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NUCLEAR REGULATORY c0MMisslOfs fj wassiscTON. D. C. 20555
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SEP 2 5 1984 b
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MEMnRANDUM FOR:
J. Davis, Director, NMSS R. DeYoung, Director, IE b v, g H. Denton, Director, NRR f
L3 G. Cunningham, Director. ELD i
r J. Felton, Director, DRR, ADM L'
J. Fouchard, Director, OPA Regional Administrators FROM:
Frank P. Gillespie, Director Division of Risk Analysis and Operations Office of Nuclear Regulatory Research
SUBJECT:
PROPOSED AMENDMENTS TO 10 CFR PART 50, APPENDIX E; CONSIDERATION OF EARTHQUAKES IN THE CONTEXT OF EMERGENCY PREPAREDNESS Please review the subject rulemaking package and provide me with your coments and concurrence by C.O.B. September 28, 1984 This package is also being sent to the Regional Administrators for infonnation and possible coment.
1.
Title:
Proposed Amendments to 10 CFR Part 50, Section 50.47, and Appendix E.
Consideration of Earthquakes in the Context of Emergency Preparedness.
2.
Task Leader:
Mike Jamgochian, DRA0, RES 3.
Task No:
RA-401-1 4.
Cognizant Individuals:
Bill Shields, ELD i
Frank Pagano, IE Leon Reiter, NRR 5.
Background:
On August 10, 198'4, the Comission ordered (CLI-84-12) that a rulemaking proceeding be initiated to consider requiring the explicit consideration of severe earthquakes in emergency response l
planning.
In response to that order we have established a task l
group which developed the enclosed Comiss on Paper.
1 i
s I
rank P. Gillespie, Director i
Division of Risk Analysis and Operations Office of Nuclear Regulatory Research
Enclosure:
Prcposed Rulemaking Package l
BELLB5-653 PDR 1
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Fori The Comissioners 1
From:
William-J. Dircks Executive Director for Operations
Subject:
PROPOSED AMENDMENTS TO 10 CFR PART 50, SECTION 50.47 AND
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APPENDIX E, CONSIDERATION OF EARTHQUAKES IN THE CONTEXT 0F EMERGENCY PREPAREDNESS
Purpose:
To obtain Comission approval for publication in the Federal Register of a proposed amendment that would not permit consideration of earthquakes in emergency planning.
Category:
This paper covers a major policy matter.
Background:
On December 8, 1981, the Comission ruled in a then-pending adjudication that its emergency planning regulations do not require consideration of potential earthquake effects on emergency plans for nuclear power reactors.
In the Matter of Southern California Edison Company, et al. (San Onofre Nuclear Generating Station, Units 2 and 3), C1.I-81-33, 14 NRC 1091 (1981). In so ruling the Comission-stated:
The Comission will consider on a generic basis whether regulations should be changed to address the potential impacts of a severe earthquake on emergency planning.
For the
- interim, the l
proximate occurrence of an accidental radio-logical release and an earthquake that could disrupt nomal emergency planning-appears sufficiently unlikely that consideration in individual licensing proceedings pending generic consideration of the matter is not warranted. 14 i
NRC at 1092.
l The Comission recently affirmed this position in the Diablo Canyon proceeding.
In the Matter of Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and Z), CLI-84-12 NRC (August 10, CONTACT:
Michael Jamgochian, RES l
443-7615 l
a
e The Comissioners 2
1984).
In this decision the Comission stated that it would initiate rulemaking "to address whether the potential for seismic impacts on emergency planning is a significant enough concern for large portions of the nation to warrant the amendment of the regulations to specifically consider those impacts." Slip Opinion at 9.
The focus of this rulemaking is to "obtain additional infomation to detemine whether, in spite of current indications to the contrary, cost-effective reductions in overall risk may be obtained by the explicit consideration of severe earthquakes in emergency response planning."
Id,.
The attached Federal Register notice is the staff's response to that Comission order.
Discussion:
Experience has shown that the ability to take protective actions throughout the plume exposure pathway (EPZ) could be hampered
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by temporary
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adverse conditions resultina from natural ohenomena such as rain, snow, floodingfr by activities in the vicinit of the plant such as major road repair.
Existing NRC regulations require that emergency plans be comprehensive and flexible enough to assure the capability to take appropriate protective action to mitigate the effects of a
, nuclear emergency under such conditio11s. Ci '!: t:::: cf in
- diti
- ::: M r:::!t '- J eartnquakes below the ]
'5afe s lu;down Eartnquake (ddti, wnich occur proximate in i
time with an unrelated accidental ralanca of nuclaar J (material fron the facility f However, emergency plans wnicn meet t1e stanoards Tn 10 CFR 50.47 and Appendix E provide reasonable assurance that appropriate protective measures can and will be taken under such circumstances.
When considering the possibilities of plant damage from seismic events, it is important to understand the severity,
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range of probabil and the potential for reactor accidents caused by events.
Three classes of seismic events are considered in this discussion.
The first class includes earthquakes of relatively (low ground motion, up to the Operating Basis Earthquake OBE).
The OBE ground motion depends on plant location.
These accelerations vary in the range of about
.05g to.109 (higher in areas of high seismicity). During an OBE, all safety-related plant systems would be expected to remain operating.
i
r-q The Comnissioners 3
The second class of events includes earthquakes with j
ground motion higher than the OBE but equal to or less -
1 than the Safe Shutdown Earthquake [, the ground motion of the SSE is typically ~ about twice that of the OBE.
Probabilities of occurrence for the SSE have typically been estimated to be on the order of one in a thousand to one in ten thousand per year.
NRC regulations require that plants be designed to achieve a safe shutdown after an SSE. Given an SSE, all seismically qualified equipment would be expected to function to bring the plant to safe shutdown.
An earthquake up to and including an SSE would be cause for an alert emergency action level classifica-tion, but would not cause failures that would result in a significant accidental release from the plant.
- Thus, although such an event would initiate certain emergency plan actions, no offsite response would be required. Only in the event of multiple unrelated failures of safety-related systems due to some undiscovered connon cause failure mechanism (s_uch as a major design error), coinci-dent with the SSE, would there be a chance of an accident which would require offsite emergency response.
The probability of these two events occurring proximately in time is very much lower than the probability of either one, perhaps on the order of one in a million per reactor year.
The final class of events includes all earthquakes with ground motion levels above the SSE.
Fragility analysis has been used to estimate the probability of failure as a function of ground motion associated with these earth-quakes. The Zion, Indian Point, and Limerick probabilistic risk assessments estimated that, in general, ground motion on the order of 0.5g to 0.75g acceleration would be l
required to damage a nuclear power plant to the extent that significant release of radioactivity could occur. Of course, some plants, such as those in high seismic regions, are desi g gg sig q ukes with ground motion this hig
. _ _ _ Mff s gFto still higher levels of groun motion.
The probability estimates for such ground accelerations are significantly less than the probability estimates for the SSE.for these plants (the Zion, Indian Point, and Limerick SSEs are.179,.15g and l
.15g respectively).
The absolute probabilities. for earthquakes at and beyond the SSE are extremely difficult to estimate and thus have large associated uncertainties.
i l
l
The Commissioners 4
Based upon the probabilistic risk assessment results, the NRC staff has considered that for most earthquakes (including some earthquakes more severe than the SSE) the power plant would not be expected to pose an imediate offsite radiological hazard. For earthquakes which would cause plant damage leading to imediate offsite radio-logical hazards, but for which there would be relatively minor offsite damage, emargency response capabilities around nuclear power plants would not be seriously affected. For earthquakes which cause more severe offsite damage, such as disabling a siren-alerting system, the earthquake itself acts as an alerting system.
For those earthquakes which cause very severe damage to both the plant and the offsite area, emergency response would have l
marginal benefit because of its impairment by offsite damage. However, the expenditure of additional resources to cope with seismically caused offsite damage may be of
\\, doubtful value considering the modest benefit in overall risk reduction which could be obtained.
It should be noted that the Federal Emergency Management
\\
gency (FEMA) reviews offsite radiological emergency lanning and preparedness to insure the adequacy of Federal, State, and local capabilities in such areas as emergency organization, alert and notification, comunica-tions, measures to protect the public, accident assess-
,pment, public education and information, and medical
- h g
y support.
- Detailed, specific assessment of potential
_w e earthquake consequences and response are not part of this g }p/yprocess related to radiological Mw emergencies.O EMA duus--
F 4 L._, Md,**1m active program of earthquake prepared-6 f
ness which includes estimates of damage and casualties, g
planning for Federal response to a major earthquake, and j
.[
hs assistance to State and local governments in their earth-quakjg 1 anning and preparedness activities. FEMA believes 1
v yd that tiese separate activities weesd complement each other I
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q in the event that a concurrent response to a major earth-quake and a serious accident at a nuclear power plant were
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After careful consideration, the staff proposes to amend W 'h 10 CFR 50.47 and 10 CFR Part 50 Appendix E to codify the F
position that the Commission has taken in the San Onofre and Diablo Canyon cases.
A new subsection (e) would be added to 10 CFR 50.47 and a sentence would be added to the
" Introduction" section of Appendix E.
While these ayg W wAy
The Commissioners 5
amendments are simple in form, the staff recognizes that they represent a significant policy determination.
The staff therefore invites comment not only on the proposed texts, but also on the fundamental question of the relationship between earthquakes and emergency planning at nuclear power facilities.
1.
Adoption of the proposed rule codifying San Onofre and Diablo Canyon.
2.
Leaving the issue open for adjudication on a case-by-case basis.
3.
Requiring by rule that emergency plans specifically address the impact of earthquakes.
FEMA is directly involved in the evaluation of offsite emergency preparedness and, therefore, would be affected by the promulgation ~of this proposed rule change.
There-fore, the NRC staff consulted with the FEMA staff during approach and in the proposed rule chang. g in the staff's the development of this paper. FEWQccur Cost Estimate:
The staff anticipates that there will be a decrease or_ 3 h tM5A W b A h,and to licensees associated with the proposed rule c increase in cost to the NRC, State and local governments U
because it is interpretative in nature.
44 gg 4 g p
Mb NA Recomendation:
That the Comission:
1.
Approve:
The publication for public coment of the proposed rule change in the Federal Register (Enclosure 1).
2.
Note:
That appropriate Congressional comittees will a.
be notified of the proposec rule.
b.
That the ACRS is being informed of the proposed ruie.
c.
That, pursuant to 51.51(d)(3) of the Comis-sion's regulations, an environmental impact statement, negative declaration, or environ-mental impact appraisal need not be prepared in connection with the subject proposed amendment because there is no substantive or significant environmental impact.
1 J
F' K'
The Commissioners 6
d.
That the Federal Register notice contains a statement that the NRC certifies that the proposed rule will not, if promulgated, have a significant economic impact on a substantial number.of small entities, pursuant to the Regulatory Flexibility Act of 1980,6605(b).
e.
That tt.e Federal Register notice contains a statement
- that, pursuant to the Paperwork Reduction Act of 1960, the NRC has made a preliminary' determination that the proposed rule does not impose new recordkeeping, information collection, or reporting requirements.
f.
That the staff will directly notify affectea applicants, licensees, State governments, and interested persons of the proposed rule.
g.
That a public announcement of the proposed rule will be made.
h.
That a Preliminary Value/ Impact Analysis has been prepared (Enclosure 2).
i.
The staff's conclusions provide the analysis called for by the Periodic and Systematic Review of the Regulations.
The criteria used were derived from Executive Order 12044, which was rescinded on February 17, 1981, by Executive Order 12291 (see memorandum dated February 27, 1981, from L. Bickwit, General Counsel to the Connission).
This approach is proposed as an interim procedure until the Comission decides what to do in response to Executive Order 12291.
William J. Dircks Executive Director for Operations
Enclosures:
1.
Federal Register Notice of Proposed Rulemaking 2.
Preliminary Value/ Impact Analysis 1
J
9-9 9
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ENCLOSURE 1 I
t
1 NUCLEAR REGULATORY COMMISSION l
10 CFR Part 50 Emergency Planning And Preparedness for Production And Utilization Facilities AGENCY:
Nuclear Regulatory Commission.
ACTION:
Notice of proposed rulemaking; statement of interim policy.
SUMMARY
In previous adjudications the Comission has ruled that its regulations do not require the consideration of potential impacts of earthquakes en emergency planning for nuclear reactor sites.
The Comission now proposes to so provide explicitly in its regulations, and states that pending completion of the rulemaking, the interpretation of its rules set out in the adjudications remains in effect.
DATES:
Coment period expires (insert date 30 days after publication).
Coments received after this date will be considered if it is practical, but assurance of consideration can be given only for comments timely filed.
ADDRESSES:
Mail coments to:
Secretary of the Comission U.S. Nuclear Regulatory Commission, Washington, D. C.
20555, ATTN:
Docketing and Service Branch. Deliver comments to: Room 1121, 1717 H. Street, N.W., Washington, D.
C., between 8:15 a.m. and 5:00 p.m. weekdays. Copies of coments received may be examined at the NRC Public Document Room at the same address.
FOR FURTHER INFORMATION CONTACT:
Michael T.
Jamgochian, Division of Risk Analysis and Operationt, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission Washington, D. C. 20555.
Telephone: (301) 443-7615.
2
[7590-01]
4 SUPPLEMENTARY INFORMATION:
i
Background
On December 8,
1981, the Conunission ruled in a then K pending adjudication that its emergency planning regulations do not require cons'deration of potential earthquake effects on emergency plans for nuclear power reactors.
In the Matter of Southern California Edison Company, et al.
(San Onofre Nuclear Generating Station, Units 2 and 3), CLI-81-33,14 NRC 1091 (1981). In so ruling ti.c Commission stated:
The Connission will consider on a generic basis whether regulations should be changed to address the potential impacts of a severe earthquake on emergency planning. For the interim, the proximate occurrence of an accidental radiological release and an earthquake that could disrupt normal emergency planning appears sufficiently unlikely that consideration in individual licensing proceedings pending generic consideration of the matter is not warranted. 14 NRC at 1092.
The Comission recently affirmed this position in the Diablo Canyon proceeding.
In the Matter of Pacific Gas and Electric Company (Diablo Canyon j
Nuclear Power Plant, Units 1 and 2), CLI-84-12, NF.L' (August 10, i
1984).
In this decision the Conunission stated that it would initiate rulemaking "to address whether the potential for seismic impacts on emergency planning is a significant enough concern for large portions of the nation to warrant the amendment of the regulations to specifically consider thos8 j
impacts."
Slip Opinion at 9.
The focus of the rulemaking was to "obtain additional information to determine whether, in spite of current indications to the contrary, cost effective reductions in overall risk may be obtained by the explicit consideration of severe earthquakes in emergency response planning." M.
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It should be noted that the Federal Emergency Management Agency (FEMA) reviews offsite radiological emergency planning and preparedness to insure the adequacy of Federal, State, and local capabilities in such areas as emergency organization, alert and notification, comunications, measures to protect the public, accident assessment, public education and infonnation, and medical support.
Detailed, specific assessment of potential earthquake consequences and response are not part of this process related to radiological emergencies.
FEMA does have, however, an active program of earthquake preparedness which includes estimates of damage and casualties, planning for Federal response to a major earthquake, and assistance to State and local governments in their earthquake planning and preparedness activities.
FEMA believes that these separate activities would complement each other in the event that a concurrent response to a major earthquake and a serious accident at a nuclear power plant were required.
i For general background on emergency planning at nuclear facilities, the public is referred to NUREG-0396, " Planning Basis for the Development of State and Local Government Radiological Ecergency Response Plans in Support of Light water Nuclear Power Plants," and NUREG-0654 / FEMA-REP-1, Rev.
1,
" Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."
The latter document, developed jointly by the NRC and FEMA, forms the basis for both NRC and FEMA regulations on emergency planning at nuclear power facilities. Also available for public inspection are the complete case records for the San Onofre and Diablo Canyon proceedings, both of which deal specifically with theearthquakef/emergencyplanninginterface.
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Statement Of Interim Policy
- 1. e Connission, in its review of the record and consideration of arguments in the Diablo Canyon proceeding, reached the view that its previous San Onofre holding was correct, i.e., that the potential impact of earthquakes-on emergency plans need not be considered. The rationale for this holding was restated in the Diablo Canyon (Slip Opinion at 4-6), and may be summarized as follows. Nuclear power plants are required to be designed to safely shut down for all earthquakes up to and including the " Safe Shutdown Earthquake," or SSE. See 10 CFR Part 50, Appendix A, General Design Criterion 2; 10 CFR Part 100 Appendix A.
