ML20235U460
| ML20235U460 | |
| Person / Time | |
|---|---|
| Issue date: | 01/23/1989 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| TASK-PINV, TASK-SE SECY-89-017, SECY-89-017-01, SECY-89-17, SECY-89-17-1, NUDOCS 8903090205 | |
| Download: ML20235U460 (92) | |
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POLICY ISSUE SECY-89-017 January 23, 1989 For: TheCommissi(onersNotation Vote)
From: Victor Stello, Jr.
Executive Director for Operations
Subject:
MARK I CONTAINMENT PERFORMANCE IMPROVEMENT PROGRAM
Purpose:
To present staff recommendations on Mark I containment performance improvements and other safety enhancements. [
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This paper covers a major policy question. (
Category:
Summary: As noted in the Integration Plan for Closure of Severe Accident Issues (SECY 88-147) and in interim reports to the Comission (SECY 87-297 and SECY 88-206), the staff has undertaken a program to determine what actions, if any, should be taken to reduce the vulnerability of containments to severe accident challenges. The containment performance improvement effort is one main element of the integrated approach to closure of severe accident issues. Staff efforts have focused initially on BWR plants with a Mark I containment. The staff has now completed its assessment of generic severe accident challenges and failure modes as well as potential improvements for plants with the Mark I containment. The assessment of Mark II, Mark III, and other containment types are the subject of parallel but separate program efforts, as discussed in SECY 88-147.
Probabilistic Risk Assessment (PRA) studies have been performed for a number of BWRs with Mark I containments.
These studies indicate that BWR Mark I risks are dominated by loss of Long-term Decay Heat Removal, Station Blackout (SBO), and Anticipated Transient Without Scram (ATWS) sequences. Although these studies do not show the BWR Mark I plants to be risk outliers as a class relative to other plant designs, they do suggest that the Mark I containment integrity could be challenged by a large scale core melt accident, principally due to its smaller size. However, estimates of containment failure likelihood under such conditions are based on analysis of complex accident conditions, where there remains a broad band of uncertainty.
Contact:
W. Beckner, RES 492-397S L. Soffer, RES hpo@
492-3916 f 3
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?,30 5 89030.2 89-017 PDC .
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2 The staff has concluded that the optimum way to reduce overall risk in BWR Mark I plants is to pursue a balanced approach utilizing accident prevention and mitigation.
Based on our assessment, the staff recommends five specific improvements for Mark I containment plants: 1) an improved hardened vent capability, 2) improved reactor pressure vessel (RPV) depressurization system reliability, 3) an alternate water supply to the reactor vessel and drywell sprtys, 4) extended emergency procedures and training and
- 5) accelerated staff actions to implement the station blackout rule. These improvements, although not representing large changes to the plants, form an integrated set which, when fully implemented, will substantially enhance the safety of Mark I plants by enhancing defense-in-depth, including improvement to containment performance.
The staff has evaluated them and found them to be cost effective. The staff proposes that orders be issued to all licensees with Mark I containments to implement these improvements.
Background:
The Reactor Safety Study (WASH-1400) found that, for the Peach Bottom BWR Mark I nuclear plant, even though the core melt probability was relatively low, the containment could be severely challenged if a large core melt occurred.
Based on this conclusion and reinforced by the anticipation of similar findings (subsequently confirmed) in the draft Reactor Risk Reference Document (NUREG-1150, February 1987) a five element program was proposed in June 1986 to enhance the performance of the BWR Mark I containment. Elements of this proposal included 1) hydrogen control, 2) containment I
drywell spray,)3) control, and containment 5 emergency venting, procedures and 4) core debris training. After l
i the initial proposal, the staff held two separate meetings I in early 1987 with researchers representing NRC contractors and industry. There was a wide range of views expressed regarding accident phenomenology as well as the efficacy of ,
the various improvements. In view of the lack of technical I consensus on the effectiveness of the proposed improvements, the staff decided to undertake additional efforts. In July 1987, the staff informed the Commission of its intention to examine the Mark I issue in the context of an integrated approach to the closure of severe accident issues.
On December 18, 1987, the staff issued a plan (SECY B7-297) l for resolving generic severe accident containment ;
performance issues for Mark I and other containment types. !
As part of the plan, a workshop was held on February 24-26, l
1988 to discuss a number of issues associated with Mark I containment challenges, failure modes and potential containment improvements with researchers, industry representatives and interested members of the public. A i major topic at the workshop was the phenomena associated j
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with containment steel;shell meltthrough. Although uncertainties remain in. estimating containment failure,.
there was'a growing consensus'that water in the containment Lmay help mitigate. risk by fission product scrubbing 'and possiblyL by preventing or delaying containment.shell melt:
- by core debris. . This is' discussed further in Enclosure 6.
L 'The. Integration Plan for: Closure of Severe Accident Issues, (SECY 88-147) characterizes the containment performance improvement effort as being one of the main elements of the integrated approach to closure of severe accident issues.
Other main elements include a) Individual Plant Examinations'(IPEs), b) improved plant operations, c) the'-
severe accident research program, d) examination of, external events, and e) a program on accident management..
The containment performance improvement program is related' to the IPE effort, and is considered complementary to it, since this effort is primarily focused on the potential
. generic vulnerabilities of specific containment classes, whereas the IPE effort is focused on plant unique vulnerabilities.
A Commission paper (SECY 88-206) dated July-15,1988 provided' a status report on the staff's. efforts regarding the Mark I containment performance evaluation. This paper reaffirmed that the risk from BWR Mark Is is low. Nevertheless, the staff proposed a program intended to further reduce overall risk in BWR Mark I: plants by pursuing a balanced approach involving accident prevention and mitigation. .A number of safety enhancements were identified which appeared attractive
-in terms of their potential risk reduction capability as well-as implementation costs.
Following that meeting the Comission requested additional-information via a staff requirements memorandum dated August'1, 1988. Responses to these questions are included as Enclosure 1.
Discussion: Probabilistic Risk Assessment (PRA) studies for BWRs .
indicate that accidents initiated by transients rather than Loss-Of-Coolant-Accidents (LOCAs) dominate the total core damage frequency estimates. The principal accident sequences for BWRs consist of Long-term Loss of Decay Heat' Removal (TW), Station Blackout (SB0) -and Anticipated L TransientWithoutScram(ATWS). WASH-1400 indicated that.
TW is the dominant core damage accident sequence for Peach Bottom. Draft NUREG-1150, however, indicated that the-dominant contribution to core melt frequency at Peach Bottom is due to Station Blackout, and estimated that TW has been greatly reduced at Peach Bottom by implementation of containment venting procedures with the assumption that said venting actions can be successfully accomplished. For those plants in which TW has been eliminated as the dominant contributor', the residual risk is largely due to-
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4-ATWS and SB0 sequences. These studies also indicate that j the estimated likelihood of core damaging accidents for existing Mark I plants is predicted to vary widely over two orders of magnitude or more. The primary containment challenges and potential failure modes for BWR Mark I containments are shown in Enclosure 2.
The. staff has examined potential Mark I containment and
. , plant improvements in the following six areas: (1) hydrogen control, (2) alternate water supply for reactor vessel injection and containment dr s, (3) containment pressure relief capability venting),
(ywell spray (4) enhanced RPV depressurization system reliability, (5) core debris controls,and(6)proceduresandtraining. Each of these was evaluated to determine their potential benefits in terms of reducing the (1) core melt frequency, (2) containment failure probability, and (3) offsite consequences.
Hydrogen Control:
Although BWR Mark Is are required to be operated with an inerted containment atmosphere, plant Technical Specifi-cations permit de-inerting to commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to plant shutdown, and do not require inerting to be completed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after plant startup, in order to permit plant personnel access. In the event of a severe accident, such as a long-term station blackout, a concern was expressed that loss of control of the valves and containment leakage could eventually lead to containment de-inerting.
Two potential improvements with regard to hydrogen control were evaluated. These were: (1)eliminationofthetwo24 hour de-inerted periods and (2) providing a backup supply of nitrogen. Since the probability of a severe accident occurring during either of the two 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> de-inerted periods is small compared to the probability of accident _
occurrence during normal operations, eliminating this time of de-inerting would not significantly reduce risk.
During a severe accident, reactor pressure is anticipated l
( to increase, releasing steam and non-condensable gases into the containment. This will increase containment pressure, preventing ingress of air. Therefore, the containment atmosphere would not become de-inerted for an extended period of time. Since offsite supplies of nitrogen could readily be obtained during this period, an onsite backup supply of nitrogen would not significantly reduce risk.
Therefore, the staff concludes that additional Mark I l
improvements to control hydrogen beyond the existing hydrogen control rule and the procedures in Revision 4 of
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5 the Emergency Procedure Guidelines would have no ~
significant benefit and are not warranted.
Alternate Water Supply for Drywell Spray / Vessel Injection An important proposed improvement would be to employ a backup or alternate. supply of water and a pumping capability that is independent of normal and emergency AC power. By connecting this source to the low pressure residual heat removal (RHR) system as well as to the existing drywell sprays, water could be delivered either into the reactor vessel or to the drywell, by use of an appropriate valving arrangement.
An alternate source of water injection into the reactor
- vessel would greatly reduce the likelihood of core melt due to station blackout or loss of long-term decay heat removal, as well as provide significant accident management capability.
Water for the drywell sprays would also provide significant mitigative capability to cool core debris, to cool the containment steel shell to delay or prevent its failure, and to scrub air borne particulate fission products from the atmosphere.
A review of some BWR Mark I facilities indicates that most plants have one or more diesel driven pumps which could be used to provide an alternate water supply. The flow rate using this backup water system may be significantly less than the design flow rate for the drywell sprays. The potential benefits of modifying the spray headers to assure a spray were compared to having the water run out of the spray nozzles. Fission product removal in the small crowded volume in which the sprays would be effective was judged to be small compared to the benefit of having a water pool on top of the core debris. Therefore, modifications to the spray nozzles are not considered warranted.
Containment Pressure Relief Capability (Venting):
Venting of the containment is currently included in BWR emergency operating procedures. The vent path external to existing containment penetrations typically consists of a ductwork system which has a low design pressure of only a few psi. Venting under high pressure severe accident conditions would fail the ductwork, release the containment atmosphere into the reactor building, and potentially contaminate or damage equipment needed for accident recovery. In addition, with the existing hardware and procedures at some plants, it may not be possible to open or to close the vent valves for some severe accident scenarios. The staff has concluded that venting, if
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plant risk. However,' venting'via a sheet. metal ductwork path, as currently implemented at some Mark I plants, is likely-to greatly-hamper or complicate post-accident recovery activities, and is therefore viewed by the staffE as yielding reduced improvements in safety. . The capability i^
to vent has long been recognized as important in reducing risk from operation of_BWR Mark I. facilities for loss of:
. long term decay heat removal events. Controlled venting '
can prevent the failure of ECCS pumps from inadequate net '
-positive suction head and re-closure of the ADS valves.
The staff agrees with this view as long as the potential '
downsides of_using the existing hardware are corrected.
i A hard pipe-vent capable.of withstanding the anticipated. !
severe accident pressure loadings would eliminate these !
disadvantages...The vent isolation. valves should also be i remotely operable from the control room and should be ' j provided with a power supply independent of normal or emergency AC power. Other changes, such as raising the ;
RCIC. turbine back pressure trip setpoint, may also be e desirable.and should be considered as part of the IPE.
This capability, in conjunction with proper operating procedures and other improvements discussed in this paper, ,
would result in greatly reducing the probability of core ;
melt due to the TW and SB0 sequences. ]
Given a core melt accident,~ venting of the wetwell would '
provide a scrubbed venting path to reduce releases of particulate fission products to the environment. Venting has been estimated to reduce the likelihood of late i containment over-pressure failure and to reduce offsite consequences-for severe accident scenarios.in which the containment shell does not fail for other reasons. Failure-of the shell due to core debris attack (shell meltthrough) would. reduce the benefits from venting'in that-it would release fission products directly.into the reactor building.
Inadvertent' venting could result in the release of normal coolant radioactivity.to the environment even when' core degradation 'is averted or vessel integrity maintained.
Measures to reduce'the probability of inadvertent venting should be considered in the vent design.
Enhanced Reactor Pressure Vessel (RPV) Depressurization System Reliability:
The Automatic Depressurization System (ADS) consists of L relief valves which can be manually operated to depres-surize the reactor coolant system. Actuation of the ADS valves requires DC power. In an extended station blackout l
after station batteries have been depleted, the ADS would not be available and the reactor would re-pressurize. With
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7 enhanced RPV depressurization system r,eliability, depressurization of the reactor coolant system would have a greater degree of assurance. Together with a low pressure alternate source of water injection into the ,
reactor vessel, the major benefit of enhanced RPV depressurization reliability would be to provide an additional source of core cooling which could significantly reduce the likelihood of high pressure severe accidents, such as from the short-term station blackout.
r Another important benefit is in the area of accident mitigation. Reduced reactor pressure would greatly reduce the possibility of core debris being expelled under high pressure, given a core melt and failure of the reactor pressure vessel. Enhanced RPV depressurization system reliability would also delay containment failure and reduce the quantity and type of fission products ultimately released to the environment. In order to increase reliability of the RPV depressurization system, assurance !
of electrical power beyond the requirements of existing regulations may be necessary as discussed later in this paper. In addition, performance of the cables needs to be reviewed for temperature capability during a severe accident.
Core Debris Controls:
Core debris controls, in the form of curbs in the drywell and/or curbs or weir walls in the torus room under the wetwell have been proposed in the past to prevent containment shell meltthrough and to retain sufficient-water to permit fission product scrubbing. However, as noted in SECY 88-206, the technical feasibility for such controls has not been established, and the design and installation costs as well as the occupational exposure during installation could be significant. There is a growing consensus that water in the containment (from an alternate supply to the drywell sprays) may help mitigate risk by fission product scrubbing and possibly by preventing or delaying containment shell melt by core debris. Research is continuing in order to confirm'and help quantify these initial conclusions.
A discussion of Mark I shell meltthrough phenomera and the current state of knowledge is included in Enclosure 6.
Emergency Procedures and Training:
A major element of the Mark I containment performance improvement evaluation involves emergency procedures and training. Current emergency operating procedures (E0Ps) are symptom-based procedures that originated from require-ments of TMI Task Action Plan item I.C.1. Plant-specific E0Ps are generally implemented based on generic Emergency
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8 L Procedure Guidelines (EPGs).. developed by the.BWR Owners
. Group.. As part of.the balanced approach-to examining potential BWR Mark I plant' improvements, both the generic.
EPGs and.the plant-specific implementation of E0Ps.and training have been examined.
NRC has recently reviewed and approved Revision 4.ofLthe BWR Owners Group EPGs (General Electric Topical Report NED0-31331, BWR Owner's Group " Emergency Procedure Guidelines, Revision 4," March 1987). Revision _4 to the BWR Owners Group EPG is a significant improvement over earlier versions.in that they continue _to be based on-symptoms, they have been simplified, and all open items from pre'tious versions have been-resolved. The BWR EPGs e extend well beyond the design bases and include many actions appropriate for severe accident management.-
The improvement to EPGs is only as good as the' plant-specific E0P implementation and the training that operators receive on use of the improved procedures. A recent staff
-safety evaluation report (Ltr.'Thadani to Grace, " Safety Evaluation of 'BWR Owners' Group - Emergency Procedure
~ Guidelines, Revision 4,' NED0-31331, March 1987," dated September 12,1988) encouraged licensees to implement Revision 4 of the EPGs.and reiterated the need for proper implementation and training of operators. Implementation of the guidelines has been voluntary, but is strongly recommended in the SER.
Impact of Existing Requirements:
As part of the balanced approach, for completeness, and to-provide a' more accurate picture of Mark I plant risk, the staff has also evaluated the impact on Mark I risk of several recent rules that have been imposed on light water reactors - the Station Blackout Rule and the ATWS Rule. As discussed earlier, PRAs typically indicate that Mark I reactor risks are dominated by TW, SB0 and ATWS sequences.
Upon implementation of-these two rules at all' Mark I plants, risk from SB0 and ATWS sequences would be expected to be substantially reduced. Implementation of the ATWS rule is' essentially complete. The staff intends to assure that, for the BWR Mark I plants, the improvements required by the SB0 rule and those being proposed in this paper-are appropriately coordinated. The response to Question #2 in Enclosure 1 provides a discussion of expected risk reductions from changes to Mark I plants as a result of ;
these rules.
