ML20043G358
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| Issue date: | 01/08/1990 |
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4 MARK I PLANT-SPECIFIC ENHANCED VENTING CAPABILI'IT REGULATORY ANALYSIS JANUARY 8,1990 l-i
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1 9006200187 900108 PDR TOPRP ENVGENE C
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Table of utntents
' 1.0 STATEMENT OF'THE PROBLEM 1
- 2. 0 OBJECTIVES 2
- 3. 0 ALTERNATIVE RESOLUTIONS 2
3.'l-Alternativo (i)-
3 3.2 Alternative (ii) 3 l
3.3 Alternative (iii) 4 3.4 Alternative (iv) 5 3.5 Alternative (v) 5 4.0 CONSEQUENCES 6
4.1 Costs and Benefits of Alternative Resolutions 7
4.1.1 Alternative (i) 7 4.1.2 Alternative (ii) 7 4.1.2.1 Value: Risk Reduction Estimates 7
4.1.2.2 Impacts: Cost Estimates 8
4.1.2.3 Valu,e-Impact Ratio.
10 4.1.3 Alternative (iii) 11 4.1.3.1 Value: Risk Reduction Estimates 11 4.1.3.2 Impacts: Cost Estimates 11 4.1.3.3 Value-Impact Ratio.
11 4.1.4 Alternative (iv) 12 4.1.4.1 Value: Risk Reduction Estimates 12 4.1.4.2 Impact: Cost Estimates.
12 4.1.4.3 Value-Impact Ratio.
12 4.1.5 Alternative (v) 13 4.1.5.1 Value: Risk Reduction Estimates 13 4.1.5.2 Impact: Cost Estimates.
13 4.1.5.3 Value-Impact Ratio.
13 4.2 Impacts on Other Requirements 16 4.2.1 Individual Plant Examination 16 4.2.2 Improved Plant Operations (IPO) 16 4.2.3 Severe Accident Research Program (SARP) 17 4.2.4 External Events.
17 4.2.5 Accident Management 17 4.3 Constraints 18 5.0 DECISION RATIONALE 18 5.1 Commission's Safety Goal 18 6.0-IMPLEMENTATION 20 6.1 Schedule for Implementation 20 6.2 Relationship to, Other Existing or Proposed Requirements 20
7.0 REFERENCES
21
'* -A - i-APPENDIX A - BACKFIT ANALYSIS Q
DRAFT Mark i Plant-Specific Enhanced Venting Capability Regulatorv Analysis 1.0 STATEMENT OF THE PROBLEM The' General Electric Company has designed and constructed seve[al Bo,iling Water Reactor (BWR) configurations with three basic containment designs designated as Mark I, Mark II, and Mark III.
Probabilistic Risk Assessment (PRA) studies have been performed for a number of BWRs with Mark I containments.
Although these PRA studies do not show the BWR Mark I plants to be risk outliers as 4
a class relative to other plant designs, they do suggest that the Mark I containment could be challenged by a large scale core melt accident, primarily due to its smaller size.
However, estimates of containment failure likelihood under such conditions are based on calculations of complex accident conditions, which contain significant uncertainty.
Draft NUREG-1150' evaluated the dominant accident sequences for five plants, one of which was a BWR Nark I.
The dominant accident sequences were identified as station blackout (TB), which includes the loss of all AC and DC power; anticipated transient without scram (TC) ; and would have included the loss of long term decay heat removal (TW) except that, for the pa.rticular plant being l.
reviewed, the likelhood of this sequence was considered _ to be e i
greatly reduced due to assumed successful venting of the containment.
While the TW sequence was not considered in NUREG-1150 to be a dominant sequence for the plant reviewed, it can be a significant contributor to overall plant risk for Mark I plants, in general.
(The subsequent June 1989 version of draf t NUREG-1150 reported similar results for Peach Bottom as was reported in the February 1987 edition.)
All BWRs with Mark I containments have a capability to vent the containment with various size lines. The largest lines usually are associated with the vent and purge system used to inert and deinert l
containment.
Venting of containment as an accident mitigative j'
action is permitted in the Emergency Operating Procedures (EOPs).
l The existing vent path uses, in part, sheetmetal ductwork from the containment isolation valves through the standby gas treatment system (SGTS) to the plant stack.
The sheetmetal ductwork is usually designed for low pressure and is expected to fail under severe accident pressures.
Failure of the ductwork would introduce the containment atmosphere to the reactor building.
This could
" Reactor Risk Reference Document",-
Drlft,'
~
l-February 1987.
i DRAFT result in harsh environmental conditions which would complicate operator accident recovery a tions within the reactM-building and could result in f ailure of etiuipment within the reactor building.
The hard pipe vent would be designed to withstand severe accident pressures, and, thus, would not fait during a TW event thereby alleviating the harsh environmental concerns in the reactor building.
This regulatory analysis is to study the cost / benefit of installing a hardened vent capability at BWRs with Mark I containments.
2.0 OBJECTIVES The staff objective is to reduce the overall risk in BWR Mark I plants by pursuing a balanced approach utilizing accident pre-vention and accident mitigation.
Most recent PRA studies indicate that an important contributor to BWR Mark I risk is the loss of long term decay heat removal (TW).
The balanced approach includes:
those features or measures that are (1) accident prevention expected to reduce the likelihood of an accident occurring or measures that the operating staff can use to control the course of an accident and return the plant to a controlled, safe state, and those features or measures that can (2) accident mitigation reduce the magnitude of radioactive releases to the environment in the event of an accident.
The proposed hardened vent capability would provide enhanced plant capabilities and procedures with regard to both accident prevention and ritigation, although for purposes of this regulatory analysis only prevention aspects associated with the TW sequence were quantified, although the quantification aspects of both were considered.
3.0 ALTERNATIVE RESOLUTIONS Plant modifications to the containment venting capability are being proposed to reduce the probability of or to mitigate the conse-quences of a severe core melt accident.
The proposed modification consists of installation of a hard pipe from the existing wetwell ventilation penetration, bypassing the ductwork to the standby gas treatment system, and going to the plant stack.
The ventilation penetration is the 18 to 24 inch penetration normally used as part of the vent and purge system for de-inerting the containment.
For the proposed modifications, the new components need not be safety-grade or safety-related.
- However, no failure of the modified system or non-safety-related component is to adversely affect any safety-related structure, system, or component required for coping with design basis accidents.
2
DRAFT 3.1 Alternative (i) 93 This alternative is the no action option, i.e. the existing venting capability remains as is.
The existing venting capability consists of venting the containment through the existing ductwork from the suppression pool to the standby gas treatment system (SGTS).