Accordingly, the probability of earthquakes large enough to cause major onsite damage that would result in a significant radiological release from the plant is extremely low, and for such large earthquakes, offsite damage could make prior offsite emergency plars premised on normal conditions marginally useful at best.
In addition, the probability of the proximate occurrence of an earthquake of substantial magnitude and a radiological release from the plant for reasons unrelated to the earthquake
- itself is even lower.
Therefore, there does not appear to exist a set of circumstances at all likely where the consideration of earthquake impacts would significantly improve the state of emergency planning at a nuclear power reactor.
.In addition, as the Connission noted in its Diablo Canyon decision (Slip.
Opinion at 5-6), emergency plans are not rigid documents which become useless if offsite conditions are less than ideal:
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i Specific consideration has been given in this case to the effects of other relatively frequent natural phenomena. The evidence includes the capability of the emergency plan to respond to disruptions in communications networks and evacuation routes as a result of fog, severe storms and heavy rain.
In the extreme, these phenomena are capable of resulting in area-wide disruptions s.imilar to some of the disruptions which may result frem an earthquake...
Thus, while no explicit consideration has been given to disruptions caused by earthquakes, the emergency plans do have considerable flexibility to handle the disruptions caused by various natural phenomena which occur with far greater frequency than do damaging earthquakes, and this implicitly includes some flexibility to handle disruptions from earthquakes as well.
Although the Commission's remarks were directed to the Diablo Canyon emergency plan, the noted flexibility is found in all nuclear power reactor emergency plans.
Such plans do address the contingency that emergency actions may need to be taken under less-than-ideal conditions and with less-than-maximum emergency response capabilities.
The ability to takc protective actions throughout the plume exposure pathway (EPZ) could be hampered dr i-- t':
c' t'e ;Mt by temporary a:! verse conditions resulting from natural phenomena such as rain, snow, W
flooding, or by activities in the vicinity of the plant such as major road repair.
Existing NRC regulations require that emergency plans be comniehensive and flexible enough to assure the capability to take appropriate protective action to mitigate the effects of a nuclear emergency under such conditions.
I'-" r ';- : c' re :: :-" t' : cer'd cre't '- d rthqua D fbelow the Safe Shutdown Earthquake (SSE), which occur proximate in time with i
{an unrelated accidental release of nuclear material from the facilit However, emergency plans which meet the standards in 10 CFR 50.47 and 1
.a--...
6
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Appendix E provide reasonable assurance that appropriate protective measures can and will be taken under such circumstances.
While the Comission intends to consider this issue carefully in this rulemaking and to weigh all arguments before reaching a final decision to be embodied in the regulations, it should be clear from the San Onofre and the existing e'M,ssion] N[RAMf^LD c
Diablo Canyon decisions that r--"'-
-' is not to consider the effects of earthquakes in emergency planning.
This interpretation of the Comission's regulations must be considered binding unless altered by the outcome of this proceeding.
All Comission adjudicatory panels should follow the' rule of these cases pending its nodification, if any, as a result of the rulemaking.
parties to adjudicatory proceedings may attempt to show "special circumstances" pursuant to 10 CFR 2.758 if they believe this interpretation of the Comission's rules should not be applied in a particular case.
Technical Infomation When considering the possibilities of plant damage from seismic events, it is important to understand the severity e' ' *
-4 r:--tr, their range of probabilitiesg and the potential for reactor accidents caused by events.
Three classes of seismic events are considered in this discussion.
Tha first class includes earthquakes of relatively low ground motion, up to the Operating Basis Earthquake (OBE). The OBE ground motion depends on plant location.
These accelerations vary in the range of about.05g to.10g (higher in areas of high seismicity).
During an OBE all safety related plant systems would be expected to remain operating.
7
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The second class of events includes earthquakes with ground motion higher than the OBE but equal to or less than the Safe Shutdown Earthquake /
(SSE); the ground motion of the SSE is typically about twice that of the OBE.
Probabilities of occurrence for the SSE have typically been estimated to be on the order of one in a thousand to one in ten thousand per year.
NRC regulations require that plants be designed to achieve a safe shutdown after an SSE.
Given an SSE, all seismically qualified equipment would be expected to function to bring the plant to safe shutdown.
An earthquake up to and including an SSE would be cause for an alert emergency action level classification, but would not cause' failures that would result in a significant accidental release from the plant.
Thus, although such an event would initiate certain emergency plan actions, no offsite response would be required. Only' in the event of multiple unrelated failures of safety related systems due to some undiscovered common cause failure mechanism (such as a major design error), coincident with the SSE, would there be a chance of an accident which would require offsite emergency response.
The probability of these two events occurring proximately in time is very much lower than the J
probability of either one, perhaps on the order of one in a million per reactor year.
The final class of events includes all earthquakes with ground motion levels. above the SSE.
Fragility analysis has been used to estimate the probability of failure as a functicn of ground motion associated with these earthquakes.
The
- Zion, Indian
- Point, and Limerick Probabilistic Risk Assessments estimated that, in general, ground motion on the order of 0.59 to 0.75g acceleration would be required to damage a nuclear power plant to the
8
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extent that significant release of radioactivity could occur. Of course, someplants,suchasthoseinhighseismdthgions,aredesignedto i
r ithstand earthquakes with ground motion this highK ^.:_,y ans ey an 2d resist damage to still higher levels, :',
The probability estimates for such ground accelerations are significantly less than -
^;
_z: for the
^
SSE for these plants (the Zion, IP, and Limerick SSEs are.179,.159, and
.15g respectively).
The absolute probabilities for earthquakes at and beyond the SSE are extremely difficult to estimate and thus have large associated uncertainties.
Based upon the probabilistic risk assessment results, the NRC staff has considered that for most earthquakes (including some earthquakes more severe than the SSE) the power pla'nt would not be expected to pose an immediate offsite radiological hazard. For earthquakes which would cause plant damage leading to immediate offsite radiological hazards but for which there would be relatively minor offsite damage, emergency response capabilities around nuclear power plants would not be seriously affected.
For earthquakes which cause more severe offsite damage, such as disabling a siren alerting system, the earthquake itself acts as an alerting system.
For those earthquakes which cause very severe damage to both the plant and the offsite area, emergency response would have marginal benefit because of its impairment by offsite damage. M te expenditure of additional resources to cope with seismically caused offsite damage may be of doubtful value considering the modest benefit in overall risk reduction which could be obtained, s
9
[7590-01]
Proposed Rule The Comission is proposing to amend 10 CFR 50.47 and 10 CFR Part 50, Appendix E to codify the position it has taken in the San Onofre and Diablo Canyon cases. A new subsection (e) would be added to 10 CFR 50.47 and a sentence would be added to the " Introduction" section of Appendix E.
While these amendments are simple in form, the Comission recognizes that they represent a significant policy deterinination. The Connission therefore invites comment not only on the proposed texts, but also on the fundamental question of the relationship between earthquakes and emergency planning at nuclear power facilities.
Comenters should, at a minimum, address the merits of three possible alternatives:
1.*
Adoption of the proposed rule codifying San Onofre and Diablo Canyon.
2.
Leaving the issue open for adjudication on a case-by-case basis.
3.
Requiring by rule that emergency plans specifically address the impact of earthquakes.
The Commission would be most assisted by coments which offer specific policy p*
$6 and technical reasons for preferring one alternative over others.
Finding Of No Significant Environmental Impact The Comission has determined under the National Environmental Policy Act of 1969, as amended, anti the Commission's regulations in Subpart A of 10 CFR Part 51, that this proposed rule, is not a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.
See 10 CFR 51.20(a)(1).
Moreover, the Comission has determined, pursuant to 10 CFR 51.32, that the
10
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proposed rule has no significant environmental impact.
This determination has been made because the Commission cannot identify any impact on the human environment associated with requiring or not requiring consideration of earthquakes in emergency planning.
Paperwork Reduction Act Statement The proposed rule contains no information collection requirements and therefore is not subject to the requirements of the paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).
Regulatory Analysis The Consnission has prepared a regulatory analysis of this proposed regulation.
The analysis examines the costs and benefits of the rule as considered by the Commission. A copy of the regulatory analysis is available for inspection and copying, for a fee, at the NRC Public Document Room,1717 H Street, N.W.,
Washington, DC.
Single copies of the analysis may be obtained from Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Consnission, Washington, DC 20555 Telephone (301) 443-7615.
Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
I 605(b), the Commission hereby certifies that this proposed rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. The proposed rule clarifies requirements for the issuance of
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an operating license for a nuc' ear power plant, licensed pursuant to Section 103 and 104b of the Atomic Energy Act of 1954, as amended, 42 U.S.C. 2133, 2134b.
The electric utility companies which own and operate nuclear power plants are dominant in their service areas and do not fall within the L
definition of a small business found in Section 3 of the Small Business Act, 15 U.S.C. 632, or within the Small Business Size Standards set forth in 13 CFR Part 121.
Accordingly, there is no significant economic impact on a substantial number of small entities under the Regulatory Flexibility Act of 1980.
List of Subje. cts in 10 CFR Part 50 Antitrust, Classified information, Fire prevention.
Part 50 Incorporation by reference, Intergovernmental relations, Nuclear Power plants and
- reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.
Pursuant to the Atomic Energy act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and section 552 and 553 of Title 5 of the United States Code, notice is hereby given that adoption of the following amendments to Title 10, Chapter I, Code of Federal Regulations, Part 50 is contemplated.
Part 50 - Domestic Licensing of Production and Utilization Facilities 1.
The authority citation for Part 50 continues to read as follows:
AUTHORITY:
Sections 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 937, i
f
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948, 953, 954, 955, 956, as amended, sec. 234 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2225, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846), unless otherwise noted.
Section 50.7 also issued under Pub. L 95-601, sec.10, 92 Stat. 2951 (42 U.S.C. 5851). - Sections 50.57(d), 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2133, 2239).
Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C.
2152).
Sections 50.80-50.8 also issued under sec.184, 68 Stat. 954, as amended (42 U...C.
2234).
Sections 50.100-50.102 also issued under sec.186, 68 Stat. 955 (42 U.S.C. 2236).
For the purposes of Sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273),
il 50.10(a)(, (b), and (c), 50.44, 50.46, 50.48, and 50.80(a) are issued under 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); [Il 50.10(b), and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C.
2201(1); and 50.55(e), 50.59(b), 50.70, 50171, 50.72, 50.73, and 50.78 are issued under sec. 1610,68 Stat.950,asamended(42U.S.C.2201(o)).
2.
A new subsection (e) is added to i 50.47 to read as follows:
50.47 Emergency Plans.
4 (e)
Offsite emergency response plans, submitted to satisfy the standards set forth in this section, need not consider the impact on emergency ;TM'.; N
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earthquakes which cause, or occur proximate in time with, an accidental release of radioactive material from the facility.
4 13
[7590-01]
3.
A sentence is added as a separate paragraph at the end of the Introduction section of Appendix E to read as follows:
1.
Introduction Neither emergency response plans nor evacuation time analyses need consider the impact of earthquakes which cause, or occur proximate in time with, an accidental release of ' adioactive material from the facility.
r (Sec. 161b, i., and o., Pub. L.83-703, 68 Stat. 948 (42 U.S.C. 2201);
Sec. 201, as amended, Pub. L.93-438, 88. Stat.1242, Pub. L. 94-79, 89 Stat.
413 (42 U.S.C. 5341))
Dated at Washington, D.C., this day of 1984.
FOR THE NUCLEAR REGULATORY COMMISSION Samuel J. Chilk Secretary of the Commission
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ENCLOSURE 2
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F
e Regulatory Analysis for Proposed Rulemaking to 10 CFR Part 50.47 and Appendix E Emergency Planning and Preparedness for Production and Utilization Facilities Statement of the Problem On December 8,1981, the Comission ruled in a then-pending adjudication that its emergency planning regulations do not require consideration of potential earthquake effects on emergency plans for nuclear power reactors.
In the Matter of Southern California Edi~ son Company, et al.
(San Onofre Nuclear Generating Station, Units 2 and 3), CLI-81-33,14 NRC 1091 (1981). In so ruling the Connission stated:
The Commission will consider on a generic basis whether regu-lations should be changed to address the potential impacts of a severe earthquake on emergency planning. For the interim, the proximate occurrence of an accidental radiological release and an earthquake that could disrupt normal emergency planning appears sufficiently unlikely that consideration in individual licensing proceedings pending generic consideration of the matter is not warranted.
14 NRC at 1092.
The Comission recently affimed this position in the Diablo Canyon proceed-ing.
In the Matter of Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), CLI-84-12, NRC (August 10,1984).
In this decision the Commission stated that it would initiate rulemaking "to address whether the potential for seismic impacts on emergency planning is a significant enough concern for large portions of the nation to warrant the amendment of the regulations to specifically consider those impacts."
Slip Opinion at 9.
The focus of this rulemaking is to "obtain additional information to determine whether, in spite of current indications to the t
~
I 2
contrary, cost effective reductions in overall-risk may be obtained by the 2
explicit consideration of severe earthquakes in emergency response planning."
}$.
Objective The objective of the proposed rule change is: to not consider the impact 4
of earthquakes on emergency preparedness.
The Commission has decided that its previous San Onofre decision was
- correct, i.e., that the potential impact of earthquakes on emergency plans need not be considered.
i l
The rationale for this decision was restated in the Diablo Canyon (Slip Opinion at 4-6), and may be summarized as follows.
Nuclear power plants are r
l required to be designed to safely shut down for all earthquakes up to and including the " Safe Shutdown Earthquake " or SSE.
See 10 CFR Part 50 Appendix A, General Design Criterion 2; 10 CFR Part 100, Appendix A.
Accord-ingly, the probability of earthquakes large enough to cause major onsite damage that would result in a significant radiological release from the plant is extremely low, and for such lar'ge earthquakes, offsite damage could make 4
prior offsite erari,ency plans premised on nomal conditions marginally useful at best.
In addition, the probability of the proximate occurrence of an earthquake of substantial magnitude and a radiological release from. the plant i
i for reasons unrelated to the earthquake itself is even lower.
Therefore.
there does not appear to exist a set of circumstances at all likely where the consideration of earthquake impacts would significantly improve the state of emergency planning at a nuclear power reactor.
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In addition, as the Commission noted in its Diablo Canyon decision (Slip Opinion at 5-6), emergency plans are not rigid documents which become useless if offsite conditions are less than ideal:
Specific consideration has been given in this case to the effects of other relatively frequent natural phenomena. The evidence includes the capability of the emergency plan to respond to disruptions in communications networks and evacuation routes as a result of fog, severe storms and heavy rain.
In the extreme, these phenomena are capable of resulting in area-wide disruptions similar to some of the disruptions which may result from an earthquake.... Thus, while no explicit consideration has been given to disruptions caused by earthquakes, the emergency plans do have considerable flexibility to handle the disruptions caused by various natural phenomena which occur with far greater frequency than do damaging earthquakes, and this implicitly includes some flexibility to handle disruptions from earthquakes as well.
Although the Commission's remarks were directed to the Diablo Canyon emergency plan, the noted flexibility is found in all nuclear power reactor emergency plans.
Such plans do address the contingency that emergency actions may need to be taken under less-than-ideal conditions and with less-than-maximum emergency response capabilities.
The ability to take protective actions throughout the plume exposure pathway (EPZ) could be hampered during the life of the plant by temporary adverse conditions resulting from ' natural phenomena such as rain, snow, flooding or by activities in the vicinity of the plant such as major road repair.
Existing NRC regulations require that emergency plans be comprehen-sive and flexible enough to assure the. capability to take appropriate protec-tive action to mitigate the effects of a nuclear emergency under such conditic1s. Similar types of adverse conditions could result from earthquakes below the Safe Shutdown Earthquake (SSE), which occur I oximate in time with
4 an unrelated accidental release of nuclear material from the facility.
However, emergency plans which meet the standards in 10 CFR 50.47 and Appendix E provide reasonable assurance that appropriate protective measures can and will be taken under such circumstances.
