ML20033E077
| ML20033E077 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/28/1990 |
| From: | Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Horton F HOUSE OF REP. |
| Shared Package | |
| ML17056A651 | List: |
| References | |
| NUDOCS 9003080312 | |
| Download: ML20033E077 (8) | |
Text
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Docket & PDRs - all Enclosures i
Remainder of Distribution: Enclosure 1 Only e
February 28.-1990 Distribution: *w/cy of incoming Docket File 50-220 JPartlow RCapra NRC/ Local PDRs*
JTaylor RMartin*
JKudrick The Honorable Frank Horton TMurley/JSniezek SVarga BBoger United States House of Representatives FMiraglia CVogan DCrutchfield Washington, D.C.
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Bev Clayton WRussell
Dear Congressman Horton:
SECY PDI-1 Rdg*
GPA/CA I am responding to your letter of February 6,1990, in which you requested that we look into some of the concerns expressed by the group " Retire Nine Mile One" regarding the Nine Mile Point Nuclear Station, Unit 1 (NMP-1).
The " Retire Nine Mile One" group raised issues concerning the use of a Mark I containment der.ign, the corrosion of certain materials comprising part of the containment (the torus), whether the refueling program should be stopped and the holding of a public hearing on NMP-1. The NRC staff has prepared the enclosed response on these issues.
In summary, I believe that these issues are being addressed appropriately by the NRC staff and the licensee for the NMP-1 facility. Consequently, I know of no reason at this time why the licensee's activities in preparation for restart should not proceed.
.The NRC staff will continue to monitor the licensee's progress in this area, and we will not make a decision on restart until the licensee's preparatory activities are essentially complete. Preparations are expected to be complete later this spring. Before NMP-1 is restarted, the licensee and the NRC staff will brief the Commission on the status of NMP-1 in a meeting that will be part of the public record.
I trust that this information will be useful to you in responding to the concerns of the " Retire Nine Mile One" group.
Sincer%}ibnal Signed By:
JamesMmg%%06 Executive Director for Operations
Enclosure:
- 1. = Response of HRC staff 2.
NRC Generic Letter 88-20, " Individual Plant Examination For Severe-Accident Vulnerabilities 10 CFR 50.54(f), " November 23, 1988 3.
NRC Generic Letter 89-16, " Installation of a Hardened Wetwell Vent," September 1, 1989 4.
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JKudrick The Honorable Frank Horton TMurley/JSniezek SVarga BBoger United States House of Representatives FMiraglia CVogan DCrutchfield Washington, D.C.
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_Bev Clayton WRussell
Dear Congressman Horton:
SECY PDI.1 Rdg*
GPA/CA I am responding to your letter of February 6,1990, in which you requested that we look into some of the concerns expressed by the group " Retire Nine Mile One" regarding the Nine Mile Point Nuclear Station, Unit 1 (NMP.1).
The " Retire Nine Mile One" group expressed concern in their letter to you dated January 8,1989, that are very similar to the concerns this group connunicated to Congressman James T. Walsh. Accordingly, we are enclosing a copy of our response to Congressman Walsh on this matter.
As we indicated to Congressman Walsh, we believe that the identified concerns are being addressed appropriately and that there is no reason at this time why the licensee's activities in preparation for restart should not proceed.
The NRC staff will continue to monitor the licensee's progress in this area, and we will not make a decision on restart until the licensee's preparatory _
activities are essentially complete.
Preparations are expected to be complete later this spring. Before NMP-1 is restarted, the licensee and the NRC staff will brief the Commission on the status of NMP.1 in a meeting that will be part of the public record.
I trust that this information will be useful to you in responding to the concerns _of the " Retire Nine Mile One" group.
Sincerely, x
James M. Taylor Executive Director t
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Dear Congressman Horton:
SECY PDI.1 Rdg*
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7 I _am responding to your letter of February 6,1990, in which you requested that we look into some of the concerns expressed by the group " Retire Nine Mile One" regarding the Nine Mile Point Nuclear Station, Unit 1 (NMP-1).
The " Retire Nine Mile One" group expressed concern in their letter to you of January 8,1989, that are very similar to the concerns this group communicated to Congressman James T. Walsh. Accordingly, we are enclosing a copy of our response to Congressman Walsh on this matter.
As we indicated to Congressman Walsh, we believe that the identified concerns are being addressed appropriately and that there is no reason at this' time why the licensee's activities in preparation for restart should not proceed.
The NRC staff will continue to monitor the licensee's progress in this area, and we will not make a decision on restart until the licensee's preparatory activities are essentially complete. Preparations are expected to be complete later this spring.
Before NMP-1 is restarted, the licensee and the NRC. staff will brief the Commission on-the status of NMP.1 in a meeting that will be part of the public record.
Itrustthatthisinformationwillbeusefultoyouinrespondingtoth/
concerns of the " Retire Nine Mile One" group.
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/[f, Executive Director for Operations Enclosu :/
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o ENCLOSURE 1 RESPONSE OF NRC STAFF TO TOPICS IN
" RETIRE NINE MILE ONE" LETTER DATED JANUARY 8, 1989 The letter of the " Retire Nine Mile One" group expresses concern about the cacability of the Mark I containment design at Nine Mile Point Unit 1 (NMP-1) to withstand the challenges of severe accidents. The issue of performance of the containment during severe accidents, in addition to other technical issues, has been a subject of the NRC's integration plan for closure of severe accident issues. Severe accider.ts are considered to be those accident scenarios that are more extensive, or severe, in their impact than the spectrum of design-basis accidents that have been analyzed, pursuant to specific regulatory u
requirements, as the basis for. licensing decisions.
The NRC has had severe accidents under consideration for many years.
Severe accident evaluations and research progressed to the point that the Commission issued a Severe Accident Policy Statement (50 FR 32138) on August 8, 1985, i
which concluded that existing plants posed no undue risk to the public.
However, based on NRC and industry experience with plant-specific probabilistic risk assessments, the NRC recognized that systematic examinations would be beneficial in identifying plant-specific vulnerabilities to severe accidents.
Accordingly, a severe accident closure implementation program was developed.
l The program includes several major elements that are directly pertinent to containment performance. One of these elements, the Individual Plant I
Examination for Severe Accident Vulnerabilities (IPE), will utilize l
probabilistic risk assessment or other systematic examination methodology to develop for each plant a better understanding of severe accident behavior and i
the specific accident sequences, including their probabilities.
Then, if i
necessary, modifications would be made to help prevent or to mitigate severe accidents.
Licensees for nuclear power plants have been requested to perform an IPE for their plant by an NRC generic letter dated November 23, 1988, and to submit the results of the IPE to NRC by August 1992. This generic letter is provided for information as Enclosure 2.
-The second major element of the severe accident closure implementation program j
to be discussed here is the containment performance improvement program (CPI).
This program is related to the IPE effort and is considered complementary to it as the CPI program is primarily focussed on the potential generic vulnerabilities of specific containment classes, whereas the IPE effort is focussed on plant-unique vulnerabilities. The CPI program includes both pressurized and boiling water reactor containment types. Based on the results of this program for i
the boiling water reactor (BWR) Mark I containment class of plants, the Commission-directed the staff to pursue Mark I enhancements on a plant-specific basis in order to account for possible unique design differences that may bear on the necessity and nature of specific safety improvements. Accordingly, the Commission concluded that the potential safety improvement candidates identified up to that point, with one exception, should be evaluated as part of the IPE Program, s
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With respect to the one exception, the Commission directed that improvements dealing with a hardened wetwell venting capability be approved, under the provisions of 10 CFR 50.59, for those licensees electing, on their own initiative, to implement them and that backfit analyses on the efficacy of requiring such improvements be undertaken for the remaining licensees, This i
issue of the hardened wetwell venting capability has been communicated to licensees by a generic letter dated September 1, 1989, which is provided as for information. The NRC staff understands from the licensee's letter of October 30, 1989 that the licensee for NMP-1 plans to install such a venting capability; however, the schedule for completing this action is still under discussion with the licensee.
The purpose of the preceding discussion is to highlight some of the present regulatory program efforts directed at assessing and improving, where needed, the performance capabilities of the Mark I containment. The statement regarding the high probability of Mark I containment failure referred to in the " Retire Nine Mile One" letter derives from the results of the WASH-1400 probabilistic risk assessment published in 1975. This assessment was based on a plant design and operating and emergency procedures as they stood at that time. As indicated by the discussion above about currently ongoing programs, much is being done to assess and, where needed, to improve the performance capabilities of the Mark I containment.
In summary, on this issue it is important to recognize that although the need has been identified to develop a severe accident closure program, which includes specific actions for the Mark I containment design, the Commission reached the conclusion in its policy statement on severe accidents in 1985 that, based on available information, the existing plants pose no undue risk to the public health and safety.
The " Retire Nine Mile One" group also asked that refueling of the NMP-1 be stopped. The licensee completed fuel loading NMP-1 on January 18, 1990, as part of its scheduled restart activities associated with a refueling outage that began in January 1988. On July 24, 1988, during this outage, the NRC l-Region ! Administrator issued a Confirmatory Action Letter (CAL) to the licensee that confirmed that certain actions would be taken and that the NRC Regional' l
Administrator's approval would be obtained before further operation of the unit. The NRC staff's concerns at that time evolved from consideration of the effectiveness that the licensee was achieving in responding to a number of issues.
One of these issues concerned the corrosion of the torus portion of the containment. This corrosion has been monitored for several years by the licensee. The licensee has prepared an assessment of the design margin remaining in the torus as a basis for supporting operation in the forthcoming fuel. cycle.
The licensee is also considering several potential corrective action approaches that it can take to resolve the problem on a long-term basis.
-The NRC staff has conducted independent measurements to confirm the licensee's measurements and currently has the licensee's response to this issue under review. The staff will require that an acceptable margin of safety fnr the torus structures be established before allowing operation in the forthcoming fuel cycle and in later fuel cycles.
t 1* The CAL confirmed the licensee's commitment to identify the root causes of such problems, to submit a restart action plan that addressed the root causes, and to submit a report about the readiness of the unit for restart.
These actions have been.taken by the licensee and have been reviewed extensively by the NRC staff.
The NRC staff has completed its review and has approved the Restart Action Plan for Nine Mile Point Unit 1.
The NRC staff concluded that the licensee's plan contains the essential elements to effect overall performance improvements.
The staff has also concluded that different aspects of the plan were being implemented with varying degrees of effectiveness.
These findings have been responded to by the licensee, and the NRC staff is monitoring the licensee's performance in this interim period as the licensee completes its planned activities to prepare the plant for restart.
The NRC staff is aware ~of the status of the issues that pertain to a decision on restart. The NRC staff has not identified any need to order the suspension of activities in preparation for restart of the plant, and the " Retire Nine Mile One" group provides no specific bases for its request in this regard. Accordingly, the staff does not plan to order suspension of the licensee's startup activities.
In response to the matter of a public hearing, it may be useful to consider that the NRC staff has previously provided an opportunity for public comment at a public meeting on the licensee's Restart Action Plan. These comments were considered and were responded to as set forth in Enclosure 4.
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November 23, 1988 Power Reactor facilitiesTo All Licensees Holding Operating Lionses anc C
SUBJECT:
INDIVIDUAL PLANT EXAMINATION TOR SEVERE ACCIDENT VULNERA8ILITIES - 10 CFR 550.54(f)
(Generic Letter No. 83-20) 1.
SUMMARY
issued on AugustIn the Commission policy statement on severe accidents in able information, that existing plants pose no undue risk to th and safety and that there is no present basis for immediate action on generic.
rulemaking or other regulatory requirements for these plants.
mission recegnizes, based on NRC and industry experience with plant-specificHow probabilistic risk assessments (PRAs), that systematic examinations are benefi -
cial in icentifying plant specific vulnerabilities to severe accidents that could be fixed with low cost improvements.
form a systematic examination to identify any plant-specific vulnerab severe accidents and report the results to the Commission.
The general purpose of this examination, defined as an Individual Plant Examina-tion (IPE), is for each utility-(1) to develop an appreciation of severe accident behavior, (2) to understand the most likely severe accident sequences that could occur at its plant, (3) to gain a more quantitative understanding of the overall probabilities of core damage and fission product releases, and (4) if necessary, to reduce the overall probabilities of core damage and fission product releases by modifying, where appropriate, hardware and procedures that would help prevent or citigate severe accidents.
It is expected that the achievement of these goals will help verify that at U.S. nuclear power plants severe core damage and large radioactive release probabilities are consistent with the Cc:aission's Safety Goal Policy Statement.
Besides the Individual Plant haminations, closure of severe accident concerns will involve future NRC and industry efforts in the areas of accident management and generic containment performance improvements.
Additianal discussion is provided in SECY-88-147 on the interrelationships among these three areas and the role they play in closure of severe accident issues for operating plants.
The portion of that document relevant to closure is pro-vided as Attachment 1. contains a list of references'of the 10COR program technical reports and also some related NRC:and NRC contractor reports.
Therefore, consistent with the stated position of the Commission and pursuant I
to'10 CFR 550.54(f), you are requested to perform an Individual Plant Examina-tion of your plant (s) for severe accident vulnerabilities and submit the results L
l to the NRC.
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2 November 23, 1988 2.
Examination Peccess The Quality and comprehensiveness of the results derived from an IPE will depend on tne vigor witn wnicn the utility applies the method of examination and on the utility's commitment to the intent of the IPE.
Furthermore, the maximum benefit from the IPE would be realized if the licensee's staff were gained from the examination becomes an integral part of plant p and training programs.
Therefore, we request each licensee to use its staff to the maximum extent possible in conducting the IPE by:
1.
Having utility engineers, who are familiar with the details of the design, controls, procedures, and system configurations, involved in-the analysis as well as in the technical review, and 2.
Formally including an independent in-house review to ensure the accuracy of the documentation packages and to validate both the IPE process and its results.
9 The NRC expects the utility's staff participating in the IPE to:
(1) Examine and understand the plant emergency procedures, design, opera-tions, maintenance, and surveillance to identify potential severe accident sequences for the plant; (2) understand the quantification of the expected secuence frequencies and unusually poor co;ntainment performance (3) determine the leading contributo and determine and develop an understanding for their underlying causes;,(4) identify any proposed plant improvements for the prevention and mitigation of severe accidents; (5) examine each of the proposed improvements, including design changes as well as changes in maintenance, operating and emergency procedures, surveillance, staffing, and training programs; and (6) identify which proposed improvements will be implemented and their schedule.
3.
External Events (Treated Separately)
Licensees are requested to proceed with the examinations only for internally initiated events (including internal flooding) at the present time.
Examina-tion of externally initiated events (i.e., internal fires, high winds / tornadoes, transportation accidents, external floods, and earthquakes) will proceed separately and on a later schedule from that of internal events (1) to peroit the identification of which external hazards need a systematic examination, (2) to permit development of simplified examination procedures, and (3) to-integrate other ongoing Commission programs that deal with various aspects of external event evaluations, such as the Seismic Design Margins Program (SDMP),
with the IPE(s) to ensure that there is no duplication of industry efforts.
Utilitics would be expected to examine and identify any plant-specific vulner-abilities to severe accidents due to externally initiated events. Therefore, while perfor. ing your IPE for internally initiated events, you should document and retain plact-specific data relevant to external events (e.g., data from plant walkdowns) such that they can be readily retrieved in a convenient form when needed for later external event analyses that may be required.
If a licensee chooses to sebmit an external event examination at this time, the staff would review it on a case-by-case basis.
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November 23, 1;ss while current staf f ef f orts are focuseo en icentifying acceptable methocs for examining external events, the staf f encourages the industry to prooose a metnocology for examining external events that nieets the inten external hazards.
logies for external ha2ure examinations.We will work with NVMARC in developing accep 4
Methods of Examination The NRC has identified three approaches that satisfy the examination requested by tnis letter.
The methods are:
1.
A PRA, provided it is at least a Level I
- and uses current methods and information, plus a containment performance analysis that follows the general guidance given in Appendix 1 to this generic letter. The staff will consider those PRAs that follow the PRA procedures described in NUREG/CR-2300, NUREG/CR 2815, or NUREG/CR 4550 to ce adeguate for perform-ing the IPE, provided the assessment considers the most current severe accident phenomenological issues (as discussed in Appendix 1) and the licensee certifies that the PRA is based on the most current design.
2.
The 10COR system analysis mothed (front end only), provided the enhancements identified in the NRC staff evaluation of the IDCOR method (to be issued snortly) are applied.
Guidance for the back-end analysis is provided in Appendix 1 and additional guidance will be issued as described in Section 11 of this generic letter.
3.
Other systematic examination methods, provided the method is described in the licensee response and is accepted by the NRC staff.
For those methods witn which the staff is not familiar, a staff review might be necessary to ensure that the methods are generally acceptable.
For the phase of the evaluation associated with core melting, release of molten core to the containment, and containment performance, the staff recognizes that for a few of the phenoment, notably associated with areas that affect containment performance, there is a wide range of views about their relative probability as well as their consequences.
