ML20245E088

Notifies of CRGR Meeting 152 on 881214.Agenda Provided

ML20245E088
Person / Time
Issue date:	12/02/1988
From:	Jordan E Committee To Review Generic Requirements
To:	Bernero R, Goldberg J, Paperiello C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), NRC OFFICE OF THE GENERAL COUNSEL (OGC)
Shared Package
ML20245E092	List: ML20245E088 ML20246F317 ML20246F327
References
RTR-REGGD-03.061, RTR-REGGD-03.062, RTR-REGGD-3.061, RTR-REGGD-3.062 NUDOCS 8812120022
Download: ML20245E088 (65)

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o UNITED STATES p, NUCLEAR REGULATORY COMMISSION

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% ,,,,4+ December 2, 1988 MEMORANDUM FOR: Robert M. Bernero, NMSS Jack R. Goldberg, 0GC Carl J. Paperiello, RIII Denwood F. Ross, RES James H. Sniezek, NRR FROM: Edward L. Jordan, Chairman Committee to Review Generic Requirements

SUBJECT:

CRGR MEETING NO. 152 The Committee to Review Generic. Requirements (CRGR) will meet on Wednesday, December 14, 1988 in Room 6507 MNBB from 10 a.m. to 5 p.m. The agenda is as follows:

10-11 a.m. M. Malsch (OGC) will brief the CRGR on staff actions in resolving public comments on the 10 CFR Part 52 rulemaking.

11-12 p.m. E. Rossi (NRR) will present for CRGR review a draft bulletin l related to thermal stresses in pressurizer piping. (A review l package will be forwarded separately.)

12-1 p.m. J. Roe (NRR) will present for CRGR review a draft generic letter related to nonconforming materials. (A review package will be forwarded separately.)

1-2:30 p.m. W. Houston (RES) will present for CRGR review a draft Commission paper which discusses staff recommendations on Mark I containment performance improvements and transmits a draft generic letter. (The draft Commission paper.is enclosed.)

2:30-4 p.m. R. Cunningham (NMSS) and B. Morris (RES) will present for follow-up CRGR review a certificate of compliance which will be incorporated in 10 CFR Part 72 rulemaking and two implementing Regulatory Guides, 3.6.1, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, and 3.6.2, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks. (The regulatory guides are enclosed. The Certificate of Compliance will be forwarded separately.)

4-5 p.m. B. Grimes (NRR) will present for CRGR review a draft Commission paper which transmits a generic letter related to contingency planning to counteract a surface vehicle threat. (A review package will be forwarded separately.)

If a CRGR member cannot attend the meeting, it is his responsibility to assure (!

'L(RC - gy RLE CENER COPYthat oraan alte.f[

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Persons making presentations to the CRGR are responsible for (1) assuring that the information required for CRGR review is provided to the Committee (CRGR Charter..- IV.B), (2) coordinating and presenting views of other offices, (3) as appropriate, assuring that other offices are represented-during the.

presentation, and (4) assuring that agenda modifications are coordinated with the CRGR contact (C. Sakenas, X24148) and others involved with the presentation. Division Directors or higher management should attend meetings addressing agenda items under their purview.

In accordance with the E00's March 29, 1984 memorandum to the Commission concerning " Forwarding of CRGR Documents to the Public Document Room (PDR),"

the review packages for items scheduled at this meeting, which contain predecisional information, will not be released to .the PDR until the NRC has considered (in a public forum) or decided the matter addressed by the information.

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Edward L. Jordan, Chairman Committee to Review Generic Requirements

Enclosures:

As stated cc w/ enclosures:

SECY V. Stello, Jr.

Distribution:

E. Jordan J. Heltemes J.-Conran.

C. Sakenas R. Fraley A. Thadani (w/o enc.) i E. Rossi (w/o enc.)

C. Berlinger (w/o enc.)

P. Kadambi (w/o enc.)

CRGR CF (w/o enc.)

CRGR SF S. Treby (w/o enc.)

J. Blaha (w/o enc.)

M. Taylor, OEDO PDR (NRG/CRGR) (w/o enc.)

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.fk For: The Commissioners From: Victor Stello, Jr.

Executive Director for Operations

Subject:

MARK 1 CONTAINMENT PERFORMANCE IMPROVEMENT PROGRAM

Purpose:

To present staff recommendations on Mark I containment performance improvements and other safety enhancements.

Category: This paper covers a major policy question.

Summary: As noted in'the Integration Plan for Closure of Severe Accident Issues (SECY 88-147) and in interim reports to the Commission (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 issets. Staff efforts 9 have focused initially on BWR plants with a Mark I contain-ment. 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.

Probabilistic Risk Assessment (PRA) studies have been performed for a number of BWRs with Mark I containments.

These studies indicate that SWR Mark I risks are dominated i by loss of Decay Heat Removal, Station Blackout, and Anticipated Transient Without Scram 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.

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.

Contact:

W. Beckrer, RES 492-3975 i

L. Sof fer, RES l 492-3916 l'

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i, 'e Based on.our assessment including the above described balanced approach, the staff recommends five specific improvements for Mark I containment plants: .1) an improved hardened vent capability, 2) improved ADS reliability, 3) an alternate water supply to the reactor vessel and drywell sprays, 4/ extended emergency procedures and training and 5) accelerated implementation of the existing ATWS and 580 i

rules. These improvements, when fully implemented, will l .substantially enhance the safety of Mark I plants, including improvement to containment performance. The staff has L evaluated them and found them to be cost effective. The staff proposes to implement these improvements through a

. generic letter to be issued to all licensees with Mark I containment plants, followed by a rulemaking to require the 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 drywell spray,)3) control, and 5 emergency containment venting, procedures 4) core After and training. debris the initial proposal, the staff held two separate meetings in early 1987 with researchers representing NRC contractors and industry. There was a wide range of views expressed l

regarding accident phenomenology as well as the efficacy of the various improvements. In view of the lack of technical 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 accifient issues.

On December 18, 1987, the staff issued a plan (SECY 87-297) 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, 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 major topic at the workshop was the phenomena associated with containment shell meltthrough as discussed in Enclosure 6. The Integration Plan for Closure of Severe Accident Issues, (SECY 88-147) characterizes the containment performance improvement effort us being one of the main elements of the intecrated approach to closure 1 of severe accident issues. Othei main elements include

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a) Individual Plant Examinations (IPEs), b) improved plant operations, c) the severe au ident 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 vulner-abilities.

A Commission paper (SECY 88-206) dated July 15, 1988 provided a status report on the staff's efforts regarding the Mark I containment. This paper reaffirmed that the risk from BWR Mar.k 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 Commission 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 (SBO), and Anticipated Transient Without Scram (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 due to ATWS and SB0 sequences. These studies also indicate that 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 or containment dryuell sprays, (3) containment pressure relief capability (ventirig), (4) enhanced ADS reliability, (5) core debris controls, and (f;) procedures p

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-and training. Each of these was evaluated to determine c

their potential benefits in terms of reducing the (1) core melt frequency. (2) containment failure probability, and l(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 stt tion lackout, 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) elimination 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.and (2) providing a backup .,upply 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, eli'minating this time of de-inerting would not significantly reduce risk.