While the Comission intends to consider this issue carefully in this rulemaking and to weigh all arguments before reaching a final decision to be embodied in the regulations, it should be clear from the San Onofre and Diablo Canyon decisions that the existing rule established by precedent is not to consider the effects of earthquakes in emergency planning.
This interpre-tation of the Commission's regulations 'must be ccN *Ored binding unless altered by the cutcome of this proceeding. All Comisc..wa adjudicatory panels should follow the rule of these cases pending its modific..ica, if any, as a result of the rulemaking.
Parties to adjudicatory proceedings reay attempt to show "special circumstances" pursuant to 10 CFR 2.758 if they believe this interpretation of the Commission's rules should not be applied in a particular Cast.
Alternatives One alternative would be to revise 10 CFR 50.47 to require that emergency plans specifically address the impact of earthquakes. The staff believes thi5 to be an appropriate alternative because of the flexibility of extf *tk emergency plans well as the very low probability of the occurrence f /,1 earthquake of substan 1 magnitude and a radiological release from the f 4 4.
Another alternative would b o adjudicate the issue on a case-by-case basis, i
The staff believes this to be an inappropriate alternative because it would be b s m ee k w, n m u k A
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b The bt,- e rule change is the extremely time consuming an best alternative available for achieving the specific regulatory objective.
Consequences NRC The staff believes that the consideration of earthquake would not improve the state of emergency planning at a nuclear power reactor.
Other Government Agencies The proposed rule change does not have a significant impact on other government agencies because it is an interpretation of existing regulations.
Industry The proposed amendment will not greatly affect the industry since licenses are required to have approved emergency plans which are r~; '
- i c = d flexible enough t -MO appropriate y;^ ~>Ms cAss be, sM.o protective suasmisto mitigate the consequences of a nuclear emergencyp Public The proposed amendment will have no effect on the public as adequate emergency preparedness at nuclear reactors will still be assured.
Impact on Other Requirements
.The proposed rule change has no impact on other NRC requirements.
Constraints No constraints have been identified that affect the implementation of the proposed rule.
6 Decision Rationale The technical decision rationale that the staff used is based on techni-cal infonnation regarding seismic events.
When considering the possibilities of plant damage from seismic events, it is important to understand the severity of seismic events, their range of probabilitiesg and tM potential for usums.
Three classes of seismic events are considered in this discussion.
The first class includes earthquakes of relatively low ground motion, up to the Operating Basis Earthquake (OBE).
The OBE ground motion depends on plant location. These accelerations vary in the range of about.05g to.10g (higher in areas of high seismicity).
During an OBE all safety related plant systems would be expected to remain operating.
The second~ class of events includes earthquakes with ground motion higher than the OBE but equal 'to or less than the Safe Shutdown Earthquakes (SSE);
the ground motion of the SSE !3 typically about twice that of the OBE. Prob-abilities of occurrence for the SSE have typically been estimated to be on the order of one in a thousand to one in ten thousand per year.
NRC regulations require that plants be designed to achieve a safe shutdown after an SSE.
Given an SSE, all seismically qualified equipment would be expected to func.
tion to bring the plant to safe shutdown.
An earthquake up to and including an SSE would be cause for an alert emergency action level classification, but would not cause failures that would result in a significant accidental release from the plant. Thus, although such an event would initiate certain emergency plan actions, no offsite response would be required.
Only in the event of multiple unrelated failures of safety related systems due to some undiscovered comon cause failure mechanism (such as a major design error), coincident with
W 9
7 the SSE, would there be a chance of an accident which would require offsite emergency response. The probability of these two events occurring proximately in time is very much lower than the probability of either one, perhaps on the order of one in a million per reactor year.
The final class of events includes all earthquakes with ground motion levels above the SSE.
Fragility analysis has been used to estimate the probability of failure as a function of ground motion associated with these earthquakes. The Zion Indian Point, and Limerick Probabilistic Risk Assess-ments estimated that, in general, ground motion on the order of 0.5g to 0.75g acceleration would be required to damage 'a nuclear power plant to the extent that significant release of radioactivity could occur. Of course, some plants, such as those in high seismic regions, are designed to withstand earthquakes with ground motion this high; they would resist damage to still h*igher levels of ground motion.
The probability estimates for such ground accelerations are significantly less than the probability estimates for the SSE for these plants (the Zion, IP, and Limerick SSEs are.17g,.15g, and.15g respectively).
The absolute probabilities for earthquakes at and beyond the SSE are extremely difficult to estimate and thus have large associated uncer-tainties.
Based upon the probabilistic risk assessment results, the NRC staff has considered that for most earthquakes (including some earthquakes more severe than the SSE) the power plant would not be expected to pose an imediate offsite radiological hazard. For earthquakes which would cause plant damage leading to immediate offsite radiological hazards but for which there would be relatively minor offsite damage, emergency response capabilities around nuclear power plants would not be seriously affected.
For earthquakes which
8 cause more severe offsite damage, such as disabling a siren alerting system, the earthquake itself acts as an alerting system.
For those earthquakes which cause very severe damage to both the plant and the offsite area, emergency response would have marginal benefit because of its impairment by offsite damage. %mtfseAO
. _.._ _., the expenditure of additional resources to cope with seis-mically caused offsite damage may be of doubtful value considering the modest
~
be'nefit in overall risk reduction which could be obtained.
It should be noted that the Federal Emergency Management Agency (FEMA) reviews offsite radiological emergency planning and preparedness to insure the adequacy of Federal, State, and local capabilities in such areas as emergency organization, alert and notification, comunications, measures to protect the public, accident assessment, public education and information, and medical support.
Detailed, specific assessment of potential earthquake consequences 1
and response are not part of this process related to radiological emergencies.
i.
FEMA does have, however, an active ' program of earthquake preparedness which includes estimates of damage and casualties, planning for Federal response to a major earthquake, ano assistance to State and local governments in their earthquake planning and preparedness activities.
FEMA believes that these separate activities would complement each other in the event that a concurrent response to a major earthquake and a serious accident at a nuclear power plant were required.
For general background on emergency planning at nuclear facilities, the public is referred to NUREG-0396, " Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants," and NUREG-06541 FEMA-REP-1, Rev.1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and i
c
~
9 Preparedness in Support of Nuclear Power Plants" The latter document, devel-oped jointly by the NRC and FEMA, forms the basis for both NRC and FEMA regulations on emergency planning at nuclear power facilities. Also available for public inspection are the complete case records for the San Onofre and Diablo Canyon proceedings, both of which deal specifically with the earth-quakes / emergency planning interface.
Implementation In order to be responsive to Commission direction, the staff has estab-lished the following schedule for publication of the proposed rule change.
Proposed Rule Published in Federal Register -10/20/84 ThiS lO 30 Day Comment Period - 11/30/84 M
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,1 NUCLE AR HE GULA10RY COMMISSION
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W ASHING 10N O r 2W.w im
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i-MEMORANDUM FOR:
Chairman Palladino Comissioner Roberts Comissioner Asselstine Comissioner Bernthal Comissioner Zech FROM:
William J. Dircks Executive Director for Operations
SUBJECT:
CONSIDERATION OF POTENTIAL COMPLICATING EFFECTS OF EARTHQUAKES ON EMERGENCY PLANNING The purpose of this memo is to infonn you of our progress and direction in preparation of the subject final rulemaking package which I plan to submit to you by early August 1985.
On December 21, 1984, the Comission published a proposed rule change to 10 CFR Part'50 that relates to Emergency Planning and Preparedness at Production and Utilization Facilities (49 FR 49640).
The proposed rule stated that neither emergency response plans nor evacuation time analyses need consider the impact of earthquakes which cause or occur proximate in time with an i
accidental release of radioactive material from a nuclear power reactor.
To date, 61 coment letters have been received.
Twenty five (25) letters favored the presulgation of the proposed rule. The majority of these letters were from utilities, consulting firms representing utilities 2 private citi-zens and the Department of Energy.
Thirty-four (34) letters opposed promulgation of the proposed rule.
Many voiced strong displeasure, shoc!, or disbelief at the position the Comission was taking in the proposed rule change. The majority of these letters were l
from private citizens, and environmental groups.
Additional input was also received from Japan France, Sweden Germany and 5
Taiwan.
All of which stated that the potential complicating effects.of earthquakes was not specifically considered in their nuclear power reactor emergency planning.
.Several issues raised in the public coments (and in particular in coments from The Union of Concerned Scientists) will require substantial technical analysis prior to going forward with promulgation of a final regulation. For a
CA
,q; The Conni.ssioners 2
' example the staff needs to:
(1) demonstrate there is sufficient data to support the finding that the complicating effects of earthquakes on emergency plans is effectively taken into account by the flexibility that exists in all emergency plans; (2) deal with the issue that defects in seismic design and quality assurance in construction can undemine the seismic strength of plant systems and structures; (3) fully evaluate the contributon of seismic events to overall core melt risks.
(4) deal with the question why emergency plans "should not consider the complicating effects of very severe earthquakes (i.e., 2 to 4 times the SSE) whose return frequency is 10E(-4) to 10E(-5) while current emergency plans concern themselves with plant accidents whose return frequency are also in the 10E(-5) to 10E(-6) range.
These _ complex analyses, which are underway, are not expected to be completed before late (July,1985.
After careful review of both the San Onofre and Diablo Canyon decisions involving the complicating effects of earthquakes on emergency planning, as well as the issues identified above, the staff is considering 3 alternative approaches:
Alternative 1:
Adoption of the proposed rule into a final rule with minor but important word changes, for example, "no additional emergency prepared-ness measures need be established to account for severe, low frequency natural phenomena than is already required in 10 CFR 50.47 and Appendix E."
' Alternative 2:
Leaving the issue open for adjudication on a case-by-case basis; accomplished by withdrawing the proposed rule or by requiring consid-eration of earthquakes.
Alternative 3:
Promulgation of a final rule which clarifies the original intent of the Commission and states that " emergency response plans shall assure that the following decisionmaking preplanning capabilities exist rela-tive to the complicating impacts of severe, low frequency natural phenomena.
The intensity of the event shall be no greater than the design basis for that event.
1.
Ability to transport necessary personnel to the plant within 12 to 24 j
hours after the event in order to augment the criginal staff to cope l
with degraded modes of plant operation.
l 2.
Ability to obtain damage estimates to the plant and to be able to comu-nicate these estimates to offsite authorities.
The infomation should be available to factor into the decision-making process, including reco-amendations for protective actions after severe, low frequency natural phenomena.
3.
Offsite authorities should consider decisionmaking preplanning that takes into account various degrees and locations of damage to the plant i
1'
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. The Comissioners 3
environs.
This shall be limited to knowing alternate routes of travel as well as establishing criteria for detennining whether to shelter, relocate or to evacuate."
Having considered all of the above, as well as all coments received, past operating reactor and emergency preparedness experience, I am leaning toward a recomendation that a final rule be promulgated which would embrace the concepts of Alternative 3.
This alternative would be a clarification and articulation of the Comissions original intent as to what is specifically required to assure the necessary flexibility to cope with the complicating effects of severe, low frequency natural phenomena on emergency planning.
The advantages of this alternative are: (1) it defines and limits scope of issues which may be open to litigation.
(2) it would be a clarification rule change, and the Comission would not be seen as changing its decisions. (3) it addresses all low frequency natural phenomena rather than singling out earthquakes.
(4)
The necessary assessments have been made for San Onofre and Diablo Canyon (although not litigated)
(5) The assessments are not difficult or expensive to accomplish.
Sincerely, William J. Dircks Executive Director for Operations cc:
SECY OGC OPE
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4 Section 13 DOMINANT CONTRIBUTORS TO SEISMIC RISK - AN APPRAISAL R. P. Kennedy *, M. K. Ravindra* and R. H. Sues
- INTRODUCTION In the last four years, seismic probabilistic risk assessment (PRA) studies on six (6) nuclear power plants have been completed and many more studies are underway.
The assessment of seismic contribution to the overall plant risk is highly sensitive to the choice of seismic hazard curves employed in the study. There is a wide dif-forence of opinion between seismic hazard analysts as to the appropriate hazard modeling parameter values for any specific site. Currently, the U.S. Nuclear Regu-latory Comission and the Electric Power Research Institute have sponsored major l
research studies to develop more real.istic seismic hazard models for the Eastern I
United States. Therefore, for the present, published seismic PRAs should be re-viewed from a viewpoint that is not fogged by the seismic hazard issues. As dis-
' ussed in this paper, such an ' appraisal of seismic PRAs is possible. A major bene-c fit of the seismic PRA is the insight it provides on the relative safety of differ-ent structures, equipment and systems in the plant and the identification of domi-nant risk contributors. Seismic PRAs have shown that structures and equipment igortant to plant safety have ground acceleration capacities well in excess of the design basis earthquake ground accelerations specified for the plants and that the contribution to the severe core damage frequency and risk from earthquakes with peak ground accelerations in the range of 0.20g to 0.30g is negligible.
'In recent years, a concern has been expressed regarding the seismic adequacy of operating or soon to be operating nuclear power plants for earthquakes of magnitudes larger than the original design basis. The results and insights gained in the seismic PRAs may be utilized in establishing the existing seismic margins in nuclear power plant structures and equipment. Admittedly, the seismic risk is highly plant-specific and is a function of the seismic hazard and the plant design basis. The sagle of PRAs may not be treated as representative of all nuclear power plants.
WStructural Mechanics Associates, Newport Beach, California 13-1 CO.
.y to derive quantitative generic conclusions about the seismic adequacy of other plants against larger earthquakes. However, the lessons learned in perfoming these and other unpublished seismic PMs as to the strengths and weaknesses of nuclear power plant structures and equipment, and the identification of dominant contribu-
~
tors -to seismic risk may be utilized in addressing the seismic design margin questions.
The purpose of this paper is to assess existing seismic margins in the structures and equipment at a selected number of nuclear power plants for which seismic PMs exist and to identify dominant contributors to seismic risk. The paper also dis-cusses,the causes for the seismic risk dominance of these contributors. Finally, it provides a list of structural and equipment items that should receive particular attention,in' any Wismic margin review. The parer has utilized the quantitative results of the seismic PMs on Zion, Indian Point Units 2 and 3. Limerick and
'MillstoneUnit3. It may be noted that several conservative assumptions were made
- i in dese PMs; for exagle, sevem structural distortion fails all the attached equipment and ispect between adjacent structures was assumed to fail all the equip-ment mounted on the floor near the impact.dTherefore, some of the seismic risk contributors identified herein map not be significant if the conservatism in the
~
PM is removed. Yet, for the purposes of planning the seismic margin reviews, this conservatism is deemed appropriate.
IDENTIFICATION OF, DOMIMNi ?
CONTR10tiTORS.
N j s i
in the seismic PM, those safety-related cogonents with median ground acceleration capacities less than 10-15 times' the plant design level ($$E) are first identified since it is these items which represent the phten'tial contributors to the seismic risk (e.g., Table 13-1). The plant safety systems analysis is next reviewed to detemine which of these componen,ts represent.the weaker links and will dominate the seismic risk. That is, those components whose failures can lead directly to plant
~damageorthesequencesof(lowcapacity)componentfailuresthatcanleadtoplant damage are identified. With this approach, the dominant contributors to seismic risk for each of the four plants being studied have been identified.
Two key assumptions are made in the PM systems analysis which should be noted and borne in mind when reviewing the analysis for identifying significant risk contri-butors. The first is that PMs have assumed that chattering of electrical relays and breaker trip am recoverable failure modes. That is, the plant operators will be able to maintain control of the plant safety systems even though these " failures" have occurred. For this reason, these low capacity failure modes (see Table 13-1) 13-2
TABLE 13-1
~
Fragilities of Key Structures and Equipment - ZION High Confidence
-Low Frequency of Failure v
Structure / Equipment A
6
'I R
U (g)
(g) 1.
Offsite Power -
0.20 0.20 0.25 0.10 Ceramic Insulators 2.
125V AC Distri-bution Panel Recoverable 0.60 0.37 0.50 0.14 Non-recoverable
> 10 x SSE 3.
125V DC Suswork Recoverable 0.60 0.37 0.50 0.14 Non-recoverable
> 10 x SSE 4.
Service Water Pumps' O.63 0.15 0.36 0.27 5.
4160V Switchgear Recoverable 0.72 0.35 0.47 0.19 Non-recoverable
> 10 X $$E 6.
480V Switchgear Recoverable 0.72 0.36 0.47 0.18 Non-Recoverable
> 10 X SSE 7.
