Analysis Should Be Added to Last Line on Page 28 of Encl Draft Generic Ltr for Improvements to BWR Mark I Containments

ML20209D535
Person / Time
Issue date:	12/05/1986
From:	Bernero R Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20209C630	List: ML20209C625 ML20209C668 ML20209D535 ML20209E077 ML20209F841 ML20209G812 ML20210P208 ML20211H552 ML20211H757 ML20211K468 ML20211K540 ML20211N263 ML20211Q265 ML20212K322 ML20212Q077
References
FOIA-87-10 NUDOCS 8704290305
Download: ML20209D535 (1)

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k MEMORANDUM FOR: Distribution

[ FROM: R. Bernero, Director Division of BWR Licensing L

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SUBJECT:

ERRATUM IN PROPOSED GENERIC LETTER FOR IMPROVEMENTS

{ TO BWR MARK I CONTAINMENTS r On page 23 of the generic letter attachment, the words " estimate for the cost benefit analysis" should be added to the last line.

R. Bernero, Director Division of BWR Licensing I

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MEMORANDUM FOR: Eric S. Beckjord, Director Office of Nuclear Regulatory Research James M. Taylor, Director Office of Inspection & Enforcement William C. Parler, Director Office of General Counsel FROM: Robert M. Bernero, Director Division of BWR Licensing

SUBJECT:

GENERIC REQUIREMENTS FOR BWR CQNTAINMENT RESPONSE TO SEVERE ACCIDENTS Enclosed for your comment's is a Draf t Generic Letter for Proposed BWR Containment Requirements which will enhance containment performance in severe accidents. The requirements specified in the Draft Generic Letter have been derived by regulatory analysis comparing the behavior of BWR containments in severe accident environments, and the benefits derived from proposed containment enhancements. The regulatory analysis is also enclosed for your comments.

Your comments and concurrence are requested by December 12, 1986. The material is scheduled to be presented to ACRS on December 9 and 12, 1986.

A CRGR review is planned for late December, 1986. The Generic Letter is scheduled to be issued for industry and public comments by the end of January 1987 after review by the CRGR and with the Commission and as final by May, 1987. Please provide your comments in time to support the schedule.

- v Robert M. Bernero, Director Division of BWR Licensing CONTACT:

M. C. Thadani X28649

Enclosures:

1. Oraft Generic Letter on Proposed BWR Severe Accident Containment Requirements

2. Regulatory Analysis cc: See Next Page ggne'7 ,,

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H. Centon R. Vollmer W. Russell T. Speis F. Miraglia T. Novak H. Thompson D. Ross, RES J. Partlow, IE E. Jordan, IE J. Scinto, 0GC J. Funches W. Houston G. Lainas G. Holahan DBL PD's DBL SC's l

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1. . i TO ALL BOILING WATER REACTOR (BWR LICENSEES AND APPLICANTS FOR BOILING WATER REACTOR LICENSES Gentlemen:

SUBJECT:

PROPOSED EWR SEVERE ACCIDENT CONTAINMENT REQUIREMENTS (GENERIC LETTER 87- )

Severe accidents dominate the risk to the public associated with nuclear power plant accidents. A fundamental objective of the Commission's Severe Accident Policy is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an accident should one occur. The Reactor Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt were low, but the containment performance following a severe accident was poor and tended to offset the benefits of low BWR core melt probabilities. Subsequent actions resulting from the TMI Action Plan have led to several plant modifications and required improvements in plant procedures to further reduce the likelihood of severe accioents. In December 1980, an industry initiative on severe accidents resulted in the formation of the Industry Degraded Core Rulemaking (10COR) group to address the concerns related to core damaging accidents. The IOCOR effort has led to industry methodology for Individual Plant Evaluations (IPEs) to search for the risk outliers and to address system reliability and containment performance on a plant specified basis. The staff has concluded, however, that for BWR containments, a set of generic requirements has been identified that moots the need to await plant specific analyses of containment performance and will lead to speedier implementation than would be possible via the IPEs. Severe accident ana!yses have indicated several areas for improvement in SWR containments which should be promptly pursued as follows:

1. Hydrocen Control Present recuirements imposed by 10 CFR Part 50.44 and the Technical Specifications shall be adhered to, no additional requirements are proposed.

2. Containment Soray All BWRs with Mark I containment shall provide at least two backup water supply systems for the containment drywell spray, one of which shall be functional during station blackout. Water to the spray system from these backup supplies shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during a severe accident scenario.

In addition, the spray nozzles shall be adjusted 50 that an evenly distributed spray pattern will be developed in the drywell whether

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water is supplied by the primary source or either of the backup sources. A flow rate on the order of 1/10 of the present flow rate is

• considered typical, the licensee shall select the flow based on an analysis of plant specific parameters. -

3. Pressure Relief The licensee shall select a pressure between design pressure and 1h times cesign pressure at which to open an exhaust path from the wetwell vapor space to the highest vent point (stack or pipe) available. This line should be capable of handling water vapor flow equivalent to 1% decay heat at the vent pressure selected without significant chance of rupture before the desired release point. The line shall be equipped with isolation valves which can be opened and reclosed by remote manual operation or by simple procedures which can be implemented during severe accident scenarios including station blackout 4 Core Oebris Control The licensee shall ensure that the water in the suppression pool in the event of torus failure is held within the confines of the torus room and the corner rooms and cannot flow out to other parts of the plant.

5. Procedures and Training The licensee shall implement emergency operating precedures and other procedures based on all significant elements appropriate to its plant of Emergency Procedure Guidelines, Revision 4.

Since these requirements are intended to be an optimized use of existing equipment it is expected that added equipment, of itself, need not meet the quality or design standards of safety related equipment. Nevertheless, modifications to or near equipment or systems which are already safety related shall not cocpromise the quality of such equipment or systems.

The equipment changes required herein shall be installed during the first refueling cutage which begins nine (9) months after the effective date of this letter. The procedures and training required shall be implemented on a schedule reviewed and approved by the NRC. Given the implementation of the generic improvements of Mark I containments there is no need for an Individual Plant Evaluation (IPE) for containment oerformance. This does not remove the need for an IPE which covers the system reliability or core melt frequency portion of the severe accident question.