Assuring the operability of the proposed improvements under severe accident conditions, including an extended period of station blackout, may require assurance of electrical power beyond the requirements of the recent Station Blackout (SB0) rule, 10 CFR 50.63. The proposed improvements have
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been coordinated with the requirements of the'SB0 rule in order not to cause an undue proliferation of power supplies, which could be countar-productive to- safety. .The
.SB0staffrule proposes that licensees by use.of;an_ alternateintending)to AC (AAC source, implement need not the provide' additional electric power supplies for the proposed
- Mark I improvements, provided that the capacity of the AAC i- 'is sufficient for the requirements of both the SB0 rule and=
the. improvements proposed here. Further details are given in Enclosures 7 and 8.
Benefit of Improvements:
The (1) an improvements that the improved hardened staff is venting capabilit recommending) include:
(2 improved {
RPVdepressurizationsystemreliability,(y[analternative 3 4 water-supply to the reactor vessel and drywell sprays, and (4) emergency procedures and training. Accelerated staff ,
review of the. implementation of the existing station blackout rule' for Mark I plants is.also planned. These ll improvements are unchanged from those indicated in the 1
~ interim report (SECY 88-206) to the Commission. ,j I
A major benefit of these improvements is that:they can- .
provide a reduction in core melt' frequency of about a factor of five to ten. With the proposed enhancements,.the coremeltfrequegcywouldbeexpectedtobereducedto about 1 to'2x10 per reactor-year. It should be noted-that'these estimates apply to internal events only.
For plants with a high.TW probability, a .large fraction of the reduction in core melt frequency is-attributable to q improved venting which, by allowing the removal of l long-term decay heat from the containment, greatly reduces "j
'the likelihood of core melt from the TW' sequence. Another reduction in core melt frequency from station blackout is' l-L
' attributable to the enhancements taken together. In'the event of station blackout, enhanced RPV depressurization reliability would permit depressurization of the reactor, availability of a low pressure backup source of water injection into the vessel would permit core cooling, while venting would allow decay heat removal from the containment.
It is important to note that under these circumstances, venting would greatly reduce the likelihood of core damage and not result in releases of fission products of any significance.
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Accident mitigation benefits are also considered to be significant. Mitigation of fission product releases would be realized for all accident sequences, including ATWS scenarios. Venting would be effective in preventing containment failure arising from slow over-pressurization.
Venting via the suppression pool would provide significant scrubbing of non-noble gas fission products by about a l 1
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I 10 factor of 10 to 100 if no containment shell failure occurs.
Water in the drywell may be effective in preventing or at least delaying failure of the containment shell by molten core debris. Finally, even if shell failure should occur, the presence of a water layer atop.the core debris combined with the drywell spray would reduce any source term releases to the environment by a factor judged to range from 2 to 10. This represents a significant enhancement of defense-in-depth.
Because of the combination of reduced core melt likelihood, reduced fission product releases due to mitigation, and possible reduction or elimination of a significant contain-ment failure mode, tM staff concludes that the overall risk reduction of the proposed improvements is in excess of one order of magnitude.
The benefitt of the ossposed enhancements in terms of their reduction in offsite risk can be calculated in terms of person-rem. Depending upon the probability of core melt due to the TW sequence the estimated reduction in risk, expressed in person-rem, for the proposed enhancements ranged from about 145 person-rem per, reactor-year to about 1330 person-rem per reactor-year, for plantg having a probability of core melt due to TW of 1x10~ per reactor-year and 1x10~4 per reactor-year, respectively. Of this total value, the risk reduction produced by lowering the likelihood of core melt due to station blackout and mitigation of ATWS accounts for a reduction of about 33 to 210 person-rem per reactor-year. For plants whose robabilit p(about 10~y per of core melt due reactor-year), to the the bulk of TW the risk sequence is high reduction can be attributed to the large reduction in the TW sequence brought about by improved venting. Additional details are provided in Enclosure 4 Finally, as noted earlier, the recommended improvements form an integrated set in the sense that they complement each other in prevention or mitigation. This results in the maximum risk reduction when all are taken together.
Summary of Costs of Improvements:
Cost estimates were made of the proposed improvements.
These are given in Enclosure 3 which provides a cost summary for all improvements that includes high and low estimates ranging from $3.1 to $1.6 million dollars.
Estimates of cost as high as $7.3M were obtained based on .
actual costs of similar improvements at an existing Mark I plant. For purposes of the regulatory analysis included in Enclosure 4, a cost range of $2.0M to $7.3M has been used. i Actual costs at many plants may be less since, as shown in i Enclosure 5, some plants already have many features of the proposed improvements.
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Conclusions:
Many of-the proposed' enhancements would require plant backfits. The' staff has examined these in light of the
-backfit rule,10 CFR 50.109. .Section (a).3 of that regulation-indicates that the Commission shall require-
-backfitting only when "there is a substantial increase.in the overall protection of the public health and safety" and "that the direct and indirect costs ... are justified in view of this increased protection".
1 In' reaching a: conclusion with respect to the first test
-_ indicated above, the~ staff considered the effect of the '
. proposed. enhancements upon reductions in core melt-frequency and improved containment performance. A major benefit of these enhancements is~in their' ability to reduce the likelihood of' core melt. Core melt frequencies for BWR
-Mark I plants prior to any of the enhancements considgred
' would be expected to range from about 1x10T to.2x10~ . per-reactor year. With the combined enhancements, core melt frequency would be reduced by about a factor of five to.
ten. : Thus,' the proposed enhancements clearly offer a substantial. reduction in core melt frequency. The core.
melt frequency reductions do not give credit for existing venting capability assumed in NUREG-1150 since the current venting capability at. plants has significant uncertainty regarding its overall effectiveness.
The increased ability to' cool core debris and to, remove-excess heat from the containment by. venting, given the occurrence of an accident, is also expected to reduce the likelihood of containment: failure, although this is not as readily. quantifiable' because of-the uncertainty in core melt progression and containment shell meltthrough phenomenology which:is discussed in Enclosure 6. . In addition, the ability to scrub _ particulate fission products by use of _ venting'through the suppression pool and by the use of a water layer atop any core debris also adds significant mitigative capability.
Since the proposed enhancements would be expected to reduce the likelihood of core melt by about a factor of five to ten, and provide significant additional' accident mitigation capability as well, the staff concludes that the proposed enhancements provide a substantial increase in the overall protection of the public health and safety as well as providing enhanced defense-in-depth.
With regard to the second or cost-benefit test required by the backfit rule, the discussion given earlier has shown that the costs of the enhancements are estimated to range from 1.6 to 3.1 million dollars per plant, although similar .
l improvements at an existing Mark I plant may. have cost about 7.3 million dollars. Both the estimated cost and the l cost associated with an existing Mark I plant were used in the cost-benefit analysis. Based on the results of a
12 survey of nine Mark I plants, the staff believes that many plants have some of these improvements already in place.
Since the estimated benefits ranged from 3.6 to 33 million dollars per plant based upon 1000 dollars per person-rem and an average remaining plant lite of 25 years for Mark I plants, the staff concludes that the proposed enhancements are generally cost beneficial.
For the reasons stated above, the staff concludes that backfit of these proposed enhancements is warranted for all Mark I plants.
Options: 1. Take no action. Pro: No further resources would be required. Con: This option would result in a situation where a number of enhancements to safety that the staff believes to be cost effective would not be implemented and closure of severe accident issues would not be obtained for Mark I plants.
- 2. Issue a generic letter. Pro: This option would be the quickest way to inform licensee. cf the staff's views and would require the least resources. Con: The generic letter can inform industry of the staff's finding, but can only request, not require, licensees to make changes to their facilities.
- 3. Issue an order. Pro: This option could be accomplished quickly and provide a regulatory requirement to implement the improvements. Con: This option could result in requests for hearings from both licensees and interveners contesting the orders which could result in lengthy delays.
A draft proposed order is included as Enclosure 7.
- 4. Initiate Rulemaking. Pro: This option would provide a regulatory basis for requiring the improvements. It is generally preferable to impose generic requirements by rule. Con: This option would require some staff resources and cause a delay in implementing the proposed improvements. A draft proposed rule is attached as Enclosure 8.
Recommendations: The proposed improvements could be implemented as a regulatory requirement either by use of orders or through rulemaking. Of these two viable options, although the stati considers that it is generally preferable to impose generic requirements by rule, it is believed that, absent a hearing, the improvements could be carried out more quickly via orders. Therefore, in order to implement these improvements which represent a substantial enhancement to the safety of Mark I plants, the staff recommends that orders be issued to require the improvements. The staff would also prepare an Environmental Assessment of venting of the containment using the improved hardware and procedures.
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Coordination: OGC has no legal objections, but views rulemaking as the preferred option due to the potential for delay beyond NRC's control if hearings are requested. The Comittee for.
Review of Generic Requirements has recomended an expedited
- rulemaking to require these improvements. The ACRS has reviewed these recommendations and will provide their comments separately.
Victor Stello, Jr.
Executive Director for Operations
Enclosures:
- 1. Response to Comission Questions
- 2. Mark I. Challenges and Relative Likelihood of Failure Modes
- 3. Sumary of Costs
- 4. Regulatory Analysis
- 5. Results of Survey of Mark I Plants
- 6. Mark I Liner Melt Status
- 7. Draft Proposed Order
- 8. Draft Proposed Rule Commissioners' comments or consent should be provided directly to the Office of the Secretary by c.o.b. Wednesday, February 8, 1989.
Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Monday, January 30, 1989, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires additional time for ,
f analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
DISTRIBUTION:
Commissioners OGC OIA GPA REGIONAL OFFICES EDO ACRS ASLBP ASLAP SECY
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ENCLOSURE 1 .
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ENCLOSURE 1 RESPONSES TO TOPICS IN STAFF REQUIREMENTS MEMORANDUM DATED AUGUST 1, 1988 In the staff requirements memorandum dated August 1, 1988, the Commission requested that the staff address the following topics with its final recommendations:
Topic 1: The relative risk from a severe core damage accident for boiling versus pressurized water reactor containment designs. 1 Staff response The relative risk from the two reactor designs is comprised of two factors: !
the likelihood (frequency) of an accident resulting in significant core damage, and the likelihood (probability) of a major containment failure after such an accident has occurred. Because both the reactor and the containment designs differ, both factors must be compared.
Regarding the first factor, a comparison of the two designs can best be made by examining the results of those probabilistic analyses already available t the staff. Core damage frequencies for BWRs and PWRs typically range from to 10~3 10 , although a strict comparison is not warranted since the studies were
, performed over 12-year period. Significant improvements in emergency operating procedures and Anticipated Transient Without Scram protection are not reflected in all of them and the assumptions and methods of the studies vary. Also, a number of the studies were intended as conservative bounding analyses. In general, however, core damage frequencies for BWRs tend to be lower than that for PWRs. The reasons for lower BWR core damage frequencies are included as part of the response to topic #2 below. It should be noted, however, that BWRs with Mark I containments generally have a slightly higher estimated core damage frequency than BWRs with Mark II and III containments.
(This is not unexpected, since the Mark II and III containments are associated withnewerreactordesigns.)
The second factor is the likelihood of containment failure, given that a severe core damage event has already taken place. The question of containment performance given a severe core damage accident is, of course, a subject of very active research, the purpose of which is to reduce the uncertainty in containment failure s
likelihood. Direct containment heating in the large dry PWR containment and liner melt-through in the Mark I BWR containment contribute to uncertainty in containment performance. <
The phenomenon of direct containment heating in PWRs was identified as the result of experiments in which simulated molten fuel debris was discharged into scaled containment volumes under high pressure. It was discovered that, in the process of melt ejection, fine droplets of molten material were sprayed throughout the airspace, leading to rapid heating of the air, both as the i result of a high rate of heat transfer from the droplets to the air and the chemical reaction of metals in the droplets with air and steam. The expected magnitude of the direct containment heating effect, and its contribution to containment failure probability, is currently a matter of considerable uncertainty.
2 The phenomenon of shell melt-through in the BWR Mark I design corresponds to the direct attack and melt-through of the drywell wall by molten core debris after vessel breach. If the drywell floor is not water covered, the molten fuel debris may flow out of the pedestal region, onto the drywell floor, and into contact with the containment wall. If the floor were water covered, it is likely that this mode of failure would be delayed or even prevented. Both scenarios are also a matter of considerable uncertainty as is discussed in ,
detail in Enclosure 6.
In summary, the likelihood of a core damage event in a BWR with a Mark I containment is believed to be less than that of a PWR, and the probability of early containment failure is likely to be higher for the Mark I design. On balance, the overall risk is believed to be generally similar.
Topic 2: The reduction in core melt frequencies resulting from the requirements of the anticipated transient without scram and the station blackout rules. The staff should address the relative benefits in risk reductions for both BWRs and PWRs, including the rationale for the differences.
Staff response It has been the staff's experience that probabilistic analyses of BWRs are generally dominated by long term loss of decay heat removal, ATWS and station
- blackout sequences. The reasons far this are very much a matter of individual plant design, but some general reasons for this can be given:
- BWRs are equipped with a large number of systems which inject water into the vessel - up to three trains of main feedwater, HPCI/HPCS, RCIC, two trains of low pressure core spray, two or more trains of LPCI, etc. This has the effect of reducing the importance of LOCA sequences.
- BWRs have not experienced the reactor coolant pump seal failures that have surfaced at many PWRs, which further reduces the importance of small LOCAs.
- BWRs can reduce primary system pressure much more easily than PWRs, which greatly increases the spectrum of events in which the low pressure systems can be used effectively.
- ATWS events provide more of a challenge to BWR containments than to PWR containments.
ATWS tends to be of greater significance in a BWR due to the significant challenge to the containment that results from an ATWS in a BWR.
i It should be noted that dominant sequence identification should be used in conjunction with core damage frequency. The existence of a dominant sequence category does not of itself mean that a safety problem exists. For example, if
4 .
3 1
a plant already had an extremely low estimated core damage frequency, the j existence of a dominant sequence category would have little significance.
Similarly, if a plant were modified such that the dominant sequences were eliminated entirely, the next highest sequence category may well become the new dominant category. It should be noted that in the NUREG-1150 study, the two BWRs were indeed dominated by station blackout and had ATWS as the next most important sequence category - but also had lower total core damage frequencies than those calculated for the three PWRs.
Although station blackout and ATWS appear in the PWR analyses, the core damage frequencies for the two BWRs (Peach Bottom and Grand Gulf) are dominated by station blackout. If station blackout were removed, ATWS would then become the new dominant category. In general, the relative importance (i.e. contrib9 tion to core damage frequency) of station blackout and ATWS sequences and consequently the benefits of reducing the frequencies of these sequences, are expected to be greater in BWRs.
The magnitude of the change in core damage frequency associated with station blackout or ATWS modifications is more difficult to estimate. Currently, there are no such before and after studies other than those done in support of the Station Blackout and ATWS regulatory analyses. Of these studies, NUREG-1109 estimatedareductionincoredamagefrequencyasgociatedwithresolutionof the station blackout issue of from 0.6 to 8 x 10- per reactor-year, for both BWRs and PWRs. Similarly, NUREG-0460 estimated a reduction in core damage frequencyassociatedwithresolgtionoftheATWSissueoffrom1to2x10,4 for BWRs, and from 4 to 8 x 10~ for PWRs. ,
Topic 3: Safety implications of a fire when diverting the fire water to the decay heat removal system.
Staff response Assuming a station blackout event and that the ability to cool the core was dependent solely upon the diesel fire pump supply, there should be no intention of compromising that sole cooling water supply by utilizing it for possible manual fire fighting.
The following factors, which exist at all plants, would mitigate the consequences of fire concurrent with station blackout.
At most plants the capacity of the diesel fire pump exceeds the core cooling requirements by several hundreds of gallons per minute. This would still permit use of one or more hand hose lines for manual fire fighting in the event of station blackout.
Automatic detection would still be available throughout the plant since each detection system is required to have its own 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery power supply.
Based upon experience, most fires in nuclear power plants are controlled and extinguished by use of manual fire extinguishers provided throughout the plant because of early detection and fire brigade response.
- - - - - - - - ~ - - - - - - - - - - _ _ - - _ - - . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
4
- Because of requirements to comply with Section III.J, of Appendix R to 10 CFR Part 50, Emergency Lighting, visual access would be assured to all essential areas of the plant for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for manual fire fighting, should a fire occur during station blackout.
Even though most nuclear power plants are located in remote areas, virtually all have mutual aid agreements with local fire departments.
These can be expected to respond with pumpers capable of drawing from the local water supply (river or lake) into the plant fire main system.