480V Motor Control Centers Recoverable 0.72 0.36 0.47 0.18 Non-recoverable
> 10 X SSE 8.
Auxiliary Building Concrete Shear Wall 0.73 0.30 0.28 0.28 9.
Refueling Water Storage Tank 0.73 0.30 0.28 0.28
- 10. Interconnecting Piping / Soil Failure Beneath Reactor Building 0.73 0.28 0.33 0.27
- 11. Ispact Between Reactor and Auxiliary Buildings 0.78*
0.28 0.41 0.46
- 12. Condensate Storage Tank,
0.83 0.28 0.29 0.32
.13-3
- 2 TABLE 13-1(Continued)
Frag 111 ties of Key Structures and Equipment - ZION High Confidence
-Low Frequency of Failure v
Structure / Equipment A
s S
l***I U
(9)
(9)
- 13. 4160V Diesel Generators Recoverable O.'86 0.35 0.37 0.26 Non-recoverable
> 10 x SSE
- 14. Crib House Collapse of Pump Enclosure Roof, t
0.86 0.24 0.27 0.37
- 15. Safety Injection Pumps 0.90 )
0.20 0.37 0.35
- 16. Containment Ventilation 0.25 Ductwork and Dampers 0.97 0.20 0.62
- 17. 125V DC Batteries and Racks 1.01 0.28 0.63 0.23
- 18. Core Geometry 1.16 0.25 0.42 0.38
- 19. Reactor Coolant System Relief Tank 1.19 0.20-0.63 0.30
- 20. 4160V Transformer 1.39 0.25
' O.60 0.34
- 21. Service Water System Buried Pipe 48",
~
1.40 0.20 0.57 0.39
- 22. CST Piping 20" 1.40 0.20 0.57 0.39
- 23. Aux 1111ary Building-Concrete Roof Diaphragm 1.40 0.31 0.33 0.49
- 24. Failure of Masonry Walls 1.70 0.50 0.:.6 0.49 t
- 25. Containment Ventilation System Fan Coolers 1.74 0.49 0.23 0.53
- 26. Cellapse of Pressurizer Enclosure Roof 1.80 0.39 0.34 O.54 Wicable only with a median lower bound of 0.74g and SU =.29 M
J o
M i
's 13-4
have not been included in the PRA systems analysis. It is clear that if these failure modes are assumed to b,e non-recoverable, these items would be contributors to the seismic risk although a' quantitative assessment is not possible without a reanalysis.of the plant safety systems.
The second assumption is that severe structural distortion and the failure of cer-tain structural members can incapacitate many equipment items and in some cases, an entire safety system. This assumption,is, of course, conservative and has the effect of asplifying the significance of certain so-called structural failures.
Tables 13-2 through 13-6 show the results of the analysis, that is, the dominant seismic contributors at each of the four plants studied. The tables.show the median ground acceleration, k, which is expected to fail the component or structure, the and 8, representing randomness and uncertainty, logarithmic standard deviations, SR g
respectively, in the median, and the acceleration level for which there is high con-fidence of a small frequency of failure (approximately 95% confidence of less than approximately 5% frequency of failure). As an example, identification of the domi-nant risk contributors at the Zion Nuclear Generating Station will be illustrated.
Table 13-2 FRAGILITIES OF DOMINANT RISK CONTRIBUTORS ZION NUCLEAR GENERATING STATION High Confidence
-Low Frequency of Failure v
A s
s l'V'I g
u Structure / Equipment (g)
(g)
- 1. Service Water Pungs 0.63 0.15 0.36 0.27 l
- 2. Auxiliary Building -
0.73 0.30 0.28 0.28 Shear Wall
- 3. Interconnecting Piping /
0.73 0.28 0.33 0.27 Soil Failure Beneath Reactor Buildinq
- 4. Crib House Collapse of 0.86 d.24 0.27 0.37 8 ump Enclosure Roof f
t' 5.125V DC Batteries 1.01 0.28 0.63 0.23 and Racks 13-5 4
tr e
r n
=
Table 13-3 FRAGILITIES OF DOMINANT RISK CONTRIBUTORS INDIAN POINT UNIT 2 High Confidence
-Low Frequency of Failure v
A B
8 l'Y'I R
U Structure / Equipment (g)
(g)
- 1. Refueling Water Storage 0.70 0.22 0.28 0.31 Tank
- 2. Unit 1 Superheater Stack 0.72 0.34 0.26 0.27
- 3. Reactor Coolant System 0.87 0.23 0.40 0.31 Pressurizer
- 4. Control Building Impact 1.00 0.30 0.30 0.37 with Unit 1 Superneater Building
- 5. Diesel Generator Fuel 1.14 0.26 0.52 0.31 011 Tanks
- 6. Battery Rooms 21 and 22 1.30 0.28 0.36 0.45 Masonry Walls
- 7. 125V DC Batteries and 1.37 0.30 0.63 0.30 Racks
- 8. Service Water System 1.40 0.20 0.57 0.39 Buried 24" Piping
- 9. Turbire Building Steel 1.40 0.30 0.26 0.56 Frame
- 10. Cable Tray Increment 1.54 0.37 0.61 0.31 Zion Table 13-1 shows the key structures and equipment items at Zion which were originally identified as potential seismic risk contributors (i.e., safety-related components with median ground acceleration capacities less than approximately 10-15 times the SSE). As discussed earlier, relay chatter and breaker trip were con-sidered recoverable failure modes in the Zion PRA. Thus, the electrical conponents listed in Table 13-1 do not contribute to the seismic risk. However, the relatively low capacities indicate that this would not be the case if chatter and trip are j
considered as non-recoverable in the plant logic analysis. The structures and equipment whose individual failures can directly cause severe core damage (assuming 13-6
that the earthquake has caused a loss of offsite power) and dominate the seismic risk are the service water pumps, auxiliary building shear wall, interconnecting piping / soil failure beneath the reactor building, pump enclosure roof and the 125V DC batteries and racks. The median capacities of these components range from 0.639 for the service water pumps to 1.019, for the 125V DC batteries and racks and the high confidence, low frequency of failure levels range from 0.23g for the batteries and racks to 0.379 for the pump enclosure roof. Thus, it is clear that there is high seismic safety margin above the plant design level of 0.179 From Table 13-1, it may also be seen that the refueling water storage tank capacity (governed by the auxiliary building shear wall failure) and the condensate storage tank capacity are comparable to the capacities of the components just discussed. These two compo-nents, however, must both lail in order for core damage to occur whereas, failure of the auxiliary building shear wall (which governs the RWST failure) can, in itself, lead to core damage. Thus, the condensate storage tank is not a dominant risk contributor.
Table.13-4 FRAGILITIES OF DOMINANT RISK CONTRIBUTORS INDIAN POINT UNIT 3 High Confidence
-Low Frequency of Failure v
A Level 8
8 Structure / Equipment (g)
R U
(g)
- 1. Refueling Water Storage 0.70 0.22 0.28 0.31 Tank
- 2. Reactor Coolant System 0.87 0.23 0.40 0.31 Pressurizer
- 3. Ciesel Generator Fuel 1.14 0.26 0.52 0.31 011 Tanks
- 4. Control Building Shear 1.20 0.16 0.23 0.63 Wall 5.125V DC Batteries and 1.29 0.28 0.59 0.31 Racks (Diesel Building)
- 6. Service Water System 1.40 0.20 0.57 0.39 Buried 24" Piping
- 7. Diesel Generator Building 1.50 0.25 0.35 0.56 Concrete Roof 13-7
~
Table 13-5 FRAGILITIES OF DOMINANT RISK CONTRIBUTORS LIMERICK GENERATING STATION High Confidence
-Low Frequency of Failure y
1 A
Level 8
8 Structure / Equipment (g)
R
,0 (g)
- 1. Reactor Internals 0.67 0.28 0.32 0.25
- 2. Reactor Enclosure and 1.05 0.31 0.25 0.42 Control Structure r
Shear Wall
- 3. Reactor Pressure Vessel' 1.25 0.28 0.22 0.55 h
- 4. Standby Liquid Control 1.33 0.27 0.19 0.62 i
Tank
- 5. Diesel Generator Heating 1.55 0.28 0.43 0.48 and Venting
- 6. 4160 to 480V Transformer 1.66 0.26 0.49 0.48 In terms of public risk, the same components discussed above are dominant. The seismic capacity of the containment is sufficiently high such that radioactive release (due to a seismic initiating event) can only occur as the result of core melt (as discussed above) followed by a late overpressure failure of the contain-ment. As mentioned previously, Table 13-2 summarizes these results.
1 ASSESSMENT OF SEISMIC RISK CONTRIBUTORS In the previous section, the dominant seismic risk contributors in the four plants being studied were identified. These risk contributors will now be grouped into generic categories and the reasons for their contribution discussed. The objective is to document the lessons learned in performing the seismic PRAs (1_, 2.) and to develop some broad guidelines for conducting seismic design margin reviews.
i 13-8
~
Table 13-6 FRAGILITIES OF DOMINANT RISK CONTRIBUTORS MILLSTONE UNIT 3 High Confidence
-Low Frequency of Failure v
A Level 8
S Structure / Equipment (g)
R g
(g)
- 1. Emergency Generator 0.88 0.20 0.46 0.30 Enclosure Building Wall Footing
- 2. Refueling Water Storage 0.88 0.30 0.36 0.30 Tank
- 3. Diesel Generator 0.91 0.24 0.43 0.30 Oil Cooler
- 4. Reactor Vessel Core 0.99 0.31 0.33 0.34 Geometry
- 5. Control Building 1.00 0.24 0.33 0.39 Diaphragm
- 6. Contro1 ~ Rod Drive 1.00 0.30 0.38 0.33 System
- 7. Service Water Pumphouse 1.30 0.24 0.49 0.39 Sliding
- 8. Engineered Safeguard 1.70 0.23 0.43 0.57 Features Building Shear Wall
- 9. Containment Crane Wall 2.20 0.39 0.38 0.62 (ContainmentBypass)
Structural Failures Structural failures have been shown to contribute to seismic risk in the plants studied. This is because the failure of a critical structure can compromise the safety of several redundant components simultaneously. In general, the risk con-tributing structural failures have median capacities and high confidence, low frequency of failure values well above the design SSE level. However, the relative contribution to seismic risk of a structural failure is a function of the capacities 13-9 7
y y
y
of 'other components that perform critical safety functions. Structural failure modes identified in these PRAs are shear wall failure, impact between buildings, failure of roof or floor slabs, and several soil failure modes.
Shear Wall Failure. A review of the ground aceleration capacities of shear wall failures identified in the risk studies indicates (Table 13-7) that the shear wall structures designed to current regulatory criteria have considerable margins against the safe shutdown earthquake peak ground acceleration. It should be pointed out that the failutt of shear wall structures was assumed to occur when inelastic deformations of the structure under seismic load were estimated to be sufficient to potentially interfere with the operability of safety-related equipment attached to the structure. These limits on inelastic energy absorption capability (ductility limits) were judged to correspond to the onset of significant structural damage.
This damage level represents a conservative lower bound on the level of inelastic structural deformation which might interfere with the operability of components (1:e., piping and equipment) attached to the structure.
Table 13-7 SHEAR WALL ACCELERATION CAPACITIES Median Ground High Confidence Acceleration
-Low Frequency SSE Capacity Acceleration Shear Wall (g)
(g)
(g)
Remarks Zion Aux. Bldg.
0.17 0.73 0.28 Service water pumps had the lowest capacity Indian Point 3 0.15 1.20 0.63 Other components had lower Control Bldg.
capacities Limerick Reactor 0.15 1.05 0.42 Enclosure and
~
Control Structure Millstone Aux.
0.17 1.40 0.39 Not in the core damage Bldg.
sequences Millstone ESF 0.17 1.70 0.57 Other components had Bldg.
lower capacities 13.
-m--
w
As mentioned earlier, shear wall failures have been included in the risk models of these PRAs not because they have low capacities. Rather, their capacities have simply been cogarable to other critical safety-related components. Even the Zion auxiliary building shear wall (the lowest capacity shear wall in Table 13-7) has a median capacity approximately four times the SSE. The high capacities of shear wall structures arise from the fact that current design criteria do not explicitly rec-ognize the inelastic energy absorption capability of such structures (1_).
Impact between Buildings. In plants founded on alluvial soil, impact between buildings was seen to occur at ground accelerations larger than the SSE accelera-tion. In fact, at Indian Point Unit 2, the dominant contributor to seismically initiated core damage and release was initially identified in the PRA as the failure of the control building due to igact with the Unit 1 superheater building, f
At the control building roof line, there initially was approximately a 1-1/2 inch space between the two buildings, thus, during a large seismic event (significantly i
exceeding the SSE), the flexibility of the upper story of the control building could result in deflections that could close this gap; the resulting impact loads could cause distortion of the steel roof and decks and could conceivably cause the 4
roof structure and decking to fall into the control room. The system analysts had modeled the loss of the control room as leading to loss of control and severe core damage. Since this scenario had a high frequency of occurrence, the plant owners 5
l decided to make structural modifications that would increase the control building
- seismic capacity and therefore, reduce the sevem core damage frequency. The 1-1/2
~
inch roof gap was increased to a minimum of 3 inches and rubber bumpers were l
provided in a portion of the gap to absorb the shock of the impact. The median acceleration capacity of the control building due to this modification increased from 0.279 to 1.0g. This exagle illustrates the potential use of seismic PRA in identifying weak links in the plant safety system and how relatively inexpensive modifications can enhance the safety. If the gap between buildings is small at the SSE level, seismic margin studies should focus attention on this aspect.
Roof / Floor Slab Failures. Failure of the crib house pug enclosure roof was l
considered to be a possible contributor to seismically-induced severe core damage at Zion. This failure is expected to result from loss of diaphragm action of the roof from shear failure initiating from the roof cut-outs. The mason that this failum was considered to be a contributor to the core damage frequency is the conservative assugtion that the entire roof would collapse and disable all the service water pumps. Sensitivity studies (3_) have shown that the conservatism of l
this assumption does not significantly affect the results of the PRA due to the i
13-11 i
i t
m-,-r
- ~.
---,~-.n.,,,,n..--+n--.m--,..
,----r----,-----v,-,--------_na
,--,-..,_,n.-,..-.,,,-
~
l existence of other low capacity components whose failures lead directly to core damage.. However, from a seismic margin standpoint, attention should be paid to situations where loss of diaphragm action may occur, due to the potential for cata-strophic results. Similarly, the control building at Millstone Unit 3 housing the reactor protection and electric power systems, control room and cable spreading area.is judged to suffer gross structural failure as a result of diaphragm failure in a floor slab perforated by openings for a stairwell and ducting.
J Foundation Failures. Foundation failures in the form of sliding, base slab uplift and wall footing failure have been important in the seismic PRAs of Zion and Millstone Unit 3.
Soil failure beneath the foundation slab of the reactor building at Zion was judged to result in the failure of interconnecting piping between the containment and other structures and also lead to impact between the reactor building and auxiliary building.
In Millstone Unit 3, the capacity of the emergency generator enclosure building which contains the emergency diesel generators and related equipment was governed by wall footing failure and sliding. The wall footing failure was judged to occur from out-of-plane wall reactions on the strip footing exerted by the laterT1 earth pressure loading-under seismic excitation. The median ground acceleracion capacity 4
for this failure mode was estimated to be 0.88g and the high confidence, low frequency of failure acceleration level for this mode was calculated to be 0.30g.
Hence, the structure has significant capacity above the design SSE level. By virtue I
of the fact, however, that it is one of the lowest capacity components and since it was conservatively assumed that the wall footing failure would lead to failure of the diesel generators (due to failure of attached piping and instruments), the wall footing failure is the most significant contributor to severe core damage for Millstone Unit 3.
Electrical Equioment Postulated seismic failures of electrical equipment (i.e., diesel generators, batteries and racks, motor control centers, and switchgears) have played an important role in the seismic PRAs. In the following, the causes for this import-ance and the conservatisms and uncertainties inherent in the fragility evaluation of electrical equipment are discussed.
/
13-12
i, Diesel Generators. In the plants studied, the diesel generators themselves have not been the low seismic capacity components. Failures have always been associated with the peripherals of diesel generators. Examples include the diesel generator j
fuel oil tank at Indian Point Units 2 and 3, the diesel generator oil cooler at Millstone Unit 3 and the diesel generator heating and venting at Limerick.