During a severe accident, reactor pressure is anticipated 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 i period of time. Since offsite supplies of nitrogen could E readily be obtained during this period, an onsite backup i- supply of nitrogen would not significantly reduce risk.

i Therefore, the staff concludes that additional Mark I i improvements to control hydrogen beyond the existing -

hydrogen control rule and the procedures in Revision 4 of ,

the Emergency Procedure Guidelines would have no significant l benefit and are not warranted.

Alternate Water Supply for Drywell Spray / Vessel Injection l 1

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 l existing drywell sprays, water could be delivered either into the reactor vessel or to the drywell, by use of an a pp r op ri d tt: valving arratigenient.

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-vessel would greatly reduce; the likelihood of core. melt due to station blackout or loss of long-term decay heat

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removal, as well as provide. significant accident management capability.

,' Vater 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 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 waterrsupply. 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 corium. .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 i few psi. Venting under high pressure severe accident

' conditions would fail the ductwork, release the containment-i.

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 properly implemented, can have a significant benefit on plant risk. However, venting via a sheet metal ductwork path,-as currently implemented at some Mark I plants, is .likely to greatly hamper or com-plicate post-accident recovery activities, and.is therefore L

viewed by the staff as yielding reduced improvements in safety. The capability to vent has long been recognized ,

as important in reducing risk from operation of BWR Mark I facilities. The staff agrees with this view as long as the potential downsides of using the existing hardware are corrected.

A hard pipe vent capeble of withstanding the anticipated severe accioent pressure loadings would eliminate these s.

' disadvantages. The vent isolation valves should also be

' remotely operable from the control room and should be

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provided.with a power suppl., independent of normal or emergency AC pbwer. Other changes, such as-raising the vent valve-operability pressure a..d/or raising the RCIC turbine back pressu.e trip setpoint, may also be desirable.

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 sequenc'es.

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 M 4te containment over-pressure failure and to-rede, W is ite consequences for severe. accident scenarios in which the -

containment shell does not fail for other reasons. Failure -

of the shell due to corium 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, such as a rupture disk, should be considered in the vent design.

Er'hanced ADS Reliability:

[ The Automatic Depressurization System (ADS)- consists of relief valves which can be operated to depressurize the reactor coolant system. Actuation of the ADS valves requires DC power. In an atended station blackout after j station batteries have been depleted, the ADS would not be i- available and the reactor would re-pressurize. With enhanced i- ADS reliability, depressurization of the reactor coolant 3- system would have a greater degree of assurance. Together s with a low pressure alternate source o.f water injection into the reactor vessel, the major benefit of enhanced ADS reliability would be to provide an additional source of core cooling which could significantly reduce the likelihood i of high pressure severe accidents, such as from the short-term 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 ADS 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 ADS, a small additional DC power source must be made available. In addition, performance of the ADS cables needs to be reviewed and the cables may need to be replaced or protected.

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[ Core Debris Controls:

Core debris controls, in the form of curbs in the drywell Y' 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-L 'has not been established, and the design and installation custs as well'as the occupational exposure during installation could be significant. The. staff intendsito pursue research programs to evaluate the need .for and .

feasibility of core debris controls. . There is a growing consensus that water in the containmerit (from an alternate l

supply to the drywell sprays) may help mitigate risk.either

~by fission product scrubbing or by preventing or delaying

'shell melt by core debris. Research is continuing in order to confirm and help ouantify these initial conclusions.

A discussion of Mark I shell melt phenomena 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 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.

HRC has recently reviewed and approved Revision 4 of the BWR Owners Grcup EPGs (General Electri.c Topical Report NE00-31331, 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 or symptoms, they have been simplified, and all open itena from previous versions have been closed. The BWR EPGs 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. Thadeni to Groce, "Safet)

Evaluation cf 'BWR Owners' Group - Energency Procedure Guidelines, Revision 4,' NED0-31331, March 1987," dated i

September 12,1900) encouraged licensees to implement

! P.evision 4 of the EPGs and reiterated the need for proper i

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implementation and training of operators. Implementation of the guidelines has been voluntary, but is strongly recommended in the SER.

n 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 plents, risk from 580 and ATWS sequences would be expected to be reduced to a low level. 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.

Benefit of Improvements:

The (1) an improvements improved hardened that theventing staff iscapability, recommending) include:

(2 improved ADS reliability, (3) an alternative water supply to the reactor vessel and drywell sprays, and (4) emergency proce-dures and. training. Accelerated implementation of the exist;ng station blackout and ATWS rules is also planned.

These improvements are unchanged from those indicated in the interim report (SECY 88-206) to the Commission.

A major benefit of these improvements is that they can l

provide a reduction in core melt frequency of about one

order of magnitude. Mark I plants with a relatively low estinated probability of core melt due to the TW sequence, suchasPeachBottom,wouldbeexegetedtohaveatotal core melt frequency of about 2x10 ,per reactor-year prior to tre above improvements and without iS ving credit for venting. Aftertheseenhancements,theegtimatedcoremelt frequency would be. reduced to about 3x10- per reactor-year.

The s taff estimates, based on some other limited PRA studies, n

that some Mark I plants have a probability of core melt that is significantly higher than the Peach Bottom plant.

For 1. hose plants, the total core melt probability, pQor. to the above enhancements, is expected to be about 1x10 per >

reactor-year. With the proposed enhancemp frequency would be reduced to about ~ perlx10 ,nts, the core melt reactor-year.

It should be noted that these estimates apply to internal L

events only.

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1 Of this reduction in core melt frequency a large fraction -)

is attributable tosimproved venting which, by allowing the removal of long-term decay heat from the containment, greatly reduces the likelihood of core melt from the TW sequence. Another reduction in core melt frequency from station blackout is attributable to the enhancements taken. j together. In the event of station blackout, enhanced ADS 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 prevent core damage and not result in releases of fission products of any significance.

Accident mitigation benefits are also considered to be significant. Venting would be effective in preventing-containment failbre 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.

The benefits of the proposed enhancements in terms of their reduction in offsite risk can be calculated in terms of person-rem. Depending upon the probability of core melt r 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 1150 person-rem per reactor-year, for plant 3

probability of core melt due to TW of 1x10-g having a

-per reactor-year and 1x10-4 per reactor-year, respectively. Of this total value, the risk reduction.produc,ed by lowering the likelihood of core melt due to station blackout accounts for a reduction of only about 33 person-rem per reactor-year.

Thus, the bulk of the risk 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 a package in the sense that they complement one another in prevention or mitigation. This results in the maximum risk reduction when all are taken together.

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.y 01 Summary of Costs of Improvements:

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Cost estimates were made of the proposed improvements.