Expected fires at a plant during station blackout would probably be localized electrical fires (most likely associated with the condition that led to the blackout). As mentioned above, these should be detected early and are expected to be extinguished by the use of manual fire extinguishers.
All plants have automatic gaseous (carbon dioxide and/or Halon) fire suppression systems protecting selected plant areas, and these would remain functional during a station blackout. In addition, most CO systems are equipped with hose stations for manual fire fighting. 2 In addition, a few early plants are located near enough towns to have connections to the town water supply as a backup to the normal plant water supply. Also, some plants have fire engine pumpers on site.
In plan'ts which do not have diesel fire pumps, this core cooling mode would not be available. Diesel fire pumps are required as backup in the event of loss of Non-Class IE electric driven pumps. Electric pumps only are permitted, however, if they are powered from the Class I busses from the emergency diesel generators. If this should be the case, there would be no diesel fire pumps, and, therefore, no fire water supply in the event of station blackout.
Topic 4 Detailed outline of the containment emergency venting operator procedures. The decision process and person responsible for the decision to vent should be clearly identified.
. Staff response The Emergency Frocedure Guidelines (EPGs) do not single out venting procedures from any other actions that the plant operators may take to respond to various plant conditions. The EPGs provide a general logic for taking certain actions and precautions based upon available information concerning the status of the plant. Venting is called upon in the BWR EPGs, Revision 4, for containment pressure control and for hydrogen control. Venting for pressure control occurs only after normal pressure control mechanisms (stancby gas treatment, suppression pool sprays and drywell sprays) have been ineffective or could not be operated.
Then venting is initiated prior to reaching the primary containment pressure limit (PCPL). The PCPL is the lowest of:
- 1. SRV operability pressure,
- 2. Containment vent valve operability pressure
- 3. Containment failure pressure, or
- 4. Reactor vent operability pressure
5 The lowest of these four is, in practice, always either #1 or #2. These assure operability of the vent and continued capability to maintain the reactor at low pressure. Maintaining the reactor at low pressure allows the use of low .
pressure systems to provide core cooling. Figure 1 illustrates the logic for containment pressure control used in Revision 4 of the BWR EPGs. ,
Venting for hydrogen control is permitted at low hydrogen concentrations provided technical specification radioactive material release limits are not violated. Venting, even with significant activity release, is permitted to prevent reaching the hydrogen deflagration limit. Venting is also called for if the hydrogen concentration cannot be determined and plant conditions indicate the likelihood of high hydrogen concentrations.
Recent team inspections at BWR plants with Mark I containments concluded that generally the licensees had paid insufficient attention to conditions for venting, the size of the vents, capacity of the vent path including valve operability to withstand pressures under severe accidents, ability to open/
close valves during station blackout, and the impact on equipment located in the reactor building should the low pressure piping fail during venting operation. Therefore, the staff has concluded that venting should be carried out using hardened vent pipe of adequate size and the procedures for initiation and control of vent process based on Rev. 4 of the BWR EPGs.
The EPGs address the technical issues related to venting. The venting guidance is based upon the analysis.and judgement of the most knowledge experts. Details of the decision process and identification of the person (s) making the decision have been left to the iiidividual licensee. We believe that the actual venting decision should be made by the senior manager on site at the time a decision is needed because he/she would have the most complete information on the plant status and offsite conditions for proper coordination of actions. The manager should also notify the NRC and state officials if time and events permit. ,
1
PRIMARY CONTAINMENT CONTROL .
Primary Containment Pressure Control Condition - Action
- Control primary containment pressure below scram setpoint using:
o CNT pressure contral s/s O SBOT .
o DW purge .
Y Operate SP sprefs.
Before PSP /SCSIP '
. ACC permitting ,
V g Operate DW spreys, gp ACC permitting V
Et1ERGENCY RPV Cannot meinteln m DEPRESSURIZATION
, p) ,p3p IS REQUIRED V
Ve t primary Befre PM i containment v
Cannot maintain m Operate SP sprefs, l .
below PCPL irresp.of ACC
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Operate DW sprefs, irresp of ACC FIGURE 1 ACC = adequate core cooling
. PSP = pressure suppression pressure
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ENCLOSURE 2 a - ---
A + I T S S R S o o R o E E A E O 0 KO N Y Y V Y N N N N SP IP RI S
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ENCLOSURE 3 I
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E!! CLOSURE 3
SUMMARY
OF COSTS Comprehensive cost estimates were made for the major containment enhancements under consideration in this paper. Table 1 provides an overview of the major costs involved in^retrofitting the proposed improvements to an existing BWR Mark I nuclear plant. In addition to the costs of physical modifications, other costs considered were: engineering and quality assurance, health physics support, anti-contamination clothing, radioac,tive waste disposal, licensee-NRC interaction, rewriting of emergency procedures, and staff training. On-site personnel radiation exposures encountered as a result of th.ese retrofits would normally be subtracted from off-site accident dosage averted (owing to the enhanced plant performance) to obtain a net exposure to the onsite and offsite population. However, these estimated radiation exposures are negligible compared to the averted accident dosage and, therefore, have not been considered in the regulatory analyses.
The approach taken by)the staff's subcontractor, Science and EngineeringAssoc plant from the BWR Mark I group of 24 units. Peach Bottom Unit 2 was chosen for consistency in terms of other calculations performed in estimating risk reduction for this project.
Major assumptions affecting cost estimates include: .
- 1. No modification requires replacement power; all work will be performed during normal plant operations or scheduled shutdowns.
- 2. Equipment, materials, and structures added to the plant will not be
! designed to meet seismic Category I requirements unless they might impact saf ety-grade equipment.
- 3. New equipment outside containment will not require harsh environment qualification unless its failure would have an-impact on qualified equipment. (See SEA Report 87-253-07-A:1 for a complete list of all assumptions).
As Table 1 indicates, the major plant improvements were subdivided into four subtasks.
Subtask 1: Emergency Power Two alternatives were developed to deliver AC and DC power to the ADS, back-up water supply and wetwell vent valves:
- 1) An additional diesel generator (much smaller than the " station" diesel)
- 2) An additional 125 volt DC battery system The estimate for the DC battery system includes the construction of a non-safety grade concrete structure to house it.
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The emergency power enhancement would be capable of providing both AC and DC power--AC power for valve operators, e.g., the containment vent valves, and DC power for ADS operation.
Subtask 2: ADS Enhancements Included in this subtask are cost estimates for logic changes, the addition of nitrogen gas bottles and cabling improvements. The logic change consists primarily of bypassing the high drywell pressure signal allowing for more complete automation for events such as a break external to the drywell. An additional nitrogen bottle was added to allow for operation of the ADS valves for up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. Existing ADS cabling is qualified for design basis accident conditions with a maximum temperature of 340*F. Low and high cost options for this enhancement consist of either wrapping the existing ADS letely cabling replacingwith the thermal tape cable with (temperature high temperature ratings mineral up to 800cable insulated F) or comp (temperature rating up to 1600 F).
Subtask 3: Back-Up Water Supply This plant enhancement involves modifying the existing Fire Protection System and Residual Heat Removal System piping to allow for a cross-tie between them.
By doing this an additional source of water will be available powered by the diesel driven fire pumps. This water can then be directed to either the containment spray system or injected directly into the reactor vessel, depending upon the accident management prerogative. To accomplish this task electrical circuit wiring and logic changes must be made to assure proper valve alignments and flow path isolation.
Subtask 4: Wetwell Vent Enhancements This objective of this improvement is to take advantage of the scrubbing ability of the wetwell. However, most Mark I wetwell vent paths outside primary containment are not capable of withstanding the pressure environment resulting from severe accidents. Cost estimates were made for a hardened pipe which would tie the wetwell vent penetration to the plant stack and simultaneously bypass the C.t work leading to and from the Standby Gas Treatment System. Electricad connections and logic changes are required to override wetwell vent line isolation signals and to provide power from the new emergency power source discussed above.
In summary, the total estimated cost for the lower bound estimate is approximately 1.6 million and 3.1 million for the upper bound estimate. These estimates are in 1988 dollars and include discounted future costs associated with long term maintenance. The NRC has determined a best estimate cost of 2.0M and has used this cost in the regulatory analysis of the improvements.
The best estimate cost was obtained by assuming the diesel generator and ADS cable wrap options (low cost), combined with the high costs of the SEA estimate for the remaining item in Table 2.
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TABLE 1-BWR MARK I POTENTIAL ENHANCEMENTS - COST
SUMMARY
(rounded 1988$)
LOW HIGH SUBTASK 1- EMERGENCY POWER SOURCE FOR ADS, BACK-UP WATER AND WETWELL VENT VALVES Option 1 ADD DIESEL GENERATOR (low case) 293,000 -
Option 2 ADDDCBATTERYSYSTEM(highcase) - 549,000 SUBTASK 2 ADS ENHANCEMENTS Option 1 WRAP EXISTING CABLING WITH HIGH TEMPERATURE 47,000 -
TAPE (low case)
Option 2 REPLACEADSCABLINGWITHMICABLE(highcase) 968,000 Options 1 & 2 LOGIC CHANGES 55,000 70,000 Options 1 & 2 ADD GAS BOTTLE 12,500 17,000 SUBTASK 3 BACK-UP WATER SUPPLY SPRAY ENHANCEMENTS ELECTRICAL. 66,000 85,000 PIPING MODIFICATIONS 473,000 603,000 SUBTASK 4 WETWELL VENT ENHANCEMENTS PIPE REPLACEMENT 400,000 509,000 ELECTRICAL . 47,000 59,000 OTHER COSTS 1 HEALTH PHYSICS SUPPORT COSTS 4,300 77,800 2 COST OF ANTI-CONTAMINATION CLOTHING 13,500 30,000 3 RADI0 ACTIVE WASTE DISPOSAL COSTS NEGLIGIBLE 4 LICENSEE COSTS FOR:
a Major Redesign Documentation 35,000 35,000 b Rewrite Procedures 11,700 11,700 c Trairing .
127,200 127,200 d) Revise Training Manual 7,700 7,700 TOTAL ESTIMATED COST PER REACTOR 1,592,900 3,149,400
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0 ENCLOSURE 4
-' ^ ' - - - -.--_-_u- _-__-.--_ _ _- - _ _ -_ _ - - . _ _
ENCLOSURE 4 REGULATORY ANALYSIS 1.dSTATEMENTOFTHEPRGBLEM Accidents which exceed those evaluated during the licensing of facilities
- (design basis accidents) have a low probability of occurrence. These.
accidents, known as " severe" accidents, could result in core damage or core melt. The General Electric Company has designed and constructed several Boiling Water Reactor (BWR) configurations with three basic containment designs designated as Mark I, Mark II, and Mark III. The BWRs with the Mark I containment design have the smallest free volume and have been considered to be most susceptible to severe accidents'which could challenge containment-integrity. The potential challenges to containment integrity were reviewed and potential enhancements were proposed to improve the probability of containment survival or to reduce the possiblity of a severe accident.
Draft NUREG-1150 I evaluated the dominant accident sequences for five plants, one of which was a BWR Mark I. The dominant accident sequences were identified asstationblackout(TB),whichincludesthelossofallACandDCpower; anticipatedtransientwithoutscram(TC);andwouldhaveincludedthelossof long term decay heat removal (TW) except that for the particular plant being reviewed this sequence was considered to be non-dominant due to assumed successful venting of the containment. For severe accidents initiated by a station blackout, all existing systems are assumed to fail due to a lack of electricity. The short term station blackout fails all AC and DC power sources immediately while the long term station blackout has immediate failure of all AC power sources and failure of all DC power sources after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
NUREG-1150, " Reactor Risk Reference Documen(", Draft, February 1987.
1
-,..m. - . . _ . __ ...._m. _m - _ _ . _ _ _ _ . _ - _ - _ _ _ _ _ _ _ _ -_____ __
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 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, principally 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.
2.0 OBJECTIVES The staff objective is to reduce overall risk in BWR Mark I plants by pursuing a balanced approach utilizing accident prevention and accident mitigation.
Most r ecent PRA studies indicate that BWR Mark I risk is dominated by loss of decay heat removal (TW), station blackout (TB), and anticipated transient with out scram (TC) sequences. The balanced approach includes: (1) accident prevention - those features or measures that are expected to reduce the
~
likelihood of an accident occurring or measures that the operating staff can use to control the course of a accident and return the plant to a controlled, safe state, and (2) accident mitigation - those features or measures that can reduce the magnitude of radioactive releases to the environment in the event of~
an accident. The containment performance improvement program would provide enhanced plant capabilities and procedures with regard to accident prevention and mitigation.
3.0 ALTERNATIVE RESOLUTIONS Plant modifications are being proposed to reduce the probability of or to mitigate the consequences of a severe core melt accident which consists of modifications to three existing plant systems. The modifications considered are (1) venting of the wetwell, (2) a backup water supply for the residual heat removal system and the containment sprays, and (3) assuring the operability of the automatic depressurization system (ADS). Other modifications were considered, such as additional hydrogen controls, but were not considered to significantly reduce either the probability of a severe accident or 2
consequences given the occurrence of a severe accident. The details of each proposed enhancement is described in this section and estimates of the I
enhancements' benefits in Section 4.
For all proposed modifications, the new components need not be safety-grade or safety-related. However, no failure of a modified system or non-safety-related co,mponent for design basis accidents is to adversely affect any safety-related structure, system, or component.
The effects of the proposed enhancements were evaluated by using a simplified containment event tree (5-CET) for station blackout events with 15 to 20 of the 107 top events used in draft NUREG-1150. The development of the S-CET and corresponding branch point split-fractions relied heavily on the data and insights generated by the draft NUREG-1150 effort. However, instead of trying to consider the entire range of possibilities and their uncertainties, the S-CET assigned best-estimate branch point probabilities. While this approach
. produced a point estimate of the risk.and does not identify the range of uncertainty in the calculations, it provides a concise and flexible model which was easily used to perform sensitivity studies. The results of the S-CET identified each specific event tree end-state and its associated probability.
These end-states were compared with the similar accident progressions from the list of Peach Bottom accident progression bins.2 The end-states from the S-CET were characterized according to the draft NUREG-1150 accident progression 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 S-CET end states were related to those identified in draft NUREG-1150, the consequences were taken directly from draft NUREG-1150 or scaling factors were applied to the draft NUREG-1150 results and interpolated to generate the consequences. The risks were then calculated by multiplying 2
Draft NUREG/CR-4551, " Evaluation of Severe Accident Risks and the Potential for Risk Reduction: Peach Bottom, Unit 2", Volume 3, Draft, May 1987.
3
the plant damage state frequency, the bin probability, and 'the consequences of !
that bin together.
To evaluate the approximate accuracy of the S-CET, the draft HUREG-1150 information related to Peach Bottom was input into the S-CET and the results compared with those of draft NUREG-1150. In all categories, the results of the S-CET compared with those in draft NUREG-1150 within about 25% accuracy, well within the uncertainty band of draft NUREG-1150. Once verified, the advanced information related to final NUREG-1150 was used to form a new base case and to evaluate the benefits of the proposed enhancements. Details of this methodology is documented in NUREG/CR-5225, "An Overview of Boiling Water Reactor Mark I Containment Venting Risk Implications".
3.1 Alternative (i)
Under this alternative no action would be taken.
3.2 Alternative (ii)
This alternative would accelerate the implementation of the existing rules (station blackout and anticipated transient without scram), without any other modifications.
3.3 Alternative (iii)
This alternative would involve alternative (ii) plus a hardened venting capability from the containment wetwell to the plant stack.
The proposed venting improvement would provide a wetwell vent path to the plant stack capable of withstanding the anticipated environmental conditions of a severe accident. This modification would include installation of a 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 ductwork and SGTS.
The hard pipe to the stack could contain a rupture disk to prevent inadvertent 4
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operation and release of radioactivity. In order to vent the wetwell, all isolation devices, except the rupture disk, need to be capable of being operated without reliance'on AC power. The emergency procedures would need to be modified to provide appropriate instructions for the operator.
3.4 Alternative (iv)
This alternative would involve alternative (ii) plus enhanced operability of the ADS.
The proposed improvement to the ADS would consist of a portable generator (and related cabling and controls) to supply DC power for actuation of the ADS valves, ensuring that ADS cables within the containment could survive the severe accident environment, and additional nitrogen gas bottles for valve operation, where necessary. Emergency procedures would need to be modified to provide appropriate instructions for the operator. The use of the ADS would reduce the probability of early containment failure from high pressure melt ejection. In addition, the corium would exit the reactor sooner, but would be cooler than when the reactor is pressurized, thus delaying the potential containment meltthrough due to corium attack. The cooler vessel also promotes plateout of non-gaseous fission products within the reactor vessel.