Typically, structural failure modes (i.e., anchor bolt failure and support failure) have been the causes for these component failures. The assumption that these I
failures are completely disabling and that failure of the peripherals will lead to failure 'of the diesel generators is conser+ative.
In the Limerick and Millstone Unit 3 risk assessment studies, the possibility of random failure of the diesel generator system to perform on demand given a seismically-induced loss of offsite power was considered. Fer Limerick, the mediar non-seismic failum frequency of the emergency power system wss considered to be 1.0x10-3 This is a significant risk contributor and dominates the seismic loss of offsite power accident sequences since the seismic capacity of the diesel generator system is relatively high (the lowest capacity component is the heating and venting
'with a 1.55g medir.n capacity). It should be pointed out, however, that if relay chatter and trip are taken as failure for the emergency power electrical components, the random failure frequency may not be dominant due to the low seismic capacity of these components. For Millstone Unit 3..the median non-seismically-induced frequency of failure of the emergency power system was taken as approximately 2x10 per demand and the median seismic capacity of the diesel generators is approximately 0.99 (governed by either the diesel building wall footing failing - 0.88g median cipacity - which is assumed to damage attached equipment; or the diesel generator oil cooler failing - 0.91g median capacity). Due to the relatively low random frequency of failure and the relatively low seismic capacity of_ the diesel generator i
system, the seismic failum is the dominant contributor for the seismically-induced loss of offsite power accident sequence for Millstone 3.
In the internal event PRAs, it has been found that the accident scenarios of random loss of offsite power along with the random failure of diesel generators, the so-called station blackout scenario, has a significant contribution to the core damage frequency. Since the frequency of seismically-induced core damage through the station blackout scenerio is generally smaller (due to the mlatively low frequency of the seismic initiating event), requiring a very high seismic capacity for the diesel generator system does not necessarily reduce public risk.
^
l i
1 13-13
Batteries and Racks. In the Zion and Indian Point Units 2 and 3 Probabilistic Safety Studies,125V DC batteries and racks were identified to be among the key i
equipment items with relatively low median ground acceleration capacities. However, their f aihees were not found to be the most significant contributors to severe core
{
' concern with the battery racks arises because they have been identified as inade-damage or release due to the dominance of other lower capacity components. The quately braced or anchored to the floor. in some of the older plants. This is an inportant point to consider in the seismic margin evaluations. To the best of our knowledge, however, all of these less rugged battery racks have now been upgraded.
s Motor Control Centers and Switchaears. Failures of these electrical components could occur either due to relay chatter and trip or due to structural failures (e.g., anchor bolt or support failures). Structural failures occur at much higher
+
ground accelerations compared to the levels at which relay chatter or trip would occur.
i In the PRAs completed to date, the system analysts have assumed that the relay a
trips caused by chattering are recoverable. Because of this assumption, these electrical equipment failures have seldom dominated the core damage or release frequencies. Although the reset of the system may be readily possible at the control room under certain circumstances, there are those relay trips which may require resetting at local panels. This raises the question of reliability of operator action under the unfamiliar and stressed conditions of a seismic event.,
Failure of operator action is equivalent to a relay failure. Yet, it may be too 4
conservative to concede operator failure because the fragilities of electrical i
equipment for the relay chatter failure mode are generally derived from sparse data. The actual capacities may be'much larger but harder to demonstrate without resorting to extensive fragility testing. Much further work is needed to properly treat the issue of relay chatter in the seismic PRAs.
l Tanks Tanks appear to possess relatively low seismic capacities. Their failure modes I
include failure of anchorage, loss of support, and tank-wall buckling. Typically, j
flat-bottom, ground mounted steel tanks have the lowest seismic capacities. The j
high confidence, low frtquency of failum acceleration levels of these tanks are, in some instances, less than 0.29 In the past, these tanks were designed using the procedure outlined in TID 7024 wherein the seismic forces were calculated based on I
the zero period accele. ration rather than the spectral acceleration. This method is clearly unconservative. However, failures of tanks have not proved to be dominant 13-14 N
--.,.--.-..,-,.,,..,...,.,,_-.,,,n..
_,n_____,,____
~
in risk calculations because failure of a single tank will not, in itself, msult in severs core damage. From a seismic margin standpoint, attention should be given to tank anchorage and the seismic criteria used in its design.
I
.Pjggs.
- Pumps have not been shown to be low capacity components except in the case of
. service water pugs at Zion. The qualification reports were not available for these pumps at the time of the Zion seismic PRA, hence, the estimated median ground acceleration capacity is conservatively low and these pumps should probably not have been identifiec as a low capacity component. In addition, data on the perform-ance of large pumps in industrial facilities during earthquakes have shown that horizontal (shaft) pugs have not suffered any damage; however, laterally unsup-ported vertical shaft pumps might potentially be vulnerable in earthquakes. Frem a margin standpoint, pumps need not be considered for further studies except t
possibly for the shaft on vertical pumps.
i
~
Reactor Pressure Vessel and Internals Reactor Pressure Vessel. In Limerick, the reactor pressure vessel is judged to fail through loss of the upper support bracket at a median ground acceleration of 1.25g.
The loss of upper support allows the vessel to move back and forth more freely
]
during an earthquake. Since them is little clearance between the steam pipes and l
the pipe whip restraints when the pipe is in the hot condition, it was assumed that all four steam lines will sever and create a LOCA significantly larger than that which the containment was designed for. In the Limerick Severe Accident Risk i
Assessment, this accident sequence dominated the risk of early fatalities.
Studies performed in the NRC-funded Lawrence Livermore National Laboratory's Load Combination Program (4) have shown that the reactor pmssure vessel supports gener-ally have very large margins against SSE since these supports are designed to with-stand large pipe break loads.
Reactor Internals. Failum to insert control rods has contributed in the risk estimates for Limerick and Millstone Unit 3.
In Limerick, this failure results from seismically-induced yielding of the annular plate that provides lateral stability to the sh'roud and the seismically-induced yielding of the shroud support 1
cylinder. It is estimated to occur at a median ground acceleration of 0.679 with a high confidence, low frequency of failum acceleration level of 0.25g. However, for this sequence to propagate into a core damage accident, failure of the standby l
13-15
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liquid control (boron injection) system must also occur. Failure of the SLC system is dominated by a non-seismic random failurt frequency of 10'2 per demand. Hence, the frequency of core melt due to inability to shut down the reactor is approxi-mately two orders of magnitude lower than the failure frequency of the control rods.
Due to the relatively low capacity of the control rod system, however, failure to shut down the reactor may still be a risk contributor (depending upon the hazard curves) since it contributes to the core damage fragility at accelerations below 0.5g. For Millstone Unit 3, failure of the scram system must occur along with failure of either the diesel generators, service water pugs or RWST in order to result in a plant damage state. Since there are lower capacity components at Millstone, the failures of which can directly lead to cora melt (assuming a seismic-induced loss of offsite power has occurred), the importance of the scram system is somewhat overshadowed. The failure of reactor internals has not generally been shown to be a dominant seismic risk contributor; however, due to the nature of the funtion it performs and the severity of accident sequences involving failure to scram it should not be overlooked in seismic margin reviews.
Systems Interactions Seismic capacities of critical equipment in a plant are predicated on the assumption that the equipment is not exposed to igact by falling debris or collapsing masonry walls. If there is unqualified equipment, block walls or ceiling fixtures in the vicinity of the critical equipment, their failures in earthquakes should be con-sidered. These system interactions have, for the most part, not been considered in past PRAs. A study is curmntly underway, however, to assess the igertance of potential systems interactions on the Indian Point Unit 3 PSS results (5.).
Ceilino Fixtures. Control room ceiling design at Indian Point Units 2 and 3 was initially such that the transite panels could fall onto the control room equipment.
Modifications to the ceiling design were iglemented thereby eliminating this problem.
Unreinforced Block Walls. In order to provide fire protection to the plant systems, unreinforced freestanding masonry block walls (fire walls) have been erected in many nuclear power plants. Collapse of these walls into critical equipment may pose a real threat to the plant safety in an earthquake. This is an exagle where an attempt to improve the safety against one kind of hazard may have reduced the over-all safety of the plant. In Indian Point Unit 2 (and Unit 3 prior to structural modification), it was observed that the seismic failure of masonry walls in the battery room was a dominant contributor to the frequencies of serious release. In 13-16
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+
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recent years, there has been an industry-wide effort to qualify and strengtheh these l
block walls. However, any seismic margin review should look into the impact of
)
seismic failures of block walls on adjacent safety-related eouissent.
Containment I
Seismic capacities of containment are generally very large such that seismic-I
. induced containment failures have not been dominant seismic risk contributors in PRAs. The early failure of containment (leading to early fatalities) used in the risk models is due to the overpressure caused by core melt accidents. The seismic capacity of the containment, generally has little bearing on its overymssure capacity.; Seismically-induced core melt accidents capable of causing early contain-ment breach are postulated to occur as a result of failures of supports of major components of the nuclear steam supply system such as the steam generator and the reactor pressure vessel. In the PRAs studied so far, these support failures in a seismic event have higher conditional frequencies of occurrence than that for the l
containment. Therefore, increasing the seismic capacity of the containment would not result in a reduction of public risk and would clearly be unnecessary.
1 i
f.19.ing Although a major enphasis is placed on the analysis and design of piping in nuclear power plants, piping systems have seldom been shown to be important contributors to seismic risk. This is the result of the stringent requirements to which piping is l
designed. Also, earthquake experience data shows that piping in industrial facili-l ties which has little or no seismic design has performed well during earthquakes (note, however, that industrial facility piping is significantly more flexible than in nuclear facilities and this may partially account for, its good performance des-pite the lack of explicit seismic design). Typically, piping frag 111 ties are calcu.
lated in the saismic PRAs on a generic basis. Plant specific analysis may reveal further conservatisms in piping design.
l Desien and Construction Errors t
Seismic PRAs have not included the effects of gross design and construction errors.
As such, the insights gained and lessons learned in these PRAs may have to be reassessed if potential for gross errors exist. Some limited sensitivity studies l
Q) have shown that only gross errors of very large magnitude may have any observ-l i,
l able influence on the seismic risk estimates. From a seismic margin standpoint, one l
f should examine the governing failure modes of each critical component to see if a gross error of plausible magnitude might affect its seismic capacity. Such 13-17
publicized errors as minor misp1acement of pipe supports, wrong grade of material used in members subjected to pure compression (i.e., governed by buckling), or
~
improper placement of anchor bolts for a column subjected to essentially compres-ston loads will not significantly reduce the capacities of critical structures and equipment so as to markedly increase the seismic risk.
Seismic-induced Goerator Errors In most seismic PRAs, the analysts have assumed that plant operators perform the needed safety functions during or inmediately after a large earthquake. However, the occurrence of an earthquake of sufficient intensity to damage reactor systems will initially disturb the perfomance of the operators and raise doubts in their minds about the performance of instrumentation and controls. The operators may not know how to respond to seismic failures - spalling concrete, falling ceiling fixtures, ruptured tanks, etc. In the Limerick Severe Accident Risk Assessment, the error rates for operator actions required during seismic accident sequences.
were increased by a factor of 10. This is a first step in recognizing the in-creased potential for operator errors during earthquakes. The topic of operator performance during earthquakes deserves much further attention.
50mARYANDC0$'LU510NS This paper has reviewed the published PRAs with a viewpoint of identifying the dominant seismic risk contributors and of documenting the insights gained in performing these PRA studies. For the PRAs examined, it was found that all components important in the plant risk analysis had high confidence, low frequency of failure levels approximately twice the plant design SSE and median ground accel-eration capacities significantly greater than this. Thus, it is clear that current seismic design practices for nuclear power plant structures and equipment are very conservative.
Based on the review conducted here, we recommand that the following items be focused on during a seismic margin review.
Active Electrical toutoment. The margin review should evaluate the ground accel-eration, at which relay chatter will occur and discern, what the consequences are in terms of relay trip. Recoverability of the relay trips should then be examined.
Particular attention must be paid to active electrical equipment in plants designed to post-1973 standards in which structures have high seismic capacities, such that electrical equipment may dominate the risk.
13 18
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Detailed Plant Walkthrouch. The plant walkthrough by experienced engineers should concentrate on the anchorage of mechanical and electrical equipment, and potential systems interactions caused by unqualified and unenchored equipment and ceiling fixtures in the vicinity of equipment critical to safety.
Masonry Wells. If there are masonry fire walls around batteries and other electri-cal equipment, their stability during earthquakes should be examined.
Tgh3,. Particular attention should be paid to the design criteria and anchorage of flat-bottom steel storage tanks, bearing in mind, that if there are redundant tanks or systems it is only necessary to be concerned with the higher capacity system.
Diesel Generators. The design criteria and qualification methods used for the diesel generator peripherals (e.g., fuel oil tanks) and their anchorage should be examined. It should be kept in mind, however, that random failure rates for the diesel system make it unwarranted to try to achieve particularly high seismic capacities for diesel gener,ators.
Structures. If the plant was designed to pre-1973 standards, the seismic design criteria should be reviewed to identify any weak spots in the structures, and the potential for adjacent buildings to impact should be examined.
Ng}$_Juggdj,. The seismic margins available in the major component supports of N555, that is, RPV, steam generator and reactor coolant pump should be assessed to assure that they are indeed large.
Reactor Scram System. The ability to insert control rods during a seismic event should be ensured by conftming that the seismic margins of reactor internals and control rod drive mechanism are large.
Containent. The seismic capacity of the containment should be reviewed to ensure that it is at least as great as the lowest capacity of the major NS$5 component
, supports. If this capacity can be demonstrated, then seismic failure of the con-tainment will not contribute to the risk. Early fatalities will be the result of early overpressure failure of the containment, thus, increasing the seismic capa-city of the containment will not reduce the risk to the pubite, a
13-1g w
e For seismic margin studies of modern plants for earthquakes below 0.39, it is concluded that it is only necessary to address relay chatter and trip, anchorage of equipment, systems interaction and flat-bottom metal tanks.
. Finally, the seismic PRAs any be made more credible by developing methods to treat relay chatter and trip, and by performing fragility tests on electrical and active sechantcal equipment.
REFERENCES
- 1. R. P. Kennedy and M. K. Ravindra, " Seismic Frag 111 ties for Nuclear Power Plant Risk Studies", Nuclear Eneineerine and Desien, Vol. 79, No.1. May,1984, pp. 47-64.
- 2. M. K. Ravindra and R. P.-Kennedy, " Lessons Learned from Seismic PRA Studies".
Proceedines of Seventh Conference on Structural Mechanics in Reactor Technoloey, Chicago, IL, Paper M 5/4, August, Issa.
- 3. M. K. Ravindra. H. Sanon, R. H. Sues, and R. O. Thrasher. Sensitivity Studies of Se< smic Risk Models, Electric Power Research Institute Palo Alto, CA, IF-350Z, June,1954.
- 4. M. K. Ravindra, R. O. Campbell, R. P. Kennedy and H. Sanon, " Assessment of Seismic Induced Pipe Break Probability in PWR Reactor Coolant Loop in Seismic Events - Probabilistic Risk Assessments". A5M Publication PVP-Vol. 79, June. 1984
- 5. R. D. Campbell and R. H. Sues, Indian Point Unit 3 System Interaction Study, Structural Mechanics Associates W ort Beach, 9a11rornia, Report No.
12901.12-R1, June,1984.
I as 13 20
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6.
NRC SE!$MIC MSIGN MARGINS PROGRAM PLM october 10, IM4 i
S1 I
l G. E. C atir,gs I
J. J. Jcn uon R. J. Sudnitz j
i t
e l
h I
l 4
k e
1 ACKNOWLEDGEMENTS l
This Program Plan was developed by the Expert Panel on Quantification of Seismic Design Margins
- in collaboration with Lawrence Livermore National Laboratory.
Other contributors were Paul Smith, who formulated an earlier version of the Plan and Abel Garcia and Don Bernreuter who provided valuable assistance in reviewing the Plan.
F
- The Panel is made up of the following members:
R.J. Sudnitz (Chaiman),
P.J. Amico, C.A. Cornell, W.J. HaII, R.P. Kennedy, and J.W. Reed.