For these issues, additional research and evaluation will be needed to help reduce the wide range of uncertainties.
Because of the concern over the ability of containments to perform well during some severe ac-cidents, the staff is conducting a Containment Performance Improvements Program.
This program complements the IPE program and is intended to focus on resolving generic containment challenges.
Licensees are expected to correct vulnerabil-ities that may be identified by their IPE results but, because of the generic Containment Performance Improvements Program that complements the IPE, the
- The PRA levels are defined as follows:
Level I - determination of core-damage frequencies based on system and numan factor evaluations; Level II - determina-tion of the physical and chemical phenomena that affect the performance of the containmer.t and other mitigating features and the behavior and release of the fission products to the environment; and Level III - determination,of the off-site transport, deposition, and health effects of fission product releases.
k 4
November 23, 1988 ments or other systems that can affect containm tion associated with the containment performance generic issues n rma-by the staff.
implement improvements before all containment perfo n developed made.
ve been Appendix 1 provides the utility with guidance t
ment performance.
Following the Appendix 1 guidance will also enable utilities!
to understand and develop strategies to minimite the challenge such severe accident phenomena may pose to the containment integri nces; recognize the role of mitigation systems while a.aiting their generic r 5.
Resolution of Unresolved safety / Generic Safety Issues (Relationsh to U51 A-45)
Because the resolution of several USI s) and GSI(s) may require an of the individual plant, it is reasona(ble to use the current I examination.
For example, Unresolved Safety Issue (USI) A 45 entitled " Shu Decay Heat Removal Requirements" had as its objective the deter the decay heat removal function at operating plants is adequate and if co beneficial improvements could be identified.
We concluded that a generic resolution to the issue (e.g.
a dedicated decay heat removal system for all plants) is not cost effective,and that resolution could only be achieved o plant specific basis.
each plant to do an examination of its decay heat remo vulnerabilities.
heat removal system and those systems used for the oth the purpose of 10ptifying severe accident vulnerabilities.
concluded that tha most officient way to resolve A 45 is to subsume it in Therefore, we have You should ensure that your IPE particularly identifies decay heat removal vulnerabilities.
from the six case studies performed for the USI A 45 program are discussed in Appendix 5 to this letter and should be considered as you c These insights duct your IPE.
during its IPE that is topically associated with any other vulnerability exists at its plant that is topically associated GSI, the staff will consider the USI or GSI resolved for a plant upon review and acceptance of the results of the IPE.
identify which USIs or GS!s it is resolving.Your IPE submittal should specifically 6.
PRA Benefits The NRC recognizes that many licensees now possess plant-specific PRAs or simila analyses.
Use of existing PRA analyses is encouraged in achieving the objectives of the IPE.
PRA analyses reflect the current state of the art regarding sev e
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3 November 23, 1988 In addition to being an acceptable method for conducting an IPEs number of potential benefits in performing PRAs on those plants without one there are a Some examples of potential additional benefits are as follows:
to justify technical spectrication changes, both i
i PRAs would also be useful in supporting other regulatory actioni design modifications).
license _ Renewals PRAs could be a basis for utilities to establish a program to ensure that risk significant components and systems are iden-tified and maintained at an acceptable level of reliability during the license renewal period.
that systematically uses the available information A PRA could be used to develop a risk management program power plant and identifies alternative combinations of design and opera-tional modifications, ranks these alternatives according to the relative l
benefits of each, and selects an optimum from the alternativet Intecrated Safety Assessment
- The staff believes that by performing a PRA a licensee would have the benefit of having developed the technical basis for an integrated assessment.
An integrated safety assessment would (1) provide integrated schedules for licensing, regulatory, and safety issues on a predictable basis, (2) evaluate licensing and generic issues on a plant-specific basis such that they are weighted against all demonstrate with its PRA that various issues that miother pen otherplantsarenotjustifiedatthatfacility,(4)ghtbeappliedto planning,h first,and (5) rank issue importance such that the most impor dealt wit This prioritization of actions benefits the licensees and the NRC by providing a rational schedule for implementation of actions and provides a basis for the possible elimination of actions determined to have low safety significance for the individual plant.
7.
Severe Accident Secuence Selection l
In performing an IPE tt is necessary to screen the severe accident sequences forthepotentiallyImportantonesandforreportingtotheNRC.
criteria to determine the potentially important functional sequences *e screening Th j
to core damage or unusually poor containment performance and should be reported that lead to the NRC with your IPE results are listed in Appendix 2.
Appendix 4 describes
- " Sequence" is used here to mean a set of faults result in the plant consequence of interest, i.e,.,either a damaged core orusually c unusually poor containment performance.
A functional sequence is a set of faulted functions that summarizes by function a set of systems faults which would result in the consequence of interest.
Functional sequences are to be i
contrasted with systemic sequences.
A systemic sequence is a set of faulted systems that summarizes by systems a set of component failures resulting in a damaged core or unusually poor containment performance.
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November 23, 1988 disposition of these sequences.the documentation needed for the accid e
It is expected that during the course of the examination or mitigation measures that could be taken to red the utility would or poor containment performance with the attendant radioactive release. cuency determination of potential benefits is plant specif The containment failure.
8.
Use of IPE Results a.
Licensee
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s After each licensee conducts a systematic search for severe accident vul l
ities in its plant (s and procedural, warra)nt implementation, it is expected t move expeditiously to correct any identified vulnerebilities that it ottarmines warrant correction.
provided consistent with the requirements of 10 CFR 50.5 Changes should also be reported in your IPE submittal (by reference to pre (see Appendix 4),submittels uncer 10 CFR 50.59 or 10 CFR 50.90) that respo;
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The NRC will evaluate licensee IPE submittals to obtain reas the licensee has adeavately analyzed the plant design and operations to discov instances of particular vulnerability to core melt or unusually poor containmen performance given a core melt accident.
Further, the NRC will assess whether the conclusions the licensee draws from the IPE regarding changes to the plant systems, components, or accident management procedures are adequate.
consideration will include both quantitative measures and nonquantitative ju The The NRC consideration may lead to one of the following assessments:
ment.
1.
If NRC consideration of all pertinent and relevant factors indicates that the plant design or operation must be changed to meet NRC regulations, then appropriate functional enhancements will be required and expected to be implemented without regard to cost except as appropriate to select among alternatives.
2.
If NRC consideration indicates that plant design or operation could be enhanced by substantial additional protection beyond NRC regulations, then appropriate functional enhancements will be recommended and supported with analysis demonstrating that the benefit of such enhancement is substantial and worth the cost to implement and maintain that enhancement, in accord-ance with 10 CFR 50.109.
3.
If NRC consideration indicates +. hat the plant design and operation meet NRC regulations, and that further safety improvements are not sibstantial or not cost effective, enhancements would not be suggested unless signifi-cant new safety information becomes available.
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Novemoer 23, 1988 9.
Accicent Manacement organizational involvement.An important aspect of severe acciden recognition of conditions or events that might lead to core e total tors and emergency teams can have a major in
{
The avail-case of a severe accicent, vents in Because the conclusions you will draw from the IPE for sever abilities (1) will depend on the credit taken for survivability of eq a severe accident environment and (2) will either depend on operators taking beneficial actions during or p,rior to the onset of severe core d on the operators not taking specific actions that would have adverse the results of your IPE will be an essential ingredient in developin accident management program for your plant.
v re At this time yo are not required to develop an accident managem integrated part of your IPE.
on this matter and are working closely with NUMARC to (
content of acceptable accident management action that will ultimately result in incorporating,any p(an)t-specific programs and 2 deemed necessary l
as a result of your IPE, into an overall severe accident management progra,m.
identify operator or other plant personnel actions that be immediately implemented in the form of emerge similar formal guidance.
such actions until a more structured t#d comprenensive a program is developed on a longer schedule, but rather to implement su immediately within the constraints of 10 CFR 50.59.
10.
Documentation of Examination Results The IPE shoold be documented in a traceable manner to provide the findings.
This can be dealt with most efficiently by a two-tier approach, first tier consists of the results of the examination, which will be report The to the NRC for review.
The second tier is the documentation of the examina itself which should be retained by the licensee for the duration of the license unless, superseded.
reporting and cocumentation. Appendix 4 contains the minimum information neces 11.
Licensee Response A document that provides additional licensee guidance for the performance IPE (both core damage and containment system performance submittals will be issued in draft form within the next few and describes the
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O November 23, 1988 Following the issuance of tne craft document tives will ce scheoulee to discuss the IPE cojectives and to answe, wo that utilities might have on ooth tne IPE generic letter and the g i r cuestions coCument.
u oance Following the completion of the workshops, the NRC, as appropria its guicence contained in the guidance documents to take int
, will revise ments received and will reissue them.
Within 60 days of receipt of the final guidance occuments, licensees are requested to submit the n com-for completing the IPEs.
The proposal should:
1.
Identify the method and approach selected for performing the 2.
Describe the method to be used, if it has not been previously for staff review (the description may be by reference), and 3.
Identify the milestones and schedules for performing the IPE ing the results to the NRC, Meetings at NRC Headquarters during the exa t
e as needed ons.
Licensees are expected to submit the IPE results within 3 years.
for severe accidents to promptly initiate the examination The Commission plant should (1) certify that the PRA meets the i in particular with respect to utility staff involvement, (2) certify that it reflects the current plant design and operation, and (3) submit the res soon as the analysis is completed but on a shorter schedule than 3 years Utilities with plants that used the initial IDCOR system analysis in the test applications are encouraged to submit their results on a shorter sche than 3 years.
This will ensure review and resolution of any items while the utility's examination team is easily accessible.
In this regard, the staff NUREG 1150 program to submit their IPEs on an exp the staf f to exercise its review and decision process for determining acc This will enable ity of the IPE, the adequacy of the licensee identification of plant spec vulnerabilities, and the associated modifications using insights and exp from NUREG-1150.
Finally, those licensees planning to perform a new Level II or Level III PRA may need more time.
additional time for such an examination.The NRC staff will consider requests for 12.
Regulatory Basis evaluation which justifies issuance of this letter is in Room.
Accordingly, all responses should be under oath or affirmation.
request for information is covered by the Office of Management and Budget un This
r 9
Neves0er 23. 1988 Clearance No. 3150 0011, which expires December 31 1989 burcen nours is 8100 person-nours cor licensee resp,onse,. The estimated average including assessment of the new requirements, searching data over a 3 year period anc analyzing the cata and preparing the required reports.
Comments on burden and duplication may ee, directed to the Office of Mana i
Management, Room 3208, New Executive Office Building,gement and lueget. Repo Washington, DC 20503.
Sincerely, i
Original signed by Dennis Crutchfield, Acting Associate Director for Projects Office of Nuclear Reactor Regulation
Enclosures:
Appendices 1 through 5 wl attachments 1 anc 2 i
O
\\
APPENDIX 1 GUIDANCE ON THE EXAMINATION OF CONTAINMENT SY (BACK-END ANALYSIS) 1.
Backcround to the environment has been widely recognized.The role of th The public safety record of nu-which relies on a set of independent barriers to fission containment and its supporting systems are one of these barriers.
The
{
design criteria are based on a set of deterministically derived challenges.
Containment i
of-coolant accident; radionuclide challenges are base 10 CFR Part 100.
floods, and tornadoes are considered.Also, criteria based on external events The margins of safety provided by such practices have been the subject of considerable research and evaluation, and these studies have shown the ability of many containment systems to survive pressure challenges of two to three times design levels.
Because of these margins, the various containment types presently used in the United States have the Capability to withstand, to varying degrees, many of the challenges presented by severe accidents.
For each type of containment, however, there remain failure mechanisms that could lead to either early or late containment failure, depending on both the accident scenarios involved and the containment types.
This appendix discusses the key phenomena and/or processes that can take place during the evolution of a severe accident and that can have an important offect on the containment behavior.
In addition, general guidance on the evaluation of containment system performance given the present state of the art of analysis of these phenomena is provided.
tion of the present containment capability.The evaluation should be a pragma and appreciation of severe accident behavior, should recognize the role mitigating systems, and should ultimately result in the develcpment of accident management procedures that could both prevent and ameliorate the consequences of some of the more probable severe accident sequences involved.
The users of this appendix are referred to Chapter 7 of Volume 1 of NUREG/CR-2300, "PRA Procedures Guide " for a more detailed description of procedures and guidance on containment performance analysis.
The additional information provided here summari2es some more recent developments in core melt phenomenology relevant to containment performance, identifies areas of uncertainty, and suggests ways of proceeding with the evaluation of containment perfomance despite uncertainties, and potential ways of improving containment performance 1
for severe accident challenges.
In this regard, the Severe Accident Prevention and Mitigation Features report (NUREG/CR-4920) summarizes insights gained from industry-sponsored PRAs, NUREG-1150, and 10COR reference plant analyses.
The report identifies plant features and operator actions that have been found to be important to either the prevention or the mitigation of severe accidents for a specific plant containment type.
The report indicates what may be important to risk and suggests potential improvements in various areas of plant design and operation.
These insights and suggestions may be helpful when conducting the IPE and when making decisions on plant improvements.
1-1
t l
i The systems analysis portion of the IPE identifies accident secuen occur as a result of an initiating event followed by failure of various syst or failure of plant personnel to resoonc correctly to the accicent.
the numoer of possible core melt accicent secuences is very large Althougn of containment system performance analyses coes not have to be as la
, the numoer nave a similar effect on the olant features that deter The i
transport of fission products, A containment event tree (CET) could provide a structured way for i
analysis of containment onenomena provided:
1.
The CET is quantified, i.e., branch point split fractions are propa for each sequence based on the most recent data base regarding imp severe accident phenomena including considerations of uncertainties the NRC/IDCOR Technical Issues," dated September (e 11, 1987).
22, 1986; November 26 1986; and March 2.
The system analysis is integrated with the containment analysis so that initiating events and system failures (resulting in core damage) that also impair containment systems are not overlooked.
3.
The duration and secuencing of the interacting events are specified s
e.g., the times at which core damage and containment failure occur,,the time of inventory depletion in particular, at related to recovery from an accident), the success or(failure of equipment or oper and the failure or cegradation of support systems that were originally available at the onset of the accident.
2.
Status of Containment Svstems Prior to Vessel Failure The role of interfaces between the s ment performance analysis (back end)ystem analysis (front-end) and the contain-tives. First, the likelihood of core damage can be influenced by the particular containment systems.
by the status of core cooling systems.Second, containment performance can be influe Thus, because the influences can flow in both directions between the system analysis (front end) and the containment p formance analysis (back end), particular attention must be given to these interfaces.
To ensure consistency within entire sequences, the aulysis should include a cross-checking sheet of the following by sequence:
(1) the sequence frequency.
(4) the containment system and reactor system availa mate source ters.
tems analyst and the source term analyst to provide added ass status of key systems is treated consistently in the front-end and back end
- analyses, Other options to ensure adequate interfaces can be used instead of the cross-checking list identified above.
In order to examine the containment performance, the status of the containment systems and related equipment prior to core melt should be determined.
The first CET nodal decision point is to determine the likelihood of whether the 1-2
4 containment is isolated, bypassed, intact, or failed (i.e., a branch fraction).
This requires analyses of (1) the pathways that could s contribute to containment-isolation failure, (2) the signals required to au matica11y isolate the cenetration, (3) the potential for generating th for all initiating events, (a) the examination of the testing and maintena procedures, and (5 moce (including com) mon moot the quantification of each containment-isolation failure e
failures).
contributors to containment pressurization.In the early phase decay heat removal systems such as sprays, fan coolers, and the su i
systems is to control the evolution of accidents that would otherwise lead t containment failure and the release of fission products to the environs.
effectiveness of the several containment decay heat removal syste The lishing the intended mitigating function should be examined to determin probability of successful performance under accident conditions.
potential intersystem dependencies as well as the id This includes and control power) or environmental conditions.considering pot If, as a result of the accident seguence, the front-line containment decay heat removal systems fail t if their effectiveness is degraded, or if the operator fails to respond in a timely manner to the accident symptoms, the containment pressure wou to increase.
In this case, some systems that were not intended to perform a safety function might be called upon to perform that role during an accident If the use of such systems is considered during the examination, their effe tiveness and probability of success for fulfilling the needed safety function should also De examined.
Part of the examination should be to determine if priate operator actions. adequate procedures exist to ensure the effective im 3.
Phenomena After Vessel Failure if adequate heat removal capability does not exist in a particular accident pressurize and eventually fail.secuence, the core will degrade and the cont vessel failure or to extend the time available for ves investigated.
the containment pressurization rate could exceed the capa mitigating systems to reject the energy associated with the severe accident phenomena encountered with vessel failure.
the molten core debris will relocate, melting through and mixing withFo materials in its path.
the accident sequence groups, a variety of important phenome challenges to containment integrity.
The guidance provided below deals with this subject at three levels.
first provides some rather general considerations regaroing the nature of The these phenomena as they impact containment (Section 3.1).