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This proposed generic letter has identified areas for improvement in BWR containments which the staff believes to be effective in reducing risk and

• which can be implemented at a reasonable cost. We welcome comments on the proposed acticns and other. t suggestions on the subject matter. The goal is to significantly reduce the[2 likelihood of containment failure given a core melt.

Sincerely, Robert M. Bernero, Director Division of BWR Licensing Office of Nuclear Reactor Regulation

Enclosure:

BWR Mark I Containment Performance During Severe Accidents i

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BWR MARK I CONTAINMENT PERFORMANCE L:ilif l" DURING SEVERE ACCIDENTS

1. 0 BACKGROUND A fundamental objective of the Commission's Severe Accident Policy of August 8,1985 is to take all reasonable steps to reduce the chances of occurrence of a severe accident and to assure substantial capability to mitigate the consequences of such an accident should one occur. The Commission also called for a balancing of accident prevention and mitigation, and special consideration of containment performance in searches for risk outliers.

Enhancements to the performance of containments in severe accidents should increase assurance of mitigation of severe accident consequences. The Reactor A

Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt were low, but the containment following a severe accident could be severely challenged and tended to offset the benefits of low BWR core melt probabilities. Subsequent actions resulting from the TMI Action Plan have led to several plant modifications and required improvements in plant procedures to further reduce the likelihood of severe accidents.

Other post TMI actions have also involved containment enhancements, particularly in the areas of isolation dependability and hydrogen control.

In concert with the Commission's policy to further reduce the chances of occurrence of severe accidents and to mitigate their consequences, an industry initiative is underway to develop a methodology for Individual Plant Evaluation (IPE) directed to search for risk outliers. The resulting approach will be applied on a plant-specific basis. The initial IPE trials

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have been made by industry, and have encompassed internal accident initiators ano systemi reliabilities to the point of estimating core melt probabilities.

Scurce terms, containment performance and offsite risks have not been considered, but have been discussed as future extensions of initial IPEs.

With respect to BWRs with Mark I, II and III type containments, the staff has reviewed these initial IPEs, historical probabilistic risk assessments and the plans for completing the search for individual plant outliers. The review has indicated that sufficient bases exist to complete the search for outliers for all such plants with respect to accident mitigation by backfitting in five areas as discussed in the subsequent sections. That is, by requiring improvements in five areas, no further evaluations of accident mitigation for BWR reactors with Mark I, II and III type containments are considered necessary.

The staff has identified five potential containment enhancements which lend themselves to generic implementation and have the potential to significantly mitigate the consequences of several severe accident sequences including station blackout and ATWS sequences. In the Policy Statement the Commission stated that the rulemaking route would generally be a preferred route to implement future severe accident related actions. However, rulemaking is extremely time consuming. The Commission's statement regarding operating reactors recognized the time element and the continued severe accident risk to public health and safety, and provided other options to dispose of the issues

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e through the conventional practice of issuing Bulletins and Orders or Generic Letters. In the interest of time the staff has chosen the Generic Letter as a preferred approach to achieve closure .of the BWR containment performance issues. Containments are required to protect the public from the consequences of accidents. The design and sizing of containment are required to assure that the containments are essentially leak-tight barriers against the uncontrolled release cf radioactivity to the environments and to assure that containment design conditions important to safety are not exceeded for as long cs 1

1 postulated accident conditions require. The containments should accommodate l with sufficient margins, the pressures and temperatures resulting from any loss-of-coolant-accident (LOCA).

Although a postulated design basis LOCA is not expected to produce more than a few percent fuel failures, an accident radiological " source term" used in in calculating offsite dose consequences is representative of a substantial core melt accident (10 CFR 100). Even for this source term, containments are designed such that calculated offsite doses are unlikely to result in an early or maior latent health hazard if the containments were to maintain their low leakage capability

• What is at issue is the capability of containments to perform a mitigating safety function as long as practicable during very low probability severe accidents, where the stress on containment may significantly exceed that of a design basis LOCA and the consequences of containment failure may be very significant. The structural integrity of BWR containments is seriously challenged for accidents with high energy release "Part 100 specifies "the expected demonstrable leak rate from the containeent", a value which is made part of each licensee's Technical Specifications.

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, 4 to the containment because in spite of the positive pressure suppression feature, they are either relatively small (Mark I & II); or have a very low design pressure (Mark III), and the likelihood of their failure in a severe accident is perceived to be higher than necessary.

Overall plant core melt probabilities for BWRs with Mark I, II and III contain-ments have been estimated to range from one in a thousand per reactor year to two in ten million per reactor year for BWR designs evaluated by the NRC and the industry. Many of these estimates have not fully included assessments of the benefits of post-TMI backfits, operator responses, or the increases in core melt probabilities arising from factors not considered in plant specific analyses such as earth-I quakes, floods and fires. Contemporary analyses break down such probabilities into classes and subclasses of accidents. The sum of the core melt probabilities for all classes and subclasses of accidents is considered to be the overall core melt pretacility. For BWRs with Mark I containments, IDCOR 1'2 has proposed the follow ng five classes of events for core melt accidents:

1 2 Individual Plant Evaluation - Peach Bottom Atomic Station, May 1986.

IDCOR Technical Report 85-3-B1, BWR Accident Sequence - Individual Plant Evaluation Methodology, April 1986.

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loss of core cooling with containment at low pressure and failure after core melt; loss of core cooling with containment failure before core melt; loss of core cooling with containment failure soon after core melt due to high containment pressure at the time of core melt; loss of core cooling with containment failure before core melt due to failure to depressurize; and containment bypass.

Our review of the IDCOR core melt probability estimates to date generally indicates that they are low. The BWR core melt frequencies of past evaluations are summarized in Table 1. ,Given a core melt, the estimates of likelihood of Mark I, II and III containment failures have been high relative to other containment types. In all of these past evaluations, little or no credit has been given to features wnich can be used with relatively modest upgrading to prevent or mitigate accidents.