Ductwork design pressure is e
usually a few psid or less.
Consequently, venting under severe accident conditions could result in failure of the ductwork and a direct release into the reactor building.
The discharge of high temperature gases over an extended period of time may pose a threat to the availability or performance of safety related equipment.
The discharge of hydrogen could result in hydrogen burns (or detonations) inside the reactor buildinq. Electrical cables, motor operators on valves, relays, and control room components may be subject to failure under these environmental conditions.
Adverse environmental conditions would complicate entry into the reactor building.
Calculations from a venting study during an anticipated transient without scram (ATWS) indicate a severe environment would be present in the reactor building during venting operations.3 It is reasonable to assume that this environment (high temperature and radiation) could hamper recovery ef forts by preventing personnel access into the reactor building if systems needed to terminate the accident needgepair.
3.2 Alternative fii)
This alternative would involve the installation of a hardened venting capability from the containment wetwell to the plant stac);.
The proposed venting improvement would provide a wetwell path to the plant stack capable of withstanding the anticipated environ-mental conditions of a severe accident. This proposed modification would include the installation of hard pipe from the outlet of an existing wetwell vent outboard containment isolation valve to the base of the plant stack.
This pipe would be routed through a new DC operated isolation valve which would bypass the existing duct-work and the standby gas treatment system (SGTS).
The emergency procedures would need to be modified to provide appropriate instructions for the operator. This alternative would mitigate the consequences of severe accidents provided there is no drywell-2NUREG/CR-5225, "An Overview of Boiling Water Reactor Mark I Containment Venting Risk Implications", dated November 1988.
- Marring, R.
M.,
" Containment Venting as a
Mitigative Technique for BWR' Mark I plant ATWS", 1986 Reactor Water Reactor Safety Meetina, Gaithersbura, Maryland, October 1986.
.. ~ _
3
c, DRAFT failure and could reduce the likelihood of core melt from the TW sequence provided the operator transfers suction of the low
- pressure injection pumps from the suppression pool to an alternate source of water, such as the condensate storage tank, before venting containment.
All releases through the vent would pass through the suppression pool and, the particulates would be scrubbed.
Given a loss of long term decay heat removal accident, this alter-native would prevent failure of the vent path inside the reactor building and would result in an elevated release.
The elevated release could reduce the of f site consequences. Since the vent path is not expected to fail inside of the reactor building, personnel would be able to repair equipment and to perform other plant recovery activities in the reactor, building.
Furthermore, there would be no harsh environmental conditions to degrade or f ail other equipment.
There is the possibility of inadvertent operation of the vent with the release of some radioactive material without any holdup time or filtration.
This alternative would not affect the releases of radioactive material for those sequences where the drywell fails, such as from corium attack, once the dryvell shell has failed.
s.3 Alternative (11 0 This alternative would involve alternative (ii) plus the instal-lation of an external filter system.
The proposed venting improvement includes the hard pipe vent discussed in alternative (ii) plus the installation of an external filter system, such as the Filtra system or the Multi Venturi Scrubbing System (MVSS).
This external filter would be installed outside of the existing facilities.
A single external filter unit could be constructed to service multiple containments with proper isolation valves.
Both the Filtra and the MVSS do not rely on AC power to perform their intended functions.
Similar to alternative (ii), the emergency procedures would need to be modified to provide appropriate instructions for the operator.
This alternative would mitigate the consequences of a severe accident and could reduce the likelihood of core melt if the operator transfers suction of the injection pumps from the suppression pool to an alternate source of water, such as the condensate storage tank, before venting containment.
With the external filter, the amount of particulate removal of the external filter would not be sensitive to the conditions in the suppression pool. No significant additional risk reduction has been estimated to. result with an external filter system in addition to the suppression pool scrubbing.
Since all particulate releases through the hardened vent (alternative 11) are scrubbed, the external filter will only provide some minimal additional scrubbing.
The external filter provides no addit ~tc*na 4
DRAPT
. benefit in core melt prevention although it would provide filtration and some holdup time for inadvertent operation of the vent.
Similar to alternative (ii), this alternative would not affect the releases of radioactive material for those sequences where the drywell fails, such as from corium attack, once the drywell shell has failed.
3.4 Alternative fiv)
This alternative would involve the use of an existing heat removal system which has not been previously been credited in a PRA or the installation of a new decay heat removal system to remove the decay heat loads from the reactor or containment.
The proposed modification would provide an additional decay heat removal system to remove heat from either the reactor or the coh-tainment, or a system which has not been previously accounted for could be used, such as the reactor water cleanup system.
If the reactor water cleanup system is used, the heat could be removed with the reactor at operating pressure or depressurized.
If a new alternate residual heat removal system is installed, the reactor would have to be depressurized and assurance would be needed that the automatic depressurization system (ADS) or the equivalent number of safety relief valves (SRVs) would be operable from the control room. Alternatively, the decay heat loads could be removed f rom the suppression pool inside containment. The emergency proce-dures would need to be modified to provide appropriate instructions for the operator.
This alternative could reduce the likelihood of core degradation.
t 3.5 Alternative (v)
This alternative would involve the removing the guidance'in the Emergency Procedure Guidelines (EPGs) which instructs the operator under certain conditions to vent the containment.
The emergency procedures would need to be modified to provide, appropriate instructions for the operator.
This alternative would not reduce core melt for the TW sequence but could delay contaminating the reactor building.
Except for failures in the drywell, the design l
of the Mark I containment is such that suppression pool bypass is l
not reasonably credible.
Given a TW sequence, no drywell failure, l
and without venting, the containment has a
relatively high
. probability of failure from over-pressurization.
The effects of-containment failure could significantly reduce the ability to 1
l 5
^
4
~
DRAFT return the plant to a safe and controlled condition and would result in an increase in risk."
4.0 CONSEQUENCES The effects of the proposed enhancement was evaluated by using qualitative and quantitative discussions.6
'The quantitative for statio(n blackout events with 15 to 20 instead of the 107 de ived using a simplified containment event tree insights were (SCET) 6 top events used in draft NUREG-1150.
(Note that the TW event is similar to a station blackout event except that the loss of injec-tion is later in the accident sequence and thus the amount of decay heat being produced is reduced.
This has the effect of stretching out the timing of the events.)
The development of the SCET relied heavily on the data and insights generated by the draf t NUREG-1150 effort.
However, instead of trying to consider the entire range
(
of possibilities and their uncertainties, the SCET assigned best-I estimate values to the branch point probabilities.
The results of the SCET identified each specific event tree end-state and its associated probability.
These end-states were compared to the similar accident progressions from the list of Peach Bottom accident progression bins.#
The end-states from the SCET were characterized according to the draft NUREG-1150 accident progres-sion bin format and then compared and assigned to the best-match accident progression bin.