The second level considers the manifestation of these phenomena in more detail within the generic high and low pressure scenarios (Sections 3.1.1 and 3.1.2).
the third level provides some specific guidance particularly re
- Finally, treatment of certain important areas of uncertainty (Section 4)garding the 1-3
~
s s
3.1 General Descriotion of the Phenomena Associat Consicerations can occur ooth in vessel ano ex vessel.The contact of m explosion) of the primary system.tne reactor vessel, it may g
'4 a on (steam I
hood to not warrant additional consideration (NU less energetic in vessel steam explosions are not unlikely a u
- smaller, on fission product release anc hydrogen generation are still n uence If the fuel-coolant interaction occurs ex vessel, as might happe fuel fell into a water filled cavity upon vessel meltthrough on.
the corium and lead to rapid pressurization
, it may disperse n.
of the corium mass and the continued dissipation of the de would lead to containment pressurization.
3 Clearly in the absence of external presence of extensive, passive heat sinks (str volume would delay the occurrence of such an event.
can also yield a chemical reaction between steam and the metallic c Fuel coolant interactions the melt, producing hyorogen and the consequent potential for burns ani explosions.
r l
of the molten corium is with the concrete floor of th on I
l action produces three challenge to containment integrity.
This inter-decomposition gives off noncondensible gases (CO, CO) (of certain com that Contribute to pressurizing the containment atmosphere.
2 on) of certain compositions decomposes and releases C0 Second, concrete H, with potential consequences ranging from be 2
I hyorogen concentrations to rapid deflagrations at high hydrogen concent Third, continued penetration of the floor can directly breach the contai r
ons.
Dounda ry.
Also, tnermal attack by the molten corium of retaining sidewalls i
could produce structurai failure within the containment causing damag systems and perhaps to failure of containment boundary.
Another type of fuel interaction is with the containment atmosphere can be postulated (e.g., station blackout Scenarios primary system remain at high pressure as)the core is melting in which the reactor vessel and to the bottom of the vessel.
vessel lower head could eventually cause the lower head toContinued could be energetically ejected from the vessel.potentially high (a Because of a the molten corium Uncertainties remain related to the effect of the following on direct containment hea (1) vessel failure, (3) the degree to which it fragments upon ejection extent to wnich a path from the lower cavity to the upper co,ntainment atmosphe (4) the degree and is obstructed, (5) the fragmented molten corium that could enter and interact witn the upper containment atmosphere, and (6) cavity gas temperature the containment atmosphere has small heat capacity, the energy in the frag-Since mented corium could rapidly transfer to the containment atmosphere, causing a 1-4 1
.,_a
,n-,,, - - -
.c
Y I
e i
rapid pressurization.
bated by any hycrogen that may et simultaneously dis l
(exotnermic) of any metallic components.
x cation integrity early in the event. factors previously mentioned, this pressu i
a nment The BwR Mark I and Mark 11 containments are normally inerted i
concensible gases such as hydrogen and oxygen released follow Therefore non-cent would pressurize the containment, but would not burn or r If the containment is deinerted, additional pressurization events i
lotes obtained from global hydrogen burn or detonations must be co the various penetrations or produce a therma
~
red.
operability of important equipment.
Even with the above limited perspective, it should be clear that melt accicent, a great deal of the phenomenological progression availability and the outcome of the fuel coolant interactions; spe ther a full cuench has been achieved and whether the resulting particula remain coolable.
cant degree would imply the occurrence of energeti s will tne presence of significant forces that would be expected to disperse culates to coolable configurations outsi e the reactor cavity.
d arti-coolability of deep corium beds of coarse particulates is the major conc Otherwise the summary of how these mechanisms interface and interact as they integrat A
an accident secuence is given below.
3.1.1 Accident Secuences - Hich Pressure Scenario The core melt sequence at high primary system pressure is often du blackout secuence.
significant contributors to risk.The high pressure scenario also represents one o volve coolant boiloff and core heatup in a steam environment.The initi pressures, the volumetric heat capacity of steam is a significant fraction of At such high energy redistribution due to natural circulation loops and the remaining cooler components of the primary system.
to be developing that as a result of this eneroy redistribution, the primar Consensus appears system pressure boundary could fail prior to the cecurrence of large scale core melt.
The location and the size of failure, however, rsm-i. vncertain.
example, concerns have been raised about the possibility of steam generat For failures and associated containment bypass.
l If the vessel lower head fails, vio' lent melt ejection could produce large scale dispersal and the direct co ment heating phenomenon mentioned previously.
in the past has not yet produced definitive results on this hsue.A significant a within the blowdown process. Concerns may also be raised about the po The presence of nydrogen arises from two comple-mentary mechanisms: (1) the metal-water reaction occurring at an accelerated pace throughout the in-vessel cors neatup/ meltdown / slump portion of the tran-sient, ano (2) the reaction between any remaining metallic components in the melt and the high speed steam flow that partly overlaps and follows the melt ejection from the reactor vessel.
large quantit-les of hydrogen into the containment volume within a sho l
15 l
l l
t 4
period (a few tens of seconds).
The implication is that the consideration o containment atmosphere compositions and associated burning I
t) nation potential becomes comolicated by a whole range of highly t regimes and large spatial gradients.
ent A recent independent review of uncertainties in.stimates of source t severe accicents by an NRC sponsored panel of experts (NUREG erms from an additional perspective on these issues and made recoismendat resolution.
In particular, "if direct containment heating or containm througn steam generator tube failure contribute importantly to risk indicate a need for a haroware modification or a procedural m this may depressuritation before primary system failure.
merits of the possibilities available would be valuable.y' The st An earl study of relative of adopting the panel recommendation and has initiated a resea r
study the effect of depressurization on the core melt progression tial benefit in preventing direct containment heating.
3.1.2 Accident Seovence -_ Low-Pressuee Scenario i
At low system pressure, decay heat redistribution due to natural ci flow (in steam) is negligible and core degradation occurs at nearl conditions.
Steam boiloff, together with any h vously released to the containment atmosphere, ydrogen generation is contin-where mix convectd.on currents coupled with concensation processes.ing is driven by natural of the reactor vessel remain relatively cold, offering the possibilit The upper internals l
ping fission product vapor and aerosols before they are released to ment atmosphere.
to significantly load the containment is small.Throughout this core significant energetic loads on the containment occurs when the molte debris plenum. penetrates the lower core support structure and slumps into the lower The outcome of this interaction cannot be predicted precisely.
a whole range of behavior must be considered in o
- Thus, events.
some steam the lower re(and hydrogen) production while the melt quickly reagglomera sion occurs. actor vessel head.
It may be possible to distinguish intermediate outc degree to which the vessel integrity is degraded.
In analyzing this phase of the accident scenario the important tasks are to determine the likelihood of l
containment failure an,d to define an envelope of corium relocation pa the containment.
for such a phenomenon as liner meltthrough,The latter is needed t Conside' ration should also be given to ex-vessel coolability as the corium c potentially interact with the concrete.
The non energetic release lower head meltthrough) and spreading upon the accessible portions o(vessel containment floor below the vessel needs to be examined.f the of variability in accessible floor area among the various designs for some PW cavity designs.
The area over which the core debris could spread is rather small given whole core melts and the resultant pool being in excess of 50 cm In the absence of water, all these configurations would yield concrete deep.
attack and decomposition of variable intensity.
In the presence of water (i.e., containment sprays), even deep pools may be considered quenchable a coolable.
barriers at the corium water interface.However, the possibility exists for in 16
i Both of these two extremes should be consiccred.
The task is to estimate the
. ell as the extent of concrete floor penetration and s the situation nas been stabilized.
core concrete interactions (dry case) would be considerably slowe!
coolable deDris configurations (wet case) because of the absence of steam pressurization.
As a final and crucial part of this scenario, one must address the combu gas effect.
This must include evaluation of the quantities and composition of e
combustiole gases released to the containment, local inerting and dei steam and CO, as well as hydrogen mixing and transport.
2 be consideration of gaseous pathways between the cavity and upper cont volume to confirm the adequacy of communication to support natural circ and recombination of combustible gases in the reactor cavity.
General Guidance on Containment Performance 4
In the approach outlined in this appendix that would ensure that the IPE process con,siders the full range of seve accidents.
accident management scheme to deal with the probab performance at each plant.
To achieve these goals, it is of vital importance to understand how reliable each of the CET estimates are, and what the factors are.
Decisions on potential improvements should be made only after appropriately considering the sources of uncertainties.
failure altogether is predicated upon recovering some containment hect rem capability.
Given that in either case pressurization develops on the time accident management. scale of many hours, feasible recovery actions cou containment response is associated with the high press a low probability of early containment failure should Unless the Similarly, for BWRs it should not be assumed that the availability of sumed.
the automatic depressurization system (A05) in an event will ensure that reactor vessel failure will always occur at low pressure A05, in some plants, depends on maintaining a req,uisite differential pres between containment and the reactor coolant systems.
For BWRs, phenomenological uncertainties are associate combustibles and the spreading of the corium on the drywell floor.
For PWRs.
these areas include the coolability behavior of deep molten corium pools and the behavior of hydrogen (and other combustibles) in the containment atmosphere.
The staff's views and guidance concerning each one of these areas is briefly summarized below.
The concerns about deep corium pools arose from experiments with top-flooded melts that exhibited crust formation and long-term isolation of the melt from the water coolant.
Such noncoolable configurations would yield continuing con-crete attack and a containment loading behavior significantly different from coolable ones.
On the other hand, it has been pointed out that small-scale 1-7
,,m..
j experiments would unrealistically not favor coolability.
as an area of uncertainty anc recommends tsat assessme The staff views this cavity (spread) area and an assumed maximum coolable cepth of 2 in excess of 25 cm botn the coolaole ano noncoolab For ceoths Along tnese lines tne IPE should cocument the geometric cetail ered.
e consio-configuration and flow paths out of the cavity, including any into it as appropriate.
s of cavity s
Witn respect to hydrogen, the staff concerns are related to compl t current understancing of hydrogen mixing and transport.
e eness of the For the larger dry containments, because of the In general combustibles slow release rates, compositions in the detonable range may e and significant spatial concentrations exist or significant steam conde In general, the containment atmosphere under such conditi occurs.
exhibit strong natural circulation currents that would tend to tendency to stratify.
However, condensation-driven circulation patterns and other potential stratification mechanisms could lim tainment volume participating in the mixing process.it the extent of the con-igniters (ice-condenser and Mark !!! plants), the buildup of combu continuin burning. g corium concrete interactions could be limited by local ignitio rom flows could limit the effectiveness of this mechanism.Howe inerting/deinerting thresholds and ignition aspects need additiona The staff recommends that, as part of the IPE, all geometri n on.
the above phenomena (i.e., heat sink distribution, circulation paths readily comprehensible form, together with repres
, ignition l
For normally inerted BWRs burns and/or explosion eve,nts in deinerted Mark I or Mark in the secondary containment building following containment failure recommends that, unless deinerting can be satisfactorily ruled out by prob The staff bility, its occurrence and consequences should be l
conbustibles in it is essential with respect to the reactor building effective ness in limiting the source term.
how the corium will spread following discharge from Mark I containments, such uncertainties impact the configuration of the cor For concrete interaction process and also the potential for drywell liner meltt It is recommended that an assessment of the debris coolability, based on able water sources, should be performed to detersine the possibility for li meltthrough.
For Mark II containments, uncertainties are associated with the retention of corium on the drywell floor (and associated corium-concrete interactions) and the extent of fuel-coolant interactions in the suppression pool.
For PWR containments, the reactor cavity configuration will influence the potential for direct attack of the liner by dispersed debris, as well as the potential for basemat failure or structural failure due to thermal attack l
The staf f recommends that the IPE document describe the detailed geom (including curbs, standoffs) of the crywell floor.
1-8
-,,s.m__
. > + - ~
---r
E As discussed earlier, a CET provides a structured way for a syste of containment pnenomena.
low-pressure secuences oeal with uncertainties discusseo earlie In general terms, and consistent with the overall IPE cojectives, the staff guioance on the approacn to the oack eno analysis can be summarized as 1.
The approach should focus on containment failure mechanisms detaileo quantifications from reference plant an plant being examined.
2.
be considered and reported.All severe accident sequences that me 3.
System / human response should be realistically integrated with pheno-menological aspects into simplified, but realistic, containment event trees for the plant being examined.
probability of recovery or other accident management proceoutesA (particularly for long-term responses).
4 take into account the expected proThe quantification of the con envelop phenomenological behavior (gression of the accident and (b) aim to implies:
i.e., account for uncertainties).
This 3
Identification of the most probable list of potential containment a.
see Table 7-1, NUREG/CR-2300). failure mechanisms applic Use of existing structural analyses to determine the ultimate pressure D.
l capability of the containment, i.e., the Quasi static internal pres-sure resulting in containment failure.
These should be modified as I
necessary to take into account any unique aspects that could substan-tially modify the range of possible failure pressures.
Use of available separate effects analyses for the other potential c.
containment failure mechanisms to determine other failure mode L
which the plant might be vulnerable.
(
As stated ea some severe accident phenomenological issues (e.g.rlier, there are
, direct containment heating and containment shell meltthrough) where research has not produced conclusive results on the challenges that these phenomena could pose to containment integrity.
strategies to deal with those severe accident issues. Consideration mus For examole, will fully quench the debris and keep it coolable an Mark I containment shell meltthrough, there is a broad agreement that the presence of water will scrub the fission products and could sub-stantially reduce the radionuclide released even if containment shell m_eltthrougn were to occur.
Utilities should be aware of these insights and experience when conducting the IPE and should develop appropriate l
strategies to deal with those phenomenological issues while awaiting their generic resolution as discussed in Section 4 of the IPE generic letter.
u 1-9
l t
Deve1coment of a plant specific probability distribution function d.
failure likelinood for the range cf failure pressures.
Any claim of decontamination factors for the secondary containmen e.
the analyses shoulo consider the possibility of no natural circula-ited hydrogen ourns causing reactor building reactor building atmosphere out into the environment.
5.
Documentation should be presented concernin other aspects of the analysis.
Any use of codes within the IPE to calcu-late accident progression up to and including the source term calculat should be described along with the circumstances under which the code used, the version of the code used, any code revisions used, the key model ing and input assumptions, and the calculated results.
6.
The insignts gained from the containment performance analysis should be factored into the utility's accident management program.
t l
1-10
l c
APPENDIX 2 i
CRITERIA FOR SELECTING IMPORTANT SEVERE i
t Seovence Selection Criteria The following screening criteria should be used to determine whic important functional ceouences* and functional failures (based on a y
?
establisnea in NUREG/CR-2300) that might lead to core damage ure containment performance should be reported to the NRC in the IPE su do not represent a threshold for vulnerability.
They this appendix are " expected""* values.
All numerical values given in 1.
Any functional sequence that contributes 1E 6*** or more per reactor year to core damage, 2.
Any functional se camage frequency,quence that contributes 5% or more to the total core 3.
equal to IE-6 per reactor year and that teacs to co which can result in a radioactive release magnitude greater than or e to the BWR-3 or PWR-4 releasa categories of WASH-1400, 4.
Functional sequences that contribute to a containment bypass fre excess of IE-7 per reactor year, or 5.
Any functional sequences that the utility determines from previous applicable PRAs or by utility engineering judgment to be important contributors to core damage frequency or poor containment performance.
"" Sequence" is used here to mean a set of faults, usually chronological, t result in the plant consequence of interest, i.e., either a damaged core or unusually poor containment performance.
A systemic sequence is a set of faulted systems that summarizes by systems a set of component failures result ing in a damaged core or unusually poor containment performance.A functional sequence is a set of faulted functions that summarizes by function a set of systems faults which would result in the consequence of interest.
- For those cases where only point estimates are generated, the licensee shall propose a suitable factor that adjusts the overall value to the "expecteo" level.
- 1E-6 denotes abbreviated scientific notation for 1 x 10 8 2-1
E l
APPENDIX 3 ACCIDENT MANAGEMENT There alreacy is an international consensus that the cause and conseq a severe core camage accident can be greatly influenced by the operator's In aceition, the acility of essential equipment to survive the environmen sulting from severe accidents is an important consideration in mitigat severe core camage accicent and managing its progression.
accidents or (2) misinform the operator.tial ecuipment can (1) inc The failure of essen-accident management strategies.The NRC has initiated a researc We intend to periodically m t
(NUMARC) to compare the results of our respective programs. eet with industry has done some preliminary work in defining the key elements of a seve,re ac However the staff management program.
of sucn a program for your plant, we are providing yo work at this time.
The main elements of an accident management program should accress:
the responses to a severe accident (1) the organizational responsibilities a alarms needed to diagnose severe ac,c(dents 2) the instrumentation, procedures, and i
and the procedures and equipment needed to accomplish the functions necessar,y to prevent and to mitigate le accidents, ano (3) the procedures and training needed for operators to be i
skilled in possible remedial actions.
Succested Elements of an Accident Management Program 1.
