ML20246P087
| ML20246P087 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1989 |
| From: | Elliott C, Hodgdon A, Wetterhahn M LIMERICK ECOLOGY ACTION, INC., NRC OFFICE OF THE GENERAL COUNSEL (OGC), PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| References | |
| CON-#389-8855 OL-2, NUDOCS 8907200108 | |
| Download: ML20246P087 (35) | |
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' COLKETED UwPC UNITED STATES OF AMERICA '89 JUL -5 A11 :17 NUCLEAR REGULATORY COMMISSION (n'
Before the At mic Safety and Licensir[hB'o'ard' In the Matter of ) Docket Nos. 50-352-OL-2
) 50-353-OL-2 Philadelphia Electric Company )
) (Severe Accident (Limerick Generating Station, ) Mitigation Design Units 1 and 2) ) Alternatives)
REPORT OF THE PARTIES In its Prehearing Conference Order of June 9, 1989, the Chairman of the Atomic Safety and Licensing Board directed the parties to the proceeding, Limerick Ecology Action, Licensee Philadelphia Electric Company and the Nuclear Regulatory Commission Staff (LEA, Licensee and the NRC Staff) to stipulate, insofar as possible, to those SAMDA's at Limerick which would be proper to consider under the May 5, 1989 Order of the Commission in this proceeding and to submit supporting memoranda on differing positions regarding other alternatives. On June 12, 1989, legal and technical representatives of the three parties and the Commonwealth of Pennsylvania met to exchange technical information, discuss possible stipulations as to the scope of the SAMDA's to be considered and tour the Limerick Generating Station. Subsequently, the parties exchanged additional information and held a number 8907200108 890603 PDR ADOCK 050003S2 g PDR kb
(* < g of telephone conversations. The following represents the report of the parties:
- 1. In accordance with the request of the Licensing Board, . Attachments 1 and 2 hereto. contain copies of the-lists drafted by the Licensee and . LEA, respectively, . for fU exchange at the Prehearing Conference. Attachment 3 hereto contains a copy.of'the May 23, 1989 NRC Staff letter to the Licensee forwarding three questions. relating.to SAMDA's. On' June.23, 1989, the Licensee submitted its response to the Staff's May 23, 1989 question, which was served.upon the Board and parties. On June'27, 1989, LEA transmitted:to the.
Licensee'and Staff a: document entitled " Supplemental' List of-Litigable Severe Accident Mitigation Alternatives for Litigation of Limerick Ecology Action, Inc. Contention on Severe Accident Mitigation Alternatives for the - Limerick
-Nuclear Generating Station." A . copy of this list is provided as Attachment 4.
- 2. After discussions -among the parties, in substi-tution for Attachment 1, the Licensee has further described and.specified the SAMDA's it is examining both in response to the remand and the Staff's May 23, 1989 letter as fol-lows:
- a. Pool Heat Removal System -
A separate in-dependent dedicated system for transferring heat from the suppression pool to the spray pond utilizing a l diesel driven 3,200 gpm pump and heat exchanger without-dependence on the Station's present electrical power or
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other systems. The diesel is cooled with water tapped ~- L L.. off'theispray pond suction line.
- b. Drywell Spray -
A new dedicated system for heat and fission' product removal .using Pool- Heat
' Removal. System described in . (a)' above. to inject water?
into the drywell. c.- ' Core Debris Control (Core Catchers) ' .Two techniques, either a basemat rubble bed or using a dry -l crucible approach, to contain the debris in a known stable condition in the containment.
- d. Anticipated' Transient Without Scram -(ATWS)
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Vent - A'large wetwell vent line to an elevated release point to remove heat added to the pool in an ATWS event and prevent overpressurization.
- e. Filtered Vent - Drywell and wetwell vents to a large filter (two types -
gravel or enhanced water - pool) to remove heat and fission products and prevent overpressurization.
~f. Large Containment Vacuum Breaker - To restore containment pressure to atmospheric level through 20" valves in certain severe accioent cases where a vacuum has been produced.