These are given in Enclosure 3 which provides a summary j for all' improvements that includes high and low estimates

! ranging from $3.1 to $1.6 million dollars. For purposes of u 4

l the regulatory analysis included in Enclosure 4, a best estimate cost of $2.0M has been used. Estimates of cost as high as $7.3M were obtained based on actual costs of similar J I

improvements at an existing Mark I plant. Actual costs at many plants may be less since, as shown in Enclosure 5 some plants already have many features of the proposed improvements.

Many of the proposed enhancements would require plant

Conclusions:

backfits. The staff has examined these in light of the L l 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".

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 t

Mark 1.plantspriortoanyoftheenhancemegtsconsidgred

would be expected to range from about 1x10- to 2x10- per reactor year. With the combined enhancements, core melt. l frequency would be reduced by about one order of magnitude.

J 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 i occurrence of an accident, is also expected to reduce the likelihood of containment failure, although this is not as

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readily quantifiable because of the uncertainty in core melt progression and shell meltthrough phenomenology which is discussed in Enclosure 6. In addition, the ability to scrub particulate fission products by use of venting through j 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 an order of magnitude,

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and provide significant additional accident mitigation capability as well, the staff concludes that the proposed i 10

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.l3l enhancements do provide a substantial increase in the L overall protection of the public health and safety. i 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 improvements'at an existing Mark I plant may have. cost about 7.3 million dollars. Based on the survey results.for nine Mark I plants, the staff believes that many plants have some of these improvements already in place. Since the estimated benefits ranged from 2.9 to 23 million dollars per reactor based upon 1000 dollars per person-rem and a-remaining plant life of 20 years, 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. 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 an.d closure of severe accident issues would not be obtained for Mark I plants.

2. Issue a generic letter. This option would be the quickest and require the least resources. However, the only basis to require these plant improvements is the Commission's severe accident and safety goals policy statements.

3. Issue an order. This option could be accomplished quickly and provide a regulatory requirement to implement the improvements. However, this type of regulatory action is not viewed as the appropriate vehicle for generic

! requirements such as the proposed improvements.

I 4. Initiate Rulemaking. This option would require some L-staff resources and cause a delay in iinplementing the p

proposed improvements. However, it would provide a firm

> regulatory basis for requiring the improvements.

! Recommendations: It is recommended that the staff issue a generic letter to all licensees with Mark I containment plants indicating L

that these improvements, as a minimum, are to be implemented in order to resolve severe accident issues for flark I plants.

Further, the Generic Letter would state that licensees should promptly start the improvements and not wait until completion of the.IPE. A draf t generic letter is attached as Enclosure 7.

Thc staff plans to follow-up the generic letter with a i rulemaking to require the improvements. As prt of the

rulemaking, the staff will prepare an Environmental Assessment of venting of the containment using the improved

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j.7 This option would represent the quickest method of-starting the implementation of the-improvements, but would also-ultimately result in. a -firm- regulatory requirement._ These; improvements do not represent major new requirements, but rather are-improvements in plant features which currently exist to various degrees at Mark I plants or have been proposed by licensees. Further,'the improvements _are' cost effective and meet the backfit tests of 10 CFR 50.109.

Coordination: OGC has no legal objections. The ACRS has reviewed these recommendations ~and will provide their comments separately.

Victor Stello, Jr.

Executive Director for Operations

Enclosures:

1. Response to Commission Questions

2. Mark I' Challenges and Relative Likelihood of Failure Modes 3.' Summary of Costs 4.-Regulatory' Analysis

5. Results of Survey of Mark I Plants

6. Mark I Liner Melt Status

7. Draft Generic letter-l I-l~ I?

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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.

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 to the staff. Figure 1 shows the most recent published results on these plants for internally initiated events.

The logarithmic average of all BWR analyses is slightly lower than that of all PWR analyses. However, as can readily be seen from Figure '1, the individual analyses vary widely (on a logarithmic scale) from the average, and the distributions of the individual core damage frequencies are quite comparable.

The differences in averages is not significant, and the concept of " average" is not very meaningful - an " average" can always be calculated, but should not be understood in the sense of meaning " typical."

It should be noted, however, that DWRs with Mark I containments generally have a higher estimated core damage frequency than BWRs with Mark 11 and III con-tainments. (This is not unexpected, since the Mark II and III containments are associated with newer reactor designs.) Moreover, although the data is rather sparse, the five BWR core damage frequencies approximately span the range of the PWR core damage frequencies. Therefore, one would expect the likelihood of a core damage event in a BWR Mark I to be not greatly different than that of most PWR designs.

The second factor is the likelihood of containment failure, given that a core-damage event has already taken place. Figure 2 shows the results of the estimates done as part of the draft NUREG-1150 project (the most recent calculations currently available).

The question of containment failure is, of course, a subject of very active research, the purpose of which is to reduce the uncertainty spread which is so evident in the large ranges shown in this figure. In addition, two phenomena l

dre treated explicitly: Direct containment heating in the large dry PWR

' containment and liner melt-through in the Mark I BWR containment.

1 p 1 1

The phenomenon of direct containment heating in PWRs was identified as the result of experiments in which simulated molten fuel debris was dischargcd 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 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.

The phenomenon of shell melt-through in the BWR Mark I design corresponds to the direct atteck and melt-through of the drywell wall by molten core debris after vessel breach. If the drywell floor is net 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 and the walls were being sprayed with water, it is less likely that this mode of failure would occur. It is possible that direct attack of the wall is not truly a mode of early failure. Very litile molten fuel may exit the vessel at the time of vessel failure. Thus, there could be an extended time period required for the fuel debris to build up within the pedestal region and eventually flow to the wall. This is therefore also a matter of considerable uncertainty.

In summary, the likelihood of a core damage event in a BWR with a Mark I containment is not greatly different than that of a PWR, and the probability of early containment f ailure is likely to be higher for the Mark I design.

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 for this are very much a ma'.ter 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 surf aced at many PWRs, which further reduces the importance of small LOCAs.

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- BWRs can reduce primary system presst:re much more easily than PWRs, which greatly increases the :pectrum of events in which the low

. pressure systems can be used ef f ectively.

- ATWS events provide more of a challenge to BWR containments than to PWR containments.

Thus, although ATWS tends to be of greater significance in a-BWR, much of the l-reason for the dominance of station blackout and ATWS is because of the lower l importance of the other sequences.

It should be r.oted that dominant sequence identification should be used in conjunction with core damage frequency. The edstence of a dominant sequence category does not of itself mean that a safety problem exists. For example, if a plant already had an extremely low estimated core damage frequency, the 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 draft NUREG-1150 study, the two BWRs were indeed dominated by station blackout and'had ATHS as the next most important sequence category - but also had much lower total core damage frequencies than those calculated for the three PWRs. The relative importances of the sequence categories can be seen in Figure 3.

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-j new dominant category. In general, the relative importance (i.e. contribution c

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. The analyses summarized in the figure above are snapshots of relative importance at a particular time. In point of fact, the two BWRs illustrated had already implemented ATWS fixes prior to these analyses, and the results reflect this.

To estimate the improvement in core damage frequency associated with the station blackout or ATWS modifications, an analysis of the plant before and af ter the modification would be necessary.