The Reactor Safety Study (WASH-1400, NUREG-75/14) indicates a conditional containment failure probability for the BWR Mark I containment reference plant (Perch Bottom) of about 90% (inferred from Table 5-3, page 81). That is, given a core melt in a BWR with a Mark I containment (Peach Bottom) there is about 90% chance of containment failure. In the November 1984, 10COR Technical Summary Report, Nuclear Power Plant Response to Severe Accidents, the estimate for Peach Bottom was about 20% (inferred from Table 10-1, page 10-6). More recently, the Vermont Yankee Containment Study provided an estimate that

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TABtf 1 - U 5. BWR PLANI-5PfCIFIC PRA STuDif5 PROGRAM REPORI CORf/ REAC10R CORE-DAMAGE EVEN15 MEDIAN, CONTAINMENT l PLANI HAMC REPORI Y[AR CONIAltMINI POWfR (MW1) FRfQufMCY PRA CON 51DERED MEAN OR CON 0lil0NAL i E511 MAIL P0lNI IAllDRE I (511 MALE PRf*ABIlliY Peach RSS WA5H-1400 1915 BWR 4/MK I 3293 3ml0'b Internal / Median Not evaluated Botton -5 Enternal Peach IDCOR lech Sussmary 1984 BWR-4 M I 3293 4n10 Internal Mean 0.2 Bottom lask 21 -5 Peach IPE IPF 1986 BWR-4 M i 3293 2x10 Internal Mean Not evaluated Bottom Millstone IRIP NUREG/CR I?83 BWR-3/MK I 1727 3x10'4 Internal Median Not evaluated 3085 4 Hillstone NUSCO Millstone 1 1986 BWR-3 M i 1727 5:10 Internal Mean Not evaluated P55 Brown, ferry IREP NUREG/CR 1982 BWR-4/MK I 3293 2x10'4 Internal Point Not evaluated 2801 Estimate Vermont VYC55 VYC55 1986 BWR-4/MK I 1593 3 10 -5 Internal Mean 0.07 Vankee Big Rock Consumers Big Rock 1981 BWR-1/ Dry 158 1:10'3 Internal / Mean 0.25 Point Point PRA 3 External Big Rock EG&G/BNL EG&G-EA- 1982 BWR-1/ Dry 158 luto Internal / Mean 0.25 Point $533 Rev. 1 External Limerick PEPC0 Limerick PRA 1981 BWR-4/MK II 3293 7 10'S Internal / Mean 1.0 External Lleerick BNL NUREG/CR- 1983 BWR-4 M 11 3293 1:10-4 Internal / Mean 1.0 3028 External Shorehae Shoreham PAA 1983 -5 LILCO BWR-4/MK !! 2436 5x10 Internal

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Shoreham BNL NUREG/CR- 1985 BWR-4/MK II 2436 In10 Internal Point Not evaluated 4050 Estleate Shoreham Shoreham IPE 1986 -5 IPE BWR-4/MK 11 2436 6 10 Internal Mean Not evaluated Susquehanna IPE IPE 1986 BWR-4/MK !! 3293 2ml0'I Internal Mean Not evaluated Grand Gulf R$5 MAP NURfG/CR- 1981 BWR-6/MK Ill 3833 410 Internal Median Not evaluated 1659 Grand Gulf Tech Susanary 1984 BWR-6/MK 111 3831 -6 IDCOR 8:10 Internal Mean Mot evaluated Task 21 GESSAR GE GESSAR !! PRA BWR-6/MK 111 3579 4mg6 Internal / Mean Not evaluated External 1*

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Vermont Yankee, a somewhat smaller BWR with Mark I containment, has a condi-tional containment failure probability of about 7%. In all these Mark I containment failure estimates tne challenges come from a spectrum of accidents including ATWS, station blackout, and ordinary transients. The principal causes of failure are overpressure and direct attack of the drywell.

The accidents of interest span a spectrue of sequences and will have a probability distribution unique to each plant. Nevertheless, because of the uncertainties in calculating the dominant accident sequences, it is prudent to consicer each principal type as the cause of large scale core melt and containment challenge.

For a BWR Mark II containment (Limerick), Brookhaven National Laboratory (BNL) estimated almost a 100% likelihood of containment failure give a core melt (ENL 33835; April 1984). The staff evaluation of the GESSAR II standard plant cesign (NUREG-0979, Supplement 4, Tables 15.1, 15.2, and 15.12) also, indicates a concitional containment failure probability of close to one for a Mark III type ccatainment. Only 10COR and GESSAR II evaluations considered containment s

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For most accidents considered, the core is postulated to melt, interact with steam, water, and the structural features in the vessel and coolant system, melt through the vessel, and attack the concrete and structural features of the lower containment. Depending on the sequence of events, the containment has the potential to fail either before or after vessel melt through. For the remainder of the accidents postulated, the containment would be bypassed, allowing radioactivity a dirert path to portions of a plant not designed to contain the releases, but with some capability to attenuate radioactivity.

BWR containments respond to heatup of the fuel in the vessel directly or indirectly. Tne direct transfer of energy is through pipe breaks, through blowdown into the suppression pool or by the aerosols generated when the core melts through the vessel. Indirectly, radiant heat is transferred through the vessel and piping. The blowdown or depressurization process, and the use of the relatively large cuantity of suppression pool water as a heat sink and fission product scrubbing cevice, act in combination with the structural capability of the containment (including penetrations) to mitigate the high temperatures, pressures and radioactivity releases in a core melt. Core melt scenarios have been identified, however, which can produce conditions that could lead to containment failures, and release of fission products to the environment without the benefit of the suppression pool scrubbing. There is strong evidence that BWR containments are capable of withstanding substantially higher stresses than those for which they have been explicitly designed and this potential containment strength can be drawn upon to demonstrate additional

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1 protection to the public at low to modest cost. The longer a containment can be expected to hold, the greater the likelihood that failure can be avoided.