14-1
EXECUTIVE
SUMMARY
Recent studies such as the 5eismic Safety Margins Research Program (SSMtP) estimate that seismically induced core melt frequencies come from earthquakes in the peak ground acceleration range from 2 to 4 times the safe shutdown earthquake
($$E) acceleration. Other studies indicate that Seismic Category I structures and PWR primary coolant piping have similar high margin anainst seismically induced
. failure. The performance of conventional power plants in past earthquakes confirm the existence of substantial ' seismic capacity in nuclear power plants. However, from a licensing perspective, there is a continuing need for consideration of the inherent quantitative seismic margins because of, among other things, the changing perceptions of the seismic hazard. A sound, practical seismic margins program, utilizing margins to failure analysis and seismic probabilistic risk assessment techniques would serve to minimize the need for changing regulatory requirements and licensing actions as estimates of.the seismic hazards change. In addition, it can provide a sound bests for confidence in the seismic capacity of nuclear power plants and serve to indicate, if necessary, places where seismic risk could be reduced.
The Seismic Design Margins Program (50MP) discussed in this Plan provides the technical basis for assessing the significance of design margins in terms of overall plant safety and will identify potential weaknesses that might have to be addressed. This, in conjunction with past studies and ongoing validation and fragility efforts, should be effective in resolving the quantification of seismic design margins issues.
A general definition of seismic design margin is empressed in terms of how much lager than the design basis earthquake an earthquake must be to compromise plant safety. In this context, margin needs to be defined at the plant, system / function, structure and component level.
The objective of 5009 is to develop the technical basis to resolve regulatory needs relating to seismic design margin (50M). There are several steps for attainment of this objective:
14-2
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o Define SOM and gather existing information t
o Assess the amount of SOM Identify generic attributes related to 50M o
o Assess the adequacy of 50M l
. Develop SOM screening guidelines, if required o
These steps may lead to generic findings or suggest the need for selected individual plant reviews, the necessity and extent of which will depend on the results of the assessment of adequacy of margins. Idhere at all possible, generic
[
attributes related to SOM will be identified so as to better focus the reviews a provide a basis for generic findings.
The above steps am addressed by the tasks in Part I of the Plan.
In Part !!,
supporting studies are suggested, many of which will be conducted independent of i
50W.
Key tasks in these supporting studies deal with the assessment of new f
infomation being developed about component and structural fragilities.
The relationship of the Part I and !! tasks is illustrated in the SDMP Flow Chart shown in Figure 1 (Page 14-11).
t in the development of this P'lan certain judgements and assumptions were made:
i That nuclear plants built to current regulations in most cases have o
significant margin above SSE but that gaps in knowledge exist. The pumose of this Plan is to close these gaps and confim margin adequacy in a quantitative way.
i That plants, systems and components can be grouped for the purpose o
of studying 50M.
That the question of adequacy of SOM will be posed and answered in o
relation to plant risk as opposed to strictly component risks but both probabilistic and deteministic techniques will be used to
~
analyse how much SOM exists.
That the NRC will need to detemine the adequacy of SOMs.
o That requiring plant-specific seismic PRAs as the principal vehicle o
for analyzing SOM is not desirable and that a less extensive plant l
(orplantgroup)reviewprocessispossibleifsuchareview process turns out to be needed.
i A schedule is presented in this Plan to complete the work in three years. This i
schedule depends on the results of the researth and IWic findings about SOM adequacy.
I To expedite the effort a three-phased approach to the Part I tasks has been developed.
Phase ! calls for a preliminary assessment of margin adequacy and guidelines development within nine months, Phase !! calls for two trial plant reviews based on these guidelines and Phase !!! is the implementation of further plant reviews and continuing studies as required.
('
14-3 e.
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Section 14 i
NRC SEISMIC DESIGN MNtGINS PROGRAM PLAN G.E. Cummings, J.J. Johnson, R.J. Budnita I'
l
~ A.
INTRODUCTION S$!!EA1 Recent studies such as the Seismic Safety Margins Research Program ($$MRP) t estimate that seismically induced core melt frequencies come from earthquakes in
[
the peak ground acceleration range from 2 to 4 times the safe shutdown earthquake i
(S$E) acceleration. Other studies indicate that Seismic Category I structures and NR primary coolant piping have similar high margin against seismically induced failure. The performance of conventional power plants in past earthquakes confim l
the entstence of substantial seismic capacity in nuclear power plants. However, from a licensing perspective, there is a continuing need for consideration of the inherent quantitative seismic margins because of, among other things, the changing l
potteptions of the seismic hazard. A sound, practical seismic margins program, utilizing margins to failure analysis and seismic probabilistic risk assessment techniques would serve to minimize the need for changing regulatory requirements j
and licensing actions as estimates of the seismic hazards change. In addition, it can provide a sound basis for confidence in the seismic capacity of nuclear power j
plants and serve to indicate, if necessary, places where seismic risk could be reduced.
The Seismic Design Margins Program ($0MP) discussed in this Plan will provide the technical basis for assessing the significance of design margins in tems of j
l overall plant safety and will identify potential weaknesses that might have to be
~ This, in conjunction with past studies and ongoing validation and addressed.
l l
fragility efforts, should be effective in resolving the quantification of seismic l
l design margins issues.
l' l
A general definition of seismic design margin ($0M) is empressed in terms of how such larger than the design basis earthquake an earthquake must be to compromise j
i 14 4 i
L J
the safety of a plant. Margin is defined at the plant level ar.d at the level of functlon/ system, structures, equipment and components.
Regulatory Needs At the June 11, 1984 joint meeting of the fetC Staff Working Group on SOMs and the tapert Panel on SOMs* there was an extensive discussion of regulatory needs.
The regulatory needs were distilled to the following:
1.
h re is a need to understand how much SOM exists. Margin in this context is expressed in terms of how much larger than the $$E an earthquake must be to compromise the safety of the plant.
2.
There is a need to create a seismic margin framework that can filter, and to some extent absort, the effects of changing knowledge and hypotheses in geology and seismology. This framework is needed to provide an engineering perspective and to avoid, when possible, over-reaction to these changes.'
3.
1 tere is a need to understand the influence of design and construction errors, systems interactions and effects of operator behavior on the seismic response of plants.
4.
There is a need for researth to understand the behavior of plants under loads induced by low-magnitude earthquakes characterized by high fre motion. quencies, short durationIt is recognized that thand highly localingd groun response of the plants may be qualitatively different for these earthquakes than for those for which the plants are designed. Them is also a need to put into perspective the significance of increased high frequencies (above 10 Hz) for larger earthquakes.
5.
There is a need to provide additional assurance concerning the validity of the models and input data now used in seismic probabilistic risk assessments (PRAs) so as to increase confidence in the validity of PRA results.
The hierarchy is that the first need is the most important and the remaining four are secondary.
The Panel is made up of the following members: A.J.Sudnits(Chairman),
P.J. Amico, C. A. Cornell, W.J. Hall, A.P. Kennedy, and J.W. Reed.
14-5
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The first Regulatory Need is too broad to serve the purpose of defining scope and should be further specified. Review of this draft plan should lead to a more focused description and highlight other potential needs. This iteration process is necessary to assure a successful, balanced, cost-effective and responsive SDMP. Desirable inputs would include a current list (last two years and in the near future) of regulatory concerns and issues relating to seismic margins. This list should be constructed by the NRC staff and updates to it made as the 50MP progresses. Typical current issues might be:
Margin implications of recently recorded high frequency features of e
earthquakes such as measured in the New Brunswick earthquake.
Margin impitcations of diffemnces found in earthquakes recorded at e
anomalous sites such as shallow satt sites.
Margin implications of recently recorded earthquake spectra on the e
design assumption that the peak vertical acceleration is 2/3 the peak horizontal acceleration.
The central regulatory issue is that the safe shutdown earthquake (SSE) used for the design of plants can be exceeded with finite probability. This exceedan;e is i
due to a variety of reasons: 1) the SSE has a finite return period and thus larger earthquakes are expected but with longer return periods and 2) the shape of design spectra can be exceeded. The basis for the adequacy of the seismic design of plants thus cannot rest on the size of the SSE alone and must also rest on there being adequate 50M.
The criteria used for plant design are known to embody SON which in most cases is believed to be large. However, this SOM primarily arises from prescriptive procedures rather than performance requirements that specify the various margins quantitatively. This means that the existence and sources of SOMs are generally known but their quantitative values are generally unknown. Since quantitative SOMs are unknown, a natural regulatory question isi "What minimum level of earthquake will compromise plant safety and where are the weakest links?"
50MP is intended to take the next step. It will quantify the earthquake levels that could compromise plant safety as part of the process of assessing 50Ms. To the extent possible this will be done by quantitative studies that will be planned to develop results with generic imp 1tcations. To the extent that this generic work may f all short, SOM screening guidelines will be developed. These guidelines 14-6
__...-__r
,.-.._..-..___,__,____.,-.,,.._.,,,,,.__.-..,__.__,,.,__.c.-
,_.-----,,7._-m__-
will be used to assess the adequacy of SOMs through various types of plant-specific reviews. Although some quantitative results on SOMs do exist, they are not based on sufficiently broad and varied studies to meet MC needs adequately.
i Strateay and Assumotions The overall strategy for the 50MP relies upon a preliminary set of conclusions that seem to be a consensus in the casuunity of knowledges:)1e experts familiar with seismic design margin issues. Perhaps the most ingwrtant consensus is a confidence that reactors designed, built and operated according to the MC's curmnt regulations in most cases possess significant margin above the SSE levels for which they have been designed -- stated another way, there is a high degree of confidence that earthquakes must be significantly larger than the SSE before they will compromise the safety of the plants.
l This confidence, which the 50W hopes to confim through specific studies (but is prepamd to fall to confim depending on hfw the studies turn out), has been taken into account in the development of the Plan. There is a conviction not only that such margin exists, bu't that it should be possible to demonstrate the existence of this margin quantitatively. Moreover, the strategy of the Plan is based on the assumption that it is possible to gr,25. the ensemble of plants into a manageable number of sub groups, characterized by similar properties, such that statements about 50M will be feasible for each sub-group separately.
If these consensus opinions and convictions are borne out by the studies contemplated, then the result will be statements about how much 50M exists for each sub-group of plants. It is recognized that the statements about 50M j
resulting from this work cannot be comprehensive -- that is, they cannot cover All issues involved in the plants' seismic responses. In particular, there are a few issues (discussed separately below) for which new research is needed before their effect on margin can be stated confidently. Nevertheless, there is the espectation that there will be groups of plants, and groups of plant attributes (suchasgroupsofstructuresorgroupsofequipmenttypes),wherestatements about 50M can be made confidently.
The Plan also rests on several answettens made as a starting point for the Plan's development. These assumptions may change as the work of the Plan evolves, and as the irsut of other knowledgeable parties is factored into the Plan.
i 14 7
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1.
We assume that both deterministic and probabilistic techniques will be used to analyse how much SOM exists, and as tools in the SM screening guidelines to be developed as part of the Plan. We also l
assume that plant risk, usine the traditional risk end-points of 1
PRAs (corewealt frequency, offsite risk, etc.) will be used as the i
figure-of-serit for determining that plant safety is compromised in the S M Analyses.
1 2.
We assume that plants, systems and c s.can be grouped j
usefully for the purpose of studying
?
3.
We assume that eutdelines w111 he required to conduct plant reviews I
in the event 50R adequacy cannet be resolved in a generic manner, j
l 4.
We assmo that requiring elant-specific seismic PRAs as the i
principal vehicle for analysing 3M at various plants is not a desirable solution to the task of finding a screening method for i
50M. We assume that an approach can and will be found involving I
less extensive analysis, although it is possible that seismic PRAs 1
may be needed for same plants to provide a piece of the required technical information.
5.
We asses that those accident seguences that are principal contributors to the seismic part of plant risk can be identified in a generic way insofar as there is any pattern identified among the
- j plants.
4.
We assume that during the execution of the 50MP the validity of I
seismic PRAs will be established sufficiently to pemit confidence
[
in the conclusions based on their use.
f i
5.
OBJECT!W TheobjectiveoftheSeismicDesignMerginProgram(50MP)istodevelopthe technical basis to resolve $0M issues. This will be accomplished through specific studies using both deterministic and probahtlistic techniques. The 50MP has several goals:
i e
To define hierarchical relationships of margin at the plant, function, system, structure, equipment and component lent.
a e
To assess the amount of SOM.
1 e
To identify generic attributes related to SOM.
4 l
T. d.i.,mi ih.
quacy of s0M's.
P e
To develop SOM screening guidelines, if necessary.
this 50W Plan is to provide a comprehensive approach (set of tasks) to address i
the 50MP objective and goals.
i i
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j
4 C.
SCOPE OF WORK 4
The scope of the 50pr has been developed in two parts. Part I encompasses the body of the Program within s1x tasks. Part !! tentative 1y idontiffes efght tasks which will provide information to help resolve 50pr issues. Part I is split into three phases.
The firat phase 1: an intensive offort of about s1x months duratton leading to a prettainary assessment of margin adequacy and a set of trial guidelines.
In Phase !! trial reviews of two plants will be accomplished.
Phase
!!! continues with further plant reviews and studies dependent on the results of Phases I and !!.
Part 1:
Assess Margin - Develop Guidelines - Trial Review of Plants Phase !
Task !.1 Assess Existing Information Task !.2 Estimate Existing Margins Task !.3 Identify Generic Attributes Task !.4 Assess Margin Adequacy Task 1.5 Develop Screening Guidelines and Methods for their Application Phase !!
Task 1.6 Conduct Trial Plant Reviews Phase !!!
Implementation of Plant Reviews and Continufng Studies Part !!:
Identification of Information Needs Task !!.1 Assess New Failure Data Task !!.2 Assess New Margins Information Task !!.3 Relate Capacity and Performance of Aelays and 8reakers During Strong Motion to Margin Issues Task !!.4 Aelate the Behavior of Operators During and !sundtately After Streep Motion to Margin Issues Task !!.$
Assess the Contribution of Design and Construction Errors to the Compromise of Safety Task !!.6 Assess Inherent Calculational Design Margin (best estimate vs.destencode)
Task !!.7 Assess the Impact of System Changes on SOM Task !!.8 Assess the !spect of Mon.1tnear $tructural Behavior on Margin Issues The relationship between these vertous tasks is 111ustrated in the 50MP Flow Chart shown in Figure 1 and the 50MP Schedule shown in Section E.
The timing is aimed at resolving NAC concerns about seismic design margins in three years. To do this effectively, two trial plant reviews will be done during the first year with further work to be defined following these trial reviews. Following completfon of 14 g
l Tasks !.1 - !.4 the NRC staff must establish, either fimly or tentatively, criteria whereby adequacy is to be judged, building on the work of Task !.4, before screening guidelines can be developed in Task !.5.
Also. NRC needs to comment on the screening guidelines developed in Task I.5 before the trial plant reviews can be conducted in Task 1.6.
The infomation needs identified in Part !!
must be available in a timely manner to allow completion of the Part I tasks.
14-10
Figure 14-1 SDMP FLOW CHART, PART I I.1 A
Assess Existing l
Information Part II I.2 I.3 Estimate Identify Existing Generic Identification of Information Margins Attributes Needs
\\
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II.1 Assess New Failure Data Phase'I, II.2 Assess New Margin Info.
I.4 II.3 Relay / Breaker Performance Develop Inputs for the II.4' Operator Behavior Determination of II.5 Design & Constr. Errors Margin Adequacy II.6 Calculational Margin II. ? Systems Margin 11.0 Nonlinear Struc. Behavior etc.
NRC Input 7
I.5 Develop Screening Guidelines y
a.{
t A
I.6 Conduct Trial Plant Phase II Reviews A
'l Implementation of Plant Reviews Phase III and Continuing Studies
(
14-11 s'
i O.
TASK DESCRIPTIONS
?
Part I: Assess Marcin - Develoo Guidelines - Review Plants Phase !-
Task I.1: Assessment of Existina Infonnation.
,Backaround. Significant information is currently available relating to SDM 1
issues although it is known that gaps in knowledge do exist. To make sure
~
that relevant existing infomation is utilized it is necessary to identify a task to assess this infomation. The information exists in a variety of sources although it is believed that the information most relevant to SDMP will be found in seismic PRAs. Nevertheless, all sources need to be reviewed.