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-1109estimatedareductionincoredamagefrequencyagsociated with resolution of 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 S issue reduction in core da of from 1 to 2 x 10~ forgage BWRs,frequency and from 4 to associated with regolution of the ATW 8 x 10- for PWRs.

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Topic 3: Safety implications of a fire when diverting the fire water to the decay heat removal system.

Staff response l

Assuming a station blackout event and that the ability to cool the core was l dependent solely upon the diesel fire pump supply, there should be no intention l 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 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.

• 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, st3uld a fire occur during station blackout.

• Even though most nuclear power plants are located in remote areas, virtually all have mutus1 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 002systems are equipped with hose stations for mariual fire fighting i

I 4

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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 plants 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 fion-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 The Detailed outline of the containment emergency venting operator procedures.

decision process and person responsible for the decision to vent should be clearly identified.

Staff response The Emergency Procedare 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 (standby gas treatment, suppression pool sprays and drywell sprays) have been ineffective or could not be operated.

Then ventin 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

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4. Reactor vent operability pressure 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 4 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.

I i' 5

.The staff considers the overall venting guidance to be appropriate given that a venting would be a decision of last resort and that no other options for }'

cooling the core'or maintaining containment integrity would exist under the conditions that. venting would be called for by the EPGs. This is not to imply that the staff views that the overall issue of containment venting has reached '

an optimal point of-resolution. Specific hardware modifications have been-considered that optimize the benefits and reduce the risks associated with.

venting. 1

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.The EPGs address the technical issues related to venting. The. venting guidance is based upon the analysis and judgement of the most knowledge experts in the industry. . However, the EPGs do not address emergency planning or political considerations. Details of the decision process and identification of'the-We person (s) making the decision have been lef t to the individual licensee.

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.

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1 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 liark I nuclear plant. In addition to the costs of physical modifications, other costs considered we. :: engineering and quality assurance, health physics support, anti-contamination clothing, radioactive waste disposal, licensee-NRC interaction, rewriting of emergency procedures, and staff training. On-site personnel radiation exposures encountered as a result of these 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 Eng>

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 i 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 safety-grade equipment.

3. 14ew 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 moreAn complete automation for events such as a break external to the drywell.

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 with thermal tape (temperature ratings up to 800*F) or comp (temperature replacing the cable with high temperature mineral insulated cable 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 i

This objective of this improvement is to take advantage of the scrubbing l

1 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 duct work leading to and from the Standby Gas Treatment System. Electrical 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 1 EMERGENCY POWER SOURCE FOR ADS, SUBTASK BACK-UP WATER AND WETWELL VENT VALVES 293,000 -

Option 1 ADD DIESEL GENLRATOR (low case) - 549,000 Option 2 ADD DC BATTERY SYSTEM (high case)

SUBTASK 2 ADS ENHANCEMENTS 47,000 -

Option 1 WRAP EXISTING CABLING WITH HIGH TEMPERATURE TAPE (low case) 968,000 Option 2 REPLACE ADS CABLING WITH MI CABLE (high case) 55,000 70,000 Options 1 & 2 LOGIC CHANGES 12,500 17,000 Options 1 & 2 ADD GAS BOTTLE SUBTASK 3 BACK-UP WATER SUPPLY SPRAY ENHANCEMENTS 85,000 66,000 ELECTRICAL 603,000 473,000 PIPING MODIFICATIONS SUBTASK 4 WETWELL VENT ENHANCEMENTS 400,000 509,000 PIPE REPLACEMENT 59,000 (

47,000 ELECTRICAL OTHER COSTS 4,300 77,800

1) HEALTH PHYSICS SUPPORT COSTS 13,500 30,000

2) COST OF ANTI-CONTAMINATION CLOTHING NEGLIGIBLE

3) RADI0 ACTIVE WASTE DISPOSAL COSTS

4) LICENSEE COSTS FOR: 35,000 35,000 a Major Redesign Documentation 11,700 11,700 b Rewrite Procedures 127,200 Training 127,200 c 7,700 i Revise Training Manual 7,700 d) 1,592,900 3,149,400 TOTAL ESTIMATED COST PER REACTOR h

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1 ENCLOSURE 4 REGULATORY ANALYSIS 1.0 STATEMENT'0F THE PROBLEM Accidentt 'thich exceed thm e evaluated during the licensing of facilities The'se accidents, (design bhsis accidents) have a low probability of occurrence.

'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 111. The BWRs with the Mark I containment design have the smallest f ree 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.

1

' Draft NUREG-1150 evaluated the dominant accident sequences for five plants, one of'which was a BWR Mark 1.

The dominant accident sequences were identified b asstationblackout(TB),whichincludesthelossofallACandDCpower;-

anticipated transient without scram (TC); and would have included the loss of long term decay heat removal (TW) except that the particular plant being reviewed considered this sequence 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 electri-city. The short term station blackout fails all AC and DC power sources

- 'INUREG-1150, " Reactor Risk Ref erence Document", Draf t, fet,ruary 1987.

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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 />.

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 like14 hood 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 miiigation.

Most recent 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 f 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, i 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, l 2

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such as additional hydrogen controls, but were not considered to significantly reduce either the probability of a severe acc L nt or consequences'given the occurrence of a severe accident. The details of each proposer' enhancement is described in this section and estimates of the enhancements' bec+ its in Section 4.

For all proposed modifications, the new components need not be safety-grade or safety-related, with two exceptions. First, the components interfacing with safety-related systems .(such as a contHnr* isolation valve) must be safety-grade. And second, no failure of a modit ied system or non-safety-related component 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 (S-CET) for station blackout events with 15 to 20 of the 107 top events used in draft NUREG-1150. (It was determined by inspection that the proposed enhancements would not have a significant effect on anticipated transients without scram). 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 j

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 1 NUREG-1150, the consequences were taken directly from draft flVREG-1150 or scaling Draft NUREG/CR-4551, " Evaluation of Severe Accident Risks and the Potential for Risk Reduction: Peach Bottom, Unit 2", Volume 3, Draft, May 1987.

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f actors were applied to the draf t NUREG-1150 results and interpolated to The risks were then calculated by multiplying the i generate the consequences.

plant damage state f requency, the bin probability, and the consequences of that bin together. {

To evaluate the approximate accuracy of the S-CET, the draft NUREG-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) l This alternative would accelerate the implementation of the existing rules l'

(station blackout and anticipated transient without scram), without any other modification::.

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 er.isting wetwell vent outboard containment isolation valve to the base of the plant stack.

l

.. w *

.This pipe would be routed through a new DC operated isolation valve which

!b 'would bypass the existing ductwork and SGTS. The hard pipe to the : tack could i- contain a rupture disk to prevent inadvertent operation and release.of. radio-activity. 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 y

i ejection. In' addition, the corium would exit the reactor sooner, but would be

- cooler than when the reactor is pressurized, thus delaying the potential 1

i containment meltthrough due to corium attack. The cooler vessel also promotes

.[ plateout of non-gaseous fission products within the reactor vessel.

L 3.5 Alternative (v) f 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 5

p.

l j

or a portable generator * .wer 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.