3.5 Alternative (v) l This alternative would involve alternative (ii) plus a backup water supply to the containment drywell sprays and as a low pressure water injection source to the reactor vessel.
The proposed improvement of the containment sprays would use an alternative water supply and pump to the residual heat removal (RHR) pump discharge line outside of the outboard containment isolation valve. At some plants, this alternate system could use an existing 1000 gpm diesel driven fire water pump or a portable generator to power an existing water pump, such as a service water pump. In addition, spool pieces, piping, and isolation valves would be necessary to cross connect the alternate water system to the RHR system.
5
_ _ . _ . _ _ _ - - _ _ _ _ . _ _ _ mm
E Modifications may be required to some of the RHR valves to permit remote manual operation without reliance upon AC power and to bypass interlocks. The emergency procedures would need to be modified to provide appropriate
. instructions for the operator.
Cool water through the sprays will condense steam and thereby reduce the potential failure of the containment b9 condensible gas over-pressure. The water will tend to scrub the non-noble gas fission products from the drywell atmosphere. The water which, runs down the inside of the steel containment shell will tend to wash and cool the shell and may prevent re-volitalization of ;
fission products. This shell cooling may reduce the potential for containment failure due to over-temperature. With water on and around the corium on the
,drywell floor, the water will cool the corium and may assist in the formation of a crust. With'a crust formed between the molten corium and the containment shell, the likelihood of containment failure by corium attack may be reduced.
The pool of water over the corium is also expected to reduce the fission products that could be released to the environment. .
The use of a diesel powered pump into the RHR system provides an additional low pressure water injection system for the reactor as a preventive feature. Thus, if the reactor is at low pressure and the alternate water system is initiated in a timely manner, the alternate water system could prevent core degradation and arrest core melt within the reactor vesse13, 3.6 Alternative (vi)
This alternative would reduce the overall risk in BWR Mark I plants by a combination of accelerated implementation of existing rules, extended emergency operating procedures and training, and potential implementation of the following hardware modifications:
3 Even for an alternate water supply system that does not provide an adequate amount of water to prevent core degradation, the alternate system would delay severe core damage and thereby increase the likelihood of recovery of a system to arrest core failure and prevent vessel failure.
6
(1) Containment drywell spray: assurance of a backup water supply to the residual heat removal (RHR) system and drywell sprays with AC l independent pumping capability, (2) Containment venting: a harder.ed containment wetwell venting capability with the ability to open and reclose the isolation valves independently of AC power, and (3) Improved reliability of the automatic depressurization system (ADS):
providing additional DC power for the solenoids, upgrade of cables, and additional nitrogen gas supply.
4.0 CONSEQUENCES 4.1 Costs and Benefits of Alternative Resolutions PRAs that the staff has available have been performed on 9 BWR Mark I plants.
These assessments are of varying quality and, in some cases, have considered both external and internal event core melt initiators. The range of core melt ~
frequencies for these units range from 3.x 10-4/ Reactor-Year (RY)4 to 8.2 x 10-6/RY5 . We have recently received the Brunswick PRA which has a core melt frequency of 2.5 x 10-6/RY. However, the staff has not completed the evalua-tion of this PRA.
WASH-1400 identified the dominant accident sequence to be the loss of long t'rm decay heat removal (TW). Peach Bottom core melt frequency due to TW is estimated to be 1x10-5/RY or less without assured venting6 .
4 Core melt frequency from A-45 studies for internal events only.
S ibid, #1.
0 The existing hardware consists of low pressure ductwork from the outboard containment isolation valve, through the standby gas treatment system (SGTS) to the plant stack. The ductwork is designed for less than one psig internal pressure while the venting procedures identify venting containment when the containment pressure is 60 psig.
7
- _ - _ _ . - - _____.-_____._____.._Jm__ _ , _ , . _ , _ _ _ _ _ _ _ _ _ _
1 .
For purposes of this regulatory analysis, it is assumed that all 24 BWR Mark I plants which would use low pressure rated ductwork as part of the containment vent path would have a core melt frequency associated with TW between 1 x ,
10~4/RY and 1 x 10-5/RY. NUREG-1150 has estimated that venting may reduce the !
probability of the TW sequence for Peach Bottom by three orders of magnitude.
The other two dominant accident sequences are station blackout (TB) and anticipated transient without scram (TC). Proper implementation and compliance with the existing station blackout and ATWS rules is ascumed to reduce the probability of the TC and TB sequences to less than 1 x 10-5/RY. Thus, for the purposes of this regulatory analysis, we assume that the plant core melt frequency for all Mark I plants would be less than 1 x 10~4/RY and probably greater than 2 x 10-5/RY.
4.1.1 Alternative (i)
This alternative would be to take no action. At least one licensee has seen the need to provide improved accident management capabilities and, thus, defense-in-depth. Because the value-inipact analysis has shown that it would b'e beneficial to implement the recommendations identified in alternative (vi) which provides for defense-in-depth for accident management, the no-action alternative is not recommended.
4.1.2 Alternative (ii) 4.1.2.1 Value: Risk Reduction Estimates For the BWR Mark I plants, the acceleration of compliance by one year represents a risk savings of approximately 1392 men-rem as identified in NUREG-1109. This could reduce the core melt frequency associated with TB sequences by an estimated 2.6 x 10-5 , however, there is no effect on the dominant accident sequence of TW and thus there would be no overall benefit wheri zompared to alternative (vi).
4.1.2.2 Impacts: Cost Estimates 8
-- - - - - _ _ - _ . . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ g
Implementation of the modifications required by the ATWS rule has been accelerated and are scheduled to be complete by the end of 1989. The station blackout rule could be accelerated to reduce the time required until compliance is achieved by possibly one year. This could be achieved by expeditious and timely staff review and approval of licensee submittals.
Thus, this could require expert staff to review the submittals. Acceleration of compliance by one year represents an estimated cost to licensees of $1.4 million.
4.1.2.3 Value-Impact Ratio The overall value-impact ratio of this alternative is about 990 man-rem j averted per million dollars. Because this alternative does not reduce the
! consequences of the dominant accident sequence (TW) nor the probability of the occurrence of the dominant accident sequence, this alternative is not recommended.
4.1.3 Alternative (iii) l 4.1.3.1 Value: Risk Reduction Estimates The alternative of installing a hardened vent capability from the contain-ment wetwell to the plant stack, in addition to accelerating the implemen-tation of the TB and TC rules, would result in a reduction in core melt frequency for TW sequences. With an independent source of power for remote operation of the valves, it would result in a reduction in core melt frequency in the range from 1x10~4 to Ix10-5 per reactor-year by reducing the contribution of the TW sequence to the total core melt frequency to be an insignificant contributor. The corresponding reduction in risk is approximately 1120 to 112 man-rem per reactor year.
9
b <,
4 . . . .
'4.1.3'.2 Impacts:. Cost-Estimates. j The estimated costs for'insta11ation of the hardened vent. system ranges 7 8 from $690,000 to $2,909,000 per plant for an estimated industry costs from $16.6'million to $69.8 million.
1 1
~
4.1.3.3lValue-ImpactRatio The overall value-impact ratio of-this alternative is from about 4060 to
'9630 man-rems averted per million' dollars. While the value-impact ratio indicates that this is'a cost effective alternative, it is not recommended
[
because it does'not provide any defense-in-depth for TB'or.TC. events.
4.1.4 Alternative (iv) 4.1.4.1.Value: Risk Reduction Estimates -
This 'alternat.ive would' provide enhanced operability of the ADS, in addition to alternative (ii). It would reduce the risk to the public by .
an estimated 5.7 man-rem per reactor year, but would not reduce the core melt frequency beyond that provided by-alternative (ii). The availability-of-the ADS would eliminate early containment over-pressure /over-temperature failure due-to direct containment heating by changing'the high pressure stationblackoutsequence(highpressuremeltejection),toa'lowpressure station blackout' sequence.
4.1.4.2 Impacts: Cost Estimates Costs estimated by Science and Engineering Associates and documented in SEA
' Report 87-253-07-A:1, dated November 1988.
8 Cost derived from information provided by Boston Edison Company (DPU 88-28, Request No. AG 13-6) and does not include costs related to Technical Specification changes revising procedures or training manual, training, or NRC costs.
10 l
- 4. i , _ . ..
The installation cost of enhancing the ADS has been estimated to range from $500,0009 to $1,993,000 10 per plant for an estimated industry cost of
$12 to.$47.8 million.
4.1.4.3 Value-Impact Ratio The overall value-impact ratio of this alternative is from about 290 to 72 man-rem averted per million dollars. Because this alternative is not cost effective, does not reduce the probability or consequences of the dominant accident sequences, and does not provide defense-in-depth, this alternative is not recommended.
4.1.5 Alternative (v) 4.1.5.1 .Value: Risk Reduction Estimates This alternative would provide a backup water supply system for the containment sprays and as an alternate low pressure water injection system forthereactorvessel,inadditiontoalternative(ii). It would provide no reduction in the probability of severe accident sequences where the reactor remains at high pressure, such as the short term station blackout scenario. However, it would delay core hoatup for the long term station blackout scenarios, i.e. where the ADS has been operating, until the safety-relief valves (SRVs) are reclosed due to high containment pressure.
The reduction in core melt frequency related to TB due to the backup water supply is estimated to be approximately 9.5x10-7 per reactor-year with a corresponding reduction in risk to the public of approximately 5.4 man-rem per reactor-year. Using the backup water supply to spray inside the containment drywell will not affect core melt frequency but could reduce the risk to the public by approximately 6 man-rem per reactor year.
4.1.5.2 Impacts: Cost Estimates 9
1 bid, #7 10 ibid, #8 11
12 The cost of this alternative is in the range of $810,000 11 to $2,438,000 with the resulting industry costs in the range of $19.4 to $58.5 million.
4.1.5.3 Value-Impact Ratio.
The overall value-impact ratio of this alternative.is from about 190 to 62 man-rem averted per million dollars. Because this alternative is not cost effective, does not reduce the probability or consequences of the ' dominant accident sequences, and.does not provide defense-in-depth, this alternative is not recommended. See Section 4.1.7 for a discussion of improved ADS reliability combined with a backup water supply.
II ibid, #7 12 ibid, #8 12 u- _ _ _ _ _ _ _--_.___.--__.-_m __.__ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _
.f- '. .
r .,
TABLE 1 CostBenefitsofAlternatives(i)-(v)
(Man-rem averted per million dollars ):
Alternative (i) .do nothing 0 Alternative'(ii) - Accelerate Rule 990 Implementation (ARI)
Alternative (iii) - ARI & venting 4060 to 9630*
72 to 290
- sprays
'* Ranges due to effects of TW frequency and two installation cost estimates.
l 13
. _ .-__-__--._-_m.u __...2&__.m_ _ _ ___ .-_____..m___-.m__- _ _ - - _._
4.1.6 Alternative (vi)
While each of the proposed improvements, individually, have some benefit in prevention or mitigation of one or more severe accident scenarios, taken together the improvements have greater benefits because of the effects of the interaction each enhancement has on the others. For example, providing containment sprays will not affect the accident sequence or . timing. Even providing an alternate low pressure reactor vessel injection capability will provide no benefit when the reactor is pressurized. Corbining this enhancement with improved ADS operability provides greater assurance that the reactor can be depressurized and thus permit operation of the alternate low pressure t reactor vessel injection system. If injection is not possible, using the ADS results in cooler corium when it does exit the vessel and thereby improves the effect of the containment sprays. Combining this enhancement with hard pipe venting of the wetwell provides assurance that the ADS valves will operate by reducing the back pressure on the valves which could, otherwise, prevent the valves from opening. Venting has been identified as the means to reduce the probability of a severe accident from the loss of long term decay heat removal.
This sequence, with venting only, has been dependent upon the operators realigning the suction of the RHR pumps to a source of water other than the suppression pool prior to venting containment. With the proposed combination of enhancements, the alternate water supply system provides an independent water system that would be available even if the RHR pumps are lost due to inadequate net positive suction head (NpSH).
4.1.6.1 Value: Risk Reduction Estimates For station blackout the reduction in risk has been determined to be 33 man-rem /RY and the reduction in the core melt frequency is anticipated to be 6.3 x 10-6/RY.
While it is reasonable to consider that these proposed enhancements would have benefits for the TC sequence, no credit is taken for those benefits.
For example, if the TC is not a full power failure to scram (i.e. all control rods remain withdrawn at their full power position) but is instead 14
~ --- - - - - - - - - - - - - _ __ ___ __
-l a partial power failure to scram (such as at Browns Ferry), the diesel driven water pump with ADS and venting may be adequate to prevent degradation indefinitely.
For those plants which have not properly eliminated TW as the dominant severe accident scenario, the benefits associated with reducing the consequences from the TW and TB sequences have been determined to range from of 1,153 man-rem /RY to 144.9 man-rem /RY for risk reduction and from 1 x 10~4/RY to 1.6 x 10-5/RY for core melt frequency reduction. These reductions correspond to an initial TW frequency of 1 x 10~4/RY, and 1 x 10-5/RY, respective'ly, and a TB frequency of 6.6 x 10-6/RY. For a plant similar to Peach Bottom with a core melt frequency of 1.8 x 10-5/RY(which includes a TW frequency of 1 x 10-5/RY), the emergency procedures and operator training is expected to reduce the core melt frequency to .
~
approximately 3.0 x 10-6/RY for a net core melt frequency reduction of approximately 1.6 x 10-5/RY.
4.1.6.2 Impacts: Cost Estimates ~
Installation of a hard pipe vent in a plant similar to Peach Bottom has been estimated to cost $690,000 13 A similar installation at Pilgrim
. has cost $2,909,000 14 . Installation of the backup water supply for containment sprays and low pressure injection into the reactor vessel at a 15 plant similar to Peach Bottom has been estimated to cost $810,'000 . A 16 similar installation at Pilgrim has cost $2,438,000 . Installation of the supplemental power supply and nitrogen gas supply at a plant similar 17 to Peach Bottom has been estimated to cost $500,000 . A similar 13 1 bid, #7 I ibid, #8 15 1 bid, #7 16 1 bid, #8, except that half of the costs are used because the spray nozzles are not to be modified.
I7 1 bid, #7 15
t 18 installation at Pilgrim has cost $1,993,000 . Together, the installation of these proposed modifications at a plant similar to Peach Bottom has been estimated to cost $2,000,000 19 .
AsimilarinstallationatPilcjrim has cost $7,340,000 20 The estimated total cost for industry (for the 24 Mark I plants)'to install the proposed enhancements ranges from $48 22 Actual total costs may be less since some million 21 to $176 million .
Mark I plants may already have some of the proposed features.
The averted cost associated with prevention and mitigation of an accident can be discussed as five separate costs: replacement power, cleanup, onsite health impacts, offsite health impacts, and offsite property damage. To estimate the costs of averting plant damage and cleanup, the reduction in accident frequency was multiplied by the discounted onsite 23 property costs. The following equations from NUREG/CR-3568 were used to make this calculation:
V = NdFU
, gp r
U = C/m [(e -rt(i))fp][1-e(t(f)-t(i))](1-e-rm) 2 where:
V = value of avoided onsite property damage gp N = number of affected facilities = 24 dF = reduction in accident frequency = 1.6 x 10-5/RY 10 ibid, #8, except half of the nitrogen supply system cost is used and the additional AC and DC capacity cost is used instead 7f the third diesel cost.
19 ibid, #7 20 ibid, #8, 16, and 18 21 1 bid, #7 22 ibid, #8 23 NUREG-CR-3568, "A Handbook for Value-Impact Assessment", December 1983, pages 3.29-3.31.
16
, f. ....
U = present value of 'onsite ' property damage 24 C := cleanup and repair costs'='$1.0 billion t(f)= years remaining until end of plant life = 25 t(i)= years before reactor begins operation = .0
~
r = discount rate = 10%
m-.v period of time over which damage costs are paid out
.(recovery period in years) = 10 Using the above values, the present value of avoided onsite property damage is estimated to be $2.23 million.
Replacement power costs can be estimated several different ways. 4 25 has used a cost of'$500,000 per day. 26 NUREG-1109 NUREG/CR-4012 lists the replacement. power costs for each nuclear power reactor by season.