If failure were to occur, however, reductions in the radioactivity released would be achieved. Actions that can be taken to prevent a catastrophic failure of containment before the fission products are acequately attenuated include sucn items as operator actions to vent the wetwell space above the suppression pool, and providing reliaole spray capability.

In a core melt accident with temperatures in excess of 5000 degrees F, fission produ' cts are released from the fuel in three general groups. The noble gases and the more volatile species of fission products are released from the fuel relatively early in a core melt accident. Later, the less volatile species are released as the fuel melts down into the vessel and combine with the in vessel structural materials. Finally, after melting through the vessel, refractory materials may be released during interactions of core debris with concrete on the floor of the containment.

The amount of radioactivity that could be released to the environment in core melt or degraded core accidents has been the subject of considerable analysis for a number of years. Present estimates (NUREG-0956) for MK I and III BWRs indicate that substantial quantities of important fission products can be released in a core melt accident and these analyses provide clues that suggest that releases can be reduced by a number of actions to enhance containment performance.

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Within the core of a contemporary BWR with MK I, II, or III containment at full power there are over five billion curies of radioactivity. Severe accident releases to the environment for a MK I, II or III have been estimated to exceed 40% of such important fission products as iodine and cesium (releases of over 300 million curies of iodine and over seven million curies of cesium for a 3458 MWt reactor).

2. 0 NEEDS AND STRATEGY FOR CONTAINMENT IMPROVEMENT Consideration of the insights drawn from previous analyses suggests that no single simple feature can be added to a BWR pressure suppression containment to provide substantial assurance that it will successfully mitigate the consequences of a large scale core melt should one occur. Rather, one must conceive of some integral approach which deals with the principal concerns.

Consider now only the Mark I containments, 24 of which are now found in licensed U.S. reactors. This analysis and development of requirements will deal first with Mark I containments because they constitute about 2/3 of the BWR copulation. Subsequent analyses will deal with Mark II and III containments.

Compared to many other U.S. reactor containments, the Mark I containment (Figure 1) has a small volume relative to the size of the reactor it contains. With a free volume of less than 300,000 cubic feet, the drywell wall is very close to the reactor and to the lower head area where molten core material would

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most likely fall from the reactor vessel. Even with a relatively high design pressure, typically about 50-60 psig, the small volume makes the Mark I containment more vulnerable to overpressure failure, given a compa*able core melt event. Any strategy to enhance Mark I containment performance must certainly consider preventing hydrogen combustion, cooling the non-condensibles in containment, and as a last resort, venting serious overpressure through some available path, where the consequence of venting is known and is preferred to the potential of uncontrolled release due to containment loss.

Should molten core material (corium) reach the drywell floor, the direct attack of the drywell becomes a serious concern. If the corium is sufficiently hot to flow with low viscosity it can easily reach the nearby wall of the drywell.

There it will attack the steel wall of the drywell between the vents or attack one cf the large steel vent passages leading down to the wetwell. The steel shell of the drywell is typically backed by a 1-2 inch construction gap filled with a plastic spacer and then by a very thick, reinforced concrete biological shield. Most analyses do not attempt to treat attack through the shell and shield mechanistically because of the complexity of the path, but it is apparent that this path to the reactor building and the ambient is not an open one, especially if some means are available to reduce the vigor of the attack by the hot corium.

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, 4 The presence of hot corium on the floor of the drywell raises other challenges.

The corium can be expected to react with the concrete floor, thus generating large quantities of non-condensible gas (including hydrogen) as well as -

voluminous aerosols carrying non-volatile but health threatening radionuclides such as lanthanum and plutonium. In addition, radiative and convective heat transfer can directly attack the steel shell'and its penetrations. Any strategy to enhance the performance of Mark I containments must seek some means to cool or quench the core debris. That strategy should also include means to cool the drywell wall to prevent overheating.

Overpressure failure of the Mark I containment may be averted even in a large scale fuel melt if debris cooling and quenching limit the amount and the temperature of the non-condensible gases in containment. Nevertheless, it is possible that pressure and temperature can build up to levels which could cause containment penetration (seal) failures or catastrophic rupture of the containment. It is desirable to have available a procedure of last resort whereby the tnreatening overpressure Can be relieved from the wetwell vapor space so that all gases released from containment will have passed through the water in the suppression pool, and thus will have been scrubbed of most non gaseous fission products. The pressure at which such relief should be taken into account must account for the ultimate strength of containment, the reliability of the valves used for venting, and backpressure effects on SRV operation. In addition, consideration should also be given to the material vented from the containment. At a minimum it will contain water

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vapor, nitrogen, unburned hydrogen, and (depending on the stage of the accident) fission product noble gases (principally Xe 133). If the path through the downtcmers and the pool water is bypassed, perhaps through a vacuum breaker line, the effluent could contain other fission products as well.

The core debris of concern includes not only the corium which melts through the pressure vessel but the large amount of aerosols which may be released and captured by the water in the suppression pool. Using Three Mile Island experience as a guide, the suppression pool water might absorb radioactive material on the order of 0.01 to 0.1 Ci/ml (0.4-4 gigaBq/ml). That water, almost I million gallons of it, would be so hot (redioactive) that it would be very desirable to see to it that it stays in the torus or at least in the torus room and immediately adjacent spaces should the torus fail. If some molten corium coes pass down through one or more of the eight drywell-to-torus vent cucts, then it would most likely cause torus rupture. In that event the water, if retained, would be available to quench the corium.

1 Finally, it is evident from all previous studies that the Mark I containment snould not be treated as a simple passive body. To be effective in mitigating core melt accidents, its features must be used by trained on-site personnel who are prepared to deal with these extreme events using the equipment at hand.

i Thus, the containment improvement strategy should include procedures and training for such accident management.