Modifications may be required to some of the RHR valves to permit remote manual The operation without reliance upon AC power and to bypass interlocks.

emergency procedures would need to be modified to provide appropriate instructions for the operator.

Cool water through the sprays will condense steam and thereby raduce the The potential failure of the containment by condensible gas over-pressure.

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.

i 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 vessel . l 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:

3Even 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

A .; re

'(1)' Containment drywell. spray: assurance of a backup water supply'to the residual heat removal (RHR) system and drywell sprays with AC:

. independent pumping capability, (2) Containment venting: a hardened 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 1 x 10~4/ Reactor-Year (RY)4 to 8.2 x 5

10-6/RY . - 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 term decay heat removal (TW). Peach Bottom core melt frequency due to TW is 6

estimated to be 1x10-5/RY or less without assured venting ,

4 Anticipated maximum core melt frequency for TW only (approximately twice the TW frequency for Pilgrim).

S ibid, #1.

6The existing hardware consists of low pressure ductwork from the outboard containment isolation valve, through the standby gas treatmer.t 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.

.t 7

j

' ' 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 I 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.

Theothertwodominantaccidentsequencesarestationblackout(TB)and anticipatedtransientwithoutscram(TC). Proper implementation and compliance with existing the station blackout and ATWS rules is assumed to reduce the probability of the TC and TB sequences to less than 1/RY. Thus, for the x 10-5 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-impact analysis has shown that it would be beneficial to implement the recommendations identified in alternative (vi) which provides for defense-in-depth for accident management, the no-action I alternative is not recommended.

l 4.1.2 Alternative (ii)

Value: Risk Reduction Estimates l

4.1.2.1 l

l For the BWR Mark I plants, the acceleration of compliance by one year represents a risk savings of approximately 1392 man-rem as identified in NUREG-1109. This could reduce the core melt frequency associated with TB l

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 l

benefit when compared to alternative (vi).

l l

8 i.

r 1

t  !

4.1.2.2 Impacts: Cost Estimates The implementation of the modifications required by the ATWS rule is scheduled to be complete by the end of 1989 snd no " acceleration" of implementation of this rule is realistic. The station blackout rule could be accelerated to reduce the time required until compliance is achieved by possibly one year. This could be achieved only by expeditious and timely staff review and approval of licensee submittals. Thus, this could require expert staff to review the submittals. This represents a NRC annual cost of $150,000 for a minimum of three years for a total cost of

$450,000. 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 994 man-rem 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.

l.

4.1.3 Alternative (iii) 4.1.3.1 Value: Risk Reduction Estimates j

The alternative of installing a hardened vent capability from the contain- 1 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 j

remote operation of the valves, it would result in a reduction in core

~4 -5 melt frequency in the range from 1x10 to 1x10 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 1153 to 112 man-rem per reactor year.

i i

9 I

l

~

4.1.3.2 Impacts: Cost Estimates 1The estimated costs for installation of.the hardened vent system ranges 7

from $690,000 to $2,909,0008 pe'r plant for an ' estimated industry costs from $16.6 million to $69.8 million, p

4.1.3.3. Value-Impact Ratio The overall value-impact ratio of this alternative is from about 3238 to 7929 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 1

j 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 station blackout sequence (high pressure melt ejection) to a low pressure station blackout sequence.

4.1.4.2 Impacts: Cost Estimates 7 Costs estimated by Science and Engineering Associates and documented in SEA Report 87-253-07-A:3, dated November 1988.

OCost derived f rom information provided by Boston Edison Company (DPU 88-28, L Request No. AG 13-6) and does not include costs related to Technical i -_________ s@sth Specification changes revising procedures or training manual, training, or NRC

The installation. cost of enhancing the ADS has been estimated to range 9 10 from $500,000 tc $1,993,000 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 228 to 57 man-rem avertet per million dollars. Because this alternative is not cost effective, doeu not reduce the probability or consequences of the dominant-accident sequences, and does not provide defense-in-depth, this alternative is not recommended.

l 4.1.5 Alternativ_e (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 for the reactor vessel, in addition to alternative (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 heatup for the long term station l

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 j

, supply is estimated to be approximately 9.5x10~ 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

k 1bic, #7 10 ibid, #8 11

c l'. 12 The cost of this alternative.is in the range of $810,00011'to $2,438,000 L 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 148 to 49 man-rem averted per million dollars. 'Because this alternative is.not cost i;

' 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.

t.

l' i

n 1

J 1,

i.

I ibio, #7

'I2 ibid, #8 12

TABLE 1 Cost Benefits of Alternatives (1) -- (v)

(Man-rem averted per million dollars )

Alternative (i) - do nothing 0 994 kiternative(ii) - Accelerate Rule Implementation (ARI)

Alternative (iii) - ARI & venting 3238 to 7929*

kiternative(iv)-ARI& ADS 208 to 57

• Alternative (v) - ARI & containment 148 to 49

• sprays

• Ranges due to effects of TW frequency and two installation cost estimates.

13 t'

4 L'

c -- -- - _ ~ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ . . _ _ _ _ _

,-. l

.., '. 1 I

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 For example, providing the interaction each enhancement has on the others.

containment sprays will not affect the accident sequence or timing. Even providing an alternate low pressure reactor vessel in'.:ction capability will provide no benefit when the reactor v pressurized. ,ombining this enhancement with improved ADS operability provides greater assurance that the reactor can be depressurized and thus permit operation of the alternate low pressure 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 l 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 j water system that would be available even if the RHR pumps are lost due to

l. inadequatenetpositivesuctionhead(NPSH).

I, 4.1.6.1 Value: Risk Reduction Estimates y

i 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 f ailure to scram (i.e. all u

control rods remain withdrawn at their full power position) but is instead

'}.

14 t

a partial power failure to scram (such as at Browns Ferry), the diesel driven water pump with ADS and venting muy 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, respectively, and a TB frequency of 6.6 x 10~0 /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.

4.1.1.2 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

.)

I4 Installation of the backup water supply for j has cost $2,909,000 .

containment sprays and low pressure injection into the reactor vessel at a

! 15 g plant similar to Peach Bottom has been estimated to cost $810,000 IO similar installation at Pilgrim has cost $2,438,000 . Installation of the supplemental power supply and nitrogen gas supply at a plant similar IU ibid, #7 I4 ibid, #8 15 ibid, #7 16 ibid, #8, except that half of the costs are used because the spray nozzles are

!= not to be modified.

J 15 i

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_____.__m_ _ , _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

i l

17 to Peach Bottom has been estimated to cost $500,000 . A similar installa-18 Together, ti.a installation of these tion at Pilgrim has cost $1,993,000 .

proposed modifications at a plant similar to Peach Bottom has been estimated to cost $2,000,000 19 A similar installation at Pilgrim has cost

$7,340,000 20 The estimated total cost for industry (for the 24 Mark 1 21

-plants) to install the proposed enhancements ranges from $48 million to $176 million 22 Actual total costs may be less since some 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 were used to property costs. The following equations from itVREG/CR-3568 make this calculation:

V = NdFU gp 2 r U = C/m [(e -rt(i))/r][1-e(t(f)-t(i)))(1.e-m) where:

g V = value of avoided onsite property damage gp I ibid, #7 I0 ibid, f 8, except half of the nitrogen supply system cost is used and the additional AC and DC capacity cost is used instead of the third diesel cost.