This process reduced the number of source terms that needed to be evaluated.
Once the SCET end-states were related to those identified in draft NUREG-1150, the consequences were taken directly from draf t NUREG-1150 results and interpolated to generate the consequences.
The risks were then calculated by multiplying the plant damage state frequency, the bin probability, and the consequences of that bin.
l To evaluate the approximate accuracy of the SCET, the draft NUREG-l 1150, information related to Peach Bottom was input into the SCET and the lesults compared with those of draft NUREG-1150.
This comparison was to determine the ability of the SCET to reasonable predict the NUREG-1150 results.
In all categories, the results of f
- NUREG/CR-5225, Addendum 1,
"An Overview of BWR Mark I Containment Venting Risk Impliciations, An Assessment of Potential l
Mark I Containment Improvements", dated June 1989.
5ibid, #4.
6ibid, #1.
Draf t NUREG/CR-4551, " Evaluation of Severe Accident Risks and 7
the Potential for Risk Reduction: Peacn Bottom, Unit 2",
Volume 3, Draft, dated May 1987.
6 l
l 4
o
[
DRAFT the SCET compared with those in draft NUREG-1150 within about 25%,
well within the uncertainty bar.d of draft NUREG-1150.
Once verified, advanced information related to the second draft NUREG-1150' was used tc, form a new base case and to evaluate the benefits of the proposed enhancements.
4.1 Costs and Lenefits of Alternativ.g_ Resolutions PRAs that the staf f have available were used to estimate the incre-mental benefit of the five alternatives discussed below.- The only 1
accident sequence that is being considered for this analysis is the loss of long term decay heat removal (TW).
This is considered to be conservative since the alternatives could have a benefitial ef fect on other sequences, although the effects are expected to be small'.
The total core melt frequency from internal events was not estimated, only the change in the plant core melt frequency'.
4.1.1 Alternative (i)
This alternative would be to take no action.
Since it is expected that the ductwoEk'would fail if the containment were vented at high 7
l pressure,.,this approach would not only jeopardize personnel, but jeopardizes the ability to regain control of the facility given the I
accident. Furthermore, based on a generic regulatory analysis" the Commission has instructed the staff to require hardened vent capa-l bility for plants for which it could be shown to be cost ef fective, therefore, the no-action alternative is not recommended.
4.1.2 Alternative (ii) 4.1.2.1 Value: Risk Reduction Estimates For those accident scenarios where t.ontainment failure results in core degradation and a severe accident, the hard pipe vent path approach could reduce or delay core degradation.
This is estimated to reduce the total core damage frequency per reactor year by the values shown in Column C of Tabic 2.
This 1
~
8NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants", Draft, June 1989.
' ibid, #4.
Sheron, B.W., Memorandum to Thadani, A.C., " Reduction in Risk from the Addition of Hardened Vents in BWR Mark I Reactors",
October 19, 1989.
" Mark I Containment Performance Improvement Program", January 23, 1989.
7 3
DRAFT represents a risk reduction in man-rems per reactor year of the values shown in Column F of Table 2.
4.1.2.2 Impacts: Cost Estimates The estimated costs for installation of the hard pipe vent path are shown in Column H of Table 2.
The averted cost associated with prevention and mitigation of an accident can be discussed as five separate costs:
replace-ment power,
- cleanup, onsite occupational health impacts, offsite health impacts, and onsite property damage.
To estimate the costs of averting plant damage and cleanup, the reduction in accident frequency was multiplied by the discounted onsite The following equations f rom NUREG/CR-3568' property costs.
were used to make this calculation:
V, = NdFU U=
(C/m)((ed'))/r ) { y,qgin] (1-e")
2 where:(Note: cited values are for Dresden Unit 2 from Table 2).
V,
= value of avoided onsite property damage N
= number of affected facilities = 1 dF
= reduction in accident frequency = 1.4 x 10'5 /RY present value of onsite property damage U
=
cleanup and ropair costs =-$1.0 billion C
= years remaining until end of plant life = 21 t(f)
=
years before reactor begins operation = 0 t(i)
=
discount rate = 10%
r
=
period of time over which damage costs are paid m
=
out (recovery period in years) = 10 Using the above values, the present value of avoided onsite property damage is estimated to be $77,660 (Table 2, Column I + 1).
Memorandum from J.G.
Partlow to T. E.
Murley, " Licensees' 12 Responses to Generic Letter 89-16 Related to Installation of Hardened Wetwell Vent", dated November 9, 1989.
uMUREG/CR-3568, "A
Handbook for Value-Impact Assessment",
December 1983, pages 3.29-3.31.
8 4
0
s^
DRAFT Replacement power costs can be estimated using NUREG/CR-
.l 4012" which lists the replacement power costs for each nuclear i
power reactor by season. Using this information for only Mark I reactors averaged over the four years of projected data and escalated-by 6% for 1989 dollars, the generic replacement power cost is $400,666 per day.-
The plant-specific replace-j ment power cost is shown in Table 3).
(NUREG-1109'S has used i
a generic cost of $500,000 per day and compares f avorable with i'
The change in public health risk associated with the instal-j lation of the proposed hardened vent system is expressed as r
total man-rem avoided exposure.
The following equations from l
NUREG/CR-3568 were used to make this calculation:
4 x R)
NT (Dp V
=
pg where:
V
= value of public health risk avoided for net-pg benefit method (S)
N
= number of affected reactors = 1 T
= average remaining lifetime of af fected facilities (years) = 21 (Table 2, Column A) j Dp
= avoided public dose per reactor-year (man-rem /RY) i
= 50.2 (Table 2, Column F)
R
= monetary equivalent of unit, dose ($/ man-rem)
= $1000 Using the above values, the avoided public health exposure is shown in the second. column to the right from Column I in Table 2 for each Mark I plant.
The occupational health risk avoided as the result.of the t
l installation of the proposed hardened vent system is expressed as man-rem avoided exposure.
The following equations from
'NUREG/CR-3568" were used to make,this calculation:
i l
"NUREG/CR-4012, " Replacement Energy Costs for Nuclear Electricity-Generating Units in the United States: 1987-1991",
Volume 2, January 1987, Table S.1, pages 2 - 5.
'5NUREG-1109, " Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station-Blackout", June 1988, page 23.
" ibid #13, pages 3.11-3.12.
f-(
" ibid #13, pages 3.16-3.17.