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. 1 3.0 NARK I MITIGATION FEATURES Af ter considering the technical factors identified in the preceding, a 5 4

eiete : strategy stands out as an effective solution to improve Mark I i containment performance to the point that there would be reasonable assurance that Mark I containments can substantially mitigate the consequences of a large-scale core melt accident. The five elements are:

1. Hydrogen Control

2. Containment Spray j 3. Pressure Relief

4. Core Debris Control

5. Procedures and Training 3.1 Hydrocen Control Under the present requirements of 10 CFR Part 50.44, all Mark I containments are requirec to have their containments inerted (with nitrogen gas) during operation. Allowance is made for a period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the beginning and at the end of the power operations cycle to operate with air in the containment to enable operators to inspect equipment in the containment for leaks etc. From time to time small, unidentified system leaks will start inside the containment during power operation. Although operators sometimes cuestion whether the 10 CFR 50.44 based technical specifications permit deinerting and entry during a cycle to investigate such leakage, they have used the 24-hour deinerted periods permitted in the technical specifications to investigate and to the extent practicable repair such leaks. Data recently l

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LI IIS presented in the Vermont Yankee Containment Study in'dicate that non-inerted '

operation at that plant amounted to 1.1% of power operations in a period of 14 years. Such a low fraction indicates a very low risk from hydrogen even now as long as inerting system power sources operate as intended The impact of the station blackout sequences for inerting system operation must be considered in any evaluation of hydrogen control for inerted containments. Taking Vermont Yankee's experience as representative, the Mark I strategy here should reaffirm the existing controls for hydrogen.

3.2 Containment Soray All Mark I containments except Oyster Creek and Nine Mile Point 1, are equipped with a dual header drywell spray system. The two spray headers are rings locatec well up in the cylindrical part of the drywell with branches holding spray nozzles pointing down at an angle. The headers are fed through each division of the RHR syster with spray operation as an alternate mode of RHR oceration. Due to the characteristically large size of RHR pumps (3,000 -

10,000 gpm) the drywell spray nas a very high flow rate. Precautions are usually incluced in operating procedures to avoid excessive use of this powerful spray system. Oyster Creek and Nine Mile Point 1 have separate dedicated spray systems. See Table 2 for a summary of key features of the 24 plants with Mark I containments.

Most plants have other systems already connected to the spray header feed lines outside of containment. They include such systems as RHR Service Water, Condensate, and in some cases bolted blind flanges which are removed to

y. ,..- E install lines for periodic containment integrated leak rate tests. Thus, it is easy for a plant to provide one or more backup supplies for the drywell spray, even in the event of a station blackout, because of the availability of fire main systems with independent pumping capability. But the available backups are all smaller systems, on the order of 10% the size of the RHR. If they were used they would probably not be able to develop sufficient header pressure for even spray flow distribution in the drywell. If there is a high assurance of drywell spray during severe accidents, even in station blackouts, a number of benefits accrue. First, the walls and penetrations of the drywell are cooled to reduce the threat of heat induced failure. Second, the drywell floor is kept flooded to provide a quenching pool for molten corium if it melts,through the reactor vessel. Third, the continuing spray cools any corium which begins to travel over the open floor toward the wall of the crywell or its vents. Fourth, the spray is available to begin washout of aerosol particles even before they pass to the suppression pool; this is another filtering and condensing mechanism which will reinforce defense-in-depth if some flow were to bypass the suppression pool.

Thus, the Mark I strategy is to replace all spray nozzles with smaller sizes (about a tenfold reduction) and to provide at least two backup water supply systems (including one for station blackout) which can be turned nn by 1

remote manual operation or by simple procedures for connection and startup. 1 1

i l

1

. -- l h.,\- \

14t

• 3.3 Pressure Relief Currently available structural analyses for Mark I containments show ultimate failure at about twice the design pressure, usually failing at the knuckle between the upper cylindrical and lower sections of the drywell. However, these analyses have not taken into account the mechanical backing which may be provided by the biological shield surrounding the drywell. The ultimate strength of the Mark I may be quite a bit higher.

Other factors may control the selection of a pressure, limit. The vent valves alreacy on containment are tested or qualified to levels about equal to design pressure and may not be reliable at pressure far above that. In addition, such high back pressures would reclose SRVs, possibly exacerbating the accicent.

The size and the durability of the vent path involves questions of accident scenario. Assuming that pressure considerations lead one to select a relief at a level on the order of design pressure, some alternatives become apparent.

First, this vent need not be the large steam escape path desired for an ATWS scenaric: for the ATWS the operator would use one or more main steam lines to the turbine bypass. With ATWS set aside, only a cecay heat level vapor flow vent capacity is needed to blunt the pressure rise. Since the containments are already designet to absorb initial sensible heat and the high, early decay heat within the design pressure, a flow equivalent to 1% of rated power at the w -- - - - - * - - -- - pw ---y a w -., ,-p -w- g -, 9 mi-- , s-- - +-- a

I l

, \u V venting pressure is sufficient. On'e porcent power is equivalent to the decay heat generation rate after ten minute's : Fcr Mark I containment, this v

translates to six to eight inches vent diameter, based on venting at about 60 psig. Figure (2) was obtained from the August 1986 IDCOR study submitted to BWROG, and gives the estimated vent diameter as a function of power level.

Tne figure can be used to determine venting capacity for ATWS. For a 2800 MWe uc en reactor ATWS power levels can range,from 30 to 40% of or 840 to 1120 MWt. A vent diameter of 40 inches or so may be needed to manage ATWS by venting the containment.

One has the choice of designing a special purpose vent for this purpose I

leading cirectly to the stack or to use existing vent valves and ducts. The i

1 staff knows of no plant which already has a high pressure vent path in place.

Given the hignly undesirable effects of the potential vent path rupturing insice the plant, the Mark I containment venting strategy is to provide a burst-resistant path with reliable valves, capable of remote manual opening and reclosing even in station blackout, to vent steam equal to 1% of rated power to the plant stack or a high point vent. The use of stack or other high point release will assure a substantial reduction of radiation doses due to post-accident venting. Figure (3) shows expected whole body dose as a function of distance for ground level and elevated release for average and acverse meteorology. The figure shows a substantial reduction in whole body deses for elevated releases.

i l

1 4 l I ..