- 3. The Parties agree that it is not necessary to further examine a large hydrogen recombiner which was
-originally suggested by the Licensee in Attachment 1 as an alternative to be examined inasmuch as the Limerick units
14 - 4 __4__ (tg . are already equipped with recombiners and larger-recombiners-would not-be expected to be' cost beneficial. 4 ~. The Licensee, LEA- and the Staff assert that the alternatives set forth in ' Paragraph 2, above, incorporate reasonable design alternatives for Limerick within the range of SAMDA's permitted by the Commission's May 5, 1989 Order. The parties will address the Board's consideration of other alternatives proposed by LEA in their separate pleadings. Respectfully submitted, Mark J. MtCe'rhahn Charles W. Elliott M Counsel for Philadelphia Counsel for Limerick Ecology. Electric' Company' ' Action Ann Hodg&on
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f) Counsel for the United States Nuclear Regulatory Commission Staff I June.30, 1989 I i _ - _ - - - - - - - - - - _ - - _ _ _ - _ - _ - - _ = - , -
q U :A n 14rl UNITED STATES OF AMERICA '89 JUL -5 m):33 l NUCLEAR REGULATORY COMMISSION I ,:% ,.,. 00cyz ,n.r :i Xf In the Matter of ) EP AC-
)
Philadelphia Electric Company ) Docket Nos. 50-352
) 50-353 (Limerick Generating Station, )
Units 1 and 2) ) CERTIFICATE OF SERVICE I hereby certify that copies of " Report of the Parties" dated June 30, 1989 in the captioned matter have been served upon the following by deposit in the United States mail this 30th day of June, 1989: Morton B. Margulies, Esq. Frederick J. Shon Chairman, Atomic Safety and Atomic Safety and Licensing Licensing Board Panel Board Panel U.S. Nuclear Regulatory Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. Jerry L. Kline Atomic Safety and Licensing Atomic Safety and Licensing Appeal Panel Board Panel U. S. Nuclear Regulatory U. S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. Jerry Harbour Joseph Rutberg, Esq. l Atomic Safety and Licensing Ann Hodgdon, Esq. l Board Panel Counsel for NRC Staff ! U.S. Nuclear Regulatory Office of General Counsel Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Atomic Safety and Licensing Edward J. Cullen, Esq. Board Panel Philadelphia Electric U. S. Nuclear Regulatory Company l Commission 2301 Market Street Washington, D.C. 20555 Philadelphia, PA 19101
l l fe - Charles W. Flliott, Esq. . ' Gregory E. Dunlap, Esq. l
'Poswistilo,:Elliott & ' Deputy.. General Counsel Elliott Commonwealth of Pennsylvania Suite'201 17th Floor Harristown II 1101 Northampton St. 333 Market Street-Easton,.PA .18042- Harrisburg, PA 17101
-Angus Love, Esq. Mr. Ralph Hippert 107 East Main Street- Pennsylvania. Emergency:
Norristown, PA 19401 Management: Agency. B151 -Transportation Safety Building Harrisburg, PA 17120 Michael B. Hirsch, Esq.- Theodore'G. . Otto, Esq.- Federal Emergency Department of' Corrections Management Agency Office of Chief Counsel 500 C' Street, S.W. P. O. Box 598 Rm. 840 Camp Hill, PA .17011 Washington, D;C. 17011
-Docketing and Service Section U. S. Nuclear-Regulatory. !
Commission Washington, D.C. 20555 Mark J. k'6tterhahn k 4
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l 1 Att. 2 ment 1 ! l, l~ ; I i f l SEVERE ACCIDENT MITIGATING DESIGN ALTERNA BEING EVALUATED BY l i PHILADELPHIA ELECTRIC COMPANY l
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o Pool Heat Removal System A separate independent dedicated system fortran I pond. Heat Removal) o Dry Well Spray (icated system for heat and fiscion product ) A new dedremoval using either external or internal water sources,
,(" Core Catchers")
o Core Debris Control i Two techniques, either a basemat rubble bed, in a known stable condition in the containment. i , o ATWS Vent A large wetwell vent line (3-5' dia) to an elev ATWS event. o Filtered Vent and wetwell vents to a large filter (two Drywell types - gravel or enhanced water pool) to remove heat and fission products. o Large H 2 Recombiner 2 from the Increased capacity recombiners to remove Containm o Large Containment Vacuum Breaker To restore containment pressure to atmosph been produced. . June 5, 1989 l l L____-__-__
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,.3/. Attachment 2
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LIMERICK ECOLOGY ACTION, INC. LIST OF PRIMARY CANDIDATES FOR SEVERE ACCIDENT MITIGATION Venting / Filter Devices Much attention has been paid in the past to add-on filtered vented containment systems-(FVCS). Such add-on systems have been proposed for a number of European plants, and have been constructed at Barseback (a Swedish ASEA-Atom BWR which is grossly similar to the Mark II design). The cost of such devices is significant, with estimates ranging from S30-5100 million (depending upon design and upon the seismic " pedigree" of the structure). L 1 An FVCS can function to both mitigate consequences (by reducing aerosol' releases to the environment) and reduce accident frequency (by reducing the frequency of loss of decay heat removal accidents). Such devices, however, cannot function I-effectively for containment pressurization phenomena which operate over short time domains, such as high pressure melt ejection / direct containment heating. Thus, the effectiveness of an FVCS is dependent at least in part on most of the core melt accidents being low pressure. sequences (i.e., the reactor vessel pressure must be low at-the time of vessel failure). l l
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Fortunately, there is an option here that requires only a modest investment and little down time (in fact, probably no down time unique to this modification; it can probably be done in the shadow of other outage work during a refueling / maintenance outage). This option involves (as part of the "Bernero fixes") installing a hard pipe vent from the drywell to the plant stack. This is the option that was utilized at Pilgrim (the so-called
" Direct Torus Venting System" or DTVS). Some plant-specific analysis will be necessary to implement this system at Limerick, but it appears promising as an option. The problem with current venting procedures (which have been in place at all operating BWRs since the NRC approved EOP Rcv. 2 in the early 1980s) is that the venting ductwork at most plants will fail at a low internal pressure (a few psi). This results in the discharge of steam to the reactor building, which can result in failure of equipment in the reactor building due to the adverse environment (humidity, steam, and, in some cases depending on the venting strategy, radiation) or due to the inability of maintenance personnel to continue to work on equipment repairs. The hard vent avoids this problem by using a pipe which can withstand the pressure without failure and conduct the steam which is vented from the wetwell airspace to the outside environment (preferable up the plant stack to take advantage of an elevated release point.
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j Containment Spray / Flooding Modifications It has been suggested that a "sure" way to reduce . consequences is to flood the drywell when core. damage is' imminent. This is by no means "sure" (e.g., what happens is, 4 high pressure. melt ejection occurs into.a pool of water - a' fairly large steam explosion seems plausible), and is probably i not necessary. I What is needed here is additional assurance of getting water onto the core debris to promote solidification and to hold down
.the aerosol source term-(in addition, if you can spray with it, you can inject water into the reactor vessel as well as long as the pressure is low). Boston Edison Company, as part of its Safety Enhancement Program (SEP) modified the drywell spray nozzles by capping six of seven spray locations on each spray head. This reduces the spray flow to a point where any reasonable source of water can be used to develop containment spray (RHR, service water, fire protection system water, etc.).
Piping interconnections may be needed between RHR and the fire protection water system and/or the service water system. This is l a relatively inexpensive task, and is consistent with the Bernero 1 fixes. Moreover, it is useful for containment spray purposes even if vessel injection is impossible due to high pressure i
Qk & (pressure reduction and source term reduction). One' final factor-may be worth consideration. Henr'y, et'al., (Fauske S Associates) have performed small-scale experiments which. provide some indication-that direct spray of water-onto simulated' core debris greatly enhances cooling of the debris
-(this may be.due to disruption of.the debris surface resulting in-enhanced cooling due . tx) the larger surface area). If these experiments are' applicable.to reactor situations, it may be
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advisable to consider running a small line from the drywell. spray rings tu) the vicinity of openings in the pedestal and orienting spray nozzles at the likely exit pathe of core debris from the pedestal. This1 design modification could support core debris cooling at the same time that the overall sprays provide
-containment cooling and fission product source' term reduction.
Containment Heat Removal Augmentation Modifications Some studies have suggested a variety of plant modifications to enhance heat removal, such as heat pipes, fan coolers, spray j coolers, extra heat exchangers, etc. Most of these schemes would either be difficult to implement (due to space limitations in the drywell and extensive modifications and additional containment penetrations) or require more R & D before they could be i _4_
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; engineered.and implemented.
One potential candidate for Limerick, however, is an i
.I augmented suppression pool cooling function. As envisioned in an I NRC-sponsored report, this system would put a heat exchanger in
'the suppression pool through which' a diesel-driven pump would
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pump water through the containment to remove heat'on a once-through basis (water from the ultimate heat sink). Spent Fuel Pool, Accident Risk Modifications Recent studies of spent fuel pool risks at a BWR (Vermont Yankee) and a PWR (H.B. Robinson) indicate that there is a potential for a self-sustaining circaloy fire to result in a large radiological release from the spent fuel pool, particularly where re-racking has taken place. This risk can be minimized at the.very least by licensing.a dry cask storage form of Independent Spent Fuel Storage Installation (ISFSI) at the site, i i and removing spent fuel from the spent fuel pool after one year's cooling. This minimizes the amount of fuel in the pool, and, if done early in plant life, eliminates the need to re-rack. Re-racking is costly, and if it can be avoided, this should be attractive to the utility. I I 5-l 1
1 i 1 l l Licensing of ISFSIs is covered by 10 CFR Part 72. We estimate that it would take from 4-6 years to prepare an , l application, file it, get it reviewed, get a construction permit / operating license, and complete construction of the first i dry storage units and bring them into operation. This should ) easily be within the reach of PECo for Limerick for Unit 2 (and perhaps for Unit 1 as well). Human Factors Modifications (Including Procedures) l Recently, General Electric Company completed an upgrade of the generic Emergency Operating Procedures (EOPs) Revision 4 for i the BWR Owners Group (BWROG). EOP Rev. 4 explicitly extends emergency procedures into the severe accident domain for the j i first time. EOP Rev. 4-based procedures have already been I implemented at Pilgrim and Shoreham. Implementation of EOP Rev. 4 has been approved by the NRC staff in a recent SER, and represents a BWROG commitment to NRC. It is also part of the Bernero fixes. We recommend its inclusion in a mitigation / prevention package. I Another human factors area of merit is in responding to seismic events. Considering the contribution of seismic events to core damage frequency (6% of core damage frequency by NRC
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e-staff estimate, 24% by PECo estimate), it is eminently reasonable that there should be plant-specific procedural guidance for operators in responding to seismic events. The procedures should explain how such events can impact the plant, point out vulnerabilities, and lay out alternative means for accomplishing safety functions. Such procedures should also include lists of relays and breakers in systems which are relied upon a primary or alternative means to perform safety functions. Relay and breaker chatter as a result of seismic events were not considered in the ! Limerick SARA seismic PRA, but have been identified elsewhere (including the NUS-performed Kuosheng PRA) as potentially important risk contributors. Further, a recommendation that the y lesign be reviewed and that chatter-vulnerable relays and recommendation that the plant design be reviewed and that chatter-vulnerable relays and breakers be replaced with more i resistant designs should be considered to reduce the vulnerability to this problem). A final human factors item that might be considered is to require PECo to expedite the control room design review (if it hasn't yet been done) and to expedite implementation of fixes for _ human engineering deficiencies (HEDs) if the review has been performed. This should help operator response to severe l -7 - l
l 1 4 l accidents in general. Seismic Modifications The Limerick PRA and the Brookhaven review of the PRA identified as an important part of the seismic risk the failure of the reactor enclosure and control structures wall at 0.90 g (the PRA estimated 1.05g). It is plausible that modifications to this wall could be identified that would sufficiently improve the fragility of this wall to reduce its importance to risk. Engineering analysis would need to be den,e to evaluate whether this is possible. I Recent studies nave identified the potential for chatter of relays and breakers to cause system failures which can contribute to seismic core damage frequency. Such failure modes were assumed in the Limerick seismic PRA (SARA) to ha fully recoverable. This is a potentially optimistic assumption. There needs to be a plant-specific assessment to evaluate whether there are any chatter-prone relays and breakers in risk significant i systems at Limerick, and, if so, to replace those components with chatter-resistant designs. In addition, it may be desirable to produce operator procedures for use following an earthquake to guide them through verification and resetting of breakers and
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- S, relays.
Reduction of Transient Initiator Frequency
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It is well known that accident sequences initiated by l transient events dominate overall core damage frequency for BWRs,: and for Limerick and Shoreham in particular. .This being so,.one straight-forward way to reduce the frequency of' severe' accidents is to reduce the frequency of transient initiating events. The NRC has been encouraging the industry to do just this, and intiastry has responded with an active. program in this regard. 4 NRC's Office for the Analysis and Evaluation of Operational Data
.(AEOD) puts out periodic reports (AEOD annual reports and Performance Indicator reports) in~which scram frequency data is provided. In. addition, a recent AEOD study of new plant performance includes such data.
This data is most often reported [ in terms of unplanned scrams per 1,000' critical hours. It can, l however, be converted to scrams per year relatively easily. In L1merick's first 4862 critica}. hours (before commercial operation), the plant experienced 4 scrams, for a scram rate of 0.82 scrams per 1,000 critical hours. Assuming this rate obtains over the course gf an average year, this works out to 5.5 scrams per year. In 1987, Limerick experienced 2 scrams in 6151 9_ l _ - . - _ - - - - - - - - - . - - - {
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1 critical. hours (a rate of 0.33/1,000~criticalThoursT. In contrast, the Limerick PRA estimated 9.08 scrams per year,.while-
.the Brockhaven review of the Limerick PRA' estimated 13.02 scrams per year. If Limerick can-maintain this performance, actual risk t
reduction'shall have occurred. compared with the level of risk predicted in the PRA. It would seem sensible, therefore, that if such a program is not already in effect that PECo be required to implement a formal scram reduction program. .This should be relatively inexpensive ! addition to.the scram analysis required by post-Salem ATWs l i
'l procedures, and should' enable lessons learned to be applied I toward avoiding future scrams in a structured manner.
In addition, given the importance of some systems to risk,
. it would seem to make sense for PECo to evaluate the potential benefits from implementation of a reliability-centered maintenance program. In such a program, reliability data are used to predict when component failures are likely to occur and to take action (i.e., preventive maintenance) to avert the failure, j
I Additional reduction of scram frequency might also be obtained by reviewing the technical specifications to see if some L s L u _ -- - -~ - - - - -
c. l l 4 l can be relaxed to avoid unnecessary shutdowns (manually-initiated shutdowns also carry with them some risk). A number of nuclear 1 power plants have used their PRA studies in such applications (among them LaSalle, Seabrook, and Byron). 1 Reactor Vessel Depressurization System Modifications The NUREG/CR-4920, Vol. 2 report identifies the modification of the ADS at Limerick as a way to reduce core damage frequency by a factor of two. If such a modification has l not been accomplished, it would make sense to include it as a i risk reduction option. Moreover, additional improvements in ADS reliability might be identified through a safety system functional inspection type of analysis. (For example, BECo implemented modification at Pilgrim to provide additional bottled nitrogen gas to nssure long-term availability of ADS during station blackout !;equences. It is also possible to modify ADS designs to permit actuation of SRVs while the containment pressure is high; some SRV designs cannot accomplish this, and the valves go closed when containment pressure rises above their design capabilities. Enhanced ADS reliability is particularly important to evoid high pressure melt ejection / direct containment heating phenotona (which could lead to early containment failure) and to take advantage of alternative sources of vessel injection.
o , y
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J i k L -Current "Best Estimate" Risk Reduction Fackage for Limerick L
- a. Implementation of a.hard-piped wetwell vent.
~b. Evaluate and implement alternative vesselfinjection/
drywellz spray-water sources (optimize what is available, and' evaluate whether other' sources could be added, similar to BECo's addition of a diesel-driven fire protection system water pump).
- c. As necessary, modify the drywell sprays to permit the use of alternative water sources, to spray.into the pedestal', and'to spray the area outside the pedestal where there are openings from the pedestal to-the drywe11.
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.d. Implement Rev. 4 of the Emergency Operating Procedures (EOPs) on a plant-specific basis, and review other pertinent procedures to upgrade their capabilities to aid in operator response to severe accidents. The goal here is to move toward accident management.
- e. . Implement reliability improvements to the ADS.
f. Implement a scram frequency reduction program. ' g. Evaluate the need to replace.any chatter-prone relays
.and breakers inJrisk significant systems with.
chatter-resistant designs. h. Evaluate the potentia: need for other fixes-(e.g., ' extra diesel generator, removal of AC/DC dependencies on venting and alternative vessel injection, diesel-driven decay heat removal pump and heat exchanger for containment heat removal, etc.). i. Evaluate potential seismic risk reduction possibilities (e.g., strengthening the reactor enclosure and control structures wall).
*j. spent fuel proof accident risk moeirication.
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e ,. - [ :. UNif ED STATES Attachnent 3
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c i NUCLEAR REGULATORY COMMISSION was m otoN o e 2o:s3 l g .x
' [, , , J "ay 23, 193g Docket Nos. 50-352/353 Mr. George A. Hunger, Jr.
Director-Licersing Philadelphia Electric Company Correspondence Control Desk P 0.. Box 7520 Philadelphia, Pennsylvania 19101
Dear fir. Hunger:
SUBJECT:
CONSIDERATION OF SE'/EPE ACCIDENT MITIGATION CESIGN ALTERNATIVES (TAC NOS. 73082) RE: LIMERICK GENEPATING STATI0f, UNITS 1 AND 2 In an opinion issued February 28, 1989, the United States Court of Appeals for the Third Circuit granted, in part, a petition for review filed by intervenor Limerick Ecology J.ction. In grenting that petition, the Court ordered the NRC to give additieral consideration to an intervenor contention asserting that, ir crder to comply with its obligations under the National Environmental Policy , Act, the agency must censider certain cesign alternatives for the mitigation of ( severe accidents at the Limerick Generating Station. In response, the Chairman of the Atomic Safety and Licensing Board Penel converad a Licensing Board to conduct additieral proceedings rela +. irs to this contention. To al'ew preparation of an flRC staff position on this issue, we request that Philadelphia Electric Company provide the additional information described in the enclosure. Please provice this ir.fermation within 30 days of receipt of this letter. This request for information is specific to one applicant and thus Office of Management and Budget clearance is not reouired under P. L. g6-511. ; t sincerely, V ,W R b e. C1 rk, Project Manager ! P 0,iect Dire torate I-2 Division of Reactor Projects :/II Office of Nuclear Reactor Peculation i
Enclosure:
Request for Additional Inferration cc w/ enclosure: See next page
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T Mr. George A. Hunger, Jr. Limerick Generating Station i Philadelphia Electric Company Units 1 & 2 I cc: Troy B. Conner, Jr., Esquire Mr. Ted Ullrich Conner and Wetterbahn Manager - Unit 2 Startup 1747 Pennsylvania Ave. , N.W. Limerick Generating Station Washington, D. C. 20006 P. O. Box A Sanatoga ,- Pennsy lvania 19464 1 Mr. Rod Krich 57-1 Philadelphia Electric Company Mr. John Doering 2301 Market Street Superintendent-Operations Phf1adelphia, Pennsylvania 19101 Limerick Generating Station P. O. Box A Mr. David Honan N2-1 Sanatoga, Pennsylvania 19464 PhiladelpM a Electric Company 2301 Market Street Thomas Gerusky, Director Philadelphia, Pennsylvania 19101 Bureau of Radiation Protection PA Dept. of Environmental Resources Mr. Graham M. Leitch, Vice President P. O. Box 2063 Limerick Generating Station Harrisburg, Pennsylvania 17120 Post Office Box A Sanatoga, Pennsylvania 19464 Single Point of Contact P. O. Box 11880 Harrisburg, Pennsylvania 17108-1880 Mr. James Linville U.S. Nuclear Regulatory Commission Mr. Philip J. Duca Region I Superintendent-Technical 475 Allendale Road Limerick Generating Station King of Prussia, PA 19406 P. O. 80x A Sanatoga, Pennsylvania 19464 Mr. Thomas Kenny Senior Resident Inspector j US Nuclear Regulatory Cossnission P. O. Box 596 Pottstown, Pennsylvania 19464 Mr. Joseph W. Gallagher , Vice President, Nuclear Services ' Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Mr. John S. Kemper Senior Vice President-Nuclear Philadelphia Electric Company 2301 Market Street l Philadelphia, Pennsylvania 19101 )
4-ENCLOSURE PEOUEST FOR ADDITIONAL INFOP.MATION LIMERICK GENERATING STATION, UNITS 1 AND 2 1. On the basis of PRA results to date, identify those accident sequences that are expected to dominate the overall mean frequency projected for severe core damage and for the significant off-site risks (i.e., projected risks of early fatalities and person-rem). It is suggested that those sequences that collectively contribute 901 to the overall mean frequency for severe core damage be identified as dominant and each described. For these dominant sequences, present the projected mean value for each, considering that three categories (i.e., internal initiations, fire initiations and earthquake initiations) will likely contribute to the overall results. 2. For the internal and fire initiated sequences, assess the potential severe accident design mitigation alternative (s), that (if put in place or installed) have a reasonable chance of reducing the projected severe core damage frequency and off-site risks and (1) which may result in a substantial increase in the overall projection of the public health and safety, and (2) which are justified by the attendant direct and indirect costs associated with putting the alternative into place. As noted, this assessment should be limited only to those internal and fire initiated secuences (exclude those sequences initiated by earthquakes over any portion of the earthquake hazard spectrum). Regarding this exclusion, it is the staff's opinion that the incremental severe accident risks due to the nuclear plant relative to all other risks that could potentially be presented by severe earthquakes (up to those large enough to cause the severe core damage accident) would be negligibly small, (i.e., so small
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.j'y '.. - that the projected risk reduction benefits attendant to seismic related plant improvaats would represent a very remote and speculative projection given the uncertain, competing risks presented to the public l off-site from the severe earthquake itself). f i f In view of the positive
- 3. Provide the results from (1) and (2) above. '
choice by PECO to maintain its PRA in a "living" status since the PRA became available, you may elect to use the PRA insights to enumerate and briefly discuss.those various alternatives considered in the interim f and/or improvements actually made to the plant design and operational procedures, that would in your judgement, serve the objectives of (2) above and have served to increase the level of public protection through either prevention and mitigation of severe accidents.
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Attachment.4 Ii SUPPLEMENTAL /
-LIST OF. LITIGABLE. SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR LITIGATION-OF LIMERICK-ECOLOGY ACTION,.INC. CONTENTION ON SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR THE LIMERICK NUCLEAR GENERATING STATION 1
b I MITIGATION ALTERNATIVE -REFERENCE _. p Liternatives Identified to ASLB/ALAB_ Mode of; operation SARA /EROL Section 7 Contentions filing of August 31, 1983 (Joint Report of LEA, Staff and Appli-cant) (J.A. pp.84,104) Procedures SARA /EROL Section 7 Contentions. filing of August 31, 1983 (Joint Report of LEA, Staff, and Appli-cant) (J.A. pp.84,104) Alternatives described in Id., at J.A. p.106 Beyea,Jan and Von Hippel,
" Nuclear Reactor Accidents:
.The value of Improved Con-tainment", Center for Energy and Environmental Studies, Princeton University (PU/ CEES Report #94), Jan. 1980 Alternatives described in Id., at-J.A. p.106 NUREG/CR-0850 Nov. 1981,
" Preliminary Assessment of Core: Melt Accidents at the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating their Effects" Filter venting of contain- LEA's Reply to Applicant and ment Staff Response to Severe Accident Risk Assessment Contentions (J.A. p. 113)
More reliable containment Id., at J.A. p. 113 heat removal subsystems Alternatives under examina- Id., at J.A. p. 118 tion in Commission severe accident research program Partial draft 6/27/89 1 CO*d 1 1 0 I ~1 *1 3 *1SIMSOd TE: ET OH1 6 8 - ei. C - t 4 n f
p . Filter vented containments Id., at J.A. p. 120 Filter venting of the- Id., at J.A. p. 122 containment-Inside NRC vol.5 no.18 (Sept. 5, 1983) Alternatives identified in Id., at'J.A. p. 122 NUREG/CR-1029, " Program Plan for the Investigation of Vent-Filtered Containment Conceptual Designs for Light Water Reactors" (Sandia, Oct. 1979) Various options for core Id., et J.A. p. 123 retention identified in NUREG/CR-2155 "A Review of the Applicability of Core Retention Concepts to Light Water Reactor Containments" (Sandia, Sept. 1981) variations of filtered- Id., at J.A. p. 123 vented containment systems
-(Proposed Policy Statement on Severe Accidents and Related Views on Nuclear '
Reactor Regulation, 48 Fed. Reg. at 16019 (April 13, 1983) Alternatives identified in Id., at J.A. p. 127 NUREG/CR-2666, Chapter 7 "Further Considerations of Mitigative Features for Specific Plants: Limerick" in PWR Severe Accident Delineation and Assessment Alternatives identified in Id., at J.A. pp. 127-8 R&D Associates reports for Contract NRC-03-83-092 strategy to address failure Id., at J.A. pp.143 mode of overpressure failure with either wetwell or drywell break (NUREG/CR-2666, p.7-6) l filtered vented containment Id., at J.A. p. 146 system (NUREG/CR-2666, p.7-9) Partial draft 6/27/89 2 zo a 11oz,,a A s I M s o e- es: st nH1 6 8 - 6 C - N F8 f
's i l
-c- upgrading performance of -Id., at J.A. p. 146 containment sprays _to cope :
with severe environmental ! conditions in accident
-filter venting strategies Id., at J.A. pp.146; 148 suggested by Work of A.S. 151 Benjamin and F.T. Harper in
" Risk Assessment of Filtered Vented containment options
.for.a BWR Mark I containment" Proceedings of the Interna-tional ANS/ ENS Topical Meet-ing on probabilistic Risk Assessment, Sept. 19C1 protection of diaphragm for Id., at J.A. p. 149 sequences that lead to containment failure caused by diaphragm failure by modifying the region under the reactor vessel (NUREG/CR-2666, p.7-12) heat removal from containment Id., at J.A. p.150 by low volume flow vent-filtered system, heat pipe or containment spray system (NUREG/CR-2666, p.7-13) increased reliability of Id., at J.A. p. 150 suppression pool cooling with system that could be driven from outside contain-ment, and closed loop heat exchange process high-volume vent-filter Id., at J.A. p.150 or high capacity sprays "if operated in a timely manner", thus requiring procedural alternatives to assure timely spray operation Partial draft 6/27/89 3
,o.g 11ox,,a isInsoa =c: 2x an1 s e - A t - N n 2'
measures to assure core Id., at J.A. p.151-2 debr.is bed coolability within pede.stal, including e.g., rubble bed, suitable flow passagen in pedestal wall, and measures identified by Swanson in " Core Melt Materials Interaction Evaluation" Annual Progress neport April 1980 to March 1981, ASAI Report No. 81-001. Alternatives identified in Id., at J.A. p. 153 NUREG/CR-3028 "A Review of the Limerick Generating Station Probabilistic Risk Assecament" Alternatives identified in NUREG/CR-3299, " Core Melt Id., at J.A. p. 153 Materials Interactions Evaluation"- Alternatives identified in Id., at J.A. p. 153 NUREG/CR-2182 " Station Blackout at Browns Ferry Unit 1 - Iodine and Noble Gas Distribution and Release" (Sept. 1982) Alternatives identified in Id., at J.A. p. 154 NUREG/CR-2672 "SBLOCA Outside Containment'at Browns Ferry Unit 1 - Accident Sequence Analysis" (November 1982) Alternatives identified in Id., at J.A. p. 154 NUREG/CR-2973 " Loss of DHR Sequences ~at Browns Ferry-Unit 1 Accident Sequence Analysis" (May 1983) Alternatives identified in R&D Monthly Project Status Id., at J.A. pp.155-173 Reports Contract NRC-03-83-092 Water cooled crucible core Id., at J.A. p. 166 retention device Partial draft 6/27/89 4 gg.g 11 o I 1 13 !
*1SIMSOd C t' : 21 3n1 6 e - 4 C -- N O C
'I a-flooded thoria rubble bed Id., at J.A. p. 166 core retention device water cooled. refractory- Id., at J.A. p. 166 tiles core retention device pebble-bed covering cooling Id., at J.A. p. 166 coils core retention device high-alumina cement covering Id., at J.A. p. 166 cooling coils core retention device magnesium dioxide covering Id., at J.A. p. 165 i
cooling coils core retention {' device-zirconium dixoide covering Id., at J.A. p. 167 cooling coils core retention device graphite covering cooling Id., at J.A. p. 167 coils core retention device I borax bath (thick layer of Id., at J.A. p. 167 borax bricks sealed in stain-less steel, covering the bottom of the reactor cavity) core retention device : heavy metal bath Id., uranium, or copper (lead, at-J.A. p. 167
)
iron oxide layer of iron Id., at J.A. p. oxide over c(ooling coils) 67 basalt concrete and basalt Id., at J.A. p. 167 rubble bed core retention device sand core retention system Id., at J.A. p. 167 iron core retention system Id., at J.A. p. 167 flooded cavity (water added Id., at J.A. p. 167 to flood entire cavity to vessel for core material to i be kept dispersed enough to ! remain quenched) ' Partial draft 6/27/89 5
, o.g 1Ao2,,3 isxnsoa z*: ET an1 se~40-NMT L_____ - - - - - -
I i: i 3 other active cooling systems Id., at J.A. p. 168 (special jackets and piping system in and around the reactor vessel with intention of retaining the molten core within the reactor vessel) i alternatives for overpressure Id., at J.A. p. 168 control from. hydrogen or hydrogen burning including oxygen exclusion, oxygen removal, oxygen dilution, igniters, fans overpressure control from Id., at J.A. p. 168 attack on concrete including special concrete composition of reactor cavity and basemat to limit release of nonconden-sible gases on core-concrete attack, and thin basemat compositon overpressure control by Id., at J.A. pp. 168-9 i venting the containment building with vent to tall stack, vent to receiver (another large, closed build!ng to provide larger total expansion volume and greater cooling) and vent to condenser-filter such as sand beds, gravel beds, scrubbers, gravel / sand, water pools, sand filters, charcoal filters, chemical scrubbers, all in various combinations overpressure control by Id., at J.A. pp. 169-170 containment heat removal with heat pipes, modified heat pipes, heat exchangers, spray coolers, fan coolers, secondary suppression pool, and more reliable residua'l heat removal system by increasing redundancy I and ruggedness of RHR system containment protection against Id., at J.A. p. 170 missiles - various structures designed to protect the , containment penetrations or walls l against flying debris or thrashing piping inside containment Partial draft 6/27/89 6 10 er 11o1773 A s 2 M s o .:s et: II sn1 se-As-unc
p
-4 special containment structures Id., at J.A. p. 170 such as underground siting of containment vessel, berm shield, double containment containment
. strength improvemen,ts of pressure ratings, increased volume of containment building, and strengthen-safety systems by-means of armor, bunkers, and heavier construction fission product removal systems Id., at J.A. p. 170' such as enhanced containment spray systems, and gas treatment system (special recirculating treatment system to remove fission products from the containment gas volume) ,
Alternatives identified in Id., at J.A. pp. 176-8 documents identified in Appen-dix A to NRC Response to FOIA 8 e432, documents 1 - 38 Alternatives identified in J.A. pp. 179-183; 189-191 LEA Contentions on the Environmental Assessment of ; Severe Accidents as Discussed In the NRC Staff DES, Supp. 1 Alternatives identified in J.A. pp.'193-256 l R&D Associates Monthly Project !
; Status Reports NRC Contract NRC 03-83-092 and other documents attached to LEA Statement of Significance of NRC Severe Accident Mitigation Systems Contract Documents to LEA Contention DES-5 Alternatives identified in J.A. pp. 233-256
" State of the Art of Reactor Containment Systems, Dominant Failure Modes, and Mitigation Opportunities", Jan. 1984 Partial draft 6/27/89 7 eo d 11o1,73 AszMsod ew ei an1 se-LC-Nnf
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . . . _ _ _ _ __ i
c
" operator action" as part of J.A. p. 237 a " containment mitigation system", defined as a cooper-ative combination of devices, subsystems, and components:
" operator action can be a part of such a system" and " operator action or modification of existing equipment can possibly perform as well as dedicated hardware in some cases and at lower cost". " State of the Art of Reactor Containment Systems, Dominant Failure Modes, and Mitigation Opportunities", Jan.
1984 Final Report, p.1-5
" operator action can play an J.A. p. 255 important role in accident mitigation providing there is enough time. Such a shrategy could potentially be much more cost effective than dedicated automatic systems with fail-safe initiating methods....[I]t is obvious that changes in current operating procedures both inside the plant...
and outside...may offer cost-effective reductions in risk. Altmatives Identified in Documents Id1Dtified to ASLB/ALAB containment heat removal NUREG-0850, " Preliminary Assess-(energy removal through containment heat removal- ment of Core Melt Accidents at active or passive) the Zion and Indian Point Nuclear Power Plants and Strategies for Mitigating Their Effects", Table 5.1, p.5-5 containment-atmosphere Id. removal (energy removal through containment-atmosphere removal- filtered vented containment systems) increased containment Id. volume (energy dilution through increased containment volume) partial draft 6/27/89 8 go.a 11ox,,a isIMsoa s ,. : zt an1 se-AC-unc
,e i- suppression of the burning of Id. hydrogen and other combustible gases -energy-release control through. suppression of burning (e.g., adding inert gases, Halon, water mists) controlled burning of hydrogen Id. and other combustible gases (energy release management through controlled burning of hydrogen and other. combustible gases, e.g., ignition systems) l core retention-devices-energy Id. release control and core mass management through core retention devices (core catchers, core ladle, cavity flooding, and active and passive cooling) missile shields- kinetic energy Id. dissipation of missiles strengthening of containment Id. structures - energy absorption enhancement through strength-ening of containment structures containment heat removal " State of the Art of Reactor alternatives such as heat containment Systems, Dominant pipes with input surface in. the drywell region and Failure Modes, and Mitigation discharge surface to the Opportunities", R&D Associates, atmosphere outside, cold water RDA-TR-127301-001, Final Report January, 1984, p.3-39 to 3-40' spray condensers in the drywell, or surface-type heat exchangers to cool suppression pool water i I Partial draft 6/27/89 9 O1 *d 1103713 is1MSOd GP: 21 ani 6 8 -- 1 C -- M M C
p. (.
', containment' venting of clean Id., at p. 3-41 steam and nitrogen.directly to surroundings and venting smaller quantities of contamin-ated steam and. gas through condensers and filter beds.
options also include those examined by Murfin,NUREG/CR-1410,
" Report of the Zion / Indian Point Study: Vol.I, 1980, Levy, " Review of Proposed Improvements, Including Filter / Vent of BWR Pressure-Suppression..." EPRI NP-1747, Ahmad, et al., NUREG/CR-2666, "PWR Severe Accident Delineation and Assessment", and Reilly,
" Conceptual Design of Alternative Core Melt Mitigation Systems for a PWR With an Ice-Condenser Containment" NUREG/CR-3068 (1982) (note that the Reilly study described as including designs suitable for the Mark II) core retention or debris Id., at p.3-41 control combustible gas control - while H2 Id.
control is provided in Mark II by deinerting containment with nitrogen, additional measures for hydrogen control may be needed (Papazoglou,NUREG/CR-3028 cited) to reduce the danger of flammability during service deinerting increased containment mass Id., at 3-42 holding capability with increaseu volume, increased pressure capability, improved pressure 4 suppression capability protection for containment Id. penetrations vent-filtered containment NUREG/CR-1029, " Program Plan options described throughout for the Investigation of Vent the document Filtered Containment Conceptual Designs for Light Water Reactors", Oct. 1979 Partial draft 6/27/89 10 g 3 .g 2,1ox,,a 1 s z ta s o a 9*: IT 3ni se-Ac-unc j
v
/
l
.O Alternatives Identified In or Suggested By Documents Published After the Denial of the LEA Contentio.D-modifications to reduce seismic NUREG-1068, " Review Insights risk on the Probabilistic Risk Assessment for the Limerick Generating Station", August 1984 safety assurance program Id., pp.8-4 to 8-5 alternatives identified in NUREG/CR-3908, " Survey of NUREG/CR-3908, " Survey of the the State of the Art in State of the Art in Mitigation Mitigation Systems", July Systems", July 1984 1984 alternatives identified in NUREG/CR-4920, " Assessment of NUREG/CR-4920, " Assessment of Severe Accident Prevention and Severe Accident Prevention Mitigation Features a BWR Mark and Mitigation Featurcs",
July 1988 II Containment Design , i NOTE: these include plant features and operator action / procedures 1 alternatives identified in NUFEG/CR-4244, " Strategies for NUREG/CR-4244, " Strategies for Implementing a Mitigation L Implementing a Mitigation Policy for Light Water Policy l for Light Water Reactors", January 1988 l Reactors", January 1988 alternatives identified in Boston Edison Co.,-" Report on Boston Edison Co., " Report Pilgrim Station Safety on Pilgrim Station Safety Enhancements", July 1987 and as Enhancements" as revised revised NOTE: these alternatives include both physical and operational plant changes supplemental containment system and other modifica- various documents of Long Island Lighting Co. for the cations as proposed and Shoreham Nuclear Power Station, installed for the Shoreham including the "Shoreham Nuclear Nuclear Power Station Power Station Probabilistic Risk Assessment With the Supplemental Containment System", February 1988; SNRC-1424 March 1988 (Response to NRC Staff Additional Questions) Partial draft 6/27/89 11 et a AAoI,7s AsIMsoa Aw: zt anA se-As-unc
a Li alternatives' suggested by NUREG-0979,. SER related to the the GESSAR II/BWR 6 " advanced final design approval of the reactor design" GESSAR II BWR/6 Nuclear Island il Design", as supplemented (1986) alternatives identified in NUREG/CR-4243, "Value/ Impact NUREG/CR-4243, "Value/ Impact Analysis for Evaluating l Analysis for Evaluating Alternative Mitigation Systems" I
-Alternative Mitigation January 1988 -{
Systems", January 1988 j i alternatives identified in NUREG-1150, " Reactor Risk ! NUREG-1150, " Reactor Risk Reference Document" 1987 ! Reference Document", 1987 operational alternatives NUREG/CR-4177, "Mangement of s
' identified.or suggested by Severe Accidents: Extending NUREG/CR-4177, " Management Plant Operating Procedures into of Severe Accidents", the Severe Accident Regime",
May 1985 May 1985 alternatives identfied in NUREG/CR-4025, " Design and NUREG/CR-4025, " Design and Feasibility of Accident Feasibility, of Accident Mitigation Systems for Light Mitigation Systems for Light Water Reactors",. August 1985 Water Reactors", August 1985 see esp., pp.3-24 to 3-77 l Partial draft 6/27/89 12 eo a llox,,a Asxmsos er: e nH1 ee-se-unc _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ . _ . _ _}}