Text

Plant TER on IPE Back-End Submittal
	ML20149L654
Person / Time
Site:	Duane Arnold
Issue date:	08/31/1995
From:	Hanry Wagage; SCIENCE & ENGINEERING ASSOCIATES, INC.
To:	; NRC
Shared Package
ML20149L652	List: ML20149L649; ML20149L654; ML20149L874;
References
	CON-NRC-05-91-068-15, CON-NRC-5-91-68-15 SCIE-NRC-218-93, NUDOCS 9608070150
	Download: ML20149L654 (30)
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i SCIE-NRC-218-93 DUANE ARNOLD ENERGY CENTER TECHNICAL EVALUATION REPORT ON THE INDIVIDUAL PLANT EXAMINATION i BACK-END SUBMITTAL Hanry A. Wagage i

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Prepaied for the U.S. Nuclear Regulatory Commission Under Contract NRC-05-91068-15 August 1995 SCIENTECH, Inc. !

11140 Rockville Pike !

Rockville, Maryland 20852 s (,o'(0 70 / Jds

A TABLE OF CONTENTS l ERER E. EXECUTIVE

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv E.1 Plant Characterization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv t

1 j E.2 1.icensee IPE Pmeess . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv E.3 Back-End Analysis ..................................iv E.4 Containment Performance Improvements (CPI) . . . . . . . . . . . . . . . . . . v i E.5 Vulnerabilities and Plant Improvements . . . . . . . . . . . . . . . . . . . . . . . v E.6 Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Review Process .....................................I

1.2 Plant Characterization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I

{ 2. CONTRACTOR REVIEW FINDINGS .......................... 3 2.1 Licensee IPE Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4

2.1.1 Completeness and Methodology . . . . . . . . . . . . . . . . . . . . . . . 3 2.1.2 Multi-unit effects and As-built, As-operated Status ........... 3 2.1.3 Licensee Participation and Peer Review ................. 4 2.2 Containment Analysis / Characterization . . . . . . . . . . . . . . . . . . . . . . 4 2.2.1 Front-end Back-end Dependencies . . . . . . . . . . . . . . . . . . . . . 4 2.2.2 Sequences with Significant Probabilities . . . . . . . . . . . . . . . . . 5 2.2.3 Failure Modes and Timing . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Containment Isolation Failure ....................... 7 Duane Amold Energy Center ii August 1995 IPE Back-End Review

i a i TABLE OF CONTENTS (cont.) l i

Entt 2.2.5 System / Human Response . . . . . . . . . . . . . . . . . . . . . . . . . . 7 ;

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2.2.6 Radionuclide Release Characterization . . . . . . . . . . . . . . . . . . 8 l

2.3 Accident Progression and Containment Performance Analysis ........ 9 2.3.1 Severe Accident Progression . . . . . . . . . . . . . . ......... 9 l 2.3.2 Dominant Contributors: Consistency with IPE Insights . . . . . . . 10 l

l 2.3.3 Charactah tion of Containment Performance .............I1 1

I 2.3.4 Impact on Equipment Behavior .....................14 1

[ 2.3.5 Uncertainty and Sensitivity Analyses . . . . . . . . . . . . . . . . . . 14 i

2.4 Reducing Probability of Core Damage or Fission Product Release . . . . . 15 j - 2.4.1 Definition of Vulnerability .......................15 4

2.4.2 Plant Improvements ............................15 2.5 Responses to CPI Program Recommendations . . . . . . . . . . . . . . . . . 16 2.6 IPE Insights, Improvements, and Commitments . . . . . . . . . . . . . . . . 17

3. CONTRACTOR OBSERVATIONS AND CONCLUSIONS . . . . . . . . . . . . . 18

4. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 Appendix Duane Arnold Energy Center iii August 1995 j IPE Back-End Review

$ E. EXECUTIVE

SUMMARY

f, SCIENTECH performed a review of the back-end portion of the Iowa Bectric Ilght and Power Company's Individual Plant Examinatinn (IPE) on the Duane Arnold Energy Center (DAEC).

i j E.1 Plant Charaderization J

! DAEC consists of a General Electric Company designed BWR-4 with a Mark I containment.