\ i I  ! -

s i

^

10 l I I I I I 40

{ . I I

' - CONTAINMENT PRESSURE = 0.5 MPa (58 psla)

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O I I I I I I I I J O 100 200 300 400 500 600 700 800

, 900 VENTING CAPACITY, Mw '

i i

l Figure 2.

Vent Size Requirement as 'a Function of Power (This figure is reproduced from IDCOR Report, " Evaluation of BUR Accident Hitigation d

l Capability Relative to proposed NRC Changes," August 1986.) 7.O 4 .

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• nEu !u i u ii w Iil n!!E EXPECTED WHOLE BODY RADIATION DOSE "'

(REMih-

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.

• FROM RELEASE OF 100% NOBLE GASES (I HOUR DECAYED AND 5 HOURS DURATION OF G __ .

3 RELEASE) FROM 3412 MWt LWR VS. DISTANCE E-~

=

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NOTES. ,

i  : ' i j gog > 1. Graphs assume one hour holdup and decay prior , '

' *to release. Greater delay in release can produce

. i000

. lower doses (e.g., as much as a factor of about

'- 30 at one mile for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of inreactor holdup i =- compared to one hour).

E.

=

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ul code calculations using revised (relative to J -.- --

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2. i s E 3. The likelihood of exceedino the

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i 3.4 Core Debris Control The strategy identified in sections 3.2 and 3.3 should provide cooling of molten corium should it come out of the reactor vessel. If molten corium does reach the drywell wall, the combination of a spray-cooled interior and a heavily backed exterior make drywell to reactor building failure unlikely.

The vents on the other hand are a remaining possible debris travel path and the torus room is pneumatically open to the reactor building. Therefore, the i

Mark I strategy is to ensure that, if the torus fails, the water in the torus

, will be retained in the torus room and the corner rooms, and will quench any corium which might reach there and limit the spread of damage by intensely radioactive material.

3.5 Procedures and Training The Emergency Procedure Guicelines, Rev. 4, now under review, have the scope and content to satisfy the needs identified in Section 2.0. The Mark I strategy then is to require that all licensees adopt all principal elements of EPG Rev. 4, and revise or modify as necessary to reflect : "s changes occasioned by 3.1 to 3.4 above.

4.0 FORMULATION OF REOUIREMENTS Based on the preceding analysis the following requirements snould be met by any BWR with Mark I containment.

4.1 Hydrocen Control 1

Present requirements imposed by 10 CFR 50.44 and the Technical Specifications shall be adhered to, no additional requirements are proposed.

===-

i

1 D ?,

liitt: ,

4.2 Containment Scriv All BWRs with Mark I containment shall provide at least two backup water supply systems for the containment drywell spray, one of which shall be functional during station blackout. Water to the spray system from these backup supplies shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during a severe accident scenario.

l In addition, the spray nozzles shall be adjusted so that an evenly distributed 4

spray pattern will be developed in the drywell whether water is supplied by the primary source or either of the backup sources. A flow rate en the order

. of 1/10 of the present flow rate is considered typical, the licensee shall i

select the flow based on an analysis of plant specific parameters.

4.3 Dressure Relief The licensee shall select a pressure between design pressure and 1 times design pressure at which to open an exhaust path from the wetwell vapor space to the highest vent point (stack or pipe) available. This line should be capable of handling water vapor flow equivalent to 1% decay heat at the vent pressure selected without significant chance of rupture before the desired release point. The line shall be equipped with isolation valves which can be opened and reclosed 3y remote manual operation or by simple procedures which can be implementeo earing severe accident scenarios including station blackout, i

4.4 Core Debris Control The licensee shall ensure that the water in the suppression pool in the event

5 tRE of torus failure is held within the confines of the torus room and the corner rooms and cannot flow out to other parts of the plant.

4.5 Procedures ano Training The licensee shall implement emergency operating procedures and other proceoures t,ased on all significant elements appropriate to its plant of Emergency Procedure Guidelines, Revision 4.

4.6 Ouality and Desion Standards Since these requirements are intended to be an optimi2ed use of existing equipment it is expected that added equipment, of itself, need not meet the quality or design standards of safety elated equipment. Nevertheless, modifications to or near equipment or sy tems which are already safety related shall not compromise the quality of such equipment or systems.

I a.7 Imolementation The equipment changes required herein shall be installed during the first refueling outage which begins nine (9) months after the erfective date of this letter. The procedures and training required shall be implemented on a schedule reviewed and approved by the NRC. Given the implementation of the generic improvements of Mark I containments there is no need for an Individual I Plant Evaluation (IPE) for containment oerformance. This does not remove the i

need for an IPE which covers the system reliability or core melt frequency portion of the severe accident question.

t

k

'b'". #*

5.0 JUSTIFICATION FOR REOUIREMENTS Tnere are two possible cases for justification of Mark I containment improvements, one on the basis that they are needea for safety and the second on the basis that they are justified backfits by cost-benefit analysis.

Examination of both be,es shows that they support the containment improvements.

5.1 Needed For Safaty Tne present General Design Criteria (GDC) set requirements for containment performance. GDC 16 - Containment Design says, "an essentially leak-tight b2rrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as costulated accident conditions recuire."

(Emphasis added). It is clear from the long application of this GDC to many cesigns that " postulated accident conditions" are design basis accident concitions, not severe accident conditions. In a similar way GDC 50 -

Containment Design Basis says, "[the containment can accommodate with sufficient margin] the pressure and temperature resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the i

determination of the peak conditions, such as the energy in the steam generators and as required by 550.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of 4

emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment

]

i responses, and (3) the conservatism of the calculational model and input i

- - -- l

t 22 parameters." The words " degradation but not total failure of emergency core cooling" clearly limit the application of this GDC to design basis accidents.

, Thus, consideration of both these GDC indicates that any mandate for change to protect against core melt accidents lies outside the requirements of the existing regulations.