I9 1 bid, #7 20 ibid, #8, 16, and 18 21 ibid, #7 22 ibid, #8

! 23NUREG-CR-3568, "A Handbook for Value-Impact Assessment", December 1983, pages

!' 3.29-3.31.

i 16

N = number of affected facilities = 24 dF

= reduction in accident frequency = 1.6 x 10-5jpy U = present value of onsite property damage 24 C = cleanup and repair costs = $1.0 t,1111on t(f)= years remaining until end of plant life = 20 t(i)= years before reactor begins operation = 0 r = discount rate = 10% and 5%

m = 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 $2.1 million (or $3.82 million with a discount rate of 5%).

Replacement power costs can be estimated several different ways.

20 25 has used a cost of $500,000 per day. NUREG/CR-4012 lists NUREG-1109 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 electric), the l 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 rep.esents a replacement power cost of

$771 million for 10 years. Thus, the present value of avoided onsite i

HUkEG/CR-2723, " Estimates of the Financial Consequences of Nuclear Power Reactor Accident", September 1982, page 10.

I 25NUREG-1109, " Regulatory /Backfit Analysis for the Resolution of Unresolved i i Safety Issue A-44, Station Blackout", June 1988, page 23. l 26NUREG/CR-4012, " Replacement Energy Costs for Nuc, lear Electricity-Generating Units in the United States: 1987-1991", Volume 2, January 1987, Table S.1, l

pages 2 - 5.

I 17 i

_ _ - _-. _ - _ = _. _ _

.~-

property damage and replacement power is estimated to be_$3.72.million (or

$6.77millionwithadiscountrateof5%).

The change in public health risk associated with the installation of the The proposed enhancements is expressed as total man-rem avoided exposure.

27 following equations from NUREG/CR-3568 were used to make this calculation:

V pg = NT (Dp x R) where:

V PH = value of public health risk avoided for net-benefitmethod($)

N = number of affected reactors = 24 T = average remaining lifetime of affected facilities (years) = 20 Dp = avoided public dose per reactor-year (man-rem /RY) = 144.9

.and 1153 R. = monetary equivalent of unit dose ($/ man-rem) = 1000 li h

ll Using the above values, the present value of avoided public health I: exposure is estimated to be $69.5 million (or $553. million using the greatest anticipated core melt frequency for the TW sequence).

4.1.1.3 Value-Impact Ratio

[

The overall value-impact ratio, not including onsite accident avoidance costs, is about 1449 man-rems averted per million dollars for those plants with a core melt frequency of 1.6 x 10-5/RY and 3144 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 i

t 27NUREG/CR-3568, "A Handbook for Value-Impact Assessment", December 1983, pages 3.11-3.12.

18

l4

, , .j - ,s -

e oi ,

~iable'2

' ' COST BENEFITS

• for Alternative (vi)

(man-remavertedpermilliondollars)

[ Cleanup, Repa'ir, Rep 1 Pwr

-No Averted Onsite Cost ** Averted Onsite Costs

'TW = 1 x 10-5 Low Ind. Costs '1449 1570 10% Di: count-

'1684 5% Discount-'

.High led. Cost 395 i

404 10% Discount 411 5% Discount- .-

, TW'= 1x 10~4:'

Low Ind. Costs 11,530 22,316 10% Discount 97,095

~5% Discount High:Ind. Cost 3144 3622 10% Discount 4139 5% Discount i

• Cost Benefit = Averted Exposure

)

(InstallationCost-AvertedOnsiteCosts)-

• • Rate discounts are not applicable when averted costs are not included in the cost beaefit ratio.

~

Conclusions:

l '. - 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.

b i

( 19 L __- - _ _ _ _ _ _ _ _ _ _ __ __ _ _ _ _ _ __ J

repair of onsite damages and replacement power which has an estimated value of $3.7 million and $6.8 million at discount rates of 10% and 5%,

respectively) were included, the overall value-impact ratio would be about 1570 and 1684 man-rem averted per million dollars, respectively, for the low core melt frequency plants. These values, which exceed the

$1000/ man-rem guidance provided by the Commission, indicate that the proposed enhancements are cost beneficial for all the BWR Mark I plants.

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 Accident Management. Each of the five programs related to Containment Performance Improvements (CPI) will be discussed briefly below28 ,

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 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, 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) 20For additional information, refer to SECY-88-147, " Integration Plan for Closure of Severe Accident Issues", dated May 25, 1988.

20

n ..

The IPO' includes consideration of the continued. improvements in the-SystematicAssessmentofLicenseePerformance(SALP) program;rcgular reviews by senior NRC staff managers to identify and evaluate those plants-that may .not be meeting NP.C and industry standards of operating performance; diagnostic team inspections; improved plant Technical. Specifications; improved operating procedures; expansion of the Emergency Operating Procedures (EOPs) to include guidance on severe accident management strateghs; 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 reliability methods to help identify potential problems early and prevent their occurrence. The IPO does not have any direct relationship to the CPI program's recommendations, except in the area of improved procedures and operator training.

I 4.2.3 Severe Accident Research Program (SARP)

! -The SARP was begun after the TM1-2 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.

t 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 21

\

i

t future consideration of external events will adversely affect the recommenda-tions of the CPI program.

4.2.5 Accident Management The accident management program is concerned with-addressing certain preparatory and recovery measures that can be taken by the plant operating and technical staff that could prevent or significantly mitigate the

-consequences of a severe accident. This includes measures taken by the plant staff to 1) prevent core damage, 2) terminate the progress of 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 with which to deal with severe ae;idents.

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' l 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. (10 CFR 50.109(a)(3))".

22 w_____-____ _ - _ _ _ _ _

Inordertoreachthis_ determination,10CFR50.109(c)setsforthninespecific factors which are to be considered in the analysis for the backfits it seeks to impose. The Commission also states in the backfit rule that "any other information relevant and material to the proposed backfit" will be considered.

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 fre backfitting the containment performance enhancements, and thet 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 f consequences of TB and TW sequences were performed based on using simplified i containment event trees and the information available from the NUREG-1150 i

program. These estimates were used to give insights, along with engineering judgement, to develop the recommendations to improve containment performance.

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 1x10-4 per reactor-year and the IREP Millstone Unit 1 PRA yielded a core

~4 melt frequency of 3 x 10 / reactor-year. The lowest core damage frequency identified from ell 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 ly10-5 per reactor-year was used in the risk i 23

?