Using this information for only Mark I reactors-averaged over the four years of projected data, escalated by 6% for 1987 dollars, and normalized for the numerically average-size reactor (in megawatts electr.ic), the
~
replacement power cost is $335,000 per day. A draft BNL report dated August 16, 1988, related to a current RES re-evaluation of the cost benefit considerations in backfit analysis,- has indicated that the replacement power cost used in regulatory analysis should be $400,000 per day. Using the $335,000 per day for the 24 Mark I plants is conservative and therefore is used here. This represents a replacement power cost of
$771 million for 10 years. Thus, the present value of avoided onsite property damage and replacement power is estimated to be $3.95 million.
54NUREG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accident", September 1982, page 10.
25NUREG-1109, " Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout", June 1988, page 23.
20 NUREG/CR-4012, " Replacement Energy Costs for Nuclear Electricity-Generating Units in the United States: 1987-1991, Volume 2, January 1987, Table S.1, pages 2 - 5.
17
__.________m__-.._ _ - - _ _ _ _ . _ _ _ - _ _ _ _ _ - _ _ _ _ . - _ _ _ _ _ . _ _ _ _ _ _ _ _
3., ' '
~
The change..in public health risk. associated with the installation of the
-proposed enhancements:is expressed as total man-rem avoided exposure. The following equations from NUREG/CR-356827 were u' sed to make this -
. calculation:
PH'
=
E (Dp x R) _
._r where:
'VPH = value of public health risk avoided for net-benefitmethod($)
N = number-of affected reactors = 24- -
T = average remaining. lifetime of affected fscilities (years) i 25
.D p = .ayoided public dose per reactor-year (man-rem /RY) = 144.9 and 1153 R- = monetary equivalent of unit dose ($/ man-rem).= 1000 - ,
Using the above values, .the present value of. avoided public health-exposure is estimated to be $86.9 million (or $692 million using the greatest anticipated core melt frequency for the T4 sequence).
4.1.6.3.Value-Impact Ratio The overall value-impact ratio, not including onsite accident avoidance costs, is about 1810. man-rems averted per million dollars for those plants with a core melt frequency of 1.6 x 10-5/RY and 3930 man-rems averted per million dollars for those plants with a core melt frequency of 1 x (
10~4/RY. If the savings to industry from accident avoidance (cleanup and
- 27 NUREG/CR-3568, "A Handbook for Value-Impact Assessment", December 1983, pages 3.11-3'.12.
18 m.-m__._ _ __ _1.._._m._. ..__.__..________m.___ -
. ;g -
.. . , 1 T'able 2 COST BENEFITS
- for Alternative (vi)
(man-rem averted'per million dollars)
No Cleanup, Repair,'Repl Pwr, Averted Onsite Cost ** Averted Onsite ' Costs (10% Discount)
- TW ='l x'10-5 Low Ind. Costs 1,810 1,970 High Ind. Cost 490 500 TW = 1x 10-4:,
. Low Ind. Costs 14,400 29,600 E.
- High Ind. Cost. 3,930 4,570
- Cost Benefit = Averted Exposure (Installation Cost - Averted Onsite Costs)
- Rate discounts are not applicable when averted costs are not included-in the cost benefit ratio.
Conclusions:
- 1. Value/ Impact is not significantly.affected by assumed value of real interest rate.
- 2. Value/ Impact is little affected (10%) by inclusion of averted onsite costs.
- 3. Value/ Impact is affected by a factor of about 4 depending on estimated industry installation costs, however the Value/ impact results-support implementation of enhancements.
4 19
____---_.m_._-______.-___2. __:.--_- A- - _ _ _ .-
l repair of onsite damages and replacement power which has an estimated l value of $3.95 million at a discount rate of 10%) were included, the ]
overall value-impact ratio would be about 1970 man-rem averted per million dollars for the low core melt frequency plants. These values, which
- exceed $1000/ man-rem, indicate that the proposed enhancements are cost beneficial for all the BWR Mark I plants. i i
1 .
4.1.7 Sensitivity Studies Table 1 indicates that alternative (iii), which primarily consists of the hardened vent system, is a cost effective alternative. The benefit of this option is largely due to a reduction in the TW core damage frequency. A range of TW probabilities was used in the analysis which is believed to cover the expected range of core damage probabilities due to TW at Mark I plants. .
Further, the cost-benefit ratios are high enough so that other plant-specific factors would not be expected to change the conclusion that the hardened vent system is cost-effective. ,
Alternative (vi) consists of alternative (iii) plus the addition of improved ADS reliability and backup water supply to the containment sprays and for vessel injection. Table 2 indicates that alternative (vi) is cost-effective, although less cost-effective than alternative (iii). This raises a question about the cost effectiveness of the incremental benefit of the ADS and backup water beyond the benefit that would be obtained with the hardened vent system alone. The incremental benefit of the ADS and backup water supply is primarily due to prevention and/or mitigation of SB0 sequences. As discussed previously, the SB0 risk reduction is estimated to be 33 man-rem /RY associated with a reduction in core melt frequency of 6.3x10-6/RY for SB0. This estimate is a single, plant-specific value determined from simplified containment event trees using assumptions from the NUREG-1150 study of Peach Bottom. Unlike the analysis of TW sequences, no sensitivities to plant-specific factors or sensitivities due to uncertainties is included.
20 4
- - - _ - - - _-m_.____. ___ ._____ ___ ________ ____ __ __
I Three major sensitivities of risk due to SB0 were examined to further investigate the incremental benefit of ADS and backup water.
I
- 1. Plant specific probability of core damage due to SB0,.
- 2. Site population, and :
- 3. Assu:nptions used in the analysis.
Estimates of SB0 core damage frequency, assuming a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping period that would result from the SB0 rule, were determined for all Mark I plants. These values ranged from 3.5x10-5/RY to 0.1x10-5/RY with an average value for Mark I !
plants of 1.0x10-5/RY. This compares to the 0.6x10-5/RY used in the previous analysis which is representative of Peach Bottom. Thus, using consideration of i SB0 frequency alone, the risk due to SB0 could be higher by a factor of five at some Mark I plants and may be at least 50% higher on average than that detennined using Peach Bottom assumptions.
The second site-specific factor considered is the effect of population on the risk.- The Peach Bottom site is typical of the most populated site. Less .
populated sites may result is risk reductions of about a factor of five due to population consideration alone. The combined effects of population and SB0 I
frequency would tend to be off setting for Peach Bottom, but depending on the plant-specific values, could result in risk due to SB0 up to five times higher than the Peach Bottom value for some Mark I plants with both a high population and a high probability of SB0 sequences.
The third sensitivity involves those due to assumptions used in the analysis. l The major sensitivity in the SB0 simplified containment event trees examined l was the effect of assumptions about liner melt. The event trees used a conditional probability of liner melt of 0.5 given a station blackout core damage event. The issue of liner melt was examined by experts during the NUREG-1150 program and no consensis was reached - hence the value of 0.5 which is an average of the differing opinions of experts. If one makes a bounding assumption that liner melt would be a certainty (probability = 1.0) given a core 1
21 i 1
melt with no water in the containment to cool the core debris, as would be the case during a 5B0. Under this assumption, the risk due to SB0 would be increased from the base value of 33' man-rem /RY to 56 man-rem /RY at a core damage frequency of 0.6x10-5 for SB0 and 93. man-rem /RY at a core damage frequency of 1.0x10-5 ,
SB0 risks could be as high as 165-280 man-rem /RY for plants with a SB0 frequency of 3.5x10-5/RY' depending an assumptions about liner melt.
The regulatory analysis did not previously examine the benefit of the proposed improvements due to prevention or mitigation of ATWS sequences. This was due to the a very low core damage frequency assumed for ATWS based on the NUREG-1150 Peach Bottom study. For Mark I plants with a higher ATWS core damage frequency, the backup water supply would provide mitigative benefits.
Given an ATWS, the containment is expected to fail and core melt would result from failure of the pumps. Assuming a failed containment, a pool of water overlying the core debris would help to scrub fission products and thus mitigate the consequences of the event. The following discussion provides an estimate of the risk reduction resulting from mitigation of ATWS events by having a backup water supply to the drywell.
The man-rem consequences of an SST1 release occurring at each plant site has oeen evaluated in NUREG/CR-2723. For the BWR Mark I plants, the man-rem consequences, averaged over all sites, was 2.6x10+7 man-rem, given an SST1 release. Based upon the radionuclides release calculations reported for a BWR l
Mark I (NUREG/CR-4624), Vol 1.) an ATWS event is expected to result in release l fractions roughly equal to one-third of an SST1. As noted in NUREG/CR-2723, Table 10, the person-rem from such a release would be about 50% that of an SST1 release. Further, the person-rem given in NUREG/CR-2723 is for an infinite radius, whereas the value-impact requires that this be estimated over 50 miles.
As noted in NUREG-1109, the total person-rem exposure within a 50 mile radius is approximately one-fourth the persen-rem exposure for an infinite radius.
Based on the above, it can be calculated that the consequences of an unmitigated ATWS event for an average BWR Mark I over a 50 mile radius is estimated to be 2.6x10+7 man-rem x 0.5x0.25 = 3.25x10+6 man-rem.
i 22 L__ _ _ _ _ _
i
., .c If the probability of an ATWS event, after the ATWS rule, is assumed to be '
2x10-5 per reactor-year, the risk prior to any improvements, is 3.25x10+6 man-rem x 2x10-5/RY=65 man-rem /RY.
Given an ATWS event, the containment is expected to fail and a core-m61t will result from failure of the pumps. Assuming liner failure, but with the presence of an overlying pool of water, fission products will be scrubbed by the overlying pool with an overall DF=3. Therefore, the consequences of an ATWS with the backup water supply present is 23 man-rem /RY, and the risk reduction benefit of water, as a mitigative effect for ATWS, is 65 - 23 = 42 man-rem /RY.
The cost-benefit of the incremental addition of ADS improvements and a backup water supply as a function of assumptions concerning SB0 sensitivities and ATWS mitigation is shown in Table 3. This table indicates that the incremental additions of ADS improvements and a backup supply of water are cost effective based on prevention of SB0 if the Mark I average SB0 frequency is used and the low cost estimate is assumed. For Mark I plants with high SB0 frequencies, the improvements are cost effective even with the high cost estimate. The additional benefits of possible ATWS mitigation further support the cost-effectiveness of the ADS and backup water supply.
4.1.8 Summary Alternative (vi) is recommended because this alternative is cost-effective and provides a substantial reduction in risk. While the improvements form a package in the sense that the benefit is maximized when the improvements are combined, the individual improvements contained in alternative (vi) have also been shown to be generally cost-effective on an incremental basis. Containment venting reduces core damage frequency due to TW sequences while the combination of venting, ADS and backup water reduces core damage frequency due to SB0 sequences.
Venting would also mitigate the consequences of core damage events by protecting the containment from failure due to long-term overpressure. The backup water supply also provides mitigation by preventing, or at least delaying the contain-ment shell from melting, by attack of the core debris, or failing that, by providing scrubbing of fission products.
23
Table 3 Cost-Benefit For Incremental Addition of ADS and Backup Water Supply (Man-remavertedpermilliondollars)
ASSUMPTION AVERTED COST COST BASE CASE
- 33 630 190 AVERAGE SB0 PROB.** 55 1,050 310 AVERAGE SB0 PROB. 93 1,780 530 HIGH LINER MELT PROB.***
AVER 2GE SB0 PROB. 135 2,580 760 HIGH LINER MELT PROB.
ATWS MITIGATION ****
HIGH SB0 PROB.***** 165 3,150 930 HIGH SB0 PROB. 280 5,344 1,580 HIGH LINER MELT PROB.
HIGH SB0 PROB. 322 6,145 1,816 HIGH LINER MELT PROB.
ATWS MITIGATION HIGH SB0 PROB. 64 1,221 361 HIGH LINER MELT PROB.
ATW MITIGATION LOW POPULATION 24
h . 9.: ~ +*
w -
NOTES FOR TABLE 3
- Based on SB0 frequency.of 6x10-6/RY and a conditional liner melt
, probability of 0.5 given a core melt.
- - Based on a SB0 frequency of 1x10-5/RY which is'an average for Mark I plants.
- *** Conditional liner melt probability of 1.0 given a core melt.
- Assumes ATWS mitigation of 42 man-rem per RY.
4
- ***** Based on a SB0 frequency of 3.5x10-5/RY.
25
________m__._ -________________m.____.__ _m___ -_---
.. , a
- 4.2 Impacts on Other Requirements There are six programs related to severe accidents. These programs are:
Individual Plant Examinations (IPE), Containment Performance Improvements (the topic of this regulatory analysis), Improved Plant Operations, Severe Accident Research Program, External Events, and Accidcat Management. Each of the five programs related to Containment Performance Improvements (CPI) will be discussed briefly below 28 ,
4.2.1 Individual Plant Examinations 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. The examination will pay specific attention to containment performance in striking a balance
. between accident prevention and consequence uitigation. It is anticipated that the IPE program may take from three to five years until the last plant has performed the IPE and incorporated the appropriate plant modifications. Since the staff has already identified cost-effective ,
j improvements that are generic, there is no need to wait for the IPE to be completed. The modifications related to the CPI program are expected to be installed in approximately 30 months.
4.2.2 Improved Plant Operations (IPO)
The IPO includes consideration of the continued improvements in the Systematic Assessment of Licensee Performance (SALP) program; regular reviews by senior NRC staff nanagers to identify and evaluate those plants that may not be meeting NRC and industry standards of operating l
performance; diagnostic team inspections; improved plant Technical l Specifications; improved operating procedures; expansion of the Emergency 28 For additional information, refer to SECY-88-147, " Integration Plan for Closure of Severe Accident Issues", dated May 25, 1988.
26
- .s
- 4 Operating Procedures (E0Ps)'to include guidance on severe accident management strategies; industry's programs to reduce tra sient and other challenges to. engineered safety feature systems; feedback from the IPE ,
program of experience and improvements in operational areas, such as maintenance and training; and continued research to evaluate the sensi- '
tivity of risk to human errors, the contribution of. management to the level of human errors, and the effectiveness of operational reliability methods to help identify potential problems early and prevent their occurrence. The IPO is related to the CPI program's recommendations since we recommend improved procedures snd operator training.
4.2.3 Severe Accident Research Program (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 required to evaluate the need for and feasibility of core debris controls. Research is also needed to 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.
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 separately. The CPI program only addresses internally initiated events and it is not anticipated that future consideration of external events will adversely affect the recommendations of the CPI program.
4.2.5 Accident Management 27
__._.-__._______________._____-___________.______.______.___mm_ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _.m_____.___ -
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 core damage 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 enhancements would provide the accident management program with additional capabi'iities to achieve their goals by providing improved hardware with which to deal with severe accidents.
4.3 Constraints The backfit rule (10 CFR 50.109) as published by the Commission on September 20, 1985 sets forth restrictions on imposing new requirements on currently licensed nuclear power plants and specifies standard procedures that must be applied to backfitting decisions. The backfit rule states:
"The Commission shall require a systematic and documented analysis pursuant to paragraph (c) of this section for backfits which it seeks to impose....(10CFR50.109(a)(2)).
"The Commission shall require the backfitting of a facility only when it determines, based on the analysis described in paragraph (c) of this section, that there is a substantial increase in the overall' protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection.(10CFR50.109(a)(3))".
In order to reach this determination, 10 CFR 50.109(c) sets forth nine specific factors which are to be considered in the analysis for the backfits it seeks to impose. These nine factors are among those discussed in the main body of this report. Appendix A provides a discussion summarizing each of these factors.
The Commission also states in the backfit rule that "any other information relevant and material to the proposed backfit" will be considered.
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.. >s This report provides additional relevant information concerning the proposed containment performance enhancements. This analysis supports a determination that a substantial increase in the protection of the public health and safety will be derived from backfitting the containment performance enhancements, and that the backfit is justified in view of the direct and indirect costs of implementing the enhancements. It is also noted that the Commission directed the NRC staff to provide all potential improvements and related recommendations to the Commission for their consideration regardless of the results of the backfit analysis.
No other constraints have been identified that affect this 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 TB and TW sequences were performed based on using simplified '
containment event trees and the information available from the NUREG-1150 program. These estimates were used to give insights, along with engineering judgement, to develop the recommendations to improve containment performance.
l A review of the available BWR Mark I PRAs provided only limited information.