I It has a MM thermal output rating of 1658 MW and a gross electric power generation

limit of 589 MW. DAEC received its operating license on February 22,1974, and began l commercial operation on February 1,1975. 'Ihe DAEC containment has an ultimate failure l pressure of 140 psig.

l E.2 Licensed) IPE Process

'Ihe DAEC IPE consists of the following four elements to meet Generic letter 88-20: )

1) front-end analysis, 2) back-end analysis, 3) consideration of safety features and plant

, improvements, and 4) IPE utility team and internal review.

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j For the back-end analyses, the IPE team used containment event trees, fault trees, and an approach developed by the Nuclear Management Resource Council. The team evaluated the j systems, phenomena, and operator actions pertinent to containment perfonnance during severe accidents.

l I E.3 Back-End Analysis a

For DAEC, the IPE team calculated a total core damage frequency (CDF) of 1.5 E-5 per l !

year. 'Ibe conditional probabilities of containment failure, given core damage, were

{. 41 percent for early failures, 29 percent for late failures, and 0.6 for bypass. The l

} probability of little or no release at DAEC was calculated to be 29 percent. The frequency ll of early, high releases was 1.3E-6 per year. These results are from Revision 3 of the

Probabilistic Safety Analysis and appear in the utility responses to NRC requests for l

additional infonnation. 'Ibe " contributor to containment failure" results, described below,

are from the original submittal [1] because the values calculated in Revision 3 are not j reported in the utility responses. [2] The utility did not report any major changes to the !'

. back-end analysis in its responses [2] and therefore the results are assumed to be valid.

'Ihe main contributor to containment failure was overpressurization (49 percent of CDF) followed by containment vent (29 percent of CDF). The drywell liner melt-through and energetic failure modes contributed less than 1 percent to the CDF. The IPE team assumed drywell liner failure to be a certainty in the accident sequences where the cavity stayed dry.

Duane Amold Energy Center iv August 1995 IPE Back-End Rwiew

E.4 Generic and Containment Perforinance Improvement Issues DAEC took the following actions in response to the recommendations of the Containment Performance Improvement Program. In addition, the utility plans to install a hardened piped vent.

Several alternate injection sources that can provide extemal water sources to the RPV or drywell sprays were included in the DAEC Emergency Operating Procedures. Altemate water injection was included in a number of accident sequences as a parar*ini method of preventing core d===y, preserving containment integrity, or flooding containment. If alternate water injection could be made perfect (failure probability of 0) for all sequences, the base CDF would be reduced by 1.8E-8 per year and the high/carly release decreased by a factor of 5.

The safety relief valves were found to be reliable for reactor pressure vessel depressurization. Alternate hardware modifications to further improve the reliability were found not to be justifiable.

The BWROG Rev. 4 EPGs were incorporated into EOPs. Iowa Electric participated in the BWROG Emergency Procedures Committee development of Rev. 4 and has been involved since then in examining potential changes.

E.5 Vulnerabilities and Plant Improvements

'Ihe DAEC defined vulnerability using the following criteria:

Are there any new or unusual means by which core damage or containment failure occurs as compared with those identified in other PRAs?

Do the results suggest that the DAEC core damage frequency would not be able to meet the NRC's safety goal for core damage?

Are there any single failures of components that lead directly to a core damage state (not including the common cause failure of multiple components of similar types)?

Rawi on the answers to the above, Iowa Electric did not find any vulnerability at DAEC.

'Ibe back-end insights gained through the conduct of this IPE, as well as suggestions for possible improvements / strategies, were:

Duane Arnold Energy Center v August 1995 IPE Bedt-End Review

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Termination (in accordance with procedures) of RPV injection when the l

containmant pressure exceeds a set limit can lead to core damage and a l subsequent containment challenge. The prudency of E...: : ng water injection to the containment under any circumstance for which core

degradation may be aggravated should be evaluated.

As an ex-vessel recovery action, the use of containment sprays and drywell 4- sprays in lieu of low-pressure coolant injection appears to be most useful in 1- response to degraded core conditions. Prioritization of injection systems could i be included in future accident management development..

Initiation of drywell sprays before RPV breach would preclude debris attack l

and failure of the drywell shell for some of the accident sequences that would

{ allow or call for spray initiation before vessel breach. (Class IIIC, Class V, j and Class ID sequence types would not allow or call for spray initiation before

vessel breach.) Consideration of changes to EOPs allowing the use of drywell

spray initiation as well as removing any ambiguity regarding the diversion of l injection sources away from the RPV when ad~=a'a core cooling is not

assured (i.e., low reactor water level) could be included as part of future j accident management development.

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Drywell sprays offer an additional alternative to controlling the drywell j temperature to avoid premature containment failure. Palavation of the i restrictions on the use of the drywell sprays in the drywell spray initiation

!' curve of the EOPs may be a possible future accident management item to develop.

EPG directions with regard to containment floodmg sequences can result in the

1. highest consequences at the earliest time. Future accident management strategies should provide guidance to the operator on prne~ ting containment

and cooling debris using methods that do not require venting of the RPV and l avoid using the drywell vent unless no other alternative exists.

I E.6 Observations l

! SCIENTECH observed the following in the DAEC IPE back end analysis:

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The IPE team appears to have re dequately a to the

recommendations of the Containment Performance Improvement l Prograin.

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Duane Arnold Energy Center vi August 1995 i

IPE Back-End Review l

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Although not explicitly stated in the submittal, the DAEC IPE team appears to ;

j have documented the radionuclide release categories appropriately for accident !

t sequences exwia: the Generic I. meter 88-20 screening criteria. l i ;

* 'Ibe DAEC appears not to have performed a thorough internal peer review of j the back-end portion of the IPE.

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'Ibe DAEC IPE team appears to have taken significant credit for the back-end i vreor actions and the back-end results were driven by these operator

! actions. For example, the operator action to flood the containment resulted in

a relatively low conditional probability (given core damage) of drywell shell i failure of 1%. Drywell venting which was performed to release combustible gas from the containment had a relatively high conditional probability of 29%.

A part of the sequences involving drywell venting resulted in early containment failure (a value not reported in the submittal).

SCIENTECH noted the following strengths in the DAEC IPE back-end analysis:

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The front-end back-end Weies appear to be treated well.

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The DAEC IPE team incorporated several plant-specific features into the .

] MAAP assessment methodology.

! * 'Ihe DAEC IPE team performed a comprehensive assessment of i

phenomenological uncertainties of severe accident progression.

l * 'Ibe descriptions of structures, systems, and components of the containment I and reactor buildings are comprehensive.

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The Summary ofInsights demonstrates a thorough knowledge of the impact of severe accidents on structures, systems, components, and operations.

I SCIENTECH noted the following weakness in the DAEC IPE back-end analysis:

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It was difficult to audit how the CETs were quantified. Quantification

! specifics were not provided in the original submittal. However, the

! querdification process and values for an important top event were provided in i the DAEC response to the staff's RAI. ' Ibis indicates that the DAEC did l employ an appropriate quantification methodology.

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1 Duane Arnold Energy Center vu August 1995 j IPE Back-End Review 4

j 1. INTRODUCTION j 1.1- Review Process

This technical evaluation repon (TER) documents the results of the SCIENTECH review of

{ the back-end ponion of the Duane Arnold Energy Center (DAEC) Individual Plant j haminatinn (IPE) submittal. [1,2] This TER was prepared to comply with the q requiranents for IPE back-end reviews of the U.S. Nucicar Regulatory Cnmmiesion (NRC)

in its contractor task ortiers, and adopts the NRC review objectives, which include the following:

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To help NRC staff determine if the IPE submittal provides the level of detail requested in the " Submittal Guidance Document," NUREG-1335

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To help NRC staff assess if the IPE submittal meets the intent of i Generic I.etter 88-20 1

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To complete the IPE Evaluation Data Summary Sheet i ~

) In July 1993 SCIENTECH delivered a draft TER for the back-end ponion of the DAEC IPE j submittal to the NRC. Rawd in part on this draft submittal, the NRC staff submitted a 2 Request for Additional Information (RAI) to Iowa Electric Iight and Power Company on

! . January 6,1995. Iowa Electric Light and Power Company responded to the RAI in a

{ document dated June 26,1995. [2] This final TER is based on the original submittal and the i response to the RAI.

l' Section 2 of the TER summarizes SCIENTECH's review and briefly describes the DAEC IPE submittal, as it penains to the work requirements outhned in the contractor task ortler.

Each ponion of Section 2 corresponds to a specific work requirement. Section 2 also outlines the insights gained, plant improvements identified, and utility commitments made as a result of the IPE. Section 3 presents SCIENTECH's overall observations and conclusions.

References are given in Section 4. The appendix contains an IPE evaluation and data summary sheet.

1.2 Plant Characterization DAEC consists of a General Electric Company designed BWR-4 with a Mark I containment.

DAEC has a designed thermal output rating of 1658 MW and a gross electric power generation limit of 589 MW. DAEC received its operating license on February 22,1974, and began commercial operation on February 1,1975.

'Ibe primary containment consists of the traditional invened light bulb steel drywell and steel torus wetwell design with a suppression pool of water typical of the Mark I design. The containment was designed for 56 psig internal pressure, has a total free volume space of 306,400 ft' (144,000 ft' in the drywell including the vent system and 162,400 ft8 above the water pool in the torus), and a minimum of 58,900 ft5 of water in the suppression pool.

Duane Arnold Energy Center 1 August 1995 IPE Back End Review

. . i The secondary containment consists of four subsystems, which are the reactor building, the building isolation control system, the standby gas treatment system, and the offgas stack. l

'Ibe secondary containment system surrounds the primary containment system and is designed ,

to provide secondary containment for the postulated loss of coolant accident. '

1he DAEC containment has an ultimate failure pressure of 140 psig.

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l Duane Arnold Energy Center 2 August 1995 IPE Back-End Review

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! 2. CONTRACTOR REVIEW FINDINGS i

I 2.1 Yh IPE Process i i

- I i LL1 Completeness and MethodoloFV-

) The DAEC IPE back end submittal appears to be annantially complete with respect to the j level of detail r== tad in NUREG-1335, and appears to meet the NRC sequence selection screening criteria described in Generic I. cater 88-20.

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'Ihe IPE methodology used is described clearly and its selection appears justified. 'Ihe ;

! approach followed is consistent with Generic I. meter GL 88-20, A,npaadix 1. l i As noted in Section 2.3.1, page 2-3, for the back-end analyses, the DAEC IPE team used cnntainment event trees (CErs), fault trees, and an approach developed by the Nuclear l

Management Resource Council (NUMARC). 'Ihe team evaluated the systems, phenomena, and operator actions pertinent to containment performance during severe accidents. I i

For the front-end back-end interface and the back-end assessment, the IPE team relied on a l set of general assumptions used in back-end PRA analyses. These are listed in l Section 4.2.3.1 of the submittal, starting on page 4-64. In addition, Section 4.2.3.2 (pages 4-70 through -72) lists known, conservative assumptions used in the back-end analysis; Section 4.2.3.3 (pages 4-72 and -73) lists potential, nonconservative assumptions.

One of the conservative assumptions was to use a containment failure curve, that was more limiting than the one calculated by Chicago Bridge and Iron (CB&I) on a plant-specific basis for DAEC and the one developed for Peach Bottom (capacities: used,140 psig; DAEC plant-specific,163 psig; and Peach Bottom,148 psig). 'Ihis decision could have resulted in slightly shorter times to containment failure than if the CB&I plant-specific DAEC curve had l been used. In addition, there might have been sequences that would have resulted in "no containment failure" if the CB&I curve had been used to assign a containment ultimate !

capacity in this analysis. One of the nonconservative assumptions was to us'e a drywell j equipment mass of 2.7 million pounds, which later appeared to the IPE team as an overecimate, resulting in additional heat sinks and longer times to reach high drywell i temperature (i.e., which affects both containment failure and revaporization source term contributions). To c>rrect this overestimmte the IPE team performed sensitivity studies with a reduced equipment inass (1 million pounds) and factored the results into the radionuclide release results.

LL2 Multi-Unit Effacts mad As hilt. As-Oper= tad Sement.

Being a single unit site, multi-unit considerations do not apply to DAEC.

'Ibe IPE team used walkdowns and an internal peer review to confirm the as-built, as-operated status of the DAEC. Introductory or general walkdnwns were performed for areas outside the containment including the reactor building, the torus room, the turbine building, the pumphouse and intake structure, and the simulator. A human error analysis Duane Arnold Energy Center 3 August 1995 IPE Back-End Review

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walkdown was performed, which covered the areas of the simulator and areas outside the l j oontrol room in which operator actions were required.

a Revision 3 of the PSA appears to reflect the plant. design and operation as of December 1994 j l

{ and the original submittal as of 1992 (RAI response, p. A-6) i LL} ficam- Panicinatian and Peer Review. j

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! Based on information in Section 5 of the submittal, which describes utility participation and )

the internal review team, it appears that the DAEC did not perform an internal peer review I of the back-end portion of the IPE. The in-house peer review team appears to have focused on the fmot-end portion only: "An iP in-house review committee was created to i review the information contained in the initial preparation of the system notebooks."

l (Section 5.2.1, page 5-4) l l Section 5.3.2, pages 5-7 through 5-46, lists the comments made during the final review and i the responses by the DAEC to the internal review team aud two review teams of the i consultant companies who participated in the IPE (ERTN Engineering and Research; Gabor, Kenton and Associates). Of 83 comments listed, five were related to the back-end analysis.

2.2 Containment Analysis / Characterization j L11 Front-end mek-and nanana,ncies, s

As noted in Section 4.5.2, page 4-124:

l 'Ibe DAEC IPE directly links the front-end to back-end portions of severe accident i sequences through directly linked event trees. 'Ibese trees convey the support state i conditions throughout the front-end and back-end trees and include considerations of j preventive or mitigative features, as well as timing considerations.

l Section 4.3.3, pages 4-80 through 4-82 of the submittal, summarizes the specific aspects of 4

the front-end back-end dependancies as follows:

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rsmi,md Failures in Front-End: 'Ihe computer carried the information on failed equipment from the front-end to the back-end. This eliminated the need i for back-end analysis of filed equipment, unless the equipment was going to

be repaired or recovered. The equipment included support systems, accident prevention systems, and mitigation systems.

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Human Errors: A check was performed to ensure that recoveries following all i

! Level 1 gances that result from human error can be justified as consistent

! with operating staff recoveries.

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RPV Status: 'Ibe CET analysis made use of the RPV pressure condition !

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resulting from the front-end analysis. ;

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Containment Status: De CET made use of the containment status resulting

! from the front-end analysis, which included whether the containment failed, i was intact, or was experiencing high-pressure conditions.

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Containment Isolation: . De containment isolation was evaluated on a i sequence-by-sequence basis using support system Wies transferred 4

from front-end analyses.

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Accident Sequence Timing: Differences in meridant sequence timing were l

i transferred with the front-end sequences, a

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nermal-Nvdraulic n .nninierte Ae.eem, nee: De thermal-hydraulic code

! analysis represented variations in timing and assumptions about subtle l sequence variations.

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Dual Usage: De analyses accounted for the dual usage of common water sources and common power sources in the front-end and back-end analyses.

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Mission Tune: He analyses considered the mission time for the entire j sequence, i.e., from initiating event to release point.

Timine of Recoverv: De analyses accounted for equipment and power recovery, ensuring no double counting.

In performing its analysis, the IPE team appears to have treated front-end back-end Meies appropriately.

W huianeae with Sienificant Probabilities.

Section 3.4.1, page 3-439, and Section 6.2, pages 6-2 through 6-8, describe sequences with a significant probability of occurrence. De DAEC IPE team identified five sequences that should be reported to the NRC according to the criteria in Generic I.etter 88-20, Appendix 2.

These five sequences had a combined CDF of 3.5E-6 per reactor year, or 44 percent of the total CDF. Table 4.6-3, page 4-198, gives a summary of the core damage accident sequence subclasses, including designators, definitions, and frequency of occurrence.

De DAEC IPE team appears to have evaluated appropriately those sequences with significant probability of occurrence.

W Failure h4adae and Timine.

Chicago Bridge and Iron evaluated the DAEC containment capacity under severe accident conditions. CB&I investigated the failures of the steel containment structure (drywell, drywell head, and torus), containment hatches, hutch seals, gea.Gons, and isolation valves.

The actual containment failure curve used was more liminng, according to DAEC, than that Duane Arnold Energy Center 5 August 1995 IPE Back-End Review

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calculated by CB&I (page 4-70). Section 4.4.4, pages 4-98 through 4-107, describes the !

following containment performance issues: l

Capacity at low temperature (below 500*F)

Capacity at intennediate temperature (between 500*F and 800'F) !

Capacity at high temperature (above 900*F) !

Capacities for high-suppression pool temperatures, high containment pressures, and for high safety relief valve discharge rates.

Figure 4.4-3, page 4-104, shows the primary containment performance regime on a plot of pressure versus temperature. DAEC has a relatively small power rating relative to the containment size. As noted on page 4-269, the containment free-volume to core-power ratios are greater than those for the Peach Bottom facility.

The containment capacity was combined with the deterministic MAAP calculations to determine the timing and location of many of the containment failure modes. Figure 4.4-2, page 4-87, schematically shows how the MAAP determinictic results compared with the ultimate containment capacity.

The mean containment ultimate failure pressure was calculated to be 140 psig (Section 4.4.4.1, page 4-99).

'Ibe large DAEC containment Weions have silicone rubber seals, which begin to fail at a i

--fed.ne of 700* F in nonsteam envimaments. Under " dry" severe accident conditions where water is not available to cool the slumped core debris, containment heatup is expected.

Under this scenario, the silicone rubber seals are expected to maintain resiliency up to !

700* F; beyond 700* F the seals are assumed to fail. Under " wet" or steam severe accident conditions where debris cooling is available, the containment environment is expected to be saturated at the maximum containment pressure (~ 150 psig or - 500* F to 600* F) and the seals are assumed to fail between 500* F to 700* F. The outer seal would be somewhat - l protected from the environment even if the inner seal failed.

At 800* F, the flange seal material is expected to deteriorate significantly. At this temperature, the yield strength of the bolts also drops to the point that yield could occur around a pressure of 88 psig. This would increase the leakage area and could create an area similar to that of a rupture.

The DAEC containment failure modes are consistent with those identified in Table 2.2 of NUREG-1335.

Duane Arnold Energy Center 6 August 1995 IPE Back-End Review

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

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2.2d Containment Isolation Failure !

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The DAEC IPE team modeled the containment isolation failure as the first node in the CET.

The fault tree for containment modeled (Section 4.6.2.4, page 4-194):

l . . . containment batches and large lines that Wwe the containment and open to

] the containment atmosphere (e.g., purge and vent lines). 'Ibe fault tree considers automatic isolation signals, