ML20035C364
| ML20035C364 | |
| Person / Time | |
|---|---|
| Issue date: | 02/18/1993 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| GL-89-16, NUDOCS 9304070122 | |
| Download: ML20035C364 (21) | |
Text
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/h 7590-01 1
tWCLEAR REGULATORY COMMISSION Installation and Operation of Hardened Vents from Suppression Pool Airspaces of Boiling Water Reactors (BWRs) with Mark I Containments i
AGENCY:
Nuclear Regulatory Commission.
ACTION:
Notice of Issuance of Generic Environmental Assessment (GEA) and Finding of No Significant Impact (FONSI):
SUMMARY
- This GEA is being issued to inform the public of the NRC's assessment of the installation of hardened wetwell venting capability. The installation of the hardened wetwell vents P
in Mark I containment plants will reduce the environmental impact of severe accidents that challenge the integrity of Mark I containments, will provide a significant improvement in safety, and will not otherwise significantly increase the environmental impacts.
The installation of such vents was recommended by NRC in Generic Letter 89-16, dated September I, 1989.
The incremental occupational radiation dose for the recommended i
operation of the hardened wetwell vent path is insignificant (unmeasurable) because the vent would be operated from the control room.
The licensees should be able to keep the small radiation doses associated with installing the hardened wetwell vent path t
within the limits of 10 CFR Part 20 and as low as is reasonably 020061
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Furthermore, the non-radiological impacts of the.
vent path will also be insignificant.
-l Alternative Use of Resources 1
i This action does not involve the use of significant' resources beyond I
the existing resources used for piping and replacement parts at all nuclear i
'l plants.
Aaencies and Persons Consulted
't The NRC staff's evaluation is based on research performed 1
by the Office of Nuclear Regulatory Research.
No other agencies or persons were consulted.
EFFECTIVE DATE:
30 days after publication Public Comments No public comments on the Draft Generic Environmental Assessment (55 FR 25916, June 25, 1990) were received. However, the NRC staff has j
made editorial changes to the text of the Generic Environmental Assessment.
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The editorial changes had no effect on the technical basis of the Generic Environmental Assessment.
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6 FOR FURTHER INFORMATION CONTACT: Hohan C. Thadani, Division of Reactor Projects I/II, Telephone (301) 504-1476, Office of Nuclear Reactor Regulation,.
U.S. Nuclear Regulatory Commission, Washington, DC 20555.
SUPPLEMENTARY INFORMATION Backaround In SECY-87-297 (Reference 1), the NRC staff presented to the-Commission its plan to evaluate generic severe accident vulnerabilities of containments. The staff's plan included a program for Containment Performance Improvement (CPI). This program was initiated to determine whether generic f
severe accident challenges to light water reactor (LWR) containments may exist that the NRC staff should assess to ascertain whether additional regulatory guidance or requirements for these containment features were
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warranted.
The staff concluded that such assessments were needed because of the relatively large uncertainty in the ability of some LWR containments (that is, Mark I containments) to successfully survive some possible severe accident challenges as indicated in draft NUREG-Il50 (Reference 2). The CPI program is intended to resolve hardware and procedural issues related to generic containment challenges. The staff presented its findings related to the Mark I CPI program to the Commission in SECY-89-017 (Reference 3), dated January 23, 1989.
In one of these findings, the staff concluded that properly implemented venting can significantly mitigate potential accident risks.
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- i The canability to vent has long been recognized as important in reducing risk at boiling water reactor (BWR) Mark I facilities. For accidents involving the loss of capability to remove long-term decay heat (TW),
-i controlled venting at pressures close to the containment pressure limits can i
i prevent (1) the long-term overpressurization and failure of the containment, (2) the failure of emergency core cooling system pumps from inadequate net-positive suction head, and (3) the failure of the automatic depressurization system (ADS) caused by the failure of ADS valves to operate.
j A vent path from wetwells of the containments exists for some Mark I facilities.
However, this vent path includes a ductwork system that has a low i
design pressure of only a few pounds per square inch.
Venting under high-pressure conditions (as would be required for accidents involving high-i pressure challenges, either before or after core melt) could cause this ductwork to fail, releasing the containment atmosphere into the reactor building (with eventual release to the environment) and potentially l
contaminating or damaging equipment needed for accident recovery.
In addition, with the existing hardware and procedures at some plants, j
1 opening or closing the vent valves during certain accident sequences may not be possible.
The inability to operate the vent path valves could result in uncontrolled release of containment atmosphere to the reactor building through I
the failed sheet metal ductwork. Therefore, venting through a sheet metal f
ductwork path, as has been implemented at some Mark I plants, most likely
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i would greatly hamper or complicate post-accident recovery activities and is viewed by the NRC staff as inadequate for minimizing the risks to the public health and. safety.
For high-pressure venting to be effective, the entire vent
system must be strengthened to withstand the expected venting pressure. On July 11, 1989, the Commission endorsed the staff's view (Reference 4) that the Mark I design should include a hardened wetwell vent from the airspace of the containment wetwell, and directed the staff to require a hardened vent capability for all Mark I plants for which the requisite modifications could be shown to be cost-effective. At that time, the Staff prepared and.on June 25. 1990 published a draft Generic Environmental Assessment concerning hardened wetwell vent path modifications. The Staff issued generic letters to Mark I licensees requesting that they install hardened wetwell vents.
All BWR Mark I licensees have voluntarily agreed to implement the actions recommended in the generic letter or, in the case of FitzPatrick plant, were
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t found by the staff to have a hardened vent path satisfying the intent of the generic letter.
The NRC has concluded, in light of both the voluntarily actions of BWR Mark I licensees and 10 CFR 51.10(d), that the issuance of the generic letter need not be supported by the GEA.
10 CFR SI.10(d) states that Commission actions " initiating or relating to adminis-f trative or judical civil or criminal enforcement actions are not subject to Section 102(?) of NEPA...[ including) any other matters covered by Appendix C to Part 2 of t his chapter."
10 CFR Part 2, Appendix C, Paragraph IV.H, "Related Adm.nistrative Actions," lists generic letters as an " administrative mechanism" that supplements its formal enforcement authority. For this h
reason, the NRC concludes that the generic letter need not be supported by a GEA. However, since a draft GEA and FONSI was published, the NRC has decided to publish a final GEA to reflect the agency's final views on the 4
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environmental impacts of installation of hardened wetwell vents at BWR plants l
with Mark I containments.
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Description of the Recommended Action:
Review of the Impacts Of Installation of the Hardened Wetwell Vents l
5 The NRC staff's safety evaluation report (Reference 5) approved Revision 4 of the Boiling Water Reactor Owners Group Emergency Procedures i
Guidelines (EPGs) that included the staff's approval for venting BWR Mark I l
containments.
This approval indicated that venting from the' wetwell with the f
existing systems could reduce the likelihood of core melt and, in extremely
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rare cases, could help avoid uncontrolled releases of radioactive materials during severe core damage accidents.
Since the issuance of Revision 4 of the t
EPGs, additional insights have indicated that the proposed venting strategy
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t has a potential to breach the vent path inside the reactor building and could have significant detrimental _ effects on (1) radiation exposure effects on personnel, (2) potential plant recovery actions, and (3) public risk. A hardened wetwell vent capable of withstanding the anticipated severe accident pressure loadings would eliminate these disadvantages of using a vent path containing sheet metal ductwork.
i The use of the containment vent to prevent a core-melt accident by l
reducing containment pressure would result in the release of very low levels f
of radioactivity associated with the reactor coolant. The reactor coolant steam would be released to the suppression pool that would retain most of the j
fission products.
In the unlikely event of a core-melt accident, venting of 2
t the wetwell airspace would provide a scrubbed venting path to significantly reduce the release of particulate and volatile fission products (radioactive materials) to the environment. Only the noble gases would escape to the i
environment without any attenuation.