Orcanization sibilities for dealing with these accidents and to ident ganizational structure.
The utility should decide which operators are to be trained to manage severe accidents or if_a separate evaluation team is to be established to direct the l
Clear lines of decisionmaking authority should be establishea. For operators.
l example, if containment venting is an option that could conceivabl during the course of an accident to prevent overpressure failure, y be considered
~
then the per-son responsible for making that decision should be clearly identified to all involved personnel.
Analyses of ultimate containment strength, the venting pressure, and the advantages, disadvantages, and potential consequences should also have been evaluated beforehand, and the decisionmakers should be propetly trained from the evaluation results to Mke an informed decision.
2.
Instrumentation and Equipment Practically every aspect of plant operation is likely to be involved in accident Coordination among the various organizational units is vital for management.
communicating the status and the control of needed equipment.
It should be clear (1) what information is needed to make decisions, (2) who is responsible 3-1
~... - - -
for obtaining the information, (3) what instruments plant personnel on to determine the status of the plant, and (4) what essential equipment is Survivability of sDecific touipment needs to be evalu ther the qualification of eouipment for design basis events is sufficient to support the assumed performance of this equipment during severe accidents.
)
of maintaining containment integrity is the main goal.For
, means Equipment needed to accomplish these functions s Heat removal from the appropriate preparations made.
All reasonable preparations to enable operators I
to recognize approaching containment failure, to assess possible remed i
and to accomplish the necessary functions should be provided.
Potentially no-i verse action should be identified and evaluated.
For example, recovery and tity of steam and hydrogen can condense the steam mixture of hydrogen. Similarly, spraying into a containment that has been vented could result in a vacuum and possible implosion.
If special equipment might be neeced to both prevent and mitigate severe ac cents, provisions might be made to ensure its timely availability.
should know where to procure the needed equipment.sibility The respon-t 3.
Procedures and Training The accident management plan should be developed to accomplish these fun for each set of the leading accident sequences despite the degraded state of the plant.
There should be consistency and smooth transition between the emer-gency operating procedures and the accident management plan.
The plan should be checked against the existing organizational structure to ensure that respon-L sibilities for managing each accident are clearly defined and the responsible l
personnel are adequately trained.
1 3-2
~
L I
APPEN0!X 4 DOCUMENTATION l
At a minimum, the following information on the IPE should be documente submitted to the NRC:
1 1.
by the provisions contained in this generic letter. Certi of the IPE and the validation of the results, inc The certification 1
sensitivity, and importance analysis.
2.
A list of all initiating events, the containment phenomena, and th states examined.
3.
All function event trees and centain' tent event trees (including quantific tion) as well as all data (including origin and method of analysis). The fault trees (or ecuivalent system failure models) for the systems identi-fied, using the criteria of Appendix 2, as main contributors to core damage or unusually poor containmnnt performance should also be provided 4
all applicable findings from the visual inspections.The suI 5.
A description of each functional secuence selected by the criteria of AD-pendix 2, including discussion of accident sequence progression, specific assumptions, and human recovery action.
6.
ability of.a large release.The estimated core damage frecuency and eacn of the leading functional sequences.The timing of significant large relea A list of analysis assumptions with their basis should be provided along with the source of uncertainties 7.
Identification of the USI(s) and G5!(s), if applicable, that have been assessed to estimate their contribution to the core damage frequency or to unusually poor containment performance.
1
?
8.
A description of the technical tasis for resolving any USI or GSI when.
applicable.
9, A list of the potential improvements, if any (including equipment changes as well as changes in maintenance, operating and emergency procedures, surveillance, staffing, and training programs) that have been selected for implementation and a schedule for their implementation or that are already implemented.
Include a discussion of the anticipated benefit as well as any drawbacks.
10.
A description of the review performed by a utility party not directly in-volved in producing the IPE to evaluate or oversee the IPE review.
11.
Documentation on the level of licensee staff involvement in the IPE.
4-1
e Retained Information The documentation pertaining to the examination that must be event trees and fault trees, current versions of t e
cable, walk through reports, and the results of the examiration.
all documents essential to an audit of the examination should be reta In general, addition, the manner in which the validity of these documents has in must be documented.
For any actions taken by the operators for which credit is allowed in the IPE, the Heensee should establish a plant proc d used by those plant staff responsible for manag e ure, to be
- occur, action.
Plant owner groups are encouraged to develop generic guideli which utilities can develop plant-specific accident management prog procedures.
r t
1 1
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D APPENDIX 5 DECAY HEAT REMOVAL VULNERABILITY INSIGHTS As cart of the Unresolved Safety issue (USI') program, six limited secoe PRAs perforted under the USI A-45 project. " Shutdown Decay Heat Removal Requirem were to assess the decay heat removal (OHR) function in existing plants
- showed that DHR related core damage risk is in a range, on some plants, where The results attentien may be warranted regarding whether or not such risks can be lowered in a cost effective manner.
related core damage risk are highly plant specific.The results also showed that the The following insights have been gained as a result of those six PRAs.
IPEs as tney relate to their search for potential core dam The in-with OHR-related severe accident secuences.Although licensees are requested iated events at the present time, insights from both interna neat removal function vulneracilities when performing the initiateo events.
Areas where such cost effective improvements might be possible were identified for severe accident sequences initiated by transients and small-break loss of-coolant accidents and were frecuently related to lack of redundancy, separation and physical protection in safety trains for internal fires, floods, sabotage, ano seismic events.
Such areas for possible improvement were particularly apparent in plant support systems.
At the support system level, there is often less redundancy, less separation and independence between trains, poorer overall general arrangement of equipment from a safety viewpoint, and much more system sharing as compared to the higher level systems.
These situations suggest the possible need to investigate corrective actions that could reduce the probability that single events such as a fire, flood, or insider sabotage could disable multiple trains (or single trains with a multiple purpose) thereby creating an inability to cool the plant.
See the following NUREG/CR reports:
- 4448,
" Shutdown Decay Heat Removal Analysis of a General Electric BWR3/
Mark I," March 1987.
- 4458,
" Shutdown Decay Heat Removal Analysis of a Westinghouse 2-Loop Pressurized Water Reactor," March 1987.
4713
" Shutdown Decay Heat Removal Analysis of a Babcock and Wilcui Pressurized Water Reactor," March 1987.
- 4762,
" Shutdown Decay Heat Removal Analysis of a Westinghouse 3-Loop Pressurized Water Reactor," March 1987.
- 4767,
" Shutdown Decay Heat Removal Analysis of a General Electric BWR4/ Mark I," July 1987.
4710
" Shutdown Decay Heat Removal Analysis of a Combustion Engineering Pressurized Water Reactor," July 1987.
5-1 l
1
Human errors were found errors of emission (e.g.to be of special significance.
The six studies mocele
, celays or failures in performing specified actions),c and it was found that in many cases the resulting risk was ve assumptions mace and to the way such errors were moceled.
Conseovently, great care is warranted in the development of hum In accition, it is likely that errors of commission are also import e s.
1 where the operator misciagnoses a situation and takes an impro is not be related to the actual, current plant situation).
per action that account will result in a more cospl&te picture of D Although such em into I
Of equal importance to human errors is the Credit that is a
+
i actions, which can have a very significant effect upon the resulting faults of batteries or diesel generators, actua Some aligning auxiliary feedwater steam and feedwater flowpath re-locally failed motor operated valves.
Considering the importance of such human methods anc assumptions used in these areas. recovery actions, Transient events that are initiated or influenced by a loss of offsite
.ere found to contribute si been issueo June 21, 1988 (gnificantly to risk.
53 FR 23203) as a resolution to 051 A-44A new rule Blackout."
Implementation of this rule will reduce the risk from su,ch even
" Station have a significant effect upon the OHR related core u
must be taken that feed and bleed operations wou However, care Quent damage.
In view of the potential benefits, significant effort might be t
justifiable in ensuring that procedures and training are actually in pla ficient to warrant credit for feed and bleed cooling,
(
i Just as the origins of DHR related risk are plant specific, the effects o rective actions are also quite plant specific and must be evaluated on a by plant basis.
In choosing which potential corrective actions to investigate in more detail, a general principle is that the modifications having potential for reducing the risk, for the lowest cost, will be those that increa the redundancy or availability of systems shared between units, tions are highly plant specific.in summary, both the DHR related risk and external causes, and the areas of support systems and hu particular significance.
Studies show that various cost-effective corrective actions may be possible to reduce OHR-related core damage risk after its has been identified.
5-2
ATTACHMENT 1 CLOSURE OF SEVERE ACCIDENT ISSUES FOR OPERATING REACTORS (Excerpted from SECY 88 147) hevior in operating light water reactors. Each progras a e-pect of severe accident behavior and may in fact result in a proposed spec pecific as-action on the part of the staff or Commission towards the regulated ind programs are critical to resolving the severe acciden and what specific steps must be taken by each licensee to achieve this resolution.
Completion of this resolution process is termed " closure" of severe accide issues.
Actions resulting from two tracks specific issues, must be taken for severe a;ccident closure.namely, generic issues and form of rulemakingsevere accident issues will be obtained when the Commissio or states whatever its required approach is.
plant-specific seve,re accident issues will be obtained when each license Closure for completed certain evaluations and implemented certain programs such that which comprise the dominant contributions to risk for each plant are identifie and that practical enhancements to the design, procedures, and operation are made such that further improvements can no longer be justified by backfit sis pursuant to 10 CFR $0.109.
However, specific plant and operational improve-ments may be identified which do not meet the backfit rule, but if implemented would significantly alter the risk profile of the plant, imp in our unoerstanding.
to the Commission with recommended action on a case-by case bas a single issue or combination of issues is achieved when the above is satisfied
. Closure of for that issue or those issues addressed, it should be noted that " closure" does not 10 ply that all severe accident acti-vities will cease.
Certain activities activities are designed to provide confirmation of prev These expected that as a result of continuing research, experience, and other activit-It is its, additional _ issues or questions regarding judgments related to severe acci-dents may arise.
basis, and are not expected to bring into question the previous c regarding closure.
The following sections describe in detail the steps that each licensee is ex-pected to complete in order to achieve severe accident closure for each of its operating reactors.
Al-1 l
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CompletinQ Inoividual plant Examinations (! pes)
The IDE oregram is intended to be "an integrated systematic approach to a examination of eacn nuclear power plant now operating or under construction for possiDie significant risk contributors (sometimes called " outliers") t mignt ce plant specific and mignt be missac absent a systematic search "
hat Each licensee is expected to perform an IPE using a staff.
tne staff expects that in many cases utilities, in the performance of thei the necessary safety improvements (confoming to th r
50.59).
However, through the review of IPE submittels, the staff may find it necessary to employ established plant-specific backfit criteria to assure that justifiable corrections are made.
For the phase of the evaluation associated with identification of dominant core melt sequences (commonly referreo to as the " front end" analysis of a decision process with respect to potential modification For the phase of the evaluation associated with core melting, release of molte core to the containment, and containment performance, the staff recognizes that there is a wide range of views about their relative proba consecuences.
For these issues additional research and evaluations will be needed to help reduce the wide range of uncertainties.
staf f is conducting a Containment Performance Improve details see Item 3 below).
This program complements the IPE program and is intended to focus on resolving generic containment challenges, including iss associated with the phenomena mentioned above.
The NRC and industry currently have ongoing research programs to address the few issues.
However, until a sufficient understanding of these phenomena is developed, eacn licensee will be faced with the need to be able to understand the potential range of probabilities and consequences associated with these issues.
Accordingly, we would expect each licensee to implement a Severe Accident Man-agement Program which provides training and guidance to their operational and technical staff on understanding and recognizing the potential consequences of these phenomena.
We do not plan to require a licensee to consider external events in its IPE at this time.
The staff is currently studying methods it would find acceptable for examining plants for severe accident vulnerabilities from external events, and will be meeting with NUMARC regarding these methods as well as the scope of an external event examination.
We expect completion of the methods develop-ment within 12 to 18 months.
Closure with respect to external events will be nel event vulnerabilities consistent with the conclusions of the described above.
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2.
Accieent Manacement i
effective severe accicent management.The staff has conclude
?eached by licensees for tneir plants will exolicitly rely on cWe also be t
actions, or on coerators not taking actions whien could adversely affect both ertain coerator tne procability anc consecuences of a severe accident.
i piementation of a Severe Accident Management Program tinue to be developed by both NRC and industry ov closure is not predicated on having a " complete" accident managem in place.
Rather, closure is based on each licensee having an Accident ment Program framework in place, that can be expanded, modified, etc. to accommodate new information as it is developed.
3.
Containment Performance Improvements As a result of concerns related to the ability of containments to withstanc generic challenges associated with severe accidents, the staff has undertake program to determine what, if any, actions should be taken to reduce the vulner-tude of releases that might result from such challenges. a Staff efforts have first focused on the BWR MARK 1 contai are primarily focused on the potential generic vulnerabilities of these contain-ments, and not plant unique vulnerabilities, which is the primary focus of the IPEs. The staff schedule calls for an interim report on BWR MARK !s to be sub mittee to the Commission in June of this year, with final recommendations due the fall of this year.
the fall of 1989.
The other types of containments are to be assessed by The IPE generic letter is now expected to be issued by July of this year, and licensees will have approximately four months to respond identifying their plan for conducting the IPEs. Following the four month period, it is expected they will commence with their IPEs.
It is further expected that any modifica-tions to Mark I containments that the staff may recommend will be available to the industry before they start their IPEs.
For the other containment types, the fact that any staff recommendations will not be available until after they have commenced with their IPEs is a concern.
However, the IPE generic letter will state that the st3ff does not expect the industry to make any major modifications to their containments until the information associated with the generic issues which affect containment performance has been developed by the staff. Hence, the industry will not be placed in a position of having to implement improvements before all containment performance decisions have been made.
4 Use_of Safety Goal in the Closure Process Thg6 staff expects to use safety goal policy and objectives, including the 10 / reactor year "large release" closure of severe accident issues. guideline, to assist in the resolution and Resolution and closure of issues are expected to be of t:vo different types, either plant unique or generic.
Safety Al-3 a
a a
-m
--m.
-a c-.
m.
o goals and objectives are to be used only for the resolution of gene i.e.,
severe accicent issues common to a cefined generic class sf pluts.
Resolution of plant unioue issues is to be accomplishea on a case by casis, viing the information cevelopec by Incivioual Plant Examination as is cescriteo in Section 1.
incorpora'.es the following, as cirected by the Co Memorancun catec November 6,1987):
(1)
Infor' nation on bow the staff proposes to implement OGC guidanc use of averted on-site costs in backfit analyses.
(2)
Whether averted off-site property damage costs should be included in a more txplicit manner in cackfit analyses.
(3)
Whether S1,000/ person-rem remains an appropriate cost / benefit criter (4)
A discussion of options for defining a "large release."
(5)
A discussion of eptions for specifying appropriate plant performance objectives.
(6) consioetations, and whether it would be accepta containnent if it met the large release criterion by prevention of core melt (cere damage) alone.
This plan will also reflect the consideration given by the staff to ACRS reco mendations amt the results of several meetings with the ACRS on this suDj
[
Resolution of severe accident pected to proceed as follows. generic issues using safety goal objectives is ex-PRA information from a variety of sourcet includ-will be used t1 make comparisons with applicable saf accordance with the implementation plan.
The staff will identify the reasons l
why particular plants appear to meet or not meet these objectives and assess these reasons in relation to current regulatory requirements.
will constitute a testing of the effectiveness of these requirements or their This assessment implementation and is expected to result in the identification of potential changes to regulatory requirements that, for some plants, would be expected to result in safety enhancements.
These, in turn, will be subject to appropriate regulatory analysis as provided in the Commission's backfit rule 10 CFR 50.109.
Those that can be shown to provide substantial safety benefit and are cost
- effective will be proposed to the Commission for backfit, possibly in the form of rulemaking.
The staff expects e at this process would have no impact on classes of plants 'or which there is reasonable assurance that safety goal ob-jectives are met.
%is expectation is based upon the intent to identify those-features of design and/or performance that are already in place at plants meeting safety goal objectives and to structure any new requirements such that they do not recuire changes or additions at these plants.
Al-4
The staff's revised Safety Goal Implementatien Plan is scheduled to Commission in August, 1988.
The first application is expected to be reflected in the staff's recommendations to the Commission in the Fall of 1988 o improvements to BWR MARK I severe accident containment performance.
5.
Summary of Closure Process on severe accidents for its plants are as follows:In summary, Complete the IPEs; identify potential improvements, evaluate and fix as appropriate.
Develop and implement a framework for an Accident Management Prog can accommodate new information as it is developed.
Implement any Commission-approved generic requirements resulting fro staff Containment Performance Improvements Program closure of containment performance generic issues.; this should constitute While programs for improved plant operations and re constitute " closure" of the severe accident issue for the plant in question.