3 .. a analysis. A typical core damage frequency, excluding TW and with' compliance with the TC and TB rules, was taken to be 8.2x10-6 per reactor-year. . Imple-mentation of the proposed recommendations will result in TW being a minor contributor and TE being a small 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 l

statement on ',' Safety Goals for the Operations of Nuclear Power Plants" (51 FR j

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.

l (2) Societal risks in life and health from nuclear power plant operation I should be comparable to or less than the risks of generating electri-i 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 prompt f atality risks resulting from other accidents to which members of the U.S. population are generally exposed.

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. s t (2) The risk to the porulation in the area near a nuclear powcr plant of cancer fatalities that might result from nuclear-power plant operation should not exceed one-tenth of one percent (0.1%) of the sum of cancer fatality risks resulting from all other causes.

Results of analyses published in draf t 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 larco imcertainties involved. Implementation of.the CPI recommendations will result in the total core damage frequency being reduced by at least an estimated one order of magnitude by reducing the two dominant sequence frequencies to below the estimated core melt frequency for the Mark I plant in NUREG-1150. Therefore the CPI recommendations meets both of the Commission's qualitative safety goals.

The Commission also stated the following regulatory objective relating to the frequency of core melt accidents at nuclear power plants.

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Severe core damage accidents can lead to more serious accidents with the 1

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 objcetive 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 plant.

With the implementation of the CPI recommendations, it is expected that the total core melt frequency can be reduced from between 1.6x10-5 to 9.7x10-5 per reactor-year. Therefore, implementing the recommendations for CPI provides increased assurance that a severe core melt accident from accidents will not 4 occur at a U.S. BWR with a Mark I containment.

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,y j Additional rationale for implementing the CPI recommendations over other alternatives is discussed as part of the value-impact ar.alysis (Section 4.1). ,

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 IWithin60 days _afterissuance'ofthegenericletter,licenseeswillsubmitto the NRC a schedule for implementing any necessary equipment and procedural modifications to meet the;performanc_ e 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 generic letter.

Other. schedules were considered; however, the staff believes the proposed implementation of the CPI recommendations can be performed with minien:

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 imple,1entation of necessary hardware items to achieve a reduction in I risk of severe accidents. Shorter or less flexible schedules would be unnecessarily bu-densome; longer schedules would delay necessary plant enhancements, potentially lower the public's perception of plant safety, and reduce the total risk benefit.

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.

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7.0 REFERENCES

-- , 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.

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-- , HUREG/CR-4012 " Replacement Energy Costs for Nuclear Electricity-Generating Units in the United States: 1987-1991", January 1987.

-- , HUREG/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, Draf t, May 1987.

-- , NUREG/CR-2723, " Estimates of the Fint.acial Consequences of Nuclear Power Reactor Accidents", September 1982.

-- , HUREG/CR-5225, "An Overview of Boiling Water Reactor Mark I Containment Venting Risk Implications", October 1988.

U.S. Atomic Energy Commission, 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 Idaho National Engineering Laboratory, " Sensitivity Results for Mark I Containment Improvements Program - RDJ-53-88", letter report to NRC, November 18, 1988.

Science and Engineering Associates, Inc., " Cost Analysis for Potential BWR Mark 1 Containment Improvements", Draf t flay 1988.

f i letter f rom Coston Edison Company, DPU 88-28, Request No. AG 13-6.

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e ENCLOSURE 5 BWR MARK I Plant Survey This appendix summarizes a portion of the prelimintey 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 11 ark 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. Inform. c.on 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 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

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 dc 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 wit.

back-up motive force provided by nitrogen filled accumulators. tio 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 It is de power for long-term possible that some of the ADS operations in station blackout conditions.

alternate power sources described previously could be arranged so that emergency

'dc power would be available to the ADS.

In summary it should ie 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 LINER FAILURE ISSUE The most difficult uncertainty issue concerning BWR/ MARK I containment performance is the potential for steel liner failure as a result of contact by molten core debris at the drywell floor location, following Reactor Pressure Vessel (RPV) failure. Failure of this liner could lead to rapid blowoewn 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 liner f ailure-depend strongly upon the outcomes of several phenomenological processes which, for the most part, result from the in-vessel phase of core melt progression. These processes, which provide the initial conditions for the ex-vessel processes, are summarized in Table I. The factors on which the liner failure issue depends include:

1. The mass, composition and temperature of the core debris within the RPV as it relocates to the lower plenum and is subsequently ejected onto the pedestal floor.

2. The pressure in the reactor vessel at RPV failure.

3. The. size and location of the RPV failure, through which the core debris passes to the drywell, and the rapidity of that passage.

4. .The interactions of the core debris with structures under the RPV, and the degree to which these structures impede and dilute the debris.

5.- The spreading of the debris across the drywell floor and its i

interaction with the concrete floor and any water present.

6. The depth of debris in contact with the liner, and the mechanism of heat transfer to the liner and from the liner into the surrounding

' structures and any water present.

" Calculations performed by Oak Ridge National Laboratory (ORNL) 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 i slow pour of the molten core debris while it is at or near its liquid temper '

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r-I' Table I - Summary of Uncertain Initial Conditions and their Potentia'l Impact Initial Condition Potential Impact Debris Ejection Rate

• Initial Energy Content Debris Depth at Contact with Liner Debris Composition " Metallic - Potential for Large Chemical Energy Release Oxidic - 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

' Debris Ejection Rate Size & Location of RPV Failure Debris / Structural Interaction Water In Drywell Debris / Liner Cooling Crust Formation Fission Product Scrubbing

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ature. 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 100 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 diversi on 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 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 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.

A potential also exists for a more catastrophic rupture of the RPV lower head leading to a massive relocation of substantial quantities of the core debris onto the drywell floor, (provided a crucible similar to what occurred at Three Mile Island can be formed, leading to a massive relocation of the core debris in a non-layered fashion into the lower plenum) challenging the liner integrity. However, the massive control rod drive and support structures are expected to retard a catastrophic failure of the RPV lower head, reducing the likelihood of massive relocation of core debris into the pedestal floor.

Following RPV failure, the melt begins t9 pour onto the pedestal floor, passing through some of the ex-vessel support stractures, 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 (containmentpressureboundary). Table 2 shows the depths 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 liner 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/ItARK 1 workshop held in Baltimore, Maryland on February 24-26, 1988, the core debris spreading processes potentially leading to liner failure were extensively debated. Some of the workshop attendees held the ,

opinion that liner failure is a very unlikely consequence of core melt-through l of the RPV, while others held the contrary opinion. Analyses and limited experimental evidence were presented by both sides to support their views. )

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Area

• Cgvered- 40% Porosity (m ) Zero Porosity In - Pedestal 29 1.0 1.9 In -

• estal + 50% Ex-Pedestal 81 0.4 0.7 In -

134 0.2 0.4 Feuestal + All Ex-Pedestal 3

• Volumeofwhole-core (ZeroPogosity)=37M Volume of Drywell Sumps = 6 m i

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a The mitigative impacts of water (if present) 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 due to water could not be supported by the insufficient data base which currently exists, (b) water would have no significant detrimental effects, (c) water could provide additional cooling for the steel liner, and (d) tests with 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 simplified debris spreading and shell heat transfer calculations have been performed by Kazimi [1] and Moody, et al [2].