9 c
L$ - _
4 DPJJT 4
x R)
Vw = NT(Dog wheret V% = value of occupational health risk due to accidents avoided ($)
N
= number of affected reactors (reactors)
=1 T
= average remaining lifetime of affected facilities (years)
Dog
= avoided occupational dose per reactor year (M:n-Rem / Reactor-Year)
R
= monetary value of unit" dose ($/ Man-Rem)
= $1000 / Man-Rem There are two types of occtipational exposure related to accidents, immediate and long-term.
The first occurs at the time of the accident and during the immediate management of the emergency. The second is a long-term exposure, presumably at significantly lower individual rates, associt.od with the cleanup and refurbishment of the damaged facility.
The best.
estimate of the immediate occupational exposure as specified in NUREG/CR-3568'8 is 1000 man-rem.
Tne best estimate of the long-term occupational exposure as specified in NUREG/CR-3568 is 20,000 man-rem.
This results in occupational exposure of 21,000 man-rem.
Using the above
- values, the present value of avoided occupational health exposure has been calculated to be approximately one to two percent of the public health risk and is not considered to be a significant contributor. Therefore, the occupational health exposures will not be considered further.
4.1.2.3 Value-Impact Ratio The valae-impact
- ratio, not including onsite accident avoidance costs, is given in Column I of Table 2 in units of man-rem averted per million dollars.
If the savings to industry from accident avoidance (cleanup and repair of onsite damages and replacement power (Table 2,
Column J) at a discount rate of 10%) were included, the overall value-impact ratio would be as shown in Column K of Table 2 in terms of man-rem averted per million dollars.
Note that the negative value (for Hope Creek) indicates that there is a monetary 181 bid $13, page 3.17-3.18.
ibid $13, page 3.18.
10
DRAFT savings to the licensee for installing the proposed modification, i.e. the estimated cost for replacement power and onsite costs exceed the estimated installation costs.
Since the value-impact ratio indicates that for most plants this is a cost effective alternative, it is recommended that all units install a
hardened vent capability unless justification can be provided otherwise.
4.1.3 hliernative-(iii) 4.1.3.1 value: Risk Reduction Estimates This alternative would provide minor additional particulate scrubbing for the hard vent.
- However, all particulate releases will have been scrubbed by the suppression pool, therefore the improvement over alternative (ii) is not expected to significantly reduce public risk (i.e. the risk reduction in man-rem per reactor year of the values shown in Column F of Table 2).
4.1.3.2 Impacts: Cost Estimates External filters have been estimated to cost $10M to $50M for the Filtra design and about $5M for the Multi-Venturi scrubber System design.
Using the same equations given in alternative (ii) above, the present value of the estimated avoided onsite property damage is shown in the column to the right of Column I of Table 2.
Similarly, the estimated replacement power cost is shown in the column to the left of Column J of Table 2 in terms of millions of d;:41ars for ten years.
- Thus, the estimated avoided onsite property damage and replacement power is shown in Column J of Table 2.
The present value of the change in the estimated public health risk associated with the installation of the hard vent and the external filter is shown in the second column to the right from Column I of Table 2.
4.1.3.3 Value-Impact Ratio The overall value-impact ratio of this alternative is in terms of man-rem averted per million dollars and is calculated from the value in Column G of Table 2 divided by the installation cost in Column H of Table 2 plus the value for the MVSS design and for the Filtra design.
If the savings to industry from accident avoidance (cleanup and repair of onsite damages and replacement power is shown in Column J of Table 2 at a discount rate of 10%) were included, the overall valu,e-impaat 11 I
r DRAFT ratio would be in terms of man-rem ave rted per million dollars and is calculated from the value i.1 Column G of Table 2 divided by the installation cost in C)1umn H of Table 2 plus the value for the MVSS and for the filtra design minus the value in Column J of Table 2.
Because this alternative is not cost effective, this alternative is nct recommended.
4.1.4 Alternative fiv) 4.1.4.1 Valuet Risk Reduction Estimates This alternative would provide a means to remove the decay heat from the reactor (or containment) and prevent core degradation.
A new system, or credit for an existing full capacity system that had not been accounted for in previous pRAs, could reduce risk to the public similar to alternative (ii), namely the value in Column F of Table 2 in terms of man-rem per reactor year.
4.1.4.2 Impact: Cost Estimates Installation of a new system has been considered as part of NUREG-1289". A new system was estimated to cost approximately
$90 million or more.
If there is the possibility of using an existing system, the' cost is expected to be lower although it has not been quantified.
4.1.4.3 Value-Impact Ratio The overall value-impact ratio of this alternative is in terms of man-rem averted per million dollars and is calculated from the value in Column G of Table 2 divided by the installation cost.
NUREG-1289 also concluded that the installation of a new decay heat removal system was not cost beneficial.
The use of another, previously unaccounted for system, could require unusual system piping line-ups, which if performed incorrectly or inappropriately, could reduce the likelihood of accident recovery with normal systems or create a new and unanalyzed accident sequence. Because this alternative is not cost effective, this alternative is not recommended.
"NUREG-1289, " Regulatory and Backfit Analysist Unresolved Safety Issue A-45, Shutdown Decay Heat Removal Requirements",
November 1988.
12
l DRAFT 4.1.5 Alternative (v) 4.1.5.1 Valuet Risk Reduction Estimates j
This alternative would remove the capability to vent the containment from the EPGs and thus the emergency operating procedures (EOPs).
While the risk was not quantified, it is expected that the risk to the public would remain constant in that both alternative (i) and this alternative would result in a ground level release, an uninhabitable reactor building, and questions regarding the survivability of the equipment in the reactor building.
This alternative only provides a marginal delay in the accident sequence.
4.1.5.2 Impact: Cost Estimates The costs associated with removing the venting option from the EPGs and EOPs and associated operator training have not been quantified.
4.1.5.3 Value-Impact Ratio The overall value-impact ratio would be negative, since the risk would not be reduced and there is
- some, though unquantified, costs'.
Because this alternative is not cost effective, this alternative is not recommended.
Table 1 - Cost Benefits of Alternatives (i)-(v)
(Man-rem averted per million dollars)
Alternative (i) - do nothing 0
Alternative (ii) - hard pipe venting Column K, Table 2 l
l Alternative (iii) - hard pipe venting
+ external filter Alternative (iv) - additional decay heat removal system l
Alternative (v) - Delete venting from EPGs
(-)
l 1
Value is to be obtained from Section 4.1.3.3
- Value is to be obtained from Section 4.1.4.3 13
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DRAFT Table 3 - Mark I Estimated Replacement Power Costs (in Dollars per Day) g Year Est. Cost Est. Cost Est. Cost Reactor Name MWe Licensed 1985$
1989$
(per year)
Browns Ferry 1 1065 1974
$202,813
$251,488
$91,792,938 Browns Ferry 2 1065 1975
$201,833
$250,273
$91,349,767 p
Browns Ferry 3 1065 1977
$196,882
$244,134
$89,108,953 Brunswick 1 821 1977_
$312,250
$387,190 $141,324,350 Brunswick 2 821-1975
$312,300
$387,252 $141,346,980 m
Cooper 778 1974
$188,400
$233,616
$85,269,840-Dresden 2 794 1970
$374,100
$463,884 $169,317,660 r
Dresden 3 794 1971
$372,000
$461,280 $168,367,200 Duane Arnold 538 1975
$108,000
$133,920
$48,880,800
(
- Fermi 2 1093 1988
$552,100
$684,604 $249,880,460
- Fitzpatrick 816 1975
$444,650
$551,366 $201,248,590 Hatch 1 776 1975'
$312,200
$387,128 $141,301,720 Hatch 2 784 1979
$312,150
$387,066 $141,279,090 Hope Creek 1067
_1986
$492,450
$610,638 $222,882,870 Millstone 1 660 1971
$346,800
$430,032 $156,961,680-Monticello 545 1971
. $143,800
$178,312
$65,083,880 Nine_ Mile Point 1 620 1969
$342,850
$425,134 $155,173,910 I
. 0yster Creek 650 1969
_$299,600
$371,504 $135,598,960 i
Peach Bottom 2 1065 1974
$484,950
$601,338 $219,488,370 Peach Bottom 3 1065 1974
$477,300
$591,852 $216,025,980-t Pilgrim 655 1972
$359,850
$446,214 $162,868,110 Quad Cities 1 789 1973
$326,750_
$4 05,-170 $14 7,~887,050
- Quad Cities 2 789 1973
$326,750-
$405,170 $147,887,050 F
-Vermont Yankee 514 1972
$264,050
$327,422 $119,509,030 Notes: 1:NUREG-1109 states that the estimated _ replacement-power cost is $500,000/ Day, Page 23.
L 2:NUREG/CR-4012 provides replacement power costs for all-plants on per plant / season basis for 1987-1991.
The aver-
^
age costs for Mark I plants is $400,666/ Day, Table S.I.
3: Inflation Rate used is 6 Percent / Year l
s t
- b OM w -
t E
15 o
L,
j r
A DRAFT 4.2 Japacts on Other Recruirements There are six programs related to severe accidents. These programs are: Individual Plant Examination (?,PE), Containment Performance Improvement (the topic of this reguletory analysis), Improved Plant Operations, Severe Accident Research Program, External Events, and Accident Management.
Each of the five programs related to Contai ment Perf ormance Improvement (CPI) will be discussed briefly below.p' 4.2.1 Individu al Fl ani_ Era:::in t*, ion The IPE involves the formulation of an integrated and systematic approach to an examination of each nuclear power plant now in operation or under construction for possible significant plant-specific vulnerabilities that might be missed without a systematic search.
Supplement 1 to Generic Letter 88-20 requested that Mark I licensees include in their IPEs an assessment of $he proposed plant improvements identified in SECY-89-017", other than the hardened vent, namely enhanced automatic depressurization system operation, and alternative low pressure water supply for injection into the reactor vessel and for containment sprays. The examination will include containment performance in striking a balance between accident prevention and consequence mitigation.
It is anticipated that the IPE program may take from three to four years until the last plant has performed the IPE.
4.2.2 Improved Plapt Operations (IPO)
The IPO includes consideration of continued improvements in the Systematic Assessment of Licensing Performance (SALP) program; regular reviews by senior NRC staff managers to identify and evaluate those plants that may not be meeting NRC and industry standards of operating performance; diagnostic team inspections; improved plant Technical Specificati:as; improved operating procedures; expansion of the proc;dures to include guidance on severe accident management strategies; i
industry's programs to reduce transient and other challenges to engineered safety feature systems; feedback from the IPE l
program of experience and improvements in operational areas, l
such as maintenance and training; and continued research to L
- 'For additional information, refer to SECY-88-147,
" Integration Plan for Closure of Severe Accident Issues", dated May 25, 1988.
22
- ~ -
ibid, $11.
16 l
l
o q
DRATT evaluate the sensitivity of risk to human errors, and the j
effectiveness of operational reliability methods to help l
identify potential problems early and prevent their occurrence.
The Ipo is related to the CPI program's recommendation since we recommend improved procedures and operator training to use the proposed hard vent system.
4.2.3 p_9 vere Accident Research Procry (SARP)
The SARP was begun after the TMI-2 accident in March 1979 to provide the Commission and the NRC staff with the technical data and analytical methodology needed to address severe accident issues.
This program has provided input to the NUREG-1150 program and to the CPI program.
Additional research is being considered, to evaluate the need for and feasibility of core debris controls.
Research will also confirm and quantify the benefits of having water in the containment to either scrub fission products or to prevent or delay shell melt through by core debris.
4.2.4 External Events The Commission's Severe Accident policy Statement does not differentiate between events initiated within the plant and externally initiated events.
Typically, external events have not been incorporated in the staff PRAs.
procedures for external events examinations are under development and the evaluation of external events will proceed as a separate part of the IPE's including their effect on containment.
The CPI program only addresses internally initiated events; however, it is not anticipated that future consideration of external events will adversely affect the recommendations of the CPI program.
4.2.5 Accident Manacement The accident management program is concerned with addressing l
certain preparatory and recovery measures that can be taken by the plant operating and technical staff that could prevent or significantly mitigate the consequences of a
severe accident.
This includes measures taken by the plant staff to
- 1) prevent core damage, 2) terminate the progress of core l
damage if it begins and retain the core within the reactor l
vessel, 3) failing that, maintain containment integrity as l
long as possible, and finally 4) minimize the consequences of offsite releases.
The CPI program recommended vent system would provide the accident management program with additional capability to achieve their goals by providing improved hardware with which to deal with a severe accident.
The
~
r 17 l
1
O i
4 i
g DRAPT procedures for using the vent should be re-examined under the Accident Management program.
4.3 p_gnstraints The plant specific imposition of a hardened vent is constrained by the guidelines of U.S.
NRC Manual Chapter 0514, "NRC Program for Management of Plant-Specific Backfitting of Nuclear Power Plants",
which is based on the backfit rule (10 CFR 50.109), as published by the Commission on September 20, 1985, and the provisions of 10 CTR 50 Appendix 0, 10 CTR 50.54(f), and 10 CTR 2.204.
No other constraints have been identified that affect thic program.
5.0 DECISION RATIONALE The evaluation of the CPI program included deterministic and probabilistic analyses.
Calculations to estimate the core damage frequency and the consequences of the TW sequence were performed based on using information available from the NUREG-ll50 program and existing PRAs.
A review of the availabl,e BWR Mark I PRAs provided only limited information.
However, the best estimate of the contribution of TW to the total plant core damage frequency in events per reactor year for each Mark I plant is shown in Table 2 Column C.
Implementation of the proposed hardened venting capability will result in TW being a minor contributor to the total core damage f requency and will result in a significant reduction in the total risk to the health and safety to the public.
5.1 Commission's Safety Goal on August 4, 1986, the commission published in the federal Recister a policy statement on " Safety Goals for the operations of Nuclear Power Plants" (51 FR 28044).
This policy statement focuses on the risks to the public from nuclear power plant operation and establishes goals that broadly define an acceptable level of radiological risk.
The discussion below address the CPI program recommendation in light of these goals.
' e two cualitative safety goals are:
1 (1)
Individual member of the public should be provided a level of protection from the consequences of l
nuclear power plant operation such that individuals bear no significant additional risk to life and health.
18
4 1
DRAPT (2)
Societal risks in life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a
significant addition to other societal risk.
The following auantitative obiectives are used in determining
'i achievement of the above safety goals:
(1)
The risk to an average individual in the vicinity
]
of a nuclear power plant of nremot fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1%) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed.
(?)
The risk to the penulation in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant operation should not exceed one-tenth of one percent (0.1%) of the sum of cancer fatality risks resulting from all other causes.
Results of analyses published in draft NUREG-1150 for the BWR Mark I (Peach Bottom Atomic Power Station, Unit 2) indicated that the Mark I plant meets the quantitative health objectives for prompt fatalities and latent cancer fatalities stated
- above, even considering the large uncertainties involved.
Implementation of the hard pipe vent capability will result in the total core damage frequency in events per reactor year being reduced by the value in Table 2 column C for each Mark I plant and the quantitative health objectives will continue to be met.
- However, the Commission also stated the following regulatory objective relating to the frequency of core melt accidents at nuclear power plants.
Severe core damage accidents can lead to more serious accidents with the potential for life-threatening offsite releases of radiation, for evacuation of members of the public, and for contamination of public property.
Apart from their health and safety consequences, such accidents can erode public confidence in the safety of nuclear power and can lead to further instability and unpredic-tability for the industry.
In order to avoid these adverse consequences, the Commission intends to continue to pursue a regulatory program that has as its objective providing reasonable assurance, giving appropriate consideration to the uncertainties involved, that d "*
~
19
4 I
l DRAFT severe core damage accident will not occur at a U.
S.
nuclear power plant.
With the implementation of the hard pipe vent capability, it is expected that the total core melt frequency can be reduced by the value in Table 2 Column C in events per reactor year by reducing the probability for one severe accident sequence, TW.
Therefore, implementing the hard pipe vent capability significantly reduces the likelihood that a severe core melt accident will occur at a U.S. BWR with a Mark I containment.
Additional rationale for implementing the hard pipe vent capability over the other alternatives is discussed as part of the value-impact analysis (Section 4.1).
This action represents the staff's position based on a comprehen-sive analysis of the containment performance improvement issues.
6.0 IMPLEMENTATION 6.1 Schedule for Imolementation Within 60 days after issuance of the backfit order, the licensee will submit to the NRC a schedule for implemetating any necessary equipment and procedural modifications to meet the performance goals and to provide adequate defense-in-depth.
All plant mo'difications are to be installed, procedures
- revised, and operators trained not later than 30 months from the issuance of the backfit order.
Other schedules were considered; however, the staff believes the proposed implementation of the hard pipe vent capability can be largely performed with minimum interf acing with containment and engineered safety feature systems and thus with the plant online.
Therefore the proposed modification is achievable without unnecessary financial burden on the licensea for plant shutdown.
The schedule allows reasonable time for the implementation of necessary hardware to achieve a reduction in the risk from TW.
Shorter or less flexible schedules would be unnecessarily burdensome.
6.2 Belationshio to other Existinc gy Proposed Requirements Several NRC programs are related to the CPI program; these are discussed in Fection 4.2.
These programs are compatible with the recommendation to install a hardened pipe venting capability.
.. ~ _
20 0
v DRAFT
7.0 REFERENCES
1.
SECY-88-147, " Integration Plan for Closure of Severe Accident Issues", May 25, 1988.
2.
" Mark I Containment Performance Improvement Program", January 23, 1989.
3.
WASH-14 00,
" Reactor Safety Study", October 1975 (also re-issued as NUREG-75/014) 4.
" Evaluation of Station Blackout Accidents at Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44", June 1988.
l S.
NUREG-1109, " Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout", June 1988.
6.
NUREG-1150, (Draft),
" Reactor Risk Reference Document",
February 1987.
7.
NUREG-1150, (Draf t), " Severe Accident Risks: An Assessment for Five US Nuclear Power Plants", June 1989.
8.
NUREG/CR-2723, " Estimates of the Financial Consequences of
~ Nuclear Power Reactor Accidents", September 1982.
9.
'NUREG/CR-3568, "A
Handbook for Value-Impact Assessment",
December 1983.
10.
" Replacement Energy Costs for Nuclear Electricity-Generating Units in the United States:
1987-
{
199\\",. January 1987.
11.
NUREC /CR-4 551, " Evaluation'of Severe Accident Risks and the Potential for Risk Reduction: Peach Bottom, Unit 2",
Volume i
3, Dralt, May 1987.
12.
NUREG/ Cit-4 624, "Radionuclide Release Calculations for Selected Severe hccident Scenarios", July 1986.
13.
NUREG/ Cit-52 2 5, "An Overview of Boiling Water Reactor Mark I Containnent Venting Risk Implications", October 1988.
L 14.
NUREG/CR-5225, Addendum 1,
"An Overview of Boiling Water Reactor Mark I Containment Venting Risk Implications, An i
Evaluation of Potential Mark I Containment Improvements", June 1989.
21
___._______________________j
r DRAFT 15.
Science and Engineering Associates, Inc. Report 87-253 A:1, " Cost Analysis' for Potential BWR Mark I Containment Improvements", November 1988.
16.
Harring, R.M., " Containment Venting as a Mitigation Technique for BWR Mark I Plant ATWS",1986 Reactor Water Safety Meetig Gaithersburo. Maryland. October 1986.
17.
- Sheron, B.W., Memorandum to Thadani, A.C., " Reduction in Risk from the Addition of Hardened Vents in BWR Mark I Reactors",
October 19, 1989.
18.
Generic Letter 88-20, Supplement No.
1,
" Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54-(f)", August 29, 1989.
19.
Generic Letter 89-16, " Installation of Hard Wetwell Vent",
September 1, 1989.
20.
Letter from Boston Edison Company, DPU 88-28, Request No. AG 13-6.
21.
Memor ndum J.G. Partlow to T.E. Murley, " Licensees' Responses to Ge1eric Letter 89-16 Related to Installation of Hardened Vent", dated November 9, 1989.
e
'S q
h-22
e DRAFT APPENDIX A BACKFIT ANAI(SIS Analysis and Determination That the Recommended Hard Pipe Vent Capability for Containment Performance Improvement Complies with the Backfit Rule 10 CFR 50.109 The Commission's existing regulations establish requirements for the design and testing of containment and containment cooling systems (10 CFR 50, Appendix A, General Design Criteria 50, 52, 53, 54, 55, 56, and 57) with respect to design basis accident conditions.
As evidenced by the accident at TMI Unit 2, accidents could progress beyond design basis considerations and result in a severe accident.
Such an accident,could pose a challenge to the integrity of containment.
Existing regulations do not require explicitly that nuclear power plant containments be designed to withstand severe accident conditions.
This issue has been studied by the staff and our consultants as part of the severe accident program for the General Electric Company Boiling Water Reactors (BWRs) with Mark I containments.
BWRs with Mark I containments have been reviewed first because of the perceived susceptibility of the Mark I containments to fail based, in part, on the small containment volume of the Mark I containment design.
Both deterministic and probabilistic analysis were performed to evaluate the loss of long term decay heat removal (TW) in challenging containment integrity and potential failure modes affecting the likelihood of core melt, reactor vessel failure, containment failure, and risk to the public health and safety.
The risk analysis shows that the risks from plants with Mark I containments is generally similar to that from plants with other containment types.
In addition, the hardened pipe vent l
capability is not needed to provide adequate protection of the public health and safety.
Rather, the proposed plant improvement will provide substantial enhancement to Mark I plant safety that is cost-effective.
The estimated benefit from implementing the proposed hard pipe vent is a reduction in the frequency of core melt due to TW and the associated reduction in risk of offsite radioactive releases.
The estimated risk reduction in terms of man-rem is shown in Column G of Table 2
and supports the Commission conclusion that implementation of the proposed improvement provides a substantial improvement in the IcVel of protection of the public health and safety.
The estimated cost to the licensee to implement the proposed safety enhancement is shown in Column H of Table 2.
This cost would be
.. ~ _
~
A-1
i e
DRATT '
primarily for the licensee to 1) assess the plant's capability, 2) install equipment to provide additional pressure relieving capability, 3) revise the emergency operating procedures, and 4) provide operator training related to mitigating the TW sequence.
The estimated value-impact ratio, not including accident avoidance costs, in terms of man-rems averted per million dollars is shown in Column I of Table 2.
If the net cost, which includes the cost savings from accident avoidance (i.e. cleanup and repair of onsite damages and replacement power following an accident), were used, the estimated overall value-impact in terms of man-rems averted per million dollars would be the value in Column K of Table 2.
These values support proceeding with the proposed hard pipe vent capability improvement.
The preceding quantitative value-impact analysis was one of the factors considered in evaluating the proposed improvements, but other factors also played a part in the decision-making process.
PRA studies performed for this issue have shown that the loss of long term decay heat removal (TW) events can be a significant contributor to core melt frequency, and, with consideration of the conditional containment failure probability, can represent an important contribution to reactor risk.
Although there are licensing requirements and guidance directed at providing a containment and support systems intended to contain any release of material from the reactore vessel, conLainment integrity may be significantly challenged under severe accident conditions.
In general, active systems required for reactor and containment heat removal are unavailable during the TW event.
Therefore, the offsite risk is higher from a severe accident that i
it is from many other accident scenarios.
The challenge to containment integrity is primarily by over-pressure for the TW events.
In addition, failure of the containment can also initiate core degradation under certain conditions.
The estimated f requency of core melt from TW events is directly proportional to the frequency of the initiating events.
The ectimate of the TW frequency for Dresden Unit 2 was based, in part, j
on information provided in draft NUREG-1150,
" Severe Accident Risks:. An Assessment for Five US Nuclear power Plants", for the Peach Bottom Atomic Power Station, Unit 2,
and other available PRAs.
This is assumed to be a realistic estimate of the core melt frequency when compliance with 10 CFR 50.63, the station blackout rule, has been achieved.
The factors discussed above support the determination that the additional defense-in-depth provided by the ability to cope with a TW event would provide a substantial increase in the overall A-2
d d) o
, -.,\\ * '
i
(
l DRAFT protection of the public health and safety, and the direct and indirect costs of implementation are justified in view of this increased protection.
Analysis of 10 CPR 50.109(c) Factors (1)
Statement of the specific obiectives that the backfit is desioned.tp achieve The objective of the proposed hard pipe vent capability is to reduce the risk from TW events by reducing the likelihood of core melt and to mitigate releases given a TW or other similar events leading to core melt.
(2)
General description of the activity required by the licensee or aonlicant in_ order to connlete the backfit In order to comply with the proposed improvement in contain-ment venting, the licensee will be required to:
Evaluate the actual capability of the existing contain-ment vent system to withstand the anticipated containment temperatures and pressures without failing any portion of the vent path to the plant stack.
Evaluate the actual capability of the existing contain6 ment vent isolation valves to be' opened and reclc.ed under anticipated containment pressures and vent flow rates during loss of long term decay heat removal severe accidents.
Determine the necessary plant modifications to assure a hard pipe vent path will be available under TW events, develop a schedule for plant modification, and submit the schedule to the NRC within 60 days from the issuance of the backfit order.
Complete the necessary modifications within 30 months from the issuance of the backfit order.
The licensee will be required to have procedures and training to cope with and recover from a TW severe accident.
These procedures should conform to Revision 4 of the BWROG Emergency Procedure Guidelines.
A-3
r o
- g DRAPT i
(3)
The notential safety imoact of chanacs in Diant or oDerational conD1exity, includino the relationshiD to Droposed and existin; t reculatory recuirements The hardened vent capability to be able to cope with the TW event should not add to plant or operational complexity because the vent is noran11y closed and not operated during normal power operation.
It does add some additional hardware to the plant, but it is a simple system.
The containment performance improvement (CPI) program is related to imple-mentation of the Commission's Severe Accident Policy Statement as defined in SECY-88-1472a.
In SECY-88-147 the various programs underway related to closure of severe accident issues were described.
Included among these was the CPI program.
Other programs described in SECY-88-147 are related to the CPI program as the fo31owing discussion indicates.
l Individual Plant Examination (IPE)
The IPE involves the formulation of an integrated and systematic approach to an examination of each nuclear power plant now in operation or under construction for
)
possible significant plant-specific risk contributors that might be missed without a
systematic search.
Supplement i to Generic Letter 88-20 requested that Mark I licensees include in their IPEs the groposed plant im-provements identified in SECY-89-017 other than the j
hardened vent, namely enhanced automatic depressurization system operation, and alternative low pressure water supply for injection into the reactor vessel and for containment sprays. The examination will pay specific attention to. containment performance in striking a balance between accident prevention and consequence mitigation.
It is anticipated that the IPE program may take from three to four years until the last plant has performed the IPE.
Improved Plant Operations (IPO)
I The IPO includos consideration of continued improvements in the Systematic Assessmert of Licensee Performance (SALP) program; regular reviews by senior NRC staff managers to identify and evaluate those plants that may j
not be meeting NRC and industry standards of operating U ibid, $21.
l 2' ibid, $11.
A-4 l
e Q
o +
DRAFT performancel diagnostic team inspectionst improved plant Technical Specifications; improved operating procedures; expansion of the Emergency Operating Procedures (EOPs) to include guidance on severe accident management strategiest industry's programs to reduce transient and other challenges to engineered safety feature systems; feedback from the IPE program of experience and improvements in operational areas, such as maintenance and trainingt and continued research to evaluate the sensitivity of risk to-human
- errors, and the ef fectiveness of operational reliability methods to help identify potential problems early and prevent their occurrence.
The IPO is related to the CPI program's recommendation since we recommend improved procedures and operator training to use the proposed hard vent system.
Severe Accident Research Program (SARP)
The SARP was begun af ter the TMI-2 accident in March 1979 to provide the Commission and the NRC staf f with the technical data and analytical methodology needed to address severs accident issues.
This program has provided input to the NUREG-1150 program and to the CPI program.
Additional research is being carried out to evaluate the need for and feasibility of core debris controls.
Research will also confirm and quantify the benefits..of having water in the containment to either scrub fission products or to prevent or delay shell melt by core debris.
Accident Management The accident management program is concerned with addressing certain preparatory and recovery measures that can be taken by the plant operating and technical staff that could prevent or significantly mitigate the consequences of a
severe accident.
This includes measures taken by the plant staff to 1) prevent core damage, 2) terminate the progress of corn dainage if it begins and retain the core within the reactor vessel, 3) failing that, maintain containment integrity as long as possible, and finally 4) minimize the consequences of offsite releases.
The CPI program recommended plant enhancement would provide the accident management program with sdditional capability to achieve their goals by providing improved hardware with which to deal with a severe accident.
The procedures for using the vent should be re-examined under the Accident Management program.
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(4)
Whether the backfit is interim or final and. if interin, the iustification for innosino the backfit on an interin basis This is the final resolution for BWRs with Mark I containments under the containment Performance Improvement program.
The proposed hardened vent capability is not an interim measure.
(5)
Potential chance in the risk to the oublic from the accidental offsite release of radioactive material Implementation of the proposed hardened vent capability is expected to result in an estimated risk reduction to the public in terms of man-rem as,shown in Column G of Table 2, over the remaining plant life.
(6)
Potential imoact on radiolocical uposure of facility t
employees The reduction in occupational exposure resulting from reduced core damage frequencies and associated post accident cleanup and repair activities has not been quantified, but could be substantial if the hardened vent prevents reactor building contamination.
The estimated total occupational exposure for installation of the hardened vent path is expected to be negligible.
No increase in occupational exposure is expected from operation and maintenance of the hardened vent system.
In fact, if the vent is ever used, it should result in a lower l
risk to employees because of the reduced potentini for vent path failure and the resulting reactor building contamination.
(7)
Installation and continuina costs associated with the backfit.
I includino the cost of facility downtime or the cost of construction delav 1
[
The plant is operating, thus there are no costis associated with construction delays.
The hardened vent path is expected to be capable of being installed with the plant operating or during normal plant
- outages, thus there are no costs associated with additional plant dowmtime.
l The estimated cost of the hardened vent system is shown in Column H of Table 2.
(8)
The estimated burden on the NRC associated with the backfit and the availability of such resources 4 N h
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DR/JT The estimated total cost for NRC review of industry submittals is $17,000 based on as estimated expenditure of 200 man-hours for review of the submittals.
(9)
Consideration of important_ aualitative f actors bearina on the need for the backfit at the Darticular facility The installation of the hardened vent will provide greater flexibility in managing accidents, other than the loss of long term decay heat removal events, and will provide defense in i
depth.
(The hardened vent provides another method of removing heat from the containment.)
(10) Statement affirmino aceropriate interoffice coordination related to the crocosed
.backfit and the olan for implenentation The proposed backfit has been developed as a cooperative effort between the Offices of Nuclear Regulatory Research and Nuclear Reactor Regulation with consultation with the office of General Counsel.
The implementation is anticipated to be handled within the office of Nuclear Reactor Regulation.
The staff has considered how this backfit should be prioritized and scheduled in light of other related regulatory activities.
Within 60 days after issuance of the backfit order, the licensee is to provide to the NRC a schedule for implementing any necessary equipment and procedural modifications to meet the performance goals and to provide adequate defense-in-depth.
All plant modifications are to be installed, procedures revised, and operators trained not later than 30 months from the issuance of the backfit order.
(11) Basis for recuirina or oermittino implementation on 'a particular schedule other schedules were considered; however, the staff believes the proposed implementation of the hard pipe vent capability can be performed with minimum interf acing with containment and engineered safety feature cystems and either with the plant online or during a normal refueling outage.
Therefore, the staf f believes the schedule is achievable without unnecessary financial burden on the licensee for plant shutdown.
The schedule allows reasonable time for the implementation of necessary hardware to achieve a reduction in the risk from TW.
Shorter or less flexible schedules would be unnecessarily burdensome.
(12) EgAgdule for staff actions involved in implementation and i
verification of imolementation of the backfit as ap.p,Iogrpg A-7 s
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DRAFT The proposed backfit is to be installed under 10 CFR 50. 59 and, thus, would require minimal staff effort.
(13) Inoortance of the oroposed backfit considered in licht of other s a Lc,ty-rel a t eql activities underway at the affected facility The proposed backfit is not expected to be directly related to any other safety-related activities that may be underway at the affected facility.
(14) StateEent of the consideration of the _orocosed clant-soecific
,backfit as a ootential ceneric backfit The staff considers that a plant-specific backfit provides a better accounting of the plant dif ferences in risk reduction, and benefits to be gained, than a generic backfit.
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