Specific issues that may arise in the future as a result of ongoing research will be treated on a case-by-case basis and will not affect the closure proce k
e Al-5 d
y -
h ATTACHMENT 2 1.IST OF REFERENCES OF THE 10COR PROGRA 10COR Reports Tech.
Report No.
Title 1.1 Safety Goal / Evaluation Implications for 10COR m
2.1 Define Initial Likely SequencesGround Rules for Industry D 3.1 3.2 Assess Dominant Sequences 3.3-Selection of Dominant Sequences 4.1 Containment Event Trees 5.1 Human Error Effects on Dominant Sequences 6.1 Risk Significant Profile for ESF and Other Equipment E
7.1 Baseline Risk Profile for Current Generation Plants 9.1 Preventive Methods to Arrest Sequences cf Events Prior to Core Damage w/ Revision 1 10.1 Containment Structural Capability of LWRs 11.1/11.5 Identifying Pathways of Fission Product Transport 11.2 11.3 Fission Product Transport in Degraded Core Accidents j
11.6 Resuspension of Deposited Aerosols 11.7 FAI Aerosol Correlation 12.1 Hydrogen Generation During Severe Core Damage Sequences 12.2 Hydrogen Combustion in Reactor Containment B 12.3 13.2-3 Evaluation of Means to Prevent, Suppress or Control Hydrogen Burning in Reactor Containments 14.1A -
Key Phenomenological Models for Assessing Explosive Steam Generation Rates 14.1B Key Phenomenological Models for Assessing Non-Explosive Steam Generation Rates 15.1 Analysis of In Vessel Core Melt Progression 15.1A In-Vessel Core Melt Progression Phenomena 15.18 In Vessel Core Melt Progression Phenomena 15.2A Effect of Core Melt Accidents on PWRs with Top Entry Instruments 15.2B Final Report on Debris coolability, Vessel Penetration, and'04bris Dispersal 15.3 Core-Concrete Interactions 16.1 Assess Available Codes, Define Use and Follow and Support Ongoing Activities 16.1A Steview of MAAP PWR and BWR Codes 16.2-3 HAAP Modular Accident Analysis Program User's Manual, Vols. I & II
-16.4 Analysis to Support MAAP Phenomenological Models 17 Equipment Survivability A2-1
t e
I:i ATTACHMENT 2 (Continued) 17.5 Draft Final Report: An Investigation of High Temperature Accident Conditions for Mark-1 Containment vessels 18.1 Evaluation of Atmospheric and Liould Pathway Dose 18,2 Completion of Conditional Complementary Cumulativt Distribution Functions 19.1 Alternate Containment Concepts 20.1 Core Retention Devices 21.1 Risk Reduction Potential 22.1 Safe Stable States 23.1 Uncertainty Studies for PB, GG, Zion, Sequoyah 23.1B Peach Bottom - Integrated Containment Analysis 23.1Z Zion ~- Integrated Containment Analysis 23.15 Sequoyah - Integr6ted Containment Analysis 23.1GG Grand Gulf - Integrated Containment Analysis 23.
MAAP Uncertainty Analysis 23.3 Containment Bypass Analysis 4
24.4 Operator Response-to Severe Accidents 85.1 10COR 85 Program Plan 85.2 Technical Support for Issue Resolution 85.3 IPEM Al Thru B2 1
IPE Applications PB, Susquehanna, Zion, Oconee, BWR User's Guide 85.4 Reassessment of Emergency Planning Requirements With Present Source Terms 85.5A Revised Source Terms 85.5B Source Terms and Emergency Planning 86.20C Verification of IPE for Oconee 86.3A2 IPE Source Term Methodology for PWRs.
i-86.3B2 IPE Source term Methodology for BWRs 86.20G Verification of IPE for Grand Gulf 86.25H Verification of IPE for Shoreham 1
A2-2
1 NRC and NRC Contractor Reports Tech. Report No.
Title NUREG-0956 Reassessment of the Technical Bases for NUREG-1032 Estimating Source Terms Evaluation of Station Blackout Accidents at t
i Nuclear Power Plants NUREG-1037 NUREG-1079 Containment Performance Working Group Report Estimates of Early Containment Loads from Core Melt Accidents NUREG-1116 A Review of the Current Understanding of the Potential for Containment Failure from In Vess Steam Explosions NUREG-1150 Volumes 1-3 NUREG-1265 Reactor Risk Reference Document Uncertainty Papers on Severe Accident Source Terms NUREG/CR-2300 PRA Procedures Guide NUREG/CR-2815 Probabilistic Safety Assessment Procedures Guide NUREG/CR-4177 Volumes 1-2 Management of Severe Accidents NUREG/CR-4458 Shutdown Decay Heat Removal Analysis of a NUREG/CR-4550 Volumes 1-4 Westinghouse 2-Loop PWR e
Analysis of Core Damage Frequency from Internal Events NUREG/CR-4551 Volumes 1-4 Evaluation of Severe Accident Risks and the NUREG/CR-4696 Potential for Risk Reduction Containment Venting Analysis for the Peach Bottom Atomic Power Station NUREG/CR-4700 Volumes 1-4 Containment Event Analysis for Postulated Severe Accidents NUREG/CR-4767 Shutdown Decay Heat Removal Analysis of a GE BWR4/ Mark I NUREG/CR-4881 Fission Product Release Characteristics into Containment Under Design Basis and Severe Accident Conditions NUREG/CR-4883 Review of Research on Uncertainties in Estimates of Source Terms from Severe Accidents in Nuclear Power Plants NUREG/CR-4920 Volumes 1-5 Assessment of Severe Accident Prevention and' Mitigation Features NUREG/CR-5132 Severe Accident Insights Report A2-3
M
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LIST Of RECENTLY ISSUED GENERIC LETTERS e.
Generic Date of Letter No.
Subject issuance issued To 89-16 INSTALLATION OF A HARDENED 09/01/89 ALL GE PLANTS WETWELL YENT (GENERIC LETTER 89-16) 88-20 GENERIC LETTER 88-20 08/29/89 ALL LICENSEES SUPPLEMENT 1 SUPPLEMENT NO. 1 (INITIATION OF THE IN0!VIOUAL HOLDING OPERATING PLANT EXAMINATION FOR SEVERE LICENSES AND i
CONSTRUCTION VULNERABILITIES 10CFRS0.54(f))
PERMITS FOR NUCLEAR POWER REACTOR FACILITIES 89-15 EMERGENCY RESPONSE DATA 08/21/89 ALL HOLDERS'0F SYSTEM GENERIC LETTER NO.-
89 15 OPERATING LICENSES OR CONSTRUCTION PERMITS FOR NUCLEAR-POWER PLANTS CORRECT ACCESSION NUMBER IS 8908220423 89-07 SUPPLEMENT 1 TO GENERIC 08/21/89 ALL LICENSEES OF LETTER 89-07, " POWER REACTOR SAFEGUARDS CONTINGENCY OPERATING PLANTS, PLANNING FOR SURFACE APPLICANTS FOR VEHICLE BOMBS" OPERATING LICENSES, AND HOLDERS OF CONSTRUCTION PERMITS 89-14 LINE-ITEMS TECHNICAL SPECIFI- 08/21/89 ALL LICENSEES OF CATION IMPROVEMENT - REMOVAL OF 3.25 LIMIT ON EXTENDING OPERATING PLANTS, APPLICANTS FOR SURVEILLANCE. INTERVALS (GENERIC LETTER _89-14)
OPERATING LICENSES, AND HOLDERS OF.
CONSTRUCTION PERMITS 89-13 GENERIC LETTER 89-13 7/18/89 LICENSEES TO Al,L SERVICE WATER SYSTEMS l
PROBLEMS AFFECTING POWER REACTORS L
SAFETY-RELATED EQUIPMENT BWRS, PWRS, AND VENDORS IN ADDITION l
TO GENERAL CODES APPLICABLE TO l
GENERIC LETTERS 89-12 GENERIC LETTER 89-12:
7/6/89 LICENSEES TO ALL OPERATOR LICENSlHG POWER REACTORS EXAMINATIONS BWRS, PWRS. AND VENDORS IN ADDITION TO GENERAL CODES APPLICA3LE TO GENERIC LETTERS I
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UNI?ED$TATES j
.c NUCLEAR REGULATORY COMMISSION j(g, j
usmNotow.o c.resss
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- September 1, 1989 70:
ALL HOLDERS OF OPERATING LICENSES FOR NUCLEAR PO dlTH MARK I CONTAlhMENTS
SUBJECT:
. INSTALLATION OF A HARDENED WETWELL VENT (GE
- As a part of a comprehensive plan for closing severe accident issues, the staff undertook a program to determine if any actions should be taken, on a generic basis, to reduce the vulnerability of BWR Mark I containments to severe accident challenges.
At the conclusion of the Mark 1 Containment Performance Improvement Program, the staff identified a number of plant modifications that suostantially enhance the plants' capability to both prevent and mitigate the consequences of severe. accidents. The improvements (2) improved reactor pressure vessel depressurization (3) an alternative water supply to the reactor vessel and drywell sprays,, and
-(4) updated emergency procedures and training.
The staff as part of that-effort also evaluated various mechanisms for implementing of these plant improvements so that the licensee and the staff efforts would result in a coordinated coherent approach to resolution of severe accident issues in accordance with. the Commission's severe accident policy.
After considering the proposed Mark I Containment Performance Program (described in SECY 89 017, January 1989), the Comission directed the staff to pursue Mark I enhancements on a plant-specific basis in order to account for possible unique design differences that may bear on the necessity and nature of specific safety-improvements.
Accordingly, the Comission concluded that the recommended safety improvements, with one exception, that is, 1
hardened watwell vent capability,(IPE) Programshould be evaluate the Individual Plant Examination With regard to the recomended plant improvement daaling with hardened vent capability, the Commission, in recognition of the circumstances and benefits associated with this modification, has directed a different approach.
Specifically, the Comission has directed the staff to approve installation of a hardened vent under the provisions of
(
10 CFR 50.59-for licensees, who on : heir own initiative, elect to incorporate l
this plant improvement.
The staff previously. inspected the design of such a system that was installed by Boston Edison Company at the Pilgrim Nuclear Power Station.
The staff found the installed system and the associated Boston Edison Company's analysis acceptable.
A copy of Boston Edison Company's description of the vent modification is enclosed for your information. For the remaining plants, the staff has been directed to initiate plant-specific backfit analyses for each of the Mark I i-plants to evaluate the efficacy of requiring the installatior: of hardened wetwell vents.
Where the backfit analysis supports imposition of that requirement, the staff is directed to issue orders for modifications to install a reliable hardened vent.
_ -m Hy-
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Generic 1.etter 89-16,
Septemter 1, 1989 L
for installation of a hardened ventThe staff believes that the available in L
plants have in place emergency proce. First, it is recognized that all affected L
certain circumstances (primarily to avoid excuding the primary cont L
pressure limit) from the wetwell airspace.
Thus, incorporation of a designated capability consistent with the objectives of the emergency proced guidelines is seen as a logical and prudent plant improvement.
reliance on pre-existing capability (non pressure-bearing vent path) which Continued may jeopardize access to vital plant areas or other equipment is an.nnec
('
complication that threatens accident management strategies.
l
- Second, implementation of reliable venting capability and procedures can reduce the likelihood of core melt from accident sequences involving loss of long-term decay heat removal by about a f actor of 10.
Reliable ventin als; beneficial, depending on plant design and capabilities,g capability is likelihood of core melt from other accident initiators, for example, stationin red blackout and anticipated transients without scram.
As a mitigation ceasure, with significant scrubbing of fission products and can resu releases even for containment failure modes not associated with pressurization i
(i.e.
linermeltthrough).
for co,nsideration of coordinated accident management str t
design capability consistent with safety objectives.
For the aforementioned.
reasons, the staff concludes that a plant modification is highly desirable and a prudent engineering solution of issues surrounding complex and uncertain pher>.mena.
Therefore, the staff strongly encourages licensees to implement requisite design changes, utilizing portions of existing systems to the greatest extent practical, under the provisions of 10 CFR 50.59.
As noted previously, for facilities not electing to voluntarily incorporate design changes, the Commission has directed the staff to perform plant-specific backfit analyses, in an effort to most accurately reflect plant specificity, the staff herein requests that each licensee provide cost estimates for implementation of a hardened vent by pipe replacement, as described in SECY 89-017.
In addition, licensees are requested to indicate the incremental. cost of installing an ac independ6nt design in comparison to'a design relying on availability of ac power.
In the absence of such information, the staff will use an estimate of $750,000. This estimate ic based on modification of prevalent existing designs to bypass the standby gas treatment system ducting and includes piping, electrical design changes, and modifications to procedures and training.
The NRC staff requests that each licensee with a Mark I plant provide notification of its plans for addressing resolution of this issue.
If the licensee elects to voluntarily proceed with plant modifications, it should be so noted, along with an estimated schedule, and no further information is necessary. Otherwise, the NRC staff requests that the above cost information be provided.
In either event, 45 days of receipt of this letter.it requests that each licensu respond within
$w
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)
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s.
Generic Letter 8916
-3 Septercer 1, 1 This request is covered by Office of Management and Budget Clearance Numbe 3150 0011, which expires December 31, 1989. The hours are.100 person hours per licensee response, estimated average burden including searching data sources, gathering and analyzing the data, and preparing the required letters.
These estimated average burden hours pertain only to the identified response-related matters and do not include the time for actual implementation of the requested actions.
other aspect of this collection of information, including suggestions reducing this burden, to the Record and Reports Management Branch, Division i
U.S. Nuclear Regulatory Commission, Washington, D.C.of Inform 20555 and to the Paperwork Redui:'cion Project (3150-0011), Office of Managemen;t and Budge Washington, D.C. 20503, if you have any questions regarding this matter, please contact the NRC Lead Project Manager, Mohan Thadant, at (301) 492-1427.
Sincerely,
\\f 6
James G. Partlow Ass 3ciate Director for Projects Office of Nuclear Reactor Regulation
Enclosures:
1.
Description of Vent Modification at the Pilgrim Nuclear Power Station 2.
List of Most Recently issued Generic Letters m..
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'..: tar BECo 88-12 6 U. 5. Nuclear Regulatory Commission August 18, 1989 Occument Control Desk Wasnington, DC 20555 License OPR-35 Occket 50-293 REVISED INFCRMATION REGARDING PILGRIM STATICN SAFETY ENWANCEWENT DROGRaM Oear Sir:
Encloses is a description of a revised design for the Direct Torus Ven (DTVS) that was described in the " Report on Pilgrim Station Safety Enhancements" dated July 1, 1987 and transmitted to the NRC with Mr. Bi letter (SECo 87-111) to Mr. Varga dated July 8. 1987.
in its entirety the Section 3.2 included in the July 1, 1987 reportThis revision Cn Maren 7. 1988 Boston Edison Company (BEco) personnel met with Mr; Russell, and Dr. Thadani and provided a tour of SEP modifications a :
informal presentation of-the quantification of competing. risks. associated w an venting the containment and conclusions drawn from these results.
This presentation provided BECo the opportunity to respond to Questions posed und
" Installation of A Direct Torus Vent System (DTVS)" in Mr.
Varga's letter to Mr. Bird of August 21, 1987 " Initial Assessment of Pilgrim Safety Enhancement Program.
a resident inspector and was included as Attachment II in NRC
- 88-12, dated May 31, 1988.
As.you are aware from plant inspections we have installed the OTVS piping and certions of related control wiring. Currently, the OTVS is isolated from the Standby Gas Treatment System ($8GTS)- by blind flanges installed in place of Valve AO-5025 and the DTVS rupture disk.
NR2 in the performance of a technical review which focused on Syst Hechanical Oesign and Structural Design issues.
The review took place on March 2-3, 1988 as-documented in NRC Inspection Report #88-07, dated May 6. 1988 and determined-the installation configuration to be acceptable.
We now plan to remove these blind flanges and proceed with' installation of Valve A0-5025 and the OTVS rupture disk.
We conclude the valve and rupture disk provide
. equivalent physical' isolation of the OTVS piping from the $8GTS and appropriately ensure the operational integrity of the 58GTS under design basis accident conditions.
Following completion of this work, we will perform a local leak rate test to verify that Valve A0-5025 is acceptably leak ti using the same method previously utilized in testing the blind flange. ght He also plantocompleteallremainingelectricalworkontheO the revised design.
--.<,. 3818
- 04-ADOCK 05000293 A
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BCSTON FD!5CN CCMPANY
'i Augusc 18, 1988 U.S. Nuclear Regulatory,Comission Page 2 i
ceserteeo in tne encloture does notCn the basis of the revised Sec aquire any chage to the Te:hnt:21
~ Scecifications and that we can proceed with installation without pri approval.
Please. feel free to contact me or Mr. J. E. Howard, of my staff at (6 849-8900 if you have any questions pertaining to the design details of i
R. G. Bird
Attachment:
Section 3.2 Revision 1 " Installation of A Direct Torus System (OTVS)"
JEH/amm/2282 Mr. D. Mcdonald, Project Manager cc:
- Division of Reactor Projects I/II Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Hail Station P1-137 Washington, D.C.