1-

The Commission spoke to the need in the Severe Accident Policy Statement of August 8, 1985

o Operating nuclear power plants require no further regulatory action

to deal with severe accident issues unless significant new safety f information arises to question whether there is adequate assurance

! of no undue risk to public health and safety.

l c In the latter event, a careful assessment shall be made of the severe accident vulnerability posed by the issue and whether this vulnerability is plant or site specific or of generic importance, o The most cost-effective options for reducing this vulnerability shall be identified and a decision shall be reached consistent with the cost-effectiveness criteria of the Commission's backfit policy as to which option or set of options (if any) are justifiable and required to be implemented.

h Y*b.

j s o

In those instances where the technical issue goes beyond current regulatory requirements, generic rulemaking will be the preferred solution. In other cases, the issue should be disposed of through the conventional practice of issuing Bulletins and Crders or Generic Letters where modifications are justified through backfit policy, or through plant-specific decisionmaking along the lines of the

, Integrated Safety Assessment Program (ISAP) conception.

From these passages it is clear that the Commission intends to deal with i

severe accident issues if there is a Question whether there is adequate assurance of no undue risk to public health and safety. As noted in Section 1.0, the Reactor Safety Study estimate of BWR Mark I containment performance

) in the face of core melt was that it had a 90% chance of failure. Many expectac that more refined analyses of risk available now would show a much 4

i lower level of severe accident risks. In many ways that expectation has been satisfied but with the Mark I containment the later results have not been so encouraging. Again as noted in Section 1.0, the IDCOR Technical Summary Report presents an inferred containment failure rate of about 2C%. The j analysis done for the smaller Vermont Yankee plant yielded a 7% estimate.

1 Considering the continuing debate on uncertainties in these estimates, it is I

fair to say that the early failure rate for Mark I containment lies in the range of 90% to 10%, perhaps in the lower rather than the upper end of that range.

O k

.y.

In discussing a plant with Mark I containment in a Congressional hearing on July 16, 1986, the Commission responded as f',llows to the question:

Question Is a 90 percent chance of failure in the event of a core meltdown an acceptable failure rate?

J Answer J

The NRC holds the position that the likelihood of core melt accidents in cny plant should be very low and, in addition, that there should be substantial assurance that the containment will mitigate the consequences of a core melt should one occur in order to ensure low risk to the public.

It is not merely a question of having low risk but of having as well the cefense-in-depth assurance of combined protection by prevention I

and mitigation. .

If we are debating in the range of 90% to 10% failure probability, even with the likelihood that it is closer to the lower figure, that is hardly

" substantial assurance that the containment will mitigate the consequences of a core melt should one cecur." There is no quantitative synonym for i

suestantial assurance but it is a defensible proposition if the range of l debate can be shifted down to something more like 10% to 1%. The Mark I 1

strategy developed in the preceding sections is not quantified but it does t

provide significant changes for the better in each of the areas of greatest 1 I

) uncertainty and significance for Mark I performance in core melt.

1

m.

't i

} Anc so it can be argued that these containment enhancements are needed for i

safety, to ensure low risk to the public by establishing substantial assurance that the containment will mitigate the consequences of a core melt should one 1 occur.

t i

j 5.2 Costs and Benefits The estimated costs of proposed action would vary substantially depending upon a specific designs of plants and ease with which performance enhancements could be incorporated. IDCOR* has presented approximate ranges of costs. The estimated costs do'not reflect any unique engineering difficulties or time available for modifications to be incorporated in plant maintenance outages.

The cost considerations incluced in the IDCOR study include the following:

o Hardware o Installation

} o Test and Maintenance o Plant Unavailability o ALARA (Exposure Costs) i o

Costs of Procedure Changes and Training and Impact of Proposed Backfits l

The cost of drywell sprays using fire pumps available at all plants was s

i

! I 1

I i

l b l

, l

! l 1

7 , - -, , . . ., ,, , , , , - - , -

.,--,,g. , . , . , , -.,,_an -

, , .. .__ . _ -. . . . _, . ,,, -- ,. ,, ... ,v n .y,,,s.,-,,

l

.lf I l i

estimated to range from 0.6 million dollars to 1.1 million dollars. The cost 4

of venting using vents of 6 to 18 inch ducts ranged from 0.1 to 1.1 mt;ibn 4

dollars. The cost of installing a short debris barrier was estimated to be 0.4 million dollars. No additional cost is expected for Mark I containment 1

hydrogen control. No significant new costs are expected for implementing the emergency procedures guidelines that the industry is pursuing as a result of TMI actions. Based on the above IDCOR estimates it would appear that the cost ,

of the proposed initiative should range from 0.7-2.2 million dollars per reactor.

To estimate the benefits of containment improvements one must estimate the averted loss. The terms necced to estimate it are: '

o FCM, Frequency of Core Melt o CCFP, Conditional Containment Failure Probability t

o Loss, the monetized cost of a large release.

Taking Frequency of Core Melt first, the containment improvements include operator training and procedures for handling the containment. Because of the

] close interaction of systems in a BWR improved procedures will undoubtedly

?

have an effect of reducing FCM. For reference, in the Reactor Safety Study, about two-thirds of all core melts were caused by the failure of containment t due to overheating which failure then caused the loss of core cooling. For l

i

" Evaluation of BWR Accident Hitigation Capability Relative to Proposed NRC

Changes, August 1986 1

1 J

__-____ a

O q r s'

-y.

kfkfl simplicity in the calculation here the reduction of FCM will not be included, i thus apparently underestimating the value of the containment features. It is l reasonable to do this because the reduction in FCM will come principally from

, the training and procedures whose costs are not included in the preceding 1

section since the BWR owners are already committed to adopting most of them.

4 The prcoer choice of a typical CCM is difficult to make. Current 10COR and

-5 NE analyses of Peach Bottom suggest FCM on the order of lx10 /yr although

~

the NRC results are dominated by station blackout sequences while the IDCOR J

1 results are dominated by ATV5. Examination of the results presented in Table l 1 indicates a number of plant specific cases where FCM ranges up to and above

~#

1xM /yr. Considering the diversity of systems and the flexibility of operation in a dWR, a FCM of 1x10 -5 /yr may well be attainable. However, the modelling differences between IDCOR and NRC results at that level and the results from other plants at the higher level suggest the choice of 1x10 /yr as the typical value for cost-benefit analysis purposes.