Results from existing programs may give qualitative if not quantitative insights into the margins issues. Decisions and supporting studies from the Systematic Evaluation Program (SEP) may help in this regard. Also, existing fragility and equipment qualification data needs to be assessed. This data may give insights related to existing margin on equipment and components if not for system and plant margin. Existing earthquake experience data must be reviewed to give insights about margins in structures as well as equipment.
[
Ten to fif teen seismic PRAs will have been performed on plants in the near.
tem. These PRAs are usually an adjunct to an internal events PRA and are a likely source of infomation for the determination of the adequacy of SDMs in risk tenns. These plants and their PRAs are also a likely source of infomation for the development of SDM screening guidelines.
The SDMP approach requires a consistent set of quantified seismic PRA results. At a minimum we anticipate the use of MtC, utility and EPRI developed seismic hazard functions which will require re-quantification of the existing seismic PRA results. The sensitivity of PRA results to this hazard function will have to be tested. We also anticipate the need to exanine closely the development of the plant logic models and fragility and uncertainty descriptions and the possible need to modify some of these relative to their characterization in the existing seismic PRAs. Any modifications will lead to a need for some requantification. Also, the development of SDM screening guidelines will require a close interrogation of the existing seismic PRAs in the fonn of sensitivity studies, uncertainty 4
f 14-12
analyses and the evaluation of various alternate corIfigurations. One of the seismic PRAs to be reviewed will be the SSM.'t? Zion Study since this is the
. most comprehensive analysis and was conducted with NRC instead of utility funds. Thus, infomation from this study is not only the most detailed but also the most readily available. Utility seismic PRAs such as were done on Millstone 3. Seabrook, Limerick, Oconee and Indian Point 2 and 3 will also give useful insights.
Ob.iective. This task will evaluate existing infomation to extract that infomation useful to SDM issues and has two main objectives:
1.
To provide infomation to be used to identify generic attributes, detemine margin adequacy, and develop screening guidelines (Tasks I.3, I.4 and I.5).
2.
To provide information to help in estimating existing margins (Task I.2). Results from Task I.2 will be used to establish what we know and don't know about existing margins.
Approach. The approach will be to review as much infomation as possible during the first phase of this Program. Further review including requantification of the existing seismic PRA's may be required later depending on Phase I findings. For Phase I, results of all studies including PRAs will be taken as stated with limited interpretation and no requantification. With this in mind, the steps in the review process will be as follows.
1.
Review existing studies relating to SOM issues, e.g. SQUG.
2.
Review existing fragility and equipment qualification data.
3.
Review existing earthquake experience data.
4.
Review existing infomation on hazards analysis, e.g. LLNL and EPRI studies.
5.
Review existing seismic PRAs (approximately 6). In Phase I, quantities of interest will be tabulated and compared on a plant, system / function, and component basis. Such quantities would include coreanelt frequency, important accident sequences, systems / functions, components and structures, and earthquake level at the median failure point. This assessment will be based on the PRA results as presented with any necessary requantification detemined after Phase I.
6.
For each of the reviews the applicability of the results to each of the Part I tasks needs to be assessed and the findings documented in i-a way useful to each task.
14-13 4
r The results of this task will Justification of Task Based on Regulatory Need.
be used by the other Part I tasks to address regulatory needs as stated for those tasks.
Relationship to Other SDi@ Tasks. The results of this task provide input to the other Part I tasks.
Milestones. See Schedule in Section E.
Details to be developed later.
Task I.2: Estimation of Existing Margig Background _.. There is a need to establish to the extent possible what the SDM is in existing plants and particularly to establish where the uncertainties Based on the infomation appear too great or where gaps in knowledge occur.
in Task I.1 an early determination of what these margins are will be made by this task to establish a base of knowledge from which the remainder of SDMP can be conducted.
The objective is to estimate the margins present in existing Ob.iective.
plants by estimating what size earthquake is necessary to compromise plant It is desired to estimate SDMs for the plant as a whole,, as well as,
safety.
at the system / function level, the structure level, and the component level.
Approach.
Identify candidate plants based on Task I.1 results for which 1.
sufficient infomation is thought to exist to enable an estimate of SDMs.
Review the analysis for each candidate plant, and develop additional 2.
specific infomation as needed.
Determining how SDM will be defined in an operational sense for the 3.
purposes of this Task, in tems of specific ground motion The definition of SDM characteristics or other physical parameters.
"SDM is given in the introduction to this SDM Plan is quite general:
expressed in tems of how much larger than the design basis earthquake an earthquake must be tn compromise the safety of a plant".
Determine the SDMs for each candidate plant, including an estimate of 4.
the uncertainties in the deterinination. The SDMs are to be estimated at the plant level and/or the level of functions / systems, structures and components as feasible.
14-14
i 5.
Determine, through existing sensitivity studies or new sensitivity studies to be carried out where needed, the extent to which the Sms calculated above depend on various assumptions, models, and generic data. Particular attention must be paid to the sensitivity of the results to the shape of the hazard curve or response or fragility 1
function curves.
6.
Document the msults.
Results of the Task. The results of this task will be estimates of the sums existing for each of a small group of candidate plants chosen because sufficient infonnation is available upon which to base such a set of estimates.
The SDMs are to be expressed where feasible at the plant, system / function, structure, and component levels. The uncertainties in the SDM estimates are to be presented and discussed, along with insights through sensitivity studies as to where the SDM estimates most depend on various assumptions, models, and generic data.
Justification of Task Based on Regulatory Need. Results of this task will partially satisfy Regulatory Needs I and Z.
4 Relationship to Other SDMP Tasks. This task will take input from Task I.l and supply input to Task 1.4.
Milestones. See schedule in Section E.
Details to be developed later.
Task I.3: Identification of Generic Attributes Background.
To make the assessment of SDM as efficient as possible, l
generic attributes need to be identified. This will help group plants or plant systems so that further review can be better focused.
Objective. The objective of this task is to identify those generic attr1Dutes of the plants studied that seem to be important contributors to plant strength, and those that appear to contain important vulnerabilities to earthquakes.
Focus should also be on identifying systems, structures r
and components which can be eliminated from further investigation.
This is to be accomplished, insofar as feasible, for broad groups of plants and/or broad groups of functions, and/or broad groups of systems j
and components.
The specific groupings are to be one result of this effort.
Approach.
l.
Identify candidate generic attributes that emerge from the studies in Task I.1 and I.2.
2.
For each identified candidate generic attribute, determine the extent to which the attribute is present in each of the plants studied.
The focus of this effort is to achieve a rough grouping of plants and/or attributes.
i 14-15.
I
Group the plants and/or attributes, taking into account the extent 3.
to which each plant group or attribute group possesses a high, Detennine the medium, or low degree of correlation and consistency.
extent to which the groupings seem ' natural' or ' forced' -- that is, whether the groupings seem to arise from some generic property of
- the plants or attributes that might be present in other similar f
plants not studied, or whether the groupings seem not to arise frore any identified generic property.
Estimate the extent to which the groups of plants or groups of
[
4.
attributes might be extendable to include other plants or attributes not specifically present within the analytical infomation set studied.
Estimate the extent to which the conclusions in (3) and (4) above 5.
depend on (are sensitive to) differences in assumptions, models, or Particular attention must be paid to whether the data used.
insights are sensitive to the hazard curves used or shapes of the fragility or response function curves.
4 medium, or low confidence, for example) groupings arrived at (h Estimate the overall confidence in the 6.
, based on the analyses in (4) and (5), and on the confidence thought to be present in the overall groupings.
Results of the Task. The results of this task will be identified generic attributes of plants studied that seem to be important contributors to plant strength, and that seem to be important aspects of plant vulnerability to -
For each identified generic attribute, the task will identify earthquakes.
and discuss the sensitivity of the conclusions to assumptions, models, and data used. An overall confidence rating will be given to the groupings specified.
Justification of Task Based on Regulatory Need. Results of this task are mainly ',o support the remaining Part I tasks.
Relationship to Other 50W Tasks. This task will take input from Task I.1 and i
supply input to Tasks I.4 and I.S.
Milestones. See schedule in Section E.
Details to be developed later.
Task I.4: Assessment of Margin Adequacy.
Based on the results of the preceding tasks, an assessment of.
Background.
margin adequacy needs to be made to determine the necessity for proceeding on to final guideline development or the plant review stage. It may be that margins will be found adequate at this stage or found adequate for some 14-16
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classes of plants, systems or components. Such a finding would eliminate or minimize the need for further. effort. Close coupling with NRC will be
~
necessary in this task since the final judgement as to adequacy of SDM must be made by the NRC. Any required additional studies are included in this task rather than in the preceding tasks.
Objective. The objectives of this task are:
1.
To assist NRC by providing results for their decisions on the adequacy of SDMs.
2.
To provide inputs to the development of SDM screening guidelines.
(Whether or not this objective is needed will dependent on NRC decisions on whether SDMs are adequate.)
Approach.
1.
Interact with NRC to finalize an approach to detemine adequacy based on the experience gained in the preceding tasks.
2.
Review and suninarize the reports from the preceding tasks.
3.
Revise this Summary-or perfom special studies as a result of the review in (2), as required by IRC.
- 4.. Interact with NRC on their decision on adequacy of SDMs, as required by NRC.
Justification of Task Based on Regulatory Needs. Regulatory Needs 1 and 2 would be addressed by this task since existing margins will have been assessed and statements about adequacy made. Also, Need 3 will partially be addressed since in reviewing the seismic risk assessments some feeling concerning the effect of design and construction errors, systems interaction and operator behavior will be made. Also, in going through the process of making statements of margin adequacy, a better feeling will be forthcoming about the validity of the mdels and input data (Regulatory Need 5).
Relationship to Other SDMP Tasks. This task will use input from Tasks I.1. I.2 and I.3 and provide input to Task I.S.
Milestones. See schedule in Section E.
Details are to be developed later.
14-17 l
j Task I.5: Development of Screenino Guidelines.
Background. The purpose of screening guidelines is to help determine how to proceed with plant specific reviews and to help assess the adequacy of SDMs in these reviews. The need to develop screening guidelines necessarily rests on NRC conclusions on the technical results developed in Task I.4.
It may be necessary to revise some of the efforts that were perfomed in Task I.4 to reflect NRC insights that resulted from their decision process on SOM adequacy or to answer various MRC questions that arise. The primary goal of SDP9 at this point is to identify screening guidelines that can be used to a
support an NRC finding that SOMs are adequate or inadequate at a plant for which no seismic PRA exists. It is anticipated that plants which do pass the r
guidelines will be judged to have adequate 50M, but those plants that fail may or may not have inadequate 50M. Additional effort will be required to detemine if the SOM is not adequate, e.g., a risk assessment.
.r
-Assuming some areas are found of questionable adequacy, the key NRC decision I
will be the priority that will be assigned to further plant or topic reviews.
Possible decision areas are plants located at sites expected to have local site amplification, plants with current or earlier seismic criteria, geographical location of plants, Westinghouse versus other PWRs, PWR versus BWR, magnitude of the' risk estimates in the existing PRAs, and so forth.
There are two major factors that inpact the development of SON screening guidelines:
NRC insights as a result of Tasks I.1, I.2. I.3 and I.4 and their e
decision process on the adequacy of SOMs.
A close examination of the differences between the plants or parts e
of plants that were and were not found to have adequate SOMs.
Close coordination with NRC will be required in this task since they play the
~
dominant role in one of these two factors. For example, it is possible to develop guidelines relating to the adequacy of SOMs as follows:
"The seismic loads used in the design of structures shall be shown to,be a factor of 2 or more times the median loads that are expected to occur assuming the occurrence of realistic earthquakes with a peak acceleration equal to the SSE."
14-18
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"The median capacity of a structure (including consideration of inelastic energy absorption) shall be shown to be 5 or more times the best estimate loads that are expected to occur assuming the occurrence of a realistic earthquake with a peak ground acceleration corresponding to the 5 x 104 per year probability level."
It is important to specify areas and forms of guidelines as completely as possible before the SDMP technical efforts in this task begin. This is because:
The area (structural response or capacity in the above example) of e
applicability of the guideline must be viewed as necessary and acceptable to NRC.
The fom of the guideline (factor of 2 or 5, etc..in the above e
examiiTe7 must also be acceptable to NRC as an appr,opriate one for determining adequacy of SDMs in the specific area (structural response in the example).
i Note that in the above examples an additional complication arises.
Specifically, most seismic design and PRA infomation available does not include the best estimate, seismic loads for the structures. These loads would need to be calculated if NRC desired a guideline of this type.
A key issue to consider in developing guidelines is current versus earlier seismic design criteria. This is bccause the SDMs for plants designed to current criteria are thought to be laq cr than for plants designed to earlier criteria. Global guidelines are yperiling as they would simplify the plant-specific SDM reviews significantly, independent of the original criteria I
used.
1-Although desirable, we anticipate that it may not be possible to develop acceptable global guidelines. One of the reasons for this is there are many i
site-and plant-specific features that have a significant impact on seismic risk even when the plants are designed to the same criteria. This is a widely recognized consideration in PRAs for internal-initiated events and seems likely to be true for seismic PRAs also. If this consideration is a significant factor in seismic PRAs then it means that design practice is as or
^
more important than design criteria. Data on the perforinance of non-nuclear 14-19
facilities in past earthquakes tends to confirm the importance of design practice in seismic vulnerability. However, global guidelines may help in grouping plants for consideretion and/or for determining which systems in plants of a certain vintage or vendor should be considered.
We thus anticipate that it may be necessary to develop SDM screening quidelines that are less global and more technical. One problem with such guidelines is that they may require significant efforts by the assessor as
. part of the plant-specific review. Recall that compliance with our example guideline would require the assessor to perform best estimate response analyses. Also, the uncertainty and variability within and between plants needs to be taken into account. If, for instance, the structures do not satisfy the guidelines but the piping does, then additional guidelines or a risk analysis may be required.
To reduce the ultimate burden on the assessor, the guidelines should be developed in such a way that they offer a spectrum of options and/or levels.
For example, for structures, one sequence of such guidelines might be the following:
1.
Structural response margins (as in the example).
2.
Margin against structural yielding.
3.
Margin against structural failure.
NRC staff and utility efforts in (1) could be used in (2) if the guidelines in (1) were not satisfied and those efforts in (1) and (2) could be used in (3) if the guidelines in (1) and (2) were not satisfied. All of this technical effort would be of use in a seismic PRA if the guidelines in (3) were not satisfied.
Ob.iective. The objective of this task is to develop SDM screening guidelines.
The purpose of these guidelines is to assess the adequacy of SDMs in plant-specific reviews and to help structure the type and priority of such reviews.
Approach.
1.
Develop a trial set of screening guidelines to be used in subsequent plant reviews and evaluations. The area and fonn should be specified in the l
guidelines as appropriate.
l 14-20 6
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2.
Develop a procedure for applying these screening gu'idelines for postulated plant reviews. It may be that certain classes of plants need no or minor review (for instance, post-1973 plants), it may be that only certain parts of plants need reviews (reactor coolant loop piping may have adequate margin), or it may be that plants can be grouped into broad classes (by vendor, A/E firm, age, etc.).
3.
Elicit _NRC guidance on the proposed guidelines and method of implementation. This may involve further work in Tasks I.2. I.3, and steps (1) and (2) of this task.
4.
Finalize guidelines.
5.
Document guidelines and review methods.
Justification of Task Based on Regulatory Needs. This task is preparatory to conducting plant reviews.
Relationship to Other 50MP Tasks. This task receives information from Task I.4 and provides infomation to Task I.6.
Milestones. See schedule in Section E.
Details are to be developed later.
Phase II h sk I.6: Conduct Trial Plant Reviews Background. To meet the Regulatory Needs concerning SOM issues, it may be necessary to review plants individually or on some sort of selective group basis. As an example of this, selective review of reactor coolant loops in a few plants within vendor categories helped in resolution of pipe restraint
, issues in the LLNL Load Combinations Research Program. The SOM reviews will be necessary if uncertainty exists concerning SOM adequacy after the completion of Task I.4.
It is now believed such uncertainty will exist because current seismig risk assessments and SEP studies show that unique plant features frequently dominate risk and therefore would need to be looked at to assure margin adequacy. To test this review concept trial plant reviews will be conducted with further reviews implemented in Phase III if found necessary by NRC.
The screening guidalines and reconnendations for their use from Task I.5 will be used to conduct these reviews. Detailed interaction with the NRC staff will be required not only in establishing the guidelines but in implementing the 14-21
reviews.~ The reviews could be done in conjunction with the utilities as was done in the Seismic Qualification of Auxiliary Feedwater Systems Program..
The objective of this task is to conduct trial plant reviews and Ob.iective.
report the results so that the adequacy of margin is established.
Accroach.
Use data from Task I.3 1 l
i process (walkdowns, PRIs,.4 and I.5 to establish a p ant rev ewsys 1.
groupings,etc.).
Catalog the important components and characteristics of each plant 2.
to be reviewed with annotation.
Do a rough assessment of each plant to establish priorities.
3.
Factors to be taken into consideration include generic attributes and work related to other NRC efforts (ISAP, licensing issues, etc.).
Do more detailed reviews if required or until NRC feels the margins 4.
issue is resolved.
5.
Document findings.
Justification of Task Based on Regulatory Needs. This task is expected to satisfy Regulatory Need I relating to the assessment of existing margin and to help satisfy Regulatory Need 2 relating to stability of the seismic review Some infomation relating to the other three Regulatory Needs should process.
also come out of this task. Of course, if 5009 were to be carried to the ultimate extreme of doing complete reviews of all plants including individual risk assessments, all Regulatory Needs would be met. At this time it is felt that reviews of all plants should not be necessary to satisfy all Regulatory Needs.
Relationship to Other 50MP Tasks. This task receives guidelines infomation from Task I.S.
Milestones. See schedule in Section E.
Details are to be developed later.
Phase III:
Implementation of Plant Specific Reviews and Continuina Studies After the trial plant reviews are completed in Phase !!, further plant specific reviews will be implemented. Although it may be necessary to do a rough assessment of all plants, detail reviews on a representative set should be 14-22
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4 sufficient. ~ Phase III of SDMP is the implementation of these additional plant reviews over a two year period. Details of how these can be implemented will be detemined during Phase I and II.
Further guideline development, risk assessment reviews including some requantification and similar work to that done in Phase I may be required to implement these plant specific reviews and to close gaps found in knowledge about margins. This continuing work would be part of the Phase III effort.
14-23
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Part II: Identification of Infomation Needs The_ following tasks are identified as being important to establishing actual seismic design margin including performance of the plant and operator immediately after the occurrence of an earthquake. These tasks are executed concurrently with i
l those of Part I and provide input to the end product, i.e. the ability to make i
statements about seismic design margin for a specific plant or for groups of I
plants: Their execution may fall under a different program either within the NRC or the Industry. However, they are presented here to emphasize their importance to the and objective.
4 Task II.lt Assessment and Development of New Failure Data.
Description.' Considerable seismic qualification testing has been perfonned using snake tables. Some infomal descriptions exist of weaknesses or failurts that were observed during testing. This is contrasted with the data
[
on the observed performance of equipment in past earthquakes where few if any failures have been reported. Obtaining these qualification test results may be difficult since they may be considered proprietary. In addition, other NRC and industry programs are gathering and generating useful data concerning fragility, e.g., Component Fragility Program, Structural Fragility Program.
4 The objective of this task is to obtain general information on failure modes and failure levels of equipment and structures as an input to NRC decisions on the adequacy of SOMs.
i The approach is to engage testing laboratories to develop reports summarizing general information on weaknesses or failures of components observed during 1
testing. Typical information sought would be year of test, general description of component, description of failure or loss of function, f
excitation description, and mounting conditions. The information from all testing laboratories would be assembled for like components and added to the existing fragility data base. Candidate components for additional testing would be identified. Note that significant effort would be expended to assure confidentiality of the data and its source.
I 14-24 t
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&stification of Task 8ased on Regulatory Needs. Regulatory Need 1 is partially addressed in the sense that this task provides limited infomation 1.
on the fragility of equipment. Accurate fragility infomation is the most important factor in the determination of the acceleration range that is the dominant contributor to seismic risk. This range is assumed to be an important factor in NRC's decisions on the adequacy of SOMs. Regulatory Need 5 is partially addressed by this task in the sense of validation of assumptions on equipment fragility.
Relationship to Other SDMP Tasks. This task provides inputs to Tasks !.2 I.3. I.4 and I.5 J
!!.2: Assessment of New Margins Infomation.
Description. The state of knowledge in the fields of seismic risk analysis, component and structural fragility, and systems behavior is evolving at a rapid pace. Many research programs and additional seismic PRAs are underway, or planned, with results expected in the next two to three years. Examples
)
include the NRC Component Fragility Program, NRC Equipment Qualification Program, NRC Category I Structures Program, EPRI Hazard Program EPRI Piping
)
and Fitting Dynamic Reliability Program, foreign research programs (HOR, Japanese French, etc.), ISAP, RMIEP, Diablo Canyon Seismic PRA, etc. These programs will provide vsluable input to the seismic design margin issue and need to' be eitplicitly recognized and included in the 50pF. The timing of i
their results is not compatible with the schedule for Task I.1, hence, an additional task is identified.
The objective of this task is to assess newly developed infomation related to seismic design margin on an ongoing basis and provide input to Tasks I.4 and 4
I.5 of the 50er. To do so requires identifying and monitoring research programs and other studies which are likely to provide infomation pertinent to the assessment of seismic design margins.
g Justification of Task Based on Regulatory Needs. Regulatory Need 1 is directly addressed by this task.
4 Regulatory Needs 2 - 5 are addressed in the same manner as the subject i
research programs address these issues.
1 14-25
F Relationship to Other SDMP Tasks. The results of Task 11.1 provide input to Tasks 1.4 and 1.5.
Task Is.3: Assessment of the Effect on Marains of Relay and Circuit Breaker Perfomance Durine 5 trona Motion Description. Recent studies have shown that seismically-induced circuit breaker failures may inhibit the proper operation of safety systems during and af ter strong motion. These failures may be caused by relay chatter in the circuit breaker electrical cystems. For example, inadvertent operation of the anti-pumping relays may lock out the circuit breaker. Failure of manual and test switches in these circuits may also be a problem. Such failures have been ignored in most seismic risk assessments to date, yet relay chatter can be caused by a relatively low intensity of ground motion. Therefore, the effect of relay chatter during strong motion may have a pronounced effect on system and plant 50M.
Since some circuit breakers, relays and circuits may be affected by this type of failure, the number of these and the generic implications need to be assessed.
The objective of this task is to estimate the influence on plant and system SOM of circuft breaker misoperation caused by relay chatter or switch malfunctions. An examination of nuclear power plant systems will be made to detemine how prevalent are circuit breaker system designs subject to strong motion failure. If large populations of systems are susceptible to these problems, effort needs to be expended on developing their fragility functions and the consequences of their failures on plant safety.
hstification of Task Based on Regulatory Needs. This task will contribute to the resolution of Regulatory Need 1.
Relationship to Other 50er Tasks. The results of this task will provide infomation for Tasks !.2 and I.3.
II.4: Assessment of the Behavior of Operators Durina and Inuiediately After Strona RDtlon Description. Concern exists that reactor operators or the displays they b
monitor may be so affected by the ground motion that they will be unable to perform their required functions. Recent experimental data from Japanese 14-26 l
tests suggests the operator may be prevented from reading and reacting to his displays at ground motions above 0.2 to 0.4 g.
These tests indicate the i'
actual level is influenced by chair design, chairs with casters being the better perfomers. Other efforts to assess the effect of earthquakes on 4
operator perfomance are being undertaken by the IRC Office of Nuclear Regulatory Research as part of their human factors msearch program.
Information from these efforts as well as results from the revisits of seismic PRAs need to be assessed. From this assessment, a feeling for the relative effect of seismic induced operator error compared to various hardware and structural failures needs to be made. The overall impact of operator behavior during and after earthquakes to SOM issues can then be understood.
Asstification of Task Based on Regulatory Needs. Regulatory Need 1 is I
directly addressed by this task.
Relationshio to Other S019 Tasks. Results from this task will provide input for Tasks !.2 and !.4.
I
!!.5: Assessment of the Effect of Desian and Construction Errors on SOM Issues b
Desc' riotion. The amount of seismic design margin at a plant is dependent not only on the design as envisioned but also as constructed in the field. Design and/or construction errors can play a significant role in this issue. Several reports in the literature have demonstrated the large effect that design errors can have on seismic risk and consequently on 50M. To date, however, no j
ger.arally accepted, practical means of including the effects of design and construction errors in a risk calculation is available. Further, no methodology or procedure exists to identify design and construction errors outside of the current QA programs. Because design or construction errors many times lead to surprises in the performance of facilities when subjected l
to an earthquake, this issue needs to be addressed in the fomat of SOMs.
i
&stification of Task Based on Regulatcry Needs. Regulatory Needs 1 and 5 are 4
addressed by this task. Regulatory Needs 2 - 4 are not addressed.
Relationshin to Other 5019 Tasks. This task provides input to Tasks !.2 and i
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i Task !!.6: Assessment of Inherent Calculstional Desian Marcin.
Description. Over the past 15 years, a significant evolution has occurred in seismic analysis and design procedures. In most cases, the evolution has been to increased design requirements introduced by the methods of seismic analysis
.and specification of the hazard and system parameters. Examples include:
Seismic design ground motion defined by US NRC Regulatory Guide (RG) e 1.60 -- three components of motion and broad-band response spectra.
Damping values defined by US NRC RG 1.61.
e Control point definition at foundation level - US Imc Standard e
Review Plan (sap) 3.7.2.
Broadened in-structure response spectra for equipment and piping e
system qualification defined by US NRC RG 1.122.
Envelope procedure for analyzing multi-supported systems such as e
piping systems -- US NRC SRP 3.g.3.
. Modal combination rules for closely-spaced modes -- US NRC RG 1.92.
e These and other requirements were introduced due to legitimate concern reganiing uncertainties in analysis methods and parameter values. However, they were introduced with little consideration of their ramifications on subsequent elements in the seismic analysis chain. It is well-recognized that conservatisms com;:ound as one moves from the seismi,c input - to soil-structure interaction (SSI) - to structure response - to-equipment and piping 1
response. This compounding was not explicitly considered in developing new i
requirements. It is a major source of seismic design margin. Quantifying these margins contributes to our ability to make definitive statements j
concerning seismic design margin on a plant-by-plant basis or for groups of j
plants. In addition, it is items such as these that may be potential I
screening guidelines.
An effective approach to this task is first to assemble existing information on quantification of calculational margin. Also, one must identify candidate seismic design criteria (methodologies, parameter values, etc.) which introduce substantial conservatism in calculated values of response. To quantify these conservatises, one must perform comparative calculations with
+
best estimate technology, the result being the margin introduced by the specified calculational procedure.
14-28
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. I 4
Justification of Task Based on Regulatory Needs. Regulatory Need 1 is partially addressed by this task, i.e. the results of this task lead to quantification of seismic design margin due to response prediction techniques. Regulatory Need 2 is addressed by demonstrating the large calculational margin which exists due to seismic analysis methodology for specific situations. Small perturbations in the definition of the ground motion should not exceed this margin.
Regulatory Needs 3 - 5 are not addressed.
Relationship to Other SOPP Tasks. Task !!.4 obtains input from Task !.1 and provides results to Tasks !.2 and I.3.
Task
'. 7: Assessment of System Chances. Such as Added Redundancy and Enhanced UDeraW onal Modes. on 50R Description. Over the past 15 years, systems design has evolved as have seismic design criteria. Thoughts have evolved concerning redundancy of components, redundancy of safety systems, isolation of cour1.. as, manual operation of portions of systems, power trains and isolation, etc. Changing NRC requirements have led to this evolution such as the implementation of fire pmtection regulations. Many of these changes may have overall plant safety consequences when considering the seismic hazard even though they were not implemented to enhance seismic safety as in Task II.6. Consequently, those plants with favorable systems aspects may be more reliable under the seismic hazard and this may lead to screening criteria for the SDMP.
The objective of this task is to identify systems aspects of nuclear power plants which lead to significant SOM and consequently constitute a screening criterion by themselves.
Justification of Task Based on Reculatory Needs. Regulatory Need 1 is
. parttally addressed by this task. Regulatory Needs 2 - 5 are not addressed.
Relationshio to Other 50pr Tasks. Task !!.7 obtains input from Task !.1 and provides results to Tasks !.2 and 1.3.
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f Task
'I.8: Assensment of the Effect of Uncertainty in Non-linear Structural genav or on som..ssues Description. In assessing seismic design margin, the realistic behavior of structums under earthquake loadings must be taken into account -- in particular, the non-linear behavior of structures. The perfomance of structures in past earthquakes has demonstrated the significant reserve capacity of ductile structures subjected to earthquake loadings and the poor estimates of behavior made by linear elastic predictive techniques.
Currently, when non-linear behavior is taken into account, it is treated by very approximate techniques. The ductility modified response spectrum technique, originally developed by Newsark, is the most extensively used approach to date. It is based on numerous studies of single-degree-of-freedom systems but lacks correlation with physically realistic structural configurations.
The problem is clearly two-fold: data acquisition on the behavior of structures and analytical modeling of the behavior. Data acquisition is partially addressed by the NRC Category I Structures Program although for limited stmeture types and scale models. Additional data for full-scale structures and of differing construction are needed. Analytical techniques, adequately benchnarked, need to be developed to pemit analysis of structures with significant non-linear behavior.
Seismic PRAs consider the range of possible earthquakes at the site and, hence, consider earthquakes substantially higher than the design level event.
Seismic PRAs quantify structural failure predictions which become an important element in seismic risk analysis and in the assessment of seismic design margin. Non-linear structure behavior dominates these predictions and requires validation.
In addition to the effect on structure forces, non-linear structure behavior has a significant impact on the input environment to subsystems (piping systems and equipment). This needs to be taken into account when estimating their capacity for seismic PRA and seismic design margin analyses purposes.
The objective of this task is to identify and quantify the margin introduced by linear or approximate calculations for structures that respond in a
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non-linear way to strong seismic excitations. The approach to executing 14-30
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this task is a combination of data acquisition (existing and new) at the structure and structure element levels and analytical development and verification of non-linear analysis techniques. Benchnarking these techniques with existing data and application to physical structures completes the effort.
Justification of Task Based on Regulatory Needs. Regulatory Need 1 is directly addressed by this task. The non-linear behavior of structurns is an important source of margin -- one which requires consideration and quantification to permiit quantitative margin statements to be made.
Regulatory Needs 2 and 4 are indirectly addressed by this task since the damage potential of earthquakes defined therein is assessed.
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50MP SCHEDULE AND DELIVERA8LES Task FY 85 FY 86 & FY 87 PART I Phase !
I.1 Assess Existing Information 7
!.2 Estimate Existing Margins 7
!.3 Identify Generic Attributes 1.4 Assess Ma nin Adequacy 7Y
!.5 Develop Screening Guidelines 7
Phase II I.6 Conduct Trial Plant Reviews 9
Y Phase !!!
t Implementation of Plant Reviews and Continuing Studies Reports as tasks l
are completed PART !!
Identification of Information Needs 7
worn to ne cone separately roisowing identification of 1,nformation needs.
Deliverables:
LLNL input to Expert Panel on Capacity & response factor data. (11/84)
Input to the Expert Panel. (2/85) i Expert Panel Report on Tasks !.1 - !.4 & Identification of Information Needs. (4/85)
Expert Panet Report of proposed guidelines and trial review procedure.
Report on results of the 2 trial plant reviews & recossendation for Phase !!! implementation.
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UNITED STATES y' e,,
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W ASHINGTON, D. C. 70555
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OCT 3 1984 v
MEMORANDUM FOR: Frank P. Gillespie, Director Division of Risk Analysis and Operations, NRR FROM:
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response, IE
SUBJECT:
PROPOSED AMENDMENT TO 10 CFR PART 50, APPENDIX E: CONSIDERATION OF EARTHQUAKES IN THE CONTEXT OF EMERGENCY PREPAREDNESS This confirms that IE office comnients were sent to Task Leader M. Jamgochian on September 28 (Pagano markup) and telephonic office concurrence furnished by Pagano on October 1, 1984.
Enclosed find a FEMA markup of the paragraph that relates to their earthquake activities which also conveys formal FEMA concurrence on the package.
r.A ard '. Jordan, Director Divisio ) of Emergency Preparedness and E ineering Response, IE
Enclosure:
As stated i
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