Based on a simplified model, Kazimi [1] has concluded that, in cases other than a catastrophic RPV f ailure, 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.

Noody, et al-[2],_ using a simple model als demonstrated the impact of heat tranfer to an overlying water pool, and cor luded that MARK I liner is nnt expected to fail by creep-rupture for severa$ days following core melt 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 compositico and ejection rate.

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

' liner melt'.through problem in an integrated methodology. The approach invnives the decomposition of the overall phenomenon into a sequence of events; to e

quantify each event independently; to express this quantification L

probabilistically, accounting for phenomenological uncertainties; and to l

integrate these results into an overall probability vs. frequency using probabilistic methods. Theofanous has also developed some analytical evidence l 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 liner melt-through potential could be significant. Another area of importance to this issue involves the flow pattern next to the liner and the associated heat transfer coefficient with the molten pool which could be affected by the 45' inclination of the liner with 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.

Recent analyses performed by a number of experts in support of the NUREG-1150 study indicate that the uncertainty in MARK I liner failure is significant and is 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 distribution shows the clear division of opinions amongst the experts, with one ,

distinct faction supporting a high probability based on assumption about the 1 5

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heat transfer processes at the liner and the initial conditions n d the other faction supporting a low probability of liner failure, based on different assumptions for the uncertain parameters. 1 In pursuit of the liner failure issue, a series of simulant tests has been conducted 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 distinct qualitative regions 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 Laboratories (SNL)y with large The WITCH-Liner scale.high tests temperature measure heat tests at Sand 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 gas 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 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.

There is litt el prospect that all of the complex uncertainties associated with BWR/ MARK I liner issue can be decided by experimental or theoretical investi-l gations in a reasonable time frame. However, there appears to be increasing analytic and experimental evidence that drywell flooding can potentially eliminate, delay, or at least mitigate the challenges associated with liner J

l failure.

References

1. M. S. Kazimi, "on the Liner Failure Potential in MARK - I BWR's", Paper Submitted to Nuclear Science Engineering (August 30,1988)

2. F. Moody, et al, " MARK - I Drywell Shell Temperature Response In A Postulated Severe Accident", Sixteenth Water Reactor Safety Information  ;

l Meeting (October 1988).

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ENCLOSURE 7 DRAFT GENERIC LETTER TO ALL MARK.1 CONTAINMENT B01 LING WATER REACTOR (BWR) LICENSEES Gentlemen:

SUBJECT:

PROPOSED SEVERE ACCIDENT REQUIREMENTS FOR PLANTS WITH MARK I CONTAINMENTS (GENERIC LETTER 88 - xx)

A fundamental objective of the Commission's Severe Accident Policy is to take i all reasonable steps tu reduce the chances of occurrence of a severe accident and to mitigate the consequences of such an accident should one occur. The Reactor Safety Study (WASH-1400) found that, for the Peach Bottom BWR Nark I.

nuclear-plant, even though the core melt probability was low, the containment could be severely challenged if. a large core melt occurred. This conclusion-was reinforced by similar findings in the draft Reactor Risk Reference Document (NUREG-1150).-

The NRC recently issued Generic Letter No. 88-xx requesting that all licensees perform Individual Plants Examinations (IPEs) to search for the risk outliers and to address system reliability and containment performance 01 a plant specific basis. The staff has concluded, hcw ier, that for BWF plants with.

Mark I containments, a set of generic requirements has beem identified that e1Pninates the need to await plant specific analyses and wili lead to speedier implementation than would be poss Therefore, these improve-ments should be promptly pursued {ble via the IPEs.

This Generic Letter deals with BWR plants with Mark I containments only.

Subsequent Generic Letters will address other types of plants. The following required actions are identified for BWR plants with Mark I containments.

L 1. Drywell Spray / Core Injection All BWRs with a-Mark I containment should provide at least one backup water supply system for the containment drywell spray which should be functional during an extended station blackout.2 Water to the spray system from this backup supply should be available by remote manual operation or by simple procedures for connection and startup which can be implemented during a severe accident scenario.

1 Additional actions to improve Mark I containment performance may be identified as part of the IPE program, and are not precluded by these requirements.

2 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 nay require an independent source of power such as a dedicated b6ttery set or a means to recharge the station batteries.

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This backup water supply system should also be capable of being diverted to the reactor vessel to provide an alternate source of water to cool the core once the vessel has been depressurized. The flow rate should be at least equivalent to provide decay heat removal of 1% of full power and all required valve realignments shall be functional during an extended station blackout.

Necessary instrumentation needed to accomplish both above functions should be operable in the expected accident conditions.

2. Containment Venting BWRMarkIcontainmentsNouldbeprovidedwithanexhaustlinefromthe wetwell vapor space to a suitable release point (e.g., plant stack) which This "hard vent" system is capable of retaining expected vent pressure.

should meet the following criteria:

a) Basic design objective should be to provide sufficient venting capacity to prevent overpressure failure of containment (sizing of vent should consider nominal 1% of full power).

b) Venting setpoint should be as high as possible and should be at or above containment design pressure. Capability of ADS valves and torus vent valves should not limit vent operability to less than containment design pressure. Current E0Ps should be revised to reflect any revised venting pressure, with suitable allowances for the time needed to implement vent actions, c) Venting capability and operator action should be available during station blackout extending 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 onset of blackout.

i d) The hardened vent path should provide a means to prevent premature or inadvertent actuation, such as a rupture disc or relief valve.

e) Vent path up to 2nd containment isolation barrier should be safety Class 2.

f) Hard vent pipe should accommodate effects of potential combustion phenomena and remain operable.

g) The hardened vent path should have a radiation monitor, alarmed in control room and operable during extended station blackout.

l h) To allow extended RCIC operability, evaluate and if possible implement, an increase in the RCIC turbine exhaust trip setpoint. (A 25 psig setpoint was increased to 46 psig at Pilgrim, as part of their improvement program.)

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3. ADS Enhancements-Licensees should examine the. automatic depressurization system (ADS) ano make modifications as required to ensure operability during severe accidents including performance during extended station blackout.

4. Procedures and Training  ;

The. licensee should implement emergency operating procedures and other procedures based on all significant elements appropriate to its plant of Emergency Procedure Guidelines, Revision 4.

5. Implementation of Existing Requirer" ts Licensees snould implement the requirements of the ATWS rule (10CFR 50.62) and the Station Blackout rule (10CFR 50.63) as expeditiously as possible.

The NRC staff will give high priority to completing any reviews required.

for implementation of these rules.

Accordingly,pursuantto50.54(f),theaddressees.arerequestedtofurnish,  ;

within 120 days, a proposed schedule for completing each of the items identified in this letter. The staff expects that-the equipment changes required herein shall be installed within 30 months after the effective date of this letter. The procedures and training required should be implemented on a schedule' reviewed and approved by the NRC.

Modifications that require staff pre-approval will be promptly reviewed. Any additional improvements, which are implemented per the criteria of 10 CFR l

50.59, must have full documentation available for possible staff inspection and j audit.

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