However, the highest core damage frequency identified was for a plant which had only been reviewed as part of the A-45 study for the TW accident sequence which was 3x10-4 per reactor-year and the IREP Millstone Unit 1 PRA yielded a core l
l melt frequency of 3 x 10'4/ reactor-year. The lowest core damage frequency identified from all dominant accident sequences was identified to be 2.5x10-6 per reactor year with successful venting. For those plants where TW is the dominant contributor to the plant core melt frequency, a range of core melt frequencies from 1x10~4 to 1x10-5 per reactor-year was used in the risk analysis. An assumed core damage frequency, excluding TW cnd with compliance with the ATWS and SB0 rules, was taken to be 8.2x10-6 per reactor-year. Sensi-tivity studies were also performed to investigate the impact of the proposed improvements for Mark I plants with higher SB0 and ATWS frequencies. Implemen-tation of the proposed recommendations will result in TW being a minor contributor 29
and SB0 being a snell contributor to the total core damage frequency.
5.1 Commission's Safety Goal On August 4, 1986, the Commission published in the. Federal Register 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 addresses the CPI program recommendations in light of these goals.
The two qualitative safety goals are:
(1) Individual member of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health.
(2) Societal risks in life and health from nuclear power plant operation should be comparable to or less than the risks of generating electri-city by viable competing technologies and should not be a significant addition to other societal risk.
The following quantitative objectives are used in determining achievement of the above safety goals:
(1) The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities that might result from reactor accidents shouldnotexceedone-tenthofonepercent(0.1%)ofthesumof I prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed.
(2) The risk to the population in the area near a nuclear power plant of cancer fatalities that might result from nuclear power plant 30
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. 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 risk criteria for prompt fatalities and latent cancer fatalities stated above, even considering the large uncertainties involved. Implementation of the CPI recommendations wi11 result in the ' total core damage frequency being reduced by about a factor of five to ten by reducing the two dominant sequence frequencies to below the estimated core melt frequency for the Mark I plant in NUREG-1150.
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 unpredictability 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 a severe core damage accident will not occur at a U.S. nuclear power i plant. 1 With the implementation of the CPI recommendations, it is expected that the i total core melt frequency can be reduced by a fu. tor of between 1.6x10-5 to 9 Jx10-5 per reactor-year. Therefore, implementing the recommendations for CPI 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 CPI recommendations over other alternatives is discussed as part of the value-impact analysis (Section 4.1).
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l This action represents the staff's position based on~a comprehensive analysis of the containment performance improvement issues.
6.0 IMPLEMENTATION 6.1 Schedule for Implementation Within 60 days after issuance of the final rule, licensees will submit 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 final rule.
Other schedules were considered; however, the staff believes the proposed implementation of the CPI recommendations can be performed with minimum interfacing with containment and engineered safety feature systems and thus with the plant online and therefore is achievable without unnecessary financial ,
burden on licensees for plant shutdown. The schedule allows reasonable time for the implementation of necessary hardware items to achieve a reduction in risk of severe accidents. Shorter or less flexible schedules would be unnecessarily burdensome.
6.2 Relationship to Other Existing or Proposed Requirements Several NRC programs are related to the CPI program; these are discussed in Section 4.2. These programs are compatible with the recommendations of the CPI program.
The electrical power requirements of the proposed improvements have been coordinated with the electrical power reliability requirements of the Station Blackout Rule (10 CFR 50.63).
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7.0 REFERENCES
i
-- , NUREG-1150, " Reactor Risk Reference Document", February 1987.
-- , NUREG-1109, " Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44, Station Blackout", June 1988.
-- , NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power-Plants, Technical Findings Related to Unresolved Safety Issue A-44", June 1988.
-- , NUREG/CR-4012, " Replacement Energy Costs for Nuclear Electricity-Ge'nerating Units in the United States: 1987-1991", January 1987.
-- , NUREG/CR-3568, "A Handbook for Value-Impact Assessment", December 1983.
-- ,NUREG/CR-4551, " Evaluation of Severe Accident Risks and the Potential for Risk Reduction: Peach Bottom, Unit 2", Volume 3, Draft, May 1987.
-- , NUREG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accidents", September 1982.
-- , NUREG/CR-4624, " Radionuclides Release Calculations for Selected Severe Accident Scenarios", July 1986.
-- , NUREG/CR-5225, "An Overview of Boiling Water' Reactor Mark I Containment Venting Risk Implications", October 1988.
U.S. Atomic Energy Comission, WASH-1400, " Reactor Safety Study", October 1975 (also re-issued as NUREG-75/014)
Victor Stello Jr, SECY-88-147, " Integration Plan for Closure of Severe Accident Issues", May 25, 1988 l
Idaho National Enginecting Laboratory, " Sensitivity Results for Mark I l
Containment Improvements Program - RDJ-53-88", letter report to NRC, November 18, 1988.
Science and Engineering Associates, Inc. Report 87-253-07-A:1," Cost Analysis j for Potential BWR Mark I Containment Improvements", November 1988.
Letter from Boston Edison Company, DPU 88-28, Request No. AG 13-6.
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I APPENDIX A BACKFIT ANALYSIS Analysis and Determination That the Recommended Safety Enhancements for Containment Performance Improvements 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 i severe accident. Such an accident could pose a challenge to the integrity of containment. Existing regulations do not require explicitly that nuclear power plant coa aine nts be designed to withstand the 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. The BWRs with Mark I containments have been reviewed first because the perceived susceptibility of the Mark I containments to fail based, in part, on the small containment volume of the fiark I containment -
design. Both deterministic and probabilistic analysis were performed to determine the dominant challenges to containment integrity and potential failure modes affecting the likelihood of core rielt, 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 improvements are not needed to provide adequate protection of the public health and safety.
The proposed plant improvements will provide substantial enhancements to Mark I plant safety that are cost-effective.
The estimated benefit from implementing the proposed plant performance improvements is a reduction in the frequency of core melt due to TB and TW events and the associated risk of offsite radioactive releases. The risk reduction for the 24 operating BWR reactors with Mark I containments is estimated to between 86,900 and 798,000 man-rem and supports the Comission conclusion implementation of the proposed safety enhancements for Mark I plants provides a substantial improvement in the level of protection of the public health and safety.
The cost to licensees to implement the proposed safety enhancements would vary depending on the existing capabilities of each plant. The costs would be primarily for licensees (1) to assess the plant's capabilities, (2) modify existing equipment to provide additional operational flexibility, (3) to retrofit plants with additional components or systems, as necessary, to meet the pro and (5)toposed providecapabilities, (4) torelated opertor training revisetothe emergency mitigating operating severe procedures, accidents.
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., .V The estimated total cost for the 24 BWRs with Mark I containments to provide the proposed safety enhancements is between $48 and $176 million. The cost per reactor would be between $1.6 and $7.3 million.
The overall value-impact ratio, not including accident avoidance costs, is estimated to between 1,810 and 3,930 man-rem averted per million dollars based
. on a range of core melt frequencies for TW and a range of installation costs.
If the net cost, which includes the cost savings from accident avoidance (i.e.,
cleanup and repair of onsite damages and replacement p'ower following an accident),wereused,theestimatedoverallvalue-impactratiowouldimprove slightly to between 1,970 and 4,570 man-rems averted per million dollars.
These values support proceeding with the proposed Mark I containment performance improvements.
The preceding quantitative value-impact analysis was one of the factors considered in evaluating the proposed improvements, but other factors also layed a part in the decision-making process. Probabilistic risk assessment p(PRA) studies performed for this issue have shown that station blackout (TB) and loss of long term decay heat removal (TW) events can be significant contributes to core melt frequency, and, with consideration of the conditional containment failure probability, TB and TW events can represent an important contribution to reactor risk. In general, active systems required for reactor and containment heat removal are unavailable during the postulated severe accidents. Therefore, the offsite risk is higher from a severe accident than it is from many other accident scenarios.
Although there are licensing requirements and guidance directed at providing a containment and support systems intended to contain any release of material from the reactor vessel, containment integrity may be significantly challenged under severe accident conditions. The challenge to containment integrity is primarily by over-pressure and over-temperature. Failure of the containment can initiate core degradation (as in the TW sequence) or can be the result of core degradation (as in the TB sequence),
The estimated frequency of core melt from TB and TW events are directly proportional to the frequency of the initiating events. Estimates of TB frequency was based on the information provided in draft HUREG-1150, " Reactor Risk Reference Document", for Peach Bottom Atomic Power Station, Unit 2. This is assumed to be a realistic estimate of the core melt freqJency for Mark I plants after the plant is in compliance with 10 CFR 50.63, the station blackout rule. For the TW sequence, a range of frequencies was used based on the PRAs available for the Mark I plants. This range is assumed to be typical of those plants where TW is the dominant contributor to core melt, as suggested,in WASH-1400.
The factors discussed above support the determination that the additional defense in-depth provided by the ability of a Mark I plant to cope with a severe accident would provide a substantial increase in the overall protection of the public health and safety, and the direct and indirect costs of implementation are justified in view of this increased protection. The staff has considered how this backfit should be prioritized and scheduled in light of j
other related regulatory activities. The proposed rule to implement these
! improvements would require notification of plans and schedules within 60 days 2
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of' the final rule and implementation of the improvement within 30 months of the final rule.
One U.S. facility, Pilgrim, has installed a hardened vent system, but it is not yet operational.
Analysis-of 10 CFR 50.109(c) Factors (1) Statement of the specific objectives that the backfit is designed to achieve.
The objective of the proposed Mark I containment performance improvements is to reduce the risk of severe accidents by reducing the likelihood of core melt and by improving the ability to mitigate the consequences in the event of a severe accident. -Specifically, the proposed improvements is for all BWRs with Mark I containments to enhance'the reliability of_the automatic depressurization system (ADS), to provide an alternate supply of water for injection into the. reactor vessel e d for operation of the containment sprays, and to provide a hardenet, vent capability from the containment wetwell to plant stack and to implement improved procedures -
and training. These modifications are to be operable during a station '
blackout by means of an additional power supply.
(2) General description of 'the activity required by the licensee or applicant in order to complete the backfit In order to comply with the proposed containment improvements, licensees will be required to
- - Evaluate the plant's actual capabilities relative to the proposed safety enhancements. This evaluation will include verifying the ability of. the ADS cables inside containment to remain operable in the anticipated environment during a severe accident
- verifying the availability of a diesel powered water pump and the capacity of the pump to provide low pressure cooling to the core with a depressurized reactor and to the drywell sprays during a severe accident ,
-- verifying the capability of the containment vent system to withstand the anticipated containment pressures without failing any portion of the vent path to the stack l
1- - Verifying the capability of the containment isolation valves to be opened and reclosed under all anticipated containment pressures and vent flow rates during severe accidents
- Determine the necessary plant modifications to comply with the proposed containment improvements, develop a schedule for plant modification, and submit the schedule to the NRC within 60 days from issuance of the final rule.
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- Complete necessary plant modifications within 30 months from issuance of the final rule / order. )
Depending on the plant's existing capability to cope with severe ac.cidents, licensees may or may not need to backfit hardware modifications, (Seeitem8 1 of this analysis for additional discussion.) Licensees will be required to have procedures and training to cope with and recover from severe accidents.
These procedures should conform to Revision 4 of the BWROG Emergency Procedure Guidelines.
(3) Potential change in the risk to the public from the accidental offsite release of radioactive material Implementation of the proposed BWR Mark I containment improvements is expected to result in an estimated total risk reduction to the public ranging from 86,900 to 798,000 man-rems over an assumed 25 years of remaining life for the 24 Mark I plants, based on the particular plant's TW frequency.
(4) Potential impact on radiological exposure of facility employees For the 24 operating BWRs with Mark I containments, the estimated total reduction in occupational exposure resulting from reduced core damage frequencies and associated past accident cleanup and repair activities is 1,500 man-rem. The estimated total occupational exposure for installation of'the proposed improvements is assured to be negligible. No increase in '
occupational exposure is expected from operation and maintenance activities associated with the proposed improvements.
(5) Installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay No Mark I containments are under construction, thus there are no costs associated with construction delays. All plant modifications are expected to be capable of being made either with the plant operating or during normal plant outages, thus there are no costs associated with additional plant downtime.
The cost of the proposed improvements has been estimated to range between
$1.6 and $3.2 million per plant with a best estimate of $2.0 million. For the 24 Mark I plants, this represents an industry cost range of $38.4 to
$76.8 million with a best estimate of $48 million. Pilgrim has instituted a Safety Enhancement Program which incorporated some of the recommended containment improvements. The Pilgrim costs for those elements identified as part of the containment performance program is $7.3 million. This would translate to an industry cost to $176 million.
(6) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements The proposed containment improvements to be able to cope with severe accidents should not add to plant or operational complexity. The 4
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}a containment performance improvement (CPI) program is related to several NRC programs as the following discussion indicates.
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- Station Blackout Rule (10 CFR 50.63) the electrical power' requirements of the proposed improvements have been coordinated with the requirements of 10 CFR 50.63 for electrical power reliability.
- IndividualPlantExamination(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 i that might be missed without a systematic search. 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 five years until the last plant has performed the IPE and incorporated the appropriate plant ;
modifications. Since the staff has already identified cost-effective improvements that are generic to BWRs with Mark I containments, there is ,
no need to wait for the IPE to.be comp %ted. The modifications related to .
the CPI program are expected to be installed approximately 30 months.
- ImprovedPlantOperations(IOP)
The IOP includes. consideration of the continued improvements in the Systematic Assessment of Licensee Performance (SALP) program; regular reviews by senior NRC staff managers to identify and evaluate those plants that may not be meeting NRC industry standards of operating performance; diagnostic team inspections; improved plant Technical Specifications; improved operatin Procedures (E0Ps)g procedures; to include guidance onexpansion of the severe accident Emergency Operating management strategies; 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 training; and continued research to evaluate the sensitivity of risk to human errors, the contribution of management to the level of human errors, and the effectiveness of operational prevent their occurrence. The IPO is related to the CPI program's recommendations since we recommend improved procedures and operating training.
- SevereAccidentResearchProgram(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 progrnm. I Additional research is needed to confirm and quantify the benefits of having water in the containment to either scrub fission products or to prevent or delay shall 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 5
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I , l 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 damage 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 enhancements would provide the accident management program with additional capabilities to achieve their goals by providing improved hardware and incorporating improved training and procedures with which to deal with severe accidents.
(7) The estimated burden on the NRC associated with the backfit and the availability of such resources The estimated total cost for NRC review of industry submittals is $0.4 million based on submittals for 24 reactors and an estimated average of 200 man-hours per reactor.
(8) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the backfit The proposed improvements are only for BWRs with Mark I containments. All reactors and containments in the category are essentially similar and thus the recommendations are applicable to all 24 Mark I plants. However, a survey of Mark I plants has identified that to varying degrees, each facility may already have some of the equipment which may.be used to satisfy the requirements of the CPI program. For example, a dedicated safe shutdown facility may provide the needed power and alternate water supply. Some plants may have a diesel driven water pump for the fire protection system or a diesel driven service water pump. A hardened vent pipe may already exist from wetwell to outside of the reactor building, leaving only a small section of the vent path requiring replacement.
(9) Whether the backfit is interim or final and, if interim, the justification for imposing the backfit on an interim basis The CPI recommendations for Mark I plants is the final resolution for containment performance except as related to the liner meltthrough issue.
Confirmatory research on this issue will continue, and is expected to show that use i,f water in the drywell will have a beneficial effect of preventing or at least delaying liner meltthrough, in addition to mitigating any consequences. The proposed improvements are not an interim measure. .
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D ENCLOSURE 5 I
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., .% i ENCLOSURE 5 l BWR MARK I Plant Survey This appendix summarizes a portion of the preliminary results obtained from the questionnaire survey of utilities with Mark I plants conducted by the Office of Nuclear Reactor Regulation. To date, information has been received from nine Mark I plants. Licensees were asked in the survey to identify plant components that they believed could be. utilized in the event of a severe accident. The objective was not to determine the safety-grade status of available equipment; rather it was to discover what alternatives exist for plant operators in terms of the total plant emergency capability. Information was available for all major areas under discussion in this paper.
ALTERNATE WATER INJECTION CAPABILITY Of the nine plants responding to the survey, six have available diesel driven fire pumps for injection into the reactor vessel. Typically, the piping is in place which will allow injection through the RHR system. Currently, many ;
plants indicate that valve alignments must be made manually to tie in the fire water system to the RHR but, once completed, the system is capable of providing between 1500 and 2500 gpm at 125 psig. The water supply is either the plant ultimate heat sink (lake or river water) or the fire water storage tank.