20555 U. S. Nuclear. Regulatory Comission Pegion 1 475 Allendale Road King of Prassia, PA 19406
-Senior NRC Resident Inspector Pilgrim Nuclear Power Station 6
F
- . ; i:
?
Attachment to 8ECo Letter 88126 9
Section -3.2 Revision 1 " Installation Of A Otreet Torus Vent Sy page: 14, 15, 16, 17, 18, 19. 19A 198 i
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-3.2 INS?Sita?1CN OF A OfREC7 70003 e:
VENT SYS?tM (Orys) e 3.2.1 Obioetive of'Denian Chance This design change provides the ability for directL venting of the torus to the main stack.
Containment venting is one core damage prevention strategy util12ed in the SWR Owners Grcup Emergency Procedure Guideline (EPGs) as previously approved by the hkC ano is re The torus vent line connecting the torus EOPs).
stack will provide an alternate vent path for implementing EOP requirements and represents a significant improvement relative to existing plant vent t
capability.
torus, apporoximately 11 decay heat can be 3.2.2 Desien Chance Deterietion This design change (Figure 3.2 1) provides a direct vent path from the torus to the main stack bypassing the Standby Gas Treatment System (56GTS.
8" line whose upstream end is connec)ted to the pipeT between primary containment isolation valves-A0-5042 A &
8.
The downstream and of the bypass is connected to the 20" main stack line downstream of SSGTS valves and AON-112.
An 8" butterfly valve (AO-5025), which can be remotely operated from the main control room, is added downstream of 8" valve AO-50428.
primary containment outboard isolation valve for-theTh direct torus vent line and will confore to NRC requirements for sealed clossi isolation valves as defined in NUREG 0800 SRP 6 2.4.
III Class 2 up to and inclusive of valve A0-5025.The new l
connections are provided upstrena and downstream of Test L
A0-5025.
The design change replaces the existing AC solenoid valve for A0-50428 with a DC solenoid valve-(powered froe essential 125 volt DC) to ensure operability without dependence on AC power.
The new isolation valve, A0-5025..is also provided with a DC solenoid powered from the redundant 125 volt DC source. Both of these valves are normally closed and fail closed on loss of electrical and pneumatic power.
One inch nitrogen lines-are added to provide nitrogen to valves AG 50428 and A0-5025.
New l
valve A0-5025 will be controlled by a remote manual key-locked control switch. During normal operation, power to the A0-5025 DC solenoid will also be disabled by removal of fuses in the wiring to the solenoid valve.
This satisfies NUREG 0800 SRP 6.2.4, Containment-Isolation System acceptance criteria for a sealed closed barrier.
An additional' fuse will be installed and restin in place to power valve status indication for A0-5025 in i
l the main control room.
1 Rev. 1 (7/25/88)
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..I NUREG 0800. SRP 6.2.4, Iten !!.6.
L*
L.
valves..Seeled closed barriers include blind flange i'
sealed closed isolatio3 valves which may be closed L
remote-manual valves.
control to assure that saaled closed is I
cannot be inadvertently opened.
cavices to seal or lock the valve closedThis includ power from esing supplied te the~ valve op,erator.or to prevent Consistent with SRP-6.2.4, valve A0-5025 will be a s closed remote manual valve und6r administ to assure that it cannot be inadvertently opened.
Administrative control-will be maintained by a key-loc i
remote manual control' switch and a fuse remove prevent power from being supplied to the valve operator In accordance with NUREG 0737. Iten II.E.4.2,7 Positio 6, A0-5025 will be sealed closed and verified as such a least every 31 days.
i' A 20" pipe will replace the existing 20" diameter duct 20" pipe to the main stack.between S8GTS valves A 1
The existing 20" diameter
- duct downstream of A0-5042A is shortened to allow fitu of the new vent line branch connection.
' downstream of valve A0-5025.A rupture disk will be L
provide a second leakage barrier.The rupture disk will The rupture disk is will be intact up to pressures equal-to or l
those which cause an automatic containment isolati during any accident conditions.
The two Primary Containment Isolation Valves (PCIVs)
A0-50428 and A0-5025 are placed in-series with the rupture disk.
No single operator error'in valve operation can activate the OTVS.
The rupture disk has a.
rupture pressure above 'the automatic containment high pressure trip point.
will ~recelve an automatic-isolation prior to disk.Thu rupture.
The inboard PCIV (AO-50428) requires physical electrical jumper. installation to open at primary containment pressure above'the automatic'high pressure trip point.
Valve A0.5025 will be closed whenever primary containment integrity is required and DC power to its solenoid control valve will be disconnected. Indication of valve position will be provided in the main control roca even with the valve power removed.
Use of the direct torus vent will be in accordance with approved EPG requirements and controlled by (ops in the same manner as other existing containment vent paths. Prior to opening the vent valves the S8GT system will be shutdown and valves AON-108 and A0N-112 (the outlet of S8GT) placed in a closed position.
-15 Rev. 1 (7/25/88)
New 8" vent pipe (8"-HBS-44), including valve AO 50
(
safety related.
Vent ptstag downstreas of A0-5025, including SSGTS discharge piping to main stack...is als safety related.-'All safety related piping will be supported as Class I.
related and will be supported as Class-II/!. Nitrogen p The interpretation of the Clu s II/I designation.through-tnis report is given below:
All Class II items which have the potential to degrade the integrity of a Class I item are analyzed. Such-Class II items do not require dependable mechanical or electrical functionality during SSE, only that all of the i
following conditions prevati:
1.
The Class II items create no missiles which impact unprotected Class I ttoms safety functions.
2.
The Class II item does not deform in a way which would degrade a Class I item.
3.
.If the Class II item falls, then the Class I item is
' protected against the full impact of all missiles-generated by the assumed failure of Class 11 items.
All electrical portions of this design are safety related except for the indicating lights on the MIMIC panel C904, the tie-ins to the annunciator, and interface with the plant computer.
3.2.3 Desien Chance Evaluation 3.2.3.1 Sys tems/Comeonents affected Containment Atmoseheric Control System (CACS)
E The torus purge exhaust line inboard isolation valve AO-50428 and the associated 8" pipe are l
.e l
the components-of the CACS affected by the design modification. With incorporation of the subject.modtfication, the CACS will depend on
- both essential AC (for valve A0-5042A) and essential DC (for AO-50428) to perform its purging function.
The new 8" torus vent line will be connected to existing 8" CACS piping between valves A0 50428 and AO-5042A.
l Rev. 1 (7/25/88) 9 v
4
(
- I
, f' Standby Cat ?reatment System (SRCTS)
The SBGTS fan outlet valves (AON-108 and AON-112), ductwork from these valves to the 20" o
line leading to the main stack, and the 20" line leading to the main stack are the components of this system affected by the-proposed change.
L Valve AON-108 is normally closed, fail-open.
l Valve AON-ll2 is normally closed, fail-closed, and these valves are provided with essential DC power and local safety related air supplies.
Primary containment Isolation Svitam (DCIS)
Valve AO-50428 is affected by the change from AC to DC power for the solenoid and by replacement of the existing air supply with nitrogen.
The addition of containment outboard isolation valve (AO-5025) will.not affect the PCIS.
Primary Containment System (PCS)
Valve A0-5025 acts as the primary containment-outboard isolation valve for the direct torus vent line and will conform to NRC requirements-for sealed closed isolation valves as defined in NUREG 0800 SRP 6.2.4.
3.2.3.2 Safety runctions of Affected-Systems /Comeonents Cantainment Atenteherie control Sviten This system has the safety function of reducing-the possibility of an energy releas9 within the primary containment from & Hydrogen-0xygen reaction following a postulated LOCA combined with degraded Core Standby Cooling. System.
Standhv Cat Treatment Svatem This system filters exhaust air from the reactor building and discharges the processed air to the main stack. The system filters particulates and todines from the exhaust I
stream in order to reduce the level of airborne contamination released to the environs via the-main stack.
The $8GTS can also filter exhaust air from the drywell and the suppression pool. Rev. 1 (7/25/88)
.v a-I Drimary Containment Isolatten Svitem
+
This. system provides timely protection again the onset and consequences of design basis accidents involving the gross release of radioactive materials from the ' primary containment by initiating automatic isolation of appropriate pipelines which. penetrate the
- rimary centain ent whenever monitored variables exceed pre-selected operational limits.
Primarv Containment Svitam The primary containment system, in conjunction with other safeguard features, limits the release of fission products in the event of a postulated design basis accident so that offsite doses do not exceed the guideline values of 10 CFR 100..
3.2.3.3 Potential-Effsett on Safety Function _s Containment ~Atmannherie Control System. Standhv Cat Treatment _ Svates. Primarv Containment Isolation Svates and Primarv Cantainment S The improvements change the A0-50428 solenoid control from AC to DC enabling it to open (from its normally-closed position) with no dependence on AC power availability.
The existing air supply to A0-50428.is being replaced-by nitrogen.
i Ductwork at the outlet of the SSGTS is replaced with pipe and the new vent line is connected to the 20" line at the outlet of the SBGTS.
1
- Addition of a new 8' vent line with containme u.
L isolation valve'A0-5025 off,the existing torus vent line could introduce n' flow path under design basis conditions that could vent the containment-directly to the stack bypassing.the SSGTS.
3.2.3.4 Analysis of Effects on Safety Functient An analysis of the effects on the safety functions of CACS, SBGTS ~PCIS and PCS for the installation of the direct-torus vent is described as follows:
The change from AC to DC control and the replacements of air with nitrogen on A0-50428 L
does not adversely affect the ability to open A0-50428 when the containment is being purged, or to isolate under accident conditions.
.)3 Rev. 1 (7/25/88) i I
t-
leading.to the anin stack do n m..
L*,. '
design basis safety. function of any of the Safety related systems.
During normal plant operations, the CACS and the 5BGTS do not use the torus 20' purge and vent line to perform their safety functions.
The containment isolation valves are in primary containment boundary inte I.
There are no adverse affects on the primary containment system by the addition of the OTVS.
t Valve A0-5025 will conform to NRC defined in NUREG 0800 S affect design basis accidents.
will be in accordance with the contain NRC and controlled by E0Ps in th as other existing containment vent paths.
effects on the torus of the new 8' piping and -
The A0-5025 have been evaluated for Mark I progr loadings, using ASME 8PVC Section III l
criteria.
The remaining piping includinq the
[
rupture disk was evaluated using ANSI B3.1 requirements.
s Ouring plant startup and shutdown (non-emergency condition) when the purge and l
vent line is in use, valve A0-5025 remains
~ closed.
In addition, the rupture disk downstream of valve A0-5025 will provide a second positive means of preventing leakage and prevent direct release up to the stack during containment purge and vent at plant startup or shutdown.
During containment high pressure conditions.
p the torus atin exhaust line is automatically isolated by the PC15.
There is no change to the entsting primary-containment isolat' on system function for A0-5042A or A0-50428.. The sealed closed position of valve A(> 5025 and the addition 41' assurance added by the rupture disk downstream will prevent any inadvertent-discharge up the stack for all design basis accident conditions.
3.2.3.5 Desian Chanas Evaluation sm ev canetunions Installation of the DTVS does not adversely-affect the safety functions of the CACS,16GTS, PCIS or the integrity of primary containsent or any other safety related systees.
_19 Rev. 1 (7/25/88)
4 l
Use of the OTVS cill be in accordanc containment venting provisions of EPGs as approved by the N K and controlled by (ops in J
^
the same manner as other existing containment vent paths.
.The OTVS provides an improved t
containment: venting capability for decay heat i
. removal which reduces-potential onsite and offsite impacts relative to the existing centainment venting capabiltty.
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w ENCLG53dE I 06DECmg Docket Nos. 50-220 50-410 Niagara Mohawk Power Corocration ATTN:
Mr. Lawrence Surknarct, III Executive Vice President Nuclear Operations 301 Plainfield Road l
Syracuse, New York 13212 i
Gentlemen:
Subject:
Point Unit 1 Received from the Public - N e
Comments ose This letter addresses the August 23, 1989,
- York, of the transcript.of the commentsto receive comments on the-Res e
n Oswego, New Based on review received by mail, the NRC staff from this meeting and of written comments through its Restart concluced that no changes are ne,eded _ to the RAP.
Assessment-Panel, ha viously noted This conclusion was pres in the NRC approval of the September 29, 1989.
RAP in a
letter dated' provides responses to the specific comments.By transm o this letter each comment.
A response has oeen proviced The responses are grouped according related to the RAP to Mile Point (20 comme (nts).
y Although the comments are provided for your information concern regarding the turning off of radiation m
, comment 35 raised a unusual events.
s letter dated October 12, 1989.
and We appreciate your cooperation.
Sfncerely.
William F. Kane, Director Division of Reactor Projects
-stachment:
As stated 0FFICIAL RECORD COPY NINE MILE POINT CCMMENT 0001.0.0 5555.M o
"O/
l Niagara Mohawk Porer' Corporation 2
cc w/
Attachment:
C. Mangan, Senior Vice President W. Hansen, Manager, Corporate Quality Assurance C. Beckham, Manager, Nuclear Quality Asst.rance Operati J. Perry. Vice President, Quality Assurance ons J. Willis General Station Superintendent C. Terry,,Vice President, Nuclear Engineering and Licen i K. Dahlberg, Unit i Station Superintencent s ng R. Smith, Unit 2 Superintendent, OperationsR. Ranca R. Abbott, Unit 2 Station Superintendent G. Wilson, Senior Attorney T. Conner, Jr., Esquire J. Keib, Esquire J. Warden, New York Consumer Protection Branch State of New York, Department of LawOf rector, Power Divisio
, State of New York Public Document Room (POR) local Public Document Room (LPOR)
Nuclear Safety-Information Center (NSIC)
NRC Resident inspector State of New York bec w/
Attachment:
Region 1 Occket Room (with concurrences)
J. Wiggins, CRP G. Meyer, ORP
- 0. Limroth, DRP R. Barkley, DRP S. Horwitz, PA0 M. Miller. SLO W. Cook, SRI - Nine Mile R. Temos,.RI - Nine Mile R. Laura,-RI - Nine Mile J. Oyer, E00 R. Capra, NRR M. Slosson, NRR R. Martin, NRR l
l
p_
b-
.e AT'aCHvENT Public Comments en tne Restart A: tion Plan 1.
Cor.ae n t : The financial viability of Niagara Mohawk Power Corporation should be evaluated as part of the Restart Action Plan (RAP),
Response
The financial viability of Niagara Mohawk is evaluated by NRC with resoect to the acility of Niagara Mohawk the-to safely operate Nine Mile Point.
During the public Commission - status -
briefing on August 2, 1939, the C0mmission questioned Niagara Mohawk with regard to its financial viability and its ability to safely operate Nine Mile Point.
Niagara Monawk assured the
{
Commission that sufficient funds are available to safely operate the units.
In summary, the Commission will Continue to monitor Niagara Mohawk's activities to ensure that financial concerns do not interfere with the safe operation of the plants, but dis-agrees that the issue requires inclusion.in the RAP.
2.
Comment:
The torus should be repaired prior to restart, Response: Both the Niagara Mohawk Power Corporation (NMPC) ard the NRC have known about the corrosion of the wall of the t0rus at Nine Mile Point Unit 1.
The tninning of the torus all wi l.1 ::e addressed in the resolution of RAP Specific Issue No. 7 c"i:r to restart.
NMPC nill be required to perform repairs to tre torus prior to restart if the torus wall nas corroced beyonc the min-
-imum requirements of the American Society of.Vechanical Engi-neers (ASME) Code governing the adequacy of this vessel.
this issue is alreacy adcressed in the RAP.
- Thus, 3.
Comment: NDE surveillance of the torus should be conducted by people other than NMPC that do not have a vested interest in keeping the plant running.
Response
The NDE surveillance of the. torus wall at Nine Mile Point Unit 1 is presently conducted by a contractor for Niagara Monawk as well as by Niagara Mohawk employees.
The results of the NDE examinations are independently reviewed by the Niagara Mohawk Quality Assurance Department and by the NRC (upon the submittal for the resolution of, RAP Specific Issue No. 7 ) ',
In addition.
Niagara Mohawk's Inservice Inspection (ISI) program, particu-larly their NDE methods, was reviewed in detail by NRC Region I during Inspection 50-220/88-31 and determined to ce acceptacle.
o i
5:
t 1
f
,;J Attachment
.Public Comments on the 2
Restart Action Plan This inspection included independent wall thickness measurements performed by the NRC.
Finally,- as an alternative confirmation of the structural strengtn of the torus wall, Niagara Mohawk will be performing an Integrated Leak Rate Test (ILRT) of the containment prior to the restart of tne plant to satisfy; the requirements of 10 CFR 50 Appenoix J.