The cost-benefit equation calls for a quantitative estimate of the CCFP before and after the containment changes. Considering the range of debate en the l present state of containments, given previously as 10% to 10%, it is t

reasonable to use 50% as the "before" 1'igure. If substantial ir.provement is achieved the exact value of the "after" figure is not important, but 5% will be used here.

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The monetized consequences of a large release coming from early containment failure can be large. Consideration should be given to counting health effects above or counting offsite economic consequences as well. Previous work

• indicates that large early releases can cause on the order of 10 persen-rem offsite exposure. Monetized at $1000 per person-rem, this gives consequences of $10 10 per event. Studies also show** that offsite economic consequences of large early releases can be up in the tens of billions of dollars or even more. On the other hand, source term studies continue and some argue that these high consequences are derived from WASH-1400 vintage source terms. For this cost-benefit calculation the health effects only consequence of 10 person-rem or 510 10 will be used. Another question arises in converting annual averted health effects into a present worth value. Since these are human health effects some argue that they should not be discounted in a cresent worth calculation. If that approach is taken then the averted l

loss per year is multiplied by the remaining years of plant operation. The other approach is to use a discount factor such as would be associated with averted economic losses. Table 2 has been prepared to illustrate the central

c. ;Nc I r , lor l 'r t. Ca d bue k. A GCE HS 1 NUREG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accicents," Appencix A, September 1982

• • NUREG/CR-3673, " Economic Risks of Nuclear Power Reactor Accidents" p.2-13, April, 1984

_ . _ . , cm._ e - - - , - - , -

L 29 Referring to Table 2, the cost is listed as $0.7-2.2 million, or in a rounder number, less than 53 million per reactor. The base calculation gives a benefit of 53 million of $12 million, indicating a balance of cost and benefit or a clear justification. The other calculations in Table 2 are a sensitivity analysis to explore the range of outcomes with different assumptions. The lower FCM case uses the frequency of core melt currently being calculated by IDCOR and NRC for Peach Bottom. The next case illustrates less improvement in containment performance, only a factor of five. The next case assumes that containment performance now is better, the 20% CCFP inferred in the IDCCR report. The " optimistic" case uses the IDCOR values for FCM and present containment performance while the " pessimistic" case assumes a relatively ;.igh FCM and CCFP.

Comparison of these estimated benefits to the range of costs indicates that these pro:osed changes are justified backfits.

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TABLE 3 COST-BENEFIT ANALYSIS COST: 50.7-2.2M BENEFIT:(1) FCM CCFP CCFP AVERTED AVERTED BEFORE AFTER LOSS /YR LOSS PRES. VALUE BASE

~4 CALCULATION 1x10 /yr 0.5 0.05 $4x105 /yr $3M/$12M

-5 4 LOWER FCM 1x10 /yr 0.5 0.05 $4x10 /yr $0.3M/$1.2M i

LESS CHANGE

~4 IN CONTAINMENT 1x10 /yr 0.5 0.1 $4x105 /yr $3M/$12M BETTER CONTAINMENT TO START 1x10 ~4' O.2 0.05 $2x105 /yr $2M/$6M "0PTIMISTIC

-5 4 CALCULATION 1x10 0.2 0.05 $2x10 /yr $0.2M/$0.6M

" PESSIMISTIC"

~4 CALCULATION 3x10 0.9 0.1 $2x105 /yr $16M/$60M T1)FCM = Frequency of Core Melt CCFP = Conditional Containment Failure Probability AVERTED LOSS PRESENT VALUE expressed as A/B where A is the averted loss per year times 8 (roughly equivalent to discount at 12%/yr rate) and B is the averted loss per year times 30 (no discount).

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l BOILING WATER REACTOR

i. PLANT INFORMATION SYSTEM i

DIVISION OF BWR LICENSING l

OFFICE OF NUCLEAR REACTOR REGU LATION i WINTER 1986 i

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~ CP DATE LICENSING ACTIVITY l DIRECTOR PROJECT DIVISION INTONMATION DISTANCE TO NGC (MILES) SITE INFORMATION

~ DOCLET NUMEER LICENSING ACTIVITY DRYWELL VOLUME (CU FT) CONTAINMENT DATA ECCS CONFIGURATION ENGINEERED SAFETY FEATURES EXCLUSION RADIUS (METERJ SITE INFORMATION FEED FUMP TYFE NSSS DATA FEEDWATER TEMF (F) NSSS DATA FUEL EUNDLES NSSS DATA FUEL CHANNEL THICLNESS (MILS) NSSS DATA HEAT GEN RATE. AVG (LW/FT) REACTOR DATA HPCI/HFCS FLOW (GFM) ECCS DATA JET FUMPS, NUMBER REACTOR DATA LEVEL 1 TRIF SETFOINT (IN) REACTOR DATA LEVEL S TRIP SETFOINT (IN) REACTOR DATA i LICENSE NUMBER LICENSING-ACTIVITY LICENSED FOWER (MWT) NSSS DATA LICENSEE / APPLICANT LICENSING ACTIVITV LICENSEE /ADFLICANT .

LICENSING ACTIVITY LOW POPULATION ZONE (METERS) SITE INFORMATION LFCI LCOP SELECTION LOGIC ENGINEERED SAFETY FEATURCS LFCI PUMPS (NUMBER) ENGINEERED SAFETY FEATURES i LFCI RATED FLOW (GFM) ECCS DATA

! LPCI RATED FRESSURE (PSID)

ECCS DATA LFCS RATED FLCW (GCM) ECCS DATA I

LPCS RATED PRESSURE (FSID) ECCS DATA NEAREST FOPULATION CENTER (NFC) SITE INFORMATICN 1  !

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$norense Pr!! 192.000 114,000 11,350 324.000 40.0

! $6scuenanna1 PK!! 239,600 !!!,900127,000 393,5C0 53.0 $3.0 i lusswomanra2 Pr !! 239,600 !!3,900127,000 393,!C0  !!.0 !3.0

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