Venting would reduce the likelihood of a late overpressure failure of the containment and would reduce offsite f
consequences for severe accidents if the containment shell does not fail.
If the shell fails because of a core debris attack (shell melt through I
by core melt released to the containment floor), venting will provide little l
benefit because fission products would be released directly into the reactor building.
However, if shell failure was delayed for a period of a few hours j
(for example, by the addition of containment spray water over the molten core debris released to the containment floor), significant scrubbing of radioactive material would still take place.
As discussed in SECY-87-297 (Reference 1), the installation of hardened I
vent to bypass the ductwork from the wetwell airspace to the plant's environment could include (1) additional isolation valves to isolate the e
ductwork from the hardened vent path and (2) radiation monitors to monitor any offsite releases of radioactive materials in case of venting. The installa-tion of a hardened vent would prevent any loss of safety equipment cwing to a n
f ailure of the vent path inside the reactor building and, in the unlikely event of core melt, would result in release only of residual fission products (not scrubbed by the suppression pool) through the stack.
Because the vent n
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path is not expected to fail inside'the reactor building, personnel would be able to repair equipment and perform other plant-recovery activities, if the levels of radiation in tne containment were not excessive.
Furthermore, because the environmental conditions in the reactor building would not be harsh, important equipment would not be expected to degrade or fail.
Estimated reductions of potential risks of severe accidents involving loss of decay heat removal capability (TW accident sequences) are shown in Table 1.
The risk reductions are shown in persons-rems per eactor year (the bases and assumptions for the staff analyses are presented in Reference 6).
The hardened wetwell vent path would also provide additional risk reduction for those accidents in which core melt occurred, because the. suppression pool would scrub the radioactive material released to the containment by core debris.
The staff estimated the costs for instaliation of the hardened.
f wetwell vent path to be about 3750,000.. Costs were also provided by the licensees for the Dresden, FitzPatrick, Millstone 1, and Dyster Creek facilities.
The NRC's and the licensees' cost estimates are summarized in-Table 1.
The estimated costs of hardened wetwell vent installations are minimal when compared to the operating expenses of the plants and.are not-excessive when compared to the significant. enhancement of safety achieved by the proposed action.
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't Environmental Impacts of Installation and Operation of Hardened Wetwell Vents u
Radiological Impacts
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The radiological impacts of installing a hardened wetwell vent system should not be significantly different from other operational modifications that occur at facilities such as reactors with MarkLI e
containments.
For example, a conceptual analysis of radiation exposures _for installation of a filtered vent at the Limerick Generating Station indicates i
that annual radiation exposures (assuming 20 years of remaining plant life) _
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would not exceed 2 man-rems per reactor-year. The small iadiation dose I
associated with this proposed plant modification will not affect the j
licensees' ability to maintain individual occupational doses within the limits of 10 CFR Part 20.
This dose is expected to meet the criteria for the requirements of as low as is reasonably achievable -(ALARA).
Each plant contains radioactive waste treatment systems that are' I
r designed to collect and process the gaseous, liquid, and solid waste that.
might contain radioactive material. The installation of 'a hardened wetwell.
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vent will not affect any waste treatment systems or their effluents under normal plant conditions or under design basis accident conditions.
Installation of the hardened wetwell vent path should not significant1y' increase the radiation dose to operating personnel-or the public. Any increased doses associated with the testing of-the additional isolation valves-f should be minimal and, in most cases, insignificant.
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r Thus, the staff has concluded that the installation of the hardened r
wetwell vent will not result in any significant increasininadiological impacts to workers or the public.
Because the operation of the wetwell vent system is postulated for f
extremely rare severe accidents, the impacts of the use of the wetwell vent t
system are discussed in terms of environmental risks. As stated previously, the venting from the wetwell airspace is intended to (1) reduce the risk of over-pressure failure of the containment and subsequent damage to. the' reactor -
core and (2) provide a scrubbed pathway (to decontaminate effluents) for containment pressure relief for rare situations involving core damage.
Table 1 lists the reductions of potential risks for all Mark I facilities-caused by venting befere core damage.
The reductions in potential risks are
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calculated to range frnm about 15 to 282 person-rems per reactor year.
for rare situations in which core damage could occur, venting could prevent
'l containment failure and unmitigated release of fission products to the
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i environment.
Venting through the suppression pool will ensure that most of the i adioactive materials, excluding noble gases, would be trapped in the suppression pool and would not be released to the environment. Therefore, venting through the wetwell would reduce the radiological risks posed by severe accidents involving core damage.
These additional benefits of wetwell venting have not been included in Table 1 results.
f On the basis of the preceding discussion, the staff concludes that no incremental radiological environmental risks will be posed by operation of_ the wetwell vent system.
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Non-radiological Impacts i
The non-radiological impacts from installing hardened wetwell vent.
systems should not be expected to be different from other operational modifications that occur during routine plant outages at facilities with -
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Mark I containments.
t No non-radiological effluents are expected to be affected from installing or using the' hardened wetwell vent. The proposed plant modification and use of hardened wetwell vent will not require any change to the national pollution discharge elimination system permit.
I Therefore, the staff concludes that the non-radiological environmental impacts of installing a hardened wetwell vent will be insignificant.
i Alternatives Considered l
To prevent or delay containment failure caused by overpressurization, the NRC staff considered the following alternatives to installation of hardened vents.
1.
The containment pressure could be relieved using the existing ductwork vent path (the "No Action" option).
2.
A hard pipe path to an external filter could be installed.
3.
An alternative means of removing the decay heat either from the i
reactor or from containment could be installed.
4.
Venting of containment could be prohibited.
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L Each of these alternatives is discussed in the following paragraphs.
Existing Ductwork Vent Path (No Action Option) i i
This alternative consists of no action and contiuued venting of the containment through the existing ductwork. However, the existing ductworks are designed to withstand a pressure of a few psi (Reference.7)
The venting i
pressures expected during some accidents will be substantially higher then the f
ductwork design presssure. Consequently, venting could result in failure of i
the ductwork and a direct release of reactor coolant steam into the reactor i
building. The discharge of this high-temperature steam and other gases over f
an extended period of time could pose a threat to the availability or performance of safety-related equipment.
In the event of core melt,'the threat would be even greater, because substantially.large amounts.of_
radioactive materials will be released with the steam to the reactor buildin5-Electrical cables, motor operators on valves, and relays could fail under these environmental conditions.
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- i Adverse environmental conditions would also complicate personnel entry
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into the reactor building. Calculations from a study that examined venting a
.during an anticipated transient without scram (ATWS) sequence indicated that a j
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severe environment (high temperature and radiation) would be present-in the
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reactor building during venting (Reference 8). The discharge of hydrogen t
under core melt conditions could result in hydrogen burns or detonations 9
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1 inside the reactor building.
This environment could hamper recovery efforts by preventing personnel access to the reactor building and preventing repair f
i of systems needed to terminate the accident.
For these reasons, the existing l
Mark I designs do not ensure an adequate reactor building environment after a severe accident to permit personnel entry to regain control of the facility and do not maximize the potential reduction in environmental risk. Thus, the staff has concluded that the no-action alternative is unacceptable.
Installation of Hard Pipe Vent to External Filter System l
i I
This alternative is the same as the installation of a hardened vent with addition of an external filter.
However, the external filter would not significantly increase removal of radioactive material because the suppression pool would remove nearly all material that could be removed by filtration.
Consequently, the additional reduction in risk caused by an external filter.
system is expected to be small.
Moreover, an external filter would not yield an incremental reduction in the core damage frequency beyond the reduction obtained with the hard pipe vent alone..
In both cases, there would be no retention of noble gases.
Externr1 filters have been estimated to cost $20 million (1982 dollars) (Reference 9) to $65 million (1987 dollars) for the Filtra design.
Because the incremental benefit is very small compared to the proposed action and the incremental cost is very high, this alternative is not r
considered practical or reasonable.
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I Installation of Other Means of Decay Heat Removal i
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In lieu of venting containment, an addition decay heat removal system i
i could be provided to remove the heat from either the reactor or the containment, or a system that has not been previously accounted for could be I
used on an ad hoc basis, such as the reactor water cleanup system.
Installation of a new system was considered in NUREG-1289 (Reference 10),
which is associated with Unresolved Safety Issue A-45, " Shutdown Decay Heat Removal Requirements." The installation of a new decay heat removal system was not found to be cost beneficial in NUREG-1289.
The use of another, previously unaccounted-for system was estimated to require unusual or unplanned system piping line-ups, which, if performed incorrectly or j
inappropriately, could reduce the likelihood of accident recovery with normal systems or create a new and unanalyzed accident sequence (Reference 11).
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Therefore, this alternative is not considered practical or reasonable.
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i No Venting of Containment This alternative would remove the guidance in Revision 4 of the Emergency Procedure Guidelines (EPGs) that instructs the operator to vent the 9
containment under certain conditions.
In the event of the -loss of long-term decay heat removal capability without drywell f ailure, the containment drywell will probably fail because of overpressurization.
The drywell failure could have a significant effect on the ability to return the plant to a safe and controlled condition and would result in an increase in risk to plant
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personnel and to the public (Reference 12). Therefore, this alternative-is
- t not practical or reasonable..
P Findina of No Sionificant Imoact l
The staff reviewed the plant-specific features in conjunction with the installation of hardened wetwell vent path modifications.
From the i
cnvironmental assessment,:the staff concluded that there are no significant
-t radiological or non-radiological impacts associated with the installation -
't of hardened wetwell vents and that such use will not. have significant adverse effects on the quality of the human environment.
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. l Table 1 i
Estimated Reduction of Potential Risks of Severe Accidents Involving Loss of Decay Heat Removal Capability (TW Accident Sequences) l Potential Risk Reduction Installation TW frequency (persons-rems per Cost t
Plant Name (per reactor-year) reactor-year)*
Imillion) i i
Browns Ferry 1 2.3 E-05 32.7 0.75*
Browns Ferry 2 2.3 E-05 32.7 0.75*
Browns Ferry 3 2.3 E-05 32.7 0.75*
i Brunswick 1 4.5 E-05 44.0 0.75*
Brunswick 2 4.5 E-05 44.0 0.75*
I Cooper 4.5 E-05 45.6 0.75*
Dresden 2 1.4 E-05 50.2 1.00(Licensee Dresden 3 1.4 E-05 50.2 1.00 Estimate)'
i Duane Arnold 4.5 E-05 55.0 0.75*
Fermi 2 4.5 E-05 192.4 0.75*
i Fitzpatrick 4.5 E-05 65.5 0.68(Licensee l
Hatch 1 4.5 E-05 39.2 0.75*
Hatch 2 4.5 E-05 39.2 0.75*
Hope Creek 6.3 E-05 281.9 0.75*
i Millstone 1 1.4 E-05 35.1 1.10 (Licensee Estimate)
Monticello 4.5 E-05 33.9 0.75*
Nine Mile Point 1 1.4 E-05 15.3 0.75*
Oyster Creek 1.4 E-05 55.4 1.50 (Licensee i
Peach Bottom 2 3.6 E-06 15.5 0.75*
Peach Bottom 3 3.6 E-06 15.5 0.75*
Pilgrim 2.3 E-05 31.2 0.75*
-i Quad Cities 1 4.5 E-05 94.1 0.75*
Quad Cities 2 4.5 E-05 94.1 0.75*
Vermont Yankee 2.3 E-05 28.9 0.75* -
l NRC estimate i
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REFERENCES 1
1 SECY-87-297, U.S. Nuclear Regulatory Commission, " Mark 1 Containment Performance Improvement Program Plan," Commission Paper from V. Stello-to NRC Commissioners, December 8, 1987.
i 2.
NUREG-1150, U.S. NRC, " Severe Accident Risks: An Assessment for Five i
U.S. Nuclear Power Plants," June 1989.
4 3.
_SECY-89-017, U.S. NRC, " Mark I Containment Performance improvement' Program," Commission Paper from V. Stello to NRC Commissioners, January 23, 1989.
lf 4.
Memorandum from S. J. Chilk to V. Stello, "SECY-89-017-Mark I Containment Performance Improvement Program," July 11, 1989.
i 5.
Letter from A. C. Thadani to D. Grace, Chairman, BWROG, " Safety f
Evaluation of BWR Owners Group - Emergency Procedures Guidelines, Revision 4, NED0-31331, March 1987," September 12, 1988.
6.
Memorandum from M. Cunningham to W. D. Beckner, " Reduction in Risk from the Addition of Hardened Vents in BWR Mark I Reactors," October 19, 1989.
i t
7.
NUREG/CR-5225, U.S. NRC, "An Overview of BWR Mark I Containment Venting.
Risk Implications," November 1988.
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8.
Harring, R. M., " Containment Venting as a Mitigation Techniqce for BWR t
Mark I Plant ATWS," 1986 Water Reactor Safety Meeting, Caithersburg, Maryland, October 1986.
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9.
K. Johansson, L. Nilsson, and A. Persson, " Design Considerations'for-Implementing a Vent-Filter System at the BARSEBACK Nuclear Power Plant,"
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International Meeting on the Thermal Nuclear Safety, Chicago, i
August 29 - September 2, 1982.
.i 10.
NUREG-1289, U.S. NRC, " Regulatory and Backfit' Analysis: Unresolved f
Safety Issue A-45, Shutdown Decay Heat Removal Requirements,"
ti November 1988.
11.
Letter from J. Dallman to J.'_Ridgely, "A Preliminary Assessment of BWR Mark II Containment Challenges, Failure Modes, and Potential.
-Improvements in Performance," May 10, 1989.
12.
NUREG\\CR-5225, Addendum I, "An Overview of BWR Mark I Containment Venting Risk Implications, An Evaluation of Potential Mark I Containment Improvements," June 1989.
t Dated at Rockville, Maryland, this 18th day of February,1993.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION-b M
Thomas E. Murley, Directo Office of Nuclear Reactor Reg ation-1 NOTE: All referenced documents are available for public inspection and k
copying for a fee in the Commission's Public Document Room at'2120 L Street,-
Lower Level, NW., Washington, DC 20555.
'l II
9
, 9.
K. Johansson, L. Nilsson, and A. Persson, " Design Considerations for j
Implementing a Vent-Filter System at the BARSEBACK Nuclear Power Plant,"
International Meeting on the Thermal Nuclear Safety, Chicago, August 29 - September 2, 1982.
10.
NUREG-1289, U.S. NRC, " Regulatory and Backfit Analysis: Unresolved Safety Issue A-45, Shutdown Decay Heat Removal Requirements,"
November 1988.
11.
Letter from J. Dallman to J. Ridgely, "A Preliminary Assessment of BWR Mark II Containment Challenges, Failure Modes, and Potential Improvements in Performance," May 10, 1989.
12.
NUREG\\CR-5225, Addendum 1, "An Overview of BWR Mark I Containment.
Venting Risk Implications, An Evaluation of Potential Mark I Containment Improvements," June 1989.
p L
y Dated at Rockville, Maryland, this 18th day of February,1993.
FOR THE U.S. NUCLEAR REGUg g gP g g N Thomas E. Murley, Directopmas E.elet Office of Nuclear Reactor Regulation j
NOTE: All referenced documents are available for public inspection and copying for a fee in the Commission's Public Document Room at 2120 L Street, Lower Level, NW., Washington, DC 20555. SEE PREVIOUS PONCURRENCE*
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