In summary, based on this degree of oversight of the NDE sur-veillance ~ on the torus wall, Niagara Mohawk's present NDE sur-veillance program on the torus is structured in accordance with all applicable NRC requirements and the NRC does not agree that reviews by additional parties are warranted, 4
Comnient:
Long-term management improvement plans should be implemented and goals achieved before restart of Unit 1.
Response: Niagara Mohawk has formulated a. Nuclear Improvement Program (NIP) to improve their overall level of performance both prior to and following the restart of Nine Mile Point Unit 1.
The N1P embodies a number of plans for long-term improvements'in manage-ment and worker performance and goals to measure the success of that program. Niagara Monawk has determined, and the NRC has agreed that these improvements are important but are not essen-tial for the safe operation of the facility following restart, provided that the management issues in the RAP - are resolved prior to restart. Thus, the issue is properly addressed NIP.
in tne 5.
Comments: RAP Section 2, page for improving management performance is to identify traini development programs for manager..suoervisor and employee intra-personal and. management skills, etc.
mented prior to restart?
Shouldn't-these be imple-Response: Above response to Comment No. 4 applies 6.
Comment:
NMPC should prove that the RAP works before restart ized.
is author-Response: The NRC's approach toward approving restart of Nine Mile Point Unit I has been structured in the following manner:
1.
Niagara Mohawk developed and implemented a Restart Action Plan (RAP).
ii. The NRC reviewed the RAP for approval.
1 Attachment - Public Comments on the 3
Restart Action Plan iii. Niagara Mohawk conducted a self-assessment of their readi-ness for restart.
iv.
NRC will review the qualfty and conclusions of the seif-assessment.
v.
NRC will conduct an Integrated Assessment Team Inspection of the. Nine Mile Point 1 organization and its ability to safely operate the facility, vi.
A decision will be made by the Region I Regional Adminis-trator regarding restart.
'{
The NRC believes that the process, as outlined, is sufficient to determine. the ef fectiveness of the RAP and the ability of Niagara Mohawk to safely operate the facility.and is consistent with the intent of this public comment.
This process has been followed successfully at other problem plants.
7.
Comment:
Nine Mile Point is unsafe based on the recent disclosures of the magnitude of the waste spill at Unit 1 plus the close to 50
.j percent failure rate of the operators at Unit 2.
l
Response
5 The waste spill in the radwaste - storage building at NMP' Unit 1 was investigated by an Augmented Inspection Team (AIT) from the NRC in August, 1989. The results of that team inspection indi-cate that the spill, while an example of ' poor operating prac-L tice, was never a ha:ard to the public health and safety and was l
properly. surveyed and controlled ' by the NMPC health physics department to ensure that the spill did not pose a threat to plant workers.
NMPC has also included the cleanup activity in i
the NIP.
The specific management problems that led to this.
condition are addressed in RAP Items 1-5.
.i The high failure rate of the licensed reactor operators at Unit l
2 during the recent requalification. examination was attributed to weaknesses in the requalification training program.
4
- However, sufficient operators had successfully passed an NRC-administered i
requalification examination to allow the plant to continue to operate with shift crews augmented by extra personnel to compen-I sate for the deficiencies noted until~ remediation could be com-pleted.
NMPC has also implemented a remedial training program to improve requalification training in the deficient areas The examination results were indicative of significant noted.
t
~ - - - ~ - -
Attachment - Public Comments on the 4
Restart Action Plan-weaknesses in the requalification program which required correc-tion but not of the scope or depth that would pose significant public health and safety concerns.
The adequacy of the operator -
training program at Unit I will be thoroughly reviewed-prior to-restart to resolve RAP Specific Issue Nos. 2 and 3.
In summary, the NRC disagrees with this comment.
8.
Comment: NRC reporting requirements / problems should be defined in the
- RAP, Response: NMPC does not have a chronic history of failing to make NRC required notifications and reports.
Based on past history, this issue is not required in the RAP as a specific technical issue requiring resolution.
NRC reporting requirements are presently outlined in 10 CFR 50.72 and 50.73 as well as the NMP Unit 1 TS.
9.
Comment:
There should be a public hearing / evidentiary proceeding prior to restart.
Response:. As specified in the Code of Federal Regulations -Title 10 Part 2.206, "Any person may file a request to institute a proceeding pursuant
- .o 52.02 to modify, suspend, or revoke a -license, or for such other action as may be proper."
Requests made in accordance with 10 CFR 2.206 will be reviewed and evaluated by the NRC.
Typically, a proceeding would not be held prior to the restart of a unit such as Nine Mile Point Unit 1, because the evaluation of the r? start of the unit is _ not a change to the operating license.
Thus, the NRC disagrees with this comment.
10, C_omment: Provide. the details of the - original problems leading to the shutdown, including the delays in that shutdown.and all subse-quent problems prolonging.that shutdown to the LPOR.
- Also, ensure that the,last several SALPs are in the Local Public Document Document Room (LPOR).
Response: The LPOR routinely receives copies of the pubitely distributed NRC correspondence, including SALPs.
The NRC will review the availability of the public documents relative to the shutdown in the LPOR and ensure the appropriate ones exist there.
A At'tachment - Public Comments on the S
Restart Action Plan
- 11. Comment:
The RAP should clearly state what verifications will be done in the presence of NRC insoectars.
Respog : As a condition of the license for Nine Mile Point Unit 1,
Niagara Mohawk is suoject to unannounced NRC inspections to determine its compliance with the assess its ability to safely operate the facility. terms of the license and to NRC has not found it necessary in this case that certain However, the licensed activities be conducted in the presence of NRC inspec-tors. Although the NRC will not require NMPC to conduct any of the activities outlined in the RAP in the presence of NRC inspectors, we will review those actions deemed necessary issure that the concerns addressed by the RAP are adequately to resolved.
- 12. Comment:
With reference to RAP Specific Issue No. 2, why didn't Niagara Mohawk establish responsibility and accountability for the main-tenance of operator licenses 20 years ago?
Response: Establishing responsibility and accountability for the mainten-ance of a program is a fundamental principle of management that The reasons for NMPC'sshould have been implemented since the inc failure to implement these management principles in the past in this area is probably attributable to deficiencies in NMPC manegement in' the Operations area, the extensive number of changes tnat have occurred in the operator training area since the TMI accicent, and poor cooperation between the Operations and Training Departments.
The RAP and
-CAls 88-13 and 38-17 were generatec precisely because of: oves-tions like these by the NRC and to resolve these types of management deficiencies.
- 13. Comment:
The spent fuel pool should be fixed prior to restart.
Response: The spent fuel pool has experienced minor leakage in recent months due to an apparent perforation in the stainless steel pool liner at a location yet to be identified.
NMPC has pro-posed a plan of action to identify and resolve this, problem sub-sequent to restart. Given the size of the-leak which has been observed, as well as the design -of the spent fuel pool, no threat to the puolic health and safety exists by delaying the repairs ~ to the pool liner until after restart. Thus, the NRC disagrees with this comment and feels that the issue is properly addressed in the RAP (Specific Issue 15) as an issue which can be resolved after restart.
a 1
{
f
. l 1
+
t
-Attachment - Public Comments on the 6
e
'g
. Restart Action Plan i
n.-
j 14.
Comment:
The radweste spill should be cleaned up prior to restart, r
i
Response
NMPC has in place a plan for the decontamination and cleanup of the radwaste storage room spill at Unit 1.
The plan involves the use of a robotic arm to remotely decontaminate the room.
Given that the spill is confined to a small, abandoned area of the plant and that there is no indication that the radioactive contamination in the room is leaking to the environment, no threat to the public or worker health and safety exists.
NMPC-is scheduled to complete the decontamination of the room by x March,1990. Therefore, the NRC disagrees with this comment and does not consider the decontamination of the room a restart x
issue. The issue will be properly addressed in the NIP.
- 15. -Comment:
A determination should be made as to how much pressure the con-tainment'a(NineMilePointUnit1canwithstand.
Response: The Nine Mile Point I containment system consists of an upper x
section called 'the drywell and a lower portion called the sup-pression chamber. Mer the Unit 1 Final Safety Analysis Report, the drywell is designed to withstand a peak 62 psig-internal
)
The suppression chamber is designed to. withstand a pressure.
peak 35 psig internal pressure.
Prior to restart, the leak i
tightness and structural integrity of the containment will be tested during a containmentlntegrated leak-rate test.
Any further analyses are not considered a restart issue.
L
- 16. Comment:
i Was' NMP Unit I considered a safe plant by the NRC in December 1987 and why, all' of the sudda1, 'ci d the NRC ask Niagara Mohawk's management to come up the a plan to resolve their 1-problems?
1 Response: Yes. Nine Mile Point was considered to be a plant which met tne conditions of its license in December 1987. ' Had W NRC believed at anytime prior to December 1987 that the operation posed a threat to public health and safety, the Unit'would have been immediately ordered to shutdown. Following the shutdown of the Unit in December 1987 due to technical problems, several other problems were identified in the areas of operator requal-ificat' ion training, control of commercial grade parts, fire bar-rier penetrations, and the operator's understanding and use-of l
emergency operating procedures. As a result, on July 24, 1988, tne NRC issued a Confirmatory Action Letter wh!ch documented Niagara Mohawk's commitment not to restart Unit 1 until correc-tive actions have been completed and the agreement of the NRC's Region I Regional Administrator was obtained.
The actions
! 4" required include a root cause assessment of why management has not been effective in recognizing and remedying problems, pre-paration of a restart action plan, and submission of a written report relative to reaainess for~ restart.
.. - --.---------.- --- n----
_ ---_ __ _ -_ _ - -- _--.- ---__.- ---. - - ~-- --.,,-
p o
]
s :
' Attachment - Public Comments on the 7
Restart Action Pla1 17, Comment:
.Why did the NRC raise a sa#ety concern regarding the scram dis-charge volume on Jure 24, ;933, but take until December 1937 to take action to resolve the issue?
Response: During a rout'ne shutdown of Browns Ferry Unit No. 3 on June 28,1980,16 of 185 control rods failed to fully insert in response to a manual scram f rom approximately - 30*.' power. All rods _were subsequently inserted within 15 minutes and no reactor damage or hazard to the public occurred.
Following an in-depth review of Boiling Water Reac;or Control Systems, short and long-1 term corrective mea:ures were identified.
Short-term corrective measures were imolemented by IE Bulletin 30-17 and an Order con-cerning these measures was issued to Nine Mile Point l_.on January 9, 1981. and modifiec Ma th 31, 1981.
A Con firma to ry Order, dated June 24, 1983, was issued to Niagara Mohawk con-cerning =long-term coreective meast.res.
Due to inspection pri-orities, the NRC staff did not inspect the scram discharge volume design for Nine Mile point Unit 1 to determine compliance with the June 24, 1983 Orcer until November 1987. As-a result of *.be inspection, two areas of deviation from the Order and the Generic Safety Evaluation, dated December 1,1980,. for the scram discharge volume were identified, These deviations had been identified by Niagara Mohawk in a January 30, 1981 letter, but 1
NMPC failed to obtain prior NRC approval for the deviations as requirec ::y the Order.
i The NRC feels this issue is properly addressed in the RAP.
letter cated October 12, 1988, the NRC staff transmitted t '. s 33 i
safety evaluation with respect to this issue and concluded that operation with the systen, for another fuel cycle o95es no undue risk to the public.
The ' staf f has evaluated Niagara-Mohawk's proposed testing program for the scram discharge. volume and determined it to be acceptable.
- 18. Comment: Niagara Monawr, is not competent to uperate atomic-power plants based upon NRC findings / fines over the last six years.
Resoonse: A purpose of the RAP and cal. 88-17 was to assure improvements in Niagara Mohawk's management of its nuclear f acilities.
While their performar.ce in the last several years has ra'isect NRC con-cerns, the NRC staf f believes that Niagara Mohawk is capable of improving.its Operation, and that a properly sccDed management
'3 improvement plan that is effectively implemented 13 an appro-priate means by which this can be accomplished.
The NRC staff must conclude t.ha t the necessary improvements have been made prior to. restart of Unit 1.
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1 Attathment - Public Comments on the Re,ttart Action Plan 3
- 19. Cp ment: Safety concerns regarcing cracking in the core spray spargers.
)
h,sponse: In accordance with examined the Unit ! oreIE Eulletin M-13. Niagara Mohawk visually scray soargers and associated piping curing the 1931 refueling outage.
As documented in their May 13. 1981 letter to the NRC, two cracks in one location were l
identified.
The cracks were evaluatec and insignificant.
cetermined to be No corrective actions were re0uired.
continued to examine NMPC has the sparger each refueling outage since 1981.
The most recent examination, ccmpleted during the cur ent
- outage, incicated that the crack length remains within the tolerance of the original crack length.
Therefore, no correc-tive actions were required.
The licensee will continue to inspect the core spray scargers and associated internal piping during future refueling outages.
Thus, this issue does not impact restart or the rap.
20.
Comment:
Will f ailed plant equipment te replaced with like-in-kind parts or by components that may be untested?
are not part of the original design and Resconse: All safety-related components which NMPC replaces due to failure or ss part of a preventive maintenance program are purchased either as a sa fe ty-rela ted c:mponent (fabricated and tested under an approved ovality assurance program) er as a co mercial grade product later subject to a quality assurance oregrn cesigned to qualify tne component for safety-related a:clica-tions.
Given the age of the f acility and tne declining rutter of supp1 1ers of safety related components in :nis country, hMPC may not necessarily replace failed components w th identical i
replacements specified oy the original design.
Instead, alter-native con'ponents may be used which are of cif ferent design, but have been demonstrated to be suitable for safety related appli-cations ano arn capable of performing the function of the orig-inal part.
This practice is commonplace in the nuclear industry and, if proporly administered, is acceptable to the NRC.
this issue is not a problem and is inappropriate for conclusionThus, in the RAP.
21.
Comment: NMP Unit I should not be restarted because it will acd to the radweste problen.'.
What will haDpen to the radwaste generated?
Resoonse: The low level racioactive waste that is generated by Nine Mile Point Unit I wii; be shipped to one of the tnree currently licensed burial sitas in the Unitka States.
However, it is the ultimate responsibility of New York State to find and develop an alternative disposal site in the near future under the provis-ions of the Low Level Waste policy Amenements Act of 1985. All
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Att4chment - Public Comments on the L
Restart Action Plan 9
high level radioactive wastes will be stored onsite, either in the spent fuel cool or if later :enstructed storage facility, until the federal governme,ntin a future onsite level develops a high-Act of 1982, waste repositery as required by the Nuclear Vaste Policy At present, there is no orchibition against the operation of any i
nuclear power plant i
the ultimate resolutionin this country due to questions regarding of the problem of disposing of high level or low level radioactive wastes.
Therefore, the NRC dis-t agrees that this issue should be included be a basis for not restarting Unit 1.
in the RAP and should 22.
Comment:
The RAP does eot reovire NMPC to concuct health studie determine the long-term health effects of Nine Mile Point. These studies should be completed prior to re s ta rt of NMP Unit 1.
Response: As stipulated by the terms of its license, NMPC has been recuired to conduct environmental monitoring of the area around the plant site since prior to the licensing of Unit I to deter-mine if there has been any accumulation of radioactive material in the environment or excessive radioactive emissions from the plant.
There is no incication from the NMPC environmental mon-itoring program that any substantial potential danger to the public from normal plant emissions ex'ists, particularly at levels that would suggest that a health study in the area is warranted.
Based upon the results of the environmental monitor-ing program to date, the NRC does not consider this activity appropriate for the Restart Action Plan.
23.
Comment:
The permanent solution to the radwaste disposal problem should be included as an item in the RAP.
Response: The response to question 21 applies.
24 Comment: Cracks in *M i concrete walls in various parts of the plant should be fixed prior to restart.
NMPC does not have a good handle on. the wall cracking. Will the plant survive a seismic event due to all the building cracks?
Response
NMPC has observed cracks in the masonary walls of several build-ings in the plant over. time.
The location of these cracks, as well as NMPC's plan of action to monitor and analyze the signif-icance of these cracks, is documented under Specific Issue No.
15 of the RAP.
The cause of the cracks in the walls appeared to be predominately due to either shrinkage during curing or tensile stress experienced due to temperature fluctuations e.
o
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Attachsent o Polic Comments on the 10 Restart Action Plan 4
h (concrete by nature has little structural
[its structural strength in tension is suppliec by the retn-strength in tension forcing rods in the Concrete) and thus has a tendency to Crack in sension and during curing).
analysis and recatr oroject on all masonary walls at Unit 1NMi correct all of the cracks noted in the plant walls.
The,e repairs ensured that the masonary walls meet their original seismic cesign criteria.
Therefore, the NRC does not consider this issue a safety problem and resolution of the issue is properly addressed in the rap.
- 25. Comment:
What is the status of the $PDS system at NMP Unit !?
Response: The Safety Parameter Disolay System was ceclared fully ope tional at Nine Mlle Point Unit 1 in June 1986 and is theref not a restart issue.
- 26. Comment:
New York State should police NMP Unit I activities.
ReJponse: By the Atomic Energy Act of 1954, as amended, and the i
t Reorganization Act of 1974 as amended, the United State Congre gave the NRC statutory authority over the regulation of nuclear factitties.
As such it is the NRC's responsibility to:
l
- 1) license the construc, tion and operation of nuclear reactors' and other nuclear facilttles, 2) license the possession, processing, handling me, and disposal of nuclear material.
- 3) develop and implement rules and regulations that ;;veM licensed nuclear activities, inspect liiensed facilities and activities, investigate nuclear incidents and allegations - con-cerning any matter regulated by the NRC, 4) conduct public hear-ings on mitters of nuclear _ and radiological safaty, environ-i mental concern, common defense and security and antitrust laws and, 5) develop effective working relationshios with the states regarding regulation of nuclear material.
Further. -to ensure adequate communication and cooperation between the NRC and states, the State Liaison Of ficer program was established the.
1976.
If New York State 50 desires, the NRC will consider statein proposals to enter into instruments of cooperation for state 1
participation in NRC inspection activities if the state's pro-gram has provisions to ensure close cooperation with the NRC.
In summary, New York State is welcome to enter into an acreement with the NRC to provide an oversignt role, but the NRC has ore-emptive federal authority in the licensing of nuclear plants.
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Attachneet - Public Comments on the 11 Restart Action Plan
- 27. Comment: The public has the perception that there isn't the slightest chance that the plant will stay closed.
Response: The Nine Mile Point Unit 1 facility was licensed under the terms of 10 CFR 50, which derives its statutory basis from the Atomic Energy Act of 1954, as amended.
Since the facility remains licensed under Part 50, the emphasis of NMPC to date has been to correct the identified management deficiencies and to attempt to restore the plant to service. As a result, the NRC's actions to date have been oriented teward determining whether the facility can be safely returned to service and is based on NMPC having corrected the management and technical deficiencies identified.
That is the purpose of our review of the Restart Action Plan.
If NMPC can not correct the deficiencies noted, then the facil-ity will remain closed.
- 28. Comment:
Monthly public meetings should be neld before restart to discuss the concerns of area citizens.
Response: The public meeting held on August 23, 1989, was an initiative on the part of the NRC to involve the public in the restart process at NMP Unit I and to gain their comments.
Additional meetings of this type would result only if there are fundamental changes in the NMPC Restart Action Plan.
NRC directly to raise safety concerns.The public is free to contact 29.
Comment: Don't restart Unit 1.
i Response: This comment was received from at least thirteen individuals during the public meeting. The response to question 27 applies.
l l
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Att'achneet - Public Comments on the 12 Restart Action Plan Public Comments en !ssues Not Related to the rap
- 30. Comment: Spent fuel pool storage / acceptability of dry cask storage.
Response: The NRC is proposing to amend its regulations to provide, as directed by the Nuclear Waste Policy Act of 1982, for the stor-age of spent fuel at the sites of power reactors, to the maximum extent practicable, without the need for additional site-specific approvals. Holders of power reactor operating licenses would be permitted to store spent fuel in casks approved by NRC under a general license.
The proposed rule contains criteria for obtaining an NRC Certificate of Compliance for sperit fuel storage casks.
The notice of proposed rule making, published in the Federal Registe en May 5,1989, solicited public com-ments by ane 19, 1989.
The NRC is currently evaluating the public comments.
In any event, this issue does not impact the restart of the facility.
- 31. Comment: The public doesn't trust Niagara Mohawk.
Response: This comment was made by four members of the public at the pub-lic meeting and represents an issue for NMPC to address.
This issue is not applicable to restart.
- 32. Comment: The public doesn't trust the NRC Response: This comment was made by three members of the public at the pub-lic meeting and is not related to the Restart Action Plan.
- 33. Comment: Why is Unit 2 allowed to operate with the same management as Unit 17 Response: While NMP Unit 2 has been categorized by NRC as a plant requir-ing close NRC scrutiny, the staff and management of Unit 2, which is separate for each unit below the site superintendent level, have performed better than the staff and management at Unit 1.
Most notably, the operators at Unit 2 have clearly dis-played a much more positive and responsive attitude than Unit 1 operators. In addition, management has displayed the ability to operate Unit 2 effectively in spite of the fact that it is significantly more complex than Unit 1 and has a staff with sig-nificantly less -operational experience with the f acility than Unit 1.
Thus, while Nine Mile Point Unit 2 remains under close NRC scrutiny, NRC senior management has determined that NMPC can safely operate Unit 2 in spite of the noted deficiencies at Unit 1.
o Attachment
- Public Comments on the 13 Restart Action Plan 34 Coment: Does the radwaste builc*ng spill 'epresent a threat to the pvblic? Why wa sn' t the NRC cou fi+c of the spill? Why cidn't the \\RC find this proolem?
3esconse: The results of the NRC A.,g ented hspecti:n T+am (AIT) conducted at Nine Mile Point indicate that the racwaste spill does not pose a threat to the puolic health and safety.
The NRC inspectors failed to icentify the spill previously because the problem occurred in the sub-basement of an abandoned The room was procerly designatedportion of the radwaste build and marked as a locked hign radiation area.
The NRC inspection program does not require all locked high radiation areas in the plant to be examined nally by the NRC inspectors due to the inter-unnecessary radiation exposure which would occur to the inspector.
Thus, the f act that the NRC did not know of the existence of this radioactivity contaminated the fact that Niagara Mohawk cid notroom was attributable largely to its loca report the incident versus a f ailure to conduct a requitec portion of the NRC inspection program.
No RAP actions as a result of this incicent are necessary.
- 35. Coment:
It is alleged that radiation process monitors are turned af f when radioactive cischarges to the environment are made :/ N" M.
Response
This allegation has been entered into the NRC allegation traci-ing system for follow up and will oe resolved prior to restart.
- 36. Coement:
~
Who is responsible for ensuring that Nine Mile point and Fit:-
Patrick have acequate emergency plans?
Response
It is the responsibility of '.he NRC to ensure that the licensees of Nine Mile Point and FittPatrick have adequate emergency plans for the staff and employees onsite and the proper notification of authorities offsite.
Verification of the adequacy of offsite planning is the responsibility of the Federal Emerg-emergency ency Management Agency (:EMA) acting as an agent for the NRC.
This delegation of responsibility was established by Presiden-tial Otrective.
- 37. Coment: The background radiation levels in an area near the plant are elevated above normal.
Response
The NRC maintains a incependent radiation monitoring program in the area around the plant.
The results of this raciation mon-itoring program are published quarterly in NUREG - 0837.
Revier of the results of that monitoring program do not incicate that there are elevated background radiation levels in any area around the plant.
)
l Attachment _Public Comments on the 14 Restart Action Plan
- 38. Comment:
Does the NRC fee schedule cose a potential ecnflict of interest?
Response: The annual funding of the NRC is estaolished and provided by Congress.
The funds received througn the fe+ schedule are deposited oirectly to the Federal Treasury.
While the NRC fee schedule does recover a significant portion of costs, the schedule is not related to the NRC's budget and doesits operating not present a conflict with the agency's role as a regulator of the utility.
- 39. Comment:
Allowing radioactive waste to be categorized as being below a
[
level of regulatory concern, and thus cacable of being disposed of as normal trash, constitutes a public health and safety l'
concern.
Re se: The NRC was mandated by the Low Level Radioactive Waste Policy Amendments Act of 1985 to establish a regulatory limit for radioactive waste which does not pose a threat to the public health and safety and thus can be disposed of by normal methods.
'The NRC has yet to establish such a regulatory limit.
When that radioactivity level is decided upon, it will be the subject of extensive review and rulemaking to determine that the limit does not pose a threat to the public health and safety.
10.
Comments: What has been cone regarding tne concerns ra i ::d
- y Ocuglas Ellison?
information regarding Niagara Mohawk?bny did the NRC oay Mr. Eilison 5 Response: The allegations initially raised by Couglas Ellison viewed by a special HRC team in inscection 50-220/36-17. As a were re-result of new allegations, two accitional inspections were recently conducted at Nine Mile Point to resolve the concerns raised.
The results of those two inspections, one regarding the technical issues and one regarding allegations of potential employee harassment and intimidation, are documented in Inspec-tion Reports 50-220/89-16 and 89-21, resoectively. While some of the concerns were partially substantiated by team, no issues which af fect the inspection identified.
the public health and safety were The matter of Mr. Ellison being paid by the NRC to provide information regarding Niagara Mohawk is the subject of in on-going investigation and thus can not be discussed in cetail at this time.
Completion of this investigation does not impact Restart Action Plan or restart of the facility, the er 1
m c
c:
l Attachment - Public Comments on the Restart Action Plan 15
- 41. Comment:
Allegation concerning an unknown NRC employee / individual im sonating an NRC employee,
[
Response: The NoC can not take any further actions with regard to indiv10ual described due to the lack of in f orma tion provided.
Although the individual stated that he was an NRC enioloyee, it is not clear that the individual was in fact employed by the NRC. However, vidual from the person making this statement,from th the individual's actions would have been totally unacceptable behavior for an NRC employee.
- 42. Comment:
I'm opposed to the operation of any nuclear power plant in this country.
Response
Congress has decreed through the Atomic Energy Act of 1954 tha atomic energy is beneficial to this country relative to the inherent risks that the energy source poses.
Thus, existing atomic cswer plants are permitted to operate in and new power stations may be constructed and licensed to oper-thi ate.
It is the NRC's responsibility from this energy source are minimized. to enture that the risks
- 43. C o.~,e n t :
(Receivect in Writing)
The audience at the public neting was not representative of the public and is biased against the
- plant, Resconse: The purpose of the public meeting was to receive comrnents on the rap. both positive and negative.
The meeting was not intenced to provide a forum to determine the level of public support for or against the restart of Unit 1.
- 44. Comment:
(Received 1.1 writing) - Do oil and coal-fired power plants that spew out the makings of acid they start up?
in have to hold hearings before Response: The NRC is not other than those powered by nuclearresponsible for the regulation of po energy.
Those facilities Environmental Protection Agency, Occupational S Administration, etc.) and may be subject to licensing actions, up to and including a public hearing, in the event that tney do not comply with tne agencies' regulations or if requested by another party in accorcance with agency.
the procedural rules of tne
T Attachment - Dublic Comments on the 16 Restart Action Plan 45.
Comment: (Received in writing) 1 have hearc that there is a " collar" around the reactor at Nine Mile point One that 15 so corredec that if the plant was turned on f all power. it would blow up.
Response: After discussions between several members of the NRC sta determine what the individual could have meant by the " collar" around the plant, it was determined tnat been referring to the torus.
The written resconse provided to tne person must have
" blow up" regardless of the operating status of NRC assured the individual that NMPC is taking actions to The resolve the torus corrosion probler' and that the NRC will y
closely follow their actions to cetermine that the problem does not pose a threat to public health and safety.
Further, the individual was informed that Niagara Mohawk will be conducting a pressurization test of the torus prior to restart of the facil-ity to ensure that it will withstand the maximum postulated pressure generated during a design basis reactor accident. Based these facts, the NRC feels that the individual's concerns upon are u'isubstantiated.
- 46. Comment:
(Received in writing) - What was the source (s) of the elevated cesium and strontium concentrations noted in local milk supplies during the late 1970's?
Resconse: The NRC conducted a review of the elevated Cs levels noted in milk samples in the area near the plant in Dil The review was conducted in response to concerns regarcing this issue which were raised by the Sierra Clvo.
The results of this review indicated that the average levels c/ Cs-137 in milk near the site were not consistently higher tnan tre rest of tre State.
TM NRC's assessment at that time was that the source Cs-137 anu.-131 concentrations in milk in the area could n precisely cetermined.
The source of the contamination could have been attributable to either reactor effluents or f allout from weapons tests, most particularly the Chinese atmospheric
- eapons tests in the late 1970's. Regardless of the source, the observed radiation levels constituted only a small the radiation dose received fraction of That small dose would also have been belowfrom natural background regulatory limits if the assumption was made that all the observed racio-even activity came from. effluents from Nine Mile Point and FitzPatrick.
51 o
Attachment- Public Comments on the
' Restart Action Plan 17
(
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- 47. Comment: (Received in writing)
Why is there a higher incidence of learning disabilities in area chilcren ::o rn ::uring the lata 1970's?
Resoonse: The NRC has no knowledge of any studies which sh incidence of than normal at any time in history. learning disabilities in a is higher into the effect of radiation on the incidence of m tion in children, no measurable ef fect on the incidence of learning disabilities in area children should be observable until facilities were atradiation doses to the public from the operation nuclea least limits.
enree er::ers of magnituce above NRC
- 48. Coment: (Received in writing) - A detailed bottom sediments or arta wetlands should be conducted by NM determine room flooding (<ent.if sadiments and biota aere af fected by the radw Response: As mentioned in the response to earlier comments, NMP required to conduct an environmental sampling program as a requirement of their license.
That sampling program involves i
collecting samples of sediments, biota, and fish from Lake Ontario as well as the surrounding land area.
Tne results of the sampling program confirm that NMPC is operating the fa.:ili i
in accordance with the ':RC's radioactive waste releas codified in 10 CFR 20.
49.
Cn.eent:
(Unrelated to Nine Mile Point Unit 1 Restart) - Don't si level waste repository in the area.
Resoonse: The siting of a low level waste repository is currently the responsibility of the State of New York.
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- GRN CRC NO: 90-0128 Exccutive Director DESC:
ROUTING:
i ENCLOSES LETTER FROM. TOM WALSH, RETIRE NINE HILE Russell, RI s
ONE RE REFUELING OF THE NINE MILE ONE AND THE SAFETY OF THE MARK I CONTAINMENT DESIGN DATE: 02/12/90 ASSIGNED TO:
CONTACT:
NRR Murley i
l -- SPECIAL INSTRUCTIONS OR REMARKS:
NRR RECEIVED: FEB. 12, 1990 ACTION:
DRP_R;YARGAt NRR ROUTING:
MURLEY/SNIEZEK PARTLOW MIRAGLIA CRUTCHFIELD GILLESPIE MOSSBURG f
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PAPER' NUMBER:
CRC-90-0128 LOGGING DATE: Feb 9 90 i
ACTION OFFICE:
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AUTHOR:
Frank Horton--Const Ref AFFILIATION:
U.S. HOUSE OF REPRESENTATIVES LETTER DATE:
Feb 6 90 FILE CODE: ID&R-5 Nine Mile 1
SUBJECT:
Refueling of the Nine Mile One and the safety of f
the Mark I containment design ACTION:
Direct Reply DISTRIBUTION:
OCA to Ack, DSB SPECIAL HANDLING: None NOTES:
Tom Walsh DATE DUE:
Feb 27 90 SIGNATURE:
DATE SIGNED:
AFFILIATION:
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coalition of citizens concerned about the safety of the Nine Mile One Nuclear Facility
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'**g-"fg,h Janua r y 8, 1989
%'"*e73 Congressman Frank Horton co.,n i 2108 Rayburn Building
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Washington, D.C.
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t.e ~.e. w Dear Cong ressman Horton RETIRE NINE MILE ONE has concerns about the safety of the Nine Mile One Nuclear Facilty that we hope you will give your immed ia t e attention.
Nine Mile One was constructed with a Mark I containment designed by General Electric. Since 1975, General Electric has reported that there is a 90% chance the Mar k I containment will fall in the event of a core accident. The Nuclear Regulatory Commission has described the Mar k I containments as " virtually certain" to fail in the event of a core accident. Nine Mile One is also showing serious signs of deterioration due to age, particularly in the torrus, that makes its safe operatica a considerable risk.
The deteriorated condition of the f acil ity coupled with a containment that is " virtually certain" to fall make Nine Mile One vulnerable to a Chernobyl type of accident that would result in considerable loss of life and force the abandonment of hundreds of square miles in Central Net. York. When these concequences are considered, di scussions of the probablity of such an event occur ring seem moot.
We are asking that you respond in the following wayst
- 1. We are requesting that you contact the Nuclear Reg ulator y
- s. opp ed.
Commission and ask that the ref ue)ing of the plant be t
The NRC has.not given Niagara Mohawk permission to restart Nine Mile One. Certainly the ref ueling of the plant prior to completion of the NRC's assessment of its worthiness to operate is prcuature.
P.O. Box $63. University Station - Syracuse, New York 13210 v
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2.
We are requesting that you sponsor a public hearing in your district on this urgent matter. There is a need for the citizens of the district to be pr ovided a forum in which to ex press their concerns about the f acility. A hearing would also l
provide an opportunity to create a record of expert testimony on this matter.
- 3. We would like to request that you make yourself available for a briefing by the Union of Concerned Scientists who have monitored the Mark One facilities throughout the countr y and whose thorough and professional research brought many of us to the opinion we now hold about Nine Mile One.
We thank you for your attention to this urgent matter and look forward to your response.
Sincerely, RETIRE NINE MILE ONE
~
by Tom Wal sh, co-che i r.
(315) 446-0435 l
_