ALTERNATE POWER SOURCE i Seven of the nine plants responding reported some form of emergency power other than the emergency diesel generators and the electrical grid. Sources of emergency electric power consisted mainly of auxiliary plant diesel generators.
Since the original plant design criteria did not call for access to these sources of power, some amount of time and effort will usually be required to press them into service. For example, most plants reported that it would take between 1 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to make power available from these sources. One plant, however, reported that their turbine generators in place to meet 10 CFR 50 Appendix R (fire protection) requirements could be accessed in 10 minutes upon demand. Long term access to the diesels was not a problem since they are not located in the reactor building and not likely to be subjected to high radiation levels. Consequently, as long as the fuel oil tanks could be replenished, power would be available.
EMERGENCY VENTING The survey indicated that three of the nine Mark I Plants currently have a vent system which could withstand significant pressurization and at the same time appear to have adequate size to accommodate the required containment depressurization rate. For example, the minimum limiting vent system design pressure of the three plants is 32.7 psig. The size of the vent paths range from a low of 8 inches in diameter to a high of 24 inches. The remaining six plants design have capacity, usually about 2 psig, or a very small size, usually around 2 inches diameter. Either of these parameters (small size or pressure capacity) makes these systems, as presently configured, less effective as l
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severe accident mitigators. From this summary it appears that most Mark I Plants will require some modification to their existing vent system in order to accommodate severe accident pressurization rates.
ADS SURVIVABILITY The' ADS system de cabling is qualified to current design basis accident criteria as specified in 10 CRF 50.49. The temperature qualification rating varies from 349 F to 366 F. Pressure ratings for ADS operation inside containment vary from 49.5 psig to 113 psig. The ADS is dc operated with back-up motive force provided by nitrogen filled accumulators. fio BWR Mark I nuclear plant responding to the survey possesses an ADS that has been qualified to operate in the temperature range symptomatic of severe accidents. In addition, none have demonstrated back-up de power for long-term ADS operations in station blackout conditions. It is possible that some of the alternate power sources described previously could be arranged so that emergency de power would be available to the ADS.
In summary it should be noted that results presented here are based on a very limited sample of BWR Mark I Plants. Although the staff believes these generalizations are accurate, a more complete elicitation of equipment availability across all plants and for specific accident sequences is required.
The Individual Plant Evaluations may provide more detailed plant specific information.
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ENCLOSURE 6 BWR/ MARK I CONTAINMENT SHELL (LINER) FAILURE ISSUE A key uncertainty concerning BWR/ MARK I containment performance is the potential for steel drywell shell (liner) failure as a result of contact by molten core debris following Reactor Pressure Vessel (RPV) failure. Early failure of the containment shell could lead to rapid blowdown of the drywell atmosphere into the reactor building and subsequently into the environment without any suppression pool scrubbing benefits.
The likelihood and timing of MARK I shell failure depend strongly upon the outcomes of several phenomenological processes which, for the most part, consist of:
- 1. In-vessel core melt. progression, the mode of reactor vessel failure and subsequent ejection of core debris onto the pedestal floor.
- 2. The spreading of the debris across the drywell floor and its interaction with the concrete floor and any water present.
- 3. The depth of debris in contact with the shell, and the mechanism of heat transfer to the shell and from the shell into the surrounding structures and any water present.
The phenomenological processes associated with core melt progression and the .
mode of reactor pressure vessel failure provide the initial conditions for the ex-vessel processes as summarized in Table 1 and discussed below.
Calculations performed by Oak Ridge National Laboratory (0RNL) using the BWRSAR code consider the control blade material, subassembly wall, and cladding to be melted early and relocated to the lower core grid plate, where they can accumulate and eventually fail it. The model assumes that the material would pour into the lower plenum and quench in the available water.
The core debris is calculated to reheat, failing the RPV penetrations, leading to a relatively slow pour of the molten core debris while it is at or near its liquid temperature. The material pouring onto the drywell floor consists, progressively, of steel, zirconium, fuel-cladding eutectic and finally fuel.
The initial pour rate is calculated at about 40-80 kg/s for several minutes followed by even slower pour rate of 10-20 kg/s as dictated by the decay heat level.
Calculations performed for the industry by Fauske and Associates using the MAAP code are based on the physical picture that blockage formation in the core leads to core debris accumulation causing steam diversion around the blockage, minimizing zirconium oxidation. Following core support structure failure, the core debris rapidly pours into the lower plenum where it is assumed not to quench in the remaining water, subsequently failing the control
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Table 1 - Summary of Uncertain Initial. Conditions and Their Potential Impact I -
. Initial Condition Potential Impact Debris Ejection Rate
- Initial Energy Content Debris Depth at Contact with Shell Debris Composition Metallic - Potential for Large Chemical Energy Release 0xidic - Favors Stable Crust Formation Debris Temperature
- Initial Energy Content Drives Heat Transfer Affects Crust Formation Pressure at RPV Failure Vessel Failure Mode Debris Ejection Rate Size & Location oi' RPV Failure Debris Ejection Rate Debris / Structural Interaction Water in Drywell Debris /Shell Cooling Crust Formation Fission Product Scrubbing
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rod drive penetrations and pouring onto the pedestal floor. The oxidic and metallic constituents are calculated to emerge as a homogeneous mixture at ;
high temperature and high zirconium metal content with an initial pour rate of {
1000 kg/s for about one minute. The subsequent melt pour is again at a low i rate of about 20 kg/s as governed by the decay heat level. Because of the assumption of substantially increased heat transfer through the vessel wall, '
the low power modules at the edge of the core never reach liquefaction temperatures and therefore remain in-vessel.
Recent analyses performed by T. G. Theofanous of the University of California, Santa Barbara (UCSB) indicate that because of large quantity of water in the reactor lower plenum, and the small interstitial spaces due to the presence of the Control Rod Drive (CRD) and Instrument Tube penetrations, the core debris is expected to be quenched, independent of debris slumping and relocation characteristics. Following the debris quenching, these calculations show that the core debris will gradually heat up (in I hour oT D c= 1750 C) resu.lting in an increase in the lower head temeprature N f hour o T 2 550*C) well above its structural failure temperature (steel loses strb8&dral !
strength at about 600-700 C), thus postulating a massive failure of the RPV lower head, while the majority of the debris is still solid. '
Following RPV failure, the melt begins to pour onto the pedestal floor, passing through some of the ex-vessel support structures, where some of it can freeze. The melt could accumulate inside the pedestal floor filling the in-pedestal sump (see Figure 1), before flowing through the personnel doorway into the ex-pedestal region, where it can contact the drywell steel liner wall (containment pressure boundary). Table 2 shows the deaths of the whole-core debris for a typical BWR/ MARK I assuming debris porosities of zero and 40%.
The value of 40% is based on the experimental observation of level swell due to gas entrainment. At 40% porosity and 50% ex-vessel floor area covered, the height of the debris would be close to the spill-over level of the downcomers. Also given is the volume of the whole-core debris (at zero porosity) and the total volume of the drywell sumps.
Provided the melt pool is deep, and the heat flux to the steel shell is sufficiently higher than the heat losses from the liner, the liner can fail due to melt through or lose its strength due to high temperature.
As part of the BWR/ MARK I workshop held in Baltimore, Maryland, on February 24-26, 1988, the core debris spreading processes potentially leading to liner failure were extensively discussed and debated. Some of the workshop attendees held the opinion that liner failure is a very unlikely consequence of core melt-through of the RPV, while others held the contrary opinion. Analyses and some experimental evidence were presented by both sides to support their views.
The potential mitigative impacts of water in the drywell floor, on debris cooling / spreading, fission product scrubbing, and steel liner cooling were also discussed and debated. It appeared that (a) debris cooling and freezing l I
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-(m ). Area Zero Porosity 40% Porosity In-pedestal 29 1. 0' 1.9 In-pedestal + 50% Ex-pedestal 81 0.4 0.7 In-pedestal + All Ex-pedestal 134 0.2 0.4 3
- VolumeofWhole-core (ZeroPogosity)=37M Volume of Drywell Sumps = 6 m I
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(' due to water could not be supported by the insufficient data base which currentlyexists,(b)waterwouldhavenosignificantdetrimentaleffects,(c) water could provide additional cooling for the steel liner, and (d) tests with i prototypic melts indicate that overlying water pools do attenuate fission product aerosol releases resulting from core debris-interactions with concrete, especially if the water is subcooled and of sufficient depth. (The maximum water depth is highly plant dependent and is-limited by the downcomer vent locations relative to the drywell floor.)
Recently debris spreading and shell heat transfer calculations have been performed by Kazimi [1], Moody, et al. [2], and Hodge, et al. [3]; while debris spreading experiments have been performed by Henry, et al. [4] and Greene,etal.[5].
Based on a simplififed model, Kazimi [1] has concluded that, in cases-other than a catastrophic RPV failure, the MARK I liner is not expected to be attacked even by pours of high superheat. This is particularly the case if the melt is mostly oxidic or when water is available in the drywell.
Moody, et al. [2], using a simple model, also demonstrated the impact of heat
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transfer to an overlying water pool, and concluded that MARK I liner is not ;
expected to fail by creep-rupture for several. days following core melt i accidents. Major assumptions of the model include: conductive heat transfer to the liner, enhanced film boiling heat transfer to the overlying pool, and the BWRSAR debris composition and ejection rate.
Hodge,etal.[3],usingtheBWRSARtogetherwiththeCONTAINcode,have ;
performed calculations of containment response to the release of molten core debris and its spread over the drywell floor. Their results indicate that debris spreading across the drywell floor to the drywell shell boundary is i expected to occur between 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after RPV failure. Furthermore, these results indicate that although the predicted debris temperatures are not high enough to cause shell melt-through, failure of the shell pressure boundary through loss of strength is possible.
Henry, et al. [4], using the results of several small scale melt-spreading experiments, concluded that the energy removal rate provided by water is significant and the quenching rate far exceeds that which could be' supplied to the debris surface by internal conduction and early in the transient far exceeds the critical heat flux value. Based on these experiments, they concluded that in the presence of water, the ejected debris will never reach the liner in a molten state, and therefore it cannot provide a challenge to the containment integrity.
In pursuit of the shell failure issue, a series of simulant tests has been conducted [5] at Brookhaven National Laboratory (BNL) to study the gross spreading of liquid metal (lead) under varying initial conditions consisting of melt mass, melt superheat and water depth. The focus on liquid metal pours was based on the BWRSAR calculations. The experimental results showed five
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. l 7
distinct qualitative region.s of geometric behavior. With no water present the molten lead spread rapidly into a very thin layer whose thickness was based on a balance between gravity and surface tension. At the other extreme, pours having low superheat into a relatively deep water pool showed much less spreading. The data suggest that the presence of water may constitute a mitigating circumstance.
Simulant experiments designed to provide understanding of heat transfer from hot, bubble-agitated liquid to a vertical metal wall are also being conducted at BNL concurrently with large scale high temperature tests at Sandia National Laboratoreis(SNL). The WITCH-Liner tests measure heat transfer into steel probes from a pool of molten steel through which gas is being bubbled.
- Heating of the melt pool is sustained inductively in the WITCH / GHOST apparatus that permits controlled bas flow through a sparger plate in the bottom of the refractory crucible. In addition, transient tests at SNL were done by. pouring ,
molten stainless steel into instrumented steel cylinders containing concrete j bottom plugs. The results of these studies show that, following an initial phase of crust formation and depletion, greatly enhanced convective heat transfer to the metal sidewall results from gas bubbling through the melt.
During the BWR/ MARK I workshop discussed earlier, T. G. Theofanous of the University of California, Santa Barbara, proposed an approach to the MARK I shell melt-through problem in an integrated methodology. The approach involves the decomposition of the overall phenomenon into a sequence of events; to quantify each event independently; to express this quantification probabilistically, accounting for phenomenological uncertainties; and to integrate these results.into an overall probability vs. frequency using probabilistic methods. Theofanous has also developed some analytical evidence that the BWR lower head will fail by creep rupture prior to the core debris remelting in the lower plenum. These ideas are being further developed in conjunction with ORNL BWRSAR. If confirmed, the impact on shell melt-through potential could be significant. Another area of importance to this issue involves the flow pattern next to the shell and the associated heat transfer coefficient with the molten pool which could be affected by the 45 inclination of the shell with the drywell floor. Data showing the impact of the 45 inclination vs. the commonly assumed angle of 90' in most analyses was presented at the 16th Water Reactor Safety Meeting (October 24-27,1988) and.will be published in February 1989.[6]
Recent analyses aerformed by a number of experts in support of the NUREG-1150 study indicate tlat the uncertainty in MARK I liner failure is significant and l
1s strongly dependent on the assumed initial conditions which were discussed earlier. In general, for large debris superheats and large metallic content debris in the absence of water on the drywell floor, steel liner failure is )
assessed to be a certainty. However, the overall failure probability i distribution shows the clear division of opinions amongst the experts, with one distinct faction supporting a high probability based on assumption about the heat transfer processes at the liner and the initial conditions and the other faction supporting a low probability of liner failure, based on different assumptions for the uncertain parameters. Nevertheless, on the
,, .g 8
average, water was found to be the most significant parameter affecting shell failure probability.
In light of the existing analyses and experiments it can be tentatively concluded that (1) the. lower vessel water and.small interstitial spaces)promotes plenum configuration (large quenching of the molten quantity core of ,
debris as it slumps into it, (2) mode of reactor pressure vessel failure will i determine the debris state and quantity, (3) low melt superheat and water on the drywell floor will likely reduce the probability of early shell failure, and (4) water will scrub firsion product releases and can therefore attenuate the source term to the environment.
It is evident that in order to achieve a successful closure of this issue, confirmatory analytic and experimental research is needed in several areas related to (a) RPV lower head failure mode, (b) effect of water on melt spreading, and (c) shell thermal attack.
Experimental and analytic studies of potential RPV failure modes are needed to better understand the vessel failure mode under low pressure conditions (200-300 psi); any experiment must be appropriately scaled and correlated to !
actual BWR configuration.
Since water has been identified as the most significant parameter affecting shell failure probability, therefore an expanded quantitative basis must be developed to confirm its effectiveness. A program to accomplish this -
includes:
- 1. Assessment of dynamic fuel-coolant interactions between the ejected core debris and water present inside the pedestal and floor sumps.
- 2. Quantitative measurement of heat transfer mechanisms to water due to molten debris spreading under water.
! 3. Continued development and verification of melt spreading models considering the impact of molten debris interactions with water and concrete for both metallic and oxidic melts.
- 4. Continued verification of fission product scrubbing effectiveness of shallow water pools prototypic of BWR/ MARK I conditions.
l
- 5. Investigation of the impact of concrete barriers (existing or to be constructed in the doorway between the in-pedestal and ex-pedestal regions) on melt spreading.
Also, there is a need to complete the ongoing experiments aimed at understanding of heat transfer mechanisms from molten pools of debris to inclined walls for development of appropriate models and/or correlations for use in shell heat transfer calculations.
Finally, the results of this confirmatory research program will be used to assess the BWR/ MARK I shell failure probability and time of potential failure, based on a comprehensive integrated analysis.
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References l
- 1. M. S. Kazimi, "On the Liner Failure Potential in MARK.- I BWR's," Paper submitted to Nuclear Science Engineering (August 30,1986).
- 2. F. Moody, et al., " MARK - I Drywell Shell Temperature Response in a Postulated Severe Accident," Sixteenth Water Reactor Safety.Information '
Meeting (October 1988).
- 3. S. Hodge, et al., " Primary Containment Response Calculations for Unmitigated Short-term Station Blackout at Peach Bottom," Oak Ridge National Laboratory Report (Nov. 28,1988).
- 4. R. E. Henry, et al., " Experiments Relating to Drywell Shell-Core Debris Interactions," Sixteenth Water Reactor Safety Information Meeting (Oct.
1988).
- 5. G. A. Greene, et al., " Experimental Studies on tielt Spreading, Bubble Heat Transfer and Coolant Layer Boiling," Sixteenth Water Reactor Safety i InformtionMeeting(Oct.1988). !
- 6. T. G. Theofanous, et al., "Towards Estimating the MARK I Liner Failure Probability," Sixteenth Water Reactor Safety Information Meeting (Oct.
1988).
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6 ENCLOSURE 7 O
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. Enclosure 7 DRAFT PROPOSED ORDER ORDER MODIFYING LICENSE I.
(Name of Licensee), (Licensee) is the holder of Operating License No, issued by the Nuclear Regulatory Commission (NRC/ Commission) on .
The license authorizes the licensee to operate (Name of Facility). The facility is a Boiling Water Reactor (BWR) located at the Licensee's site in which utilizes a Mark I containment.
II.
Probabilistic Risk Assessment (PRA) studies have been performed for a number of 1 Boiling Water Reactors with Mark I containments. These studies indicate that, although the risk from the BWR Mark I plant is low, containment integrity could be challenged if a large scale core melt accident were to occur, principally due to the smaller size of the containment. The studies which have been performed l
indicate that BWR Mark I plant risks are dominated by Loss of Decay Heat Removaf (TW), Station Blackout (SBO) and Anticipated Transient Without Scram (ATWS) '
seevences. The staff has concluded that the optimum way to reduce everall risk in BWR Mark I plants is to pursue a balanced approach utilizing accident prevention and mitigation. Based on this assessment, potential improvements have been identified in the following specific areas which, when implemented, will substantially enhance the overall safety of Mark I plants:
(1) Containment Pressure Relief Capability (Venting).
The capability to vent the containment has long been recognized as important in reducing risk from operation of BWR Mark I facilities for Loss of Long Term Decay Heat Removal events. However, the vent path external to most of the existing containment penetrations typically consists of a ductwork system which has a low design pressure of only a few pounds per square inch (psi). Venting under high pressure severe accident conditions would fail the ductwork, release the containment atmosphere into the reactor building, and potentially contaminate or damage equipment needed for accident recovery. Furthermore, with the l
l existing hardware and procedures at some plants, it may not be possible to open or close the vent valves for some severe accident scenarios. A hard pipe vent capable of withstanding the anticipated severe accident pressure loadings would eliminate or minimize the consequences of these disadvantages.
Other changes, in conjunction with proper operating procedures, would l result in greatly reducing the probability of core melt due to the Loss of Decay of Heat Removal (TW) and Station Blackout (SBO) sequences.
..____.-________m__ . _ _ _ _ _ _ _ _ _ _
(2) Reactor Pressure Vessel (RPV) Depressurization System Reliability.
The automatic depressurization system (ADS) consists of relief valves which can be manually operated to depressurize the reactor coolant system.
Actuation of these valves aquires DC power. In an extended station blackout after station batteries have been depleted, the ADS valves would not be available and the reactor would repressurize. With enhanced RPV depressurization system reliability, depressurization of the reactor coolant system would have a greater degree of assurance. Together with a low pressure alternate source of water injection into the reactor vessel, the major benefit of enhanced RPV depressurization reliability would be to provide an additional source of core cooling which could significantly reduce the likelihood of high pressure severe accidents, such as from the short-tenn station blackout.
Another important benefit is in the area of accident mitigation. Reduced reactor pressure would greatly reduce the possibility of core debris being expelled under high pressure, given a core melt and failure of the reactor pressure vessel. Use of the RPV depressurization would also delay containment failure and reduce the quantity and type of fission products ;
ultimately released to the environment. In order to increase reliability of the RPV depressurization system assurance of electrical power beyond the requirements of existing regulations may be necessary. In addition, performance of the depressurization system valves needs to be reviewed for temperature capability during a severe accident. -
(3) Alternate Water Supply for Drywell Spray / Vessel Injection.
An important proposed improvement would be to employ a backup or alternate supply of water and a pumping capability that is independent of normal and emergency AC power. By connecting this source to the low pressure residual heat removal (RHR) systems as well as to the existing drywell sprays, water could be delivered either into the reactor vessel or to the i
drywell, by use of an appropriate valving arrangement.
An alternate source of water injection into the reactor vessel would greatly reduce the likelihood of core melt due to station blackout or loss of long-term decay heat removal, as well as provide significant accident management capability.
Water for the drywell sprays would also provide significant mitigative capability to cool core debris, to cool the containment liner to delay or prevent failure, and to scrub airborne particulate fission products from j the atmosphere.
(4) Emergency Procedures and Training.
A major element of the Mark I containment performance improvement evaluation involves emergency procedures and training. Proper operator i actions can preclude milder events from progressing to core damage or core meltdown accidents, and can greatly mitigate the consequences of severe accidents. Since a variety of unusual conditions can be present, particularly for beyond design bases events, emergency training and 2
, symptom based procedures are essential for guiding the operator to those actions which provide the greatest measure of protection to the public.
NRC has recently reviewed and approved Revision 4 of the BWR Owners Group EPGs (General Electric Topical Report NED0-31331, BWR Owner's Group l
" Emergency Procedure Guidelines, Revision 4," March 1987). Revision 4 to i the BWR Owners Group EPGs is a significant improvement over earlier versions in that they continue to be based on symptoms, they have been simplified, and all open items from previous versions have been closed.
The BWR EPGs extend well beyond the design bases and include many actions appropriate for severe accident management. Since operator actions affect the risk for all severe accident scenarios, implementation of procedures based upon Revision 4 to the EPGs is important for maximizing overall risk reduction.
III.
Improvements in the above mentioned areas can provide a reduction in core melt frequency of about a factor of five to ten. Furthermore, accident mitigation benefits are also considered to be significant. Mitigation of fission product releases would be realized for all accident sequences, including ATWS. Venting would be effective in preventing containment failure arising from slow over-pressurization. Venting via the suppression pool would provide significant scrubbing of non-noble gas. fission products by about a factor of 10 to 100 if no containment shell failure occurs. Water in the drywell may be effective in preventing or at least delaying failure of the shell by molten core debris. Finally, even if shell failure should occur, the presence of a water layer atop the core debris combined with the drywell spray would reduce any source term releases to the environment by a factor judged to range from 2 to 10. In sum, improvements in these areas would result in reduced core melt likelihood, reduced fission product releases due to mitigation, possible reduction or elimination of a significant containment failure mode, and provide a substantial increase in the overall protection of the public health and safety.
IV.
Accordingly, in view of the foregoing, and pursuant to sections 103, 161b.,
1611., 1610. and 182 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR & 2.204 and 10 CFR Part 50, it is hereby ordered that the Licensee:
A. Within 30 months of the date of this Order: ;
- 1. Provide its BWR Mark I containment with an exhaust line from the wetwell vapor space to a suitable release point (e.g., plant stack). The basic design objective shall be to provide sufficient venting capacity to prevent long-term overpressure failure of containment. This "hard vent" system shall meet the following criteria:
a) The vent shall be sized such that under conditions of 1) constant heat input at a rate equal to 1% of rated thermal power, and 2)
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containment pressure equal to the primary containment pressure limit, the exhaust flow through the vent is sufficient to prevent the containment pressure from increasing. .
l b) Capability of RPV depressurization system valves, torus vent valves, or other equipment shall not limit venting capability to less than containment design pressure.
c) The venting capability shall be available during severe accident conditions and for a period up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of a station blackout.
d) The hardened vent shall include a means to prevent inadvertent actuation.
e) The vent path up to and including the second containment isolation barrier shall be designated safety Class 2.
f) The hard vent path shall be capable of withstanding, without loss of functional capability, expected venting conditions and the effects of potential combustion phenomena, g) The hardened vent path shall have a radiation monitor, alarmed and indicating in the control room and functional during extended station blackout. .
- 2. Examine the reactor pressure vessel (RPV) depressurization system and make modifications to ensure its functional capability during severe accidents and during extended station blackout conditions. As a minimum, the following shall be provided: 1) the capability of the RPV depressurization system cables and components to withstand, without loss of functional capability, the environment in the containment during a severe accident prior to vessel feilure, and 2) an alternate power supply system capable l of opening and maintaining open a sufficient number of depressurization system valves to depressurize the reactor and to maintain the reactor depressurized for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of station blackout.
Any sources of electrical power required to assure that operability of the backup water supply, containment venting systems, and RPV depressurization system during an extended station blackout should be coordinated with the ,
requirements of 10 CFR 50.63, as follows: !
a) Thoselicenseeswhochoosetoimplementtherequ{rementsof10CFR 50.63 by the use of an alternate AC (AAC) source need not provide any additional power supplies to comply with the provisions of this section, provided that the capacity, capability, and duration of the AAC can be shown to meet the requirements of both 10 CFR 50.63 and this section.
1 As defined in Regulatory Guide 1.155, Section 3.3.5.
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b) Those licensees who choose to implement the requirements of 10 CFR 50.63 solely by means of a coping analysis must provide additional power supplies of sufficient capacity and reliability to assure the
, operability of the backup water supply, containment venting system and RPV depressurization systems during an extended station blackout.
, 3,. Provide at least one water supply system for the containment drywell spray which shall be functional during an extended station blackout. An extended station blackout is defined as loss of all normal and emergency AC power and loss of DC power due to depletion of station batteries. Operability of controls and valves durIng such an event may require an independent source of power such as a dedicated battery set or a means to recharge the station batteries. Water to the spray system from this supply shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during severe accident conditions.
The water supply system shall also be capable of delivering water to the reactor vessel once the vessel has been depressurized. The mass flow rate shall be equal to or greater than the boiling rate which would occur under depressurized, saturated conditions with a constant heat input rate equal to 1% of rated thermal power.
All valve realignments or other actions necessary to realize this capability shall be reasonably achievable during an extended station blackout. Instrumentation needed to realize these capabilities shall be functional in the expected accident conditions and should, as a minimum, include [to be determined].
- 4. ImplementproceduresbasedonEmergencyProcedureGuidelineg(EPGs) developedbytheBWROwners'Ggoup. Revision 4 to the EPGs as modified by the staff safety evaluation shall be used as the basis for the procedures.
- 5. Provide certification to the NRC upon implementation of Items 1 through 4, above.
B. Within 60 days, submit to the Director of Nuclear Reactor Regulation plans for implementation of the above improvements and a schedule for implementation.
The licensee or any person adversely affected by this Order may request a hearing within 30 days of the date of this Order. A request for hearing should ~
be clearly marked as a " Request for Hearing" and shall be addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555, with copies to the Assistant General Counsel for Enforcement at the same address, the Regional 2 BWR Owner's Group Emergency Procedures Guidelines, Revision 4, NED0-31331 March, 1987.
3 1.ctter from A. Thadani to D. Grace " Safety Evaluation at BWR Owner's Group Emergency Procedure Guidelines" dated September 12, 1988.
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Administrator, Region , and the NRC Resident Inspector, at (Plants affected).
l If a person other than the licensee requests a hearing, that person shall set l forth with particularity the manner in which the petitioner's interest is adversely affected by this Order and should address the criteria set forth in 10CFR2.714(d).
If a hearing is requested, the Commission will issue an Order designating the time and place of the hearing. If a hearing is held, the issue to be considered shall be whether this Order should be sustained. Upon the failure to answer or request a hearing within the specific time, this Order shall be final without further proceedings.
FOR THE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, Director Office of Nuclear Reactor Regulation Dated this day of
,1989 0
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k ENCLOSURE 8 i
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3 Enclosure 8 j DRAFT PROPOSED RULE SECTION 50.XX - SEVERE ACCIDENT REQUIREMENTS FOR BOILING-WATER REACTORS HAVING MARK I CONTAINMENTS (a) Applicability The requirements of this section apply to all boiling water reactors (BWR) having Mark I containments. '
(b) Requirements (1) Backup Water Suppy for Drywell Spray / Core Injection l
All licensees having BWRs with Mark I containments shall provide at least one water supply system for the containment drywell spray which shall be functional during an extended station blackout.1 Water to the spray system from this supply shall be available by remote manual operation or by simple procedures for connection and startup which can be implemented during severe accident conditions.
The water supply system shall also be capable of delivering water to the reactor vessel once the vessel has been depressurized. The mass flow rate shall be equal.to or greater than the boiling rate which would occur under depressurized, saturated conditions with a constant heat input rate equal to 1% of rated thermal power. All valve realignments or other actions necessary to realize this capability shall be achievable during an extended station blackout.
1 .An extended station blackout is defined as loss of all normal and emergency AC power and loss of DC power due to depletion of station batteries. Operability of contrcls and valves during such an event may require an independent source of power such as a dedicated battery set or a means to-recharge the station batteries.
- _ ' _ - __- _._.-e = - . . . -
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Instrumentation needed to realize this capability shall be functional in the expected accident conditions and should, as a minimum, include [to be determined].
(2) Containment Venting For BWR plants with a Mark I containment, an exhaust line which is capable of withstanding expected venting conditions shall be provided from the wetwell vapor space to a suitable release point (e.g., plant stack). The basic design objective shall be to provide sufficient venting capacity to prevent long-term overpressure failure of containment. This "hard vent" system shall meet the following criteria:
(1) The vent shall be sized such that under conditions of 1) constant heat input at a rate equal to 1% of rated thermal power, and 2) containment pressure equal to the primary containment pressure limit, the exhaust flow through the vent is sufficient to prevent the containment pressure from increasing.
(ii) Capability of RPV depressurization system valves, torus vent valves, or other equipment should not limit venting capability to less than containment design pressure.
(iii) The venting capability shall be available during severe accident conditions and for a period up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of a station blackout.
(iv) The hardened vent shall include a means to prevent inadvertent actuation.
(v) The vent path up to and including the second containment isolation barrier shall be designated safety Class 2.
1 (vi) The hard vent path shall accommodate effects of potential combustion phenomena and remain functional.
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(vii) The hardened vent path shall have a radiation monitor, alarmed and indicating in the control room and functional during extended station blackout.
(3) Reactor Pressure Yessel Depressurization Capability -
All licensees having BWRs with Mark I containments shall examine the reactor pressure vessel (RPV) depressurization system and make modifica-tions to ensure its functional capability during severe accidents and
'during extended station blackout conditions. As a minimum, the following shall be provided: 1) the capability of the RPV depressurization system cables and components to withstand, without loss of functional capability, the environment in the containment during a severe accident prior to vessel failure; and 2) an alternate power supply system capable of opening and maintaining open a sufficient number of RPV depressurization system valves to depressurize the reactor and to maintain the reactor depressurized for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the onset of station blackout. Coordination of this requirementwiththeStationblackoutRule(10CFR50.63)isdiscussedin Section(c)(2)below.
(4) Procedures and Training All licensees having BhRs with Mark I containments shall implement procedures based on Emergency Procedures Guidelines (EPGs) developed by 2
l the BWR Owner's Group. Revision 4 to the EPG's as modified by the staff safety evaluation3 shall be used as the basis for the procedures.
(c) Implementation (1) Schedule 2 BWR Owner's Group Emergency Procedure Guidelines, Revision 4, NED0-31331, March, 1987. !
3 Letter from A. Thadani to D. Grace " Safety Evaluation of BWR Owners Group Emergency Procedure Guidelines" dated September 12, 1988.
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, d.
All licensees having BWRs with Mark I containments shall submit their anticipated schedule within 60 days after a final rule is issued which identifies any actions taken and those needed to be taken to comply with
- he requirements of this section. The requirements of this section shall be fully implemented within .30 months after a final rule is issued, and
. each licensee shall certify to the NRC that the requirements have been implemented.
(2) Co-ordination with requirements of the Station Blackout Rule
)
(10 CFR 50.63) 1 Any sources of electrical power required to assure the operability of the backup water supply, containment venting system, and RPV depressurization system during an extended station blackout, as reqefred in part (b) above, should be coordinated with the requirements of 10 CeR 50.63, as follows:
(i) Those licensees who choose to implement the. requirements of 10 CFR 50.63 by the use of an alternate AC (AAC)4 source need not provide any additional power supplies to comply with the provisions of this section, provided that the capacity, capability, and duration of the AAC can be shown to meet the requirements of both 10 CFR 50.63 and this section.
(ii) Those licensees who choose to implement the requirements of 10 CFR 50.63 solely by means of a coping analysis, must provide additional power supplies of sufficient capacity and reliability to assure the operability of the backup water supply, containment venting system and RPV depressurization systems during an extended station blackout.
4 As defined in Rt.gulatory Guide 1.155, Section 3.3.5.
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UNITFD STATES NUCLEAR REGULATORY COMMISSION
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5' ;j WASHIN GTON, D.C. 20555
%, , , , , , $ March 2, 1989 OFFICE OF THE SECRETARY MEMORANDUM FOR: Betsy Shelburne, Chief Public Document Room
\
THRU: Sandy Sho f Correspon ecords Branch FROM: [J'kk ew Bates, Chief O perations Branch
SUBJECT:
RELEASE OF SECY-89-017 TO THE PDR Attached for placement in the PDR is an advance copy of SECY-89-017 " Mark I Containment Performance Improvement Program." The request for release was made by the staff and all Commissioners concur.
Attachment:
As stated cc: DCS - P1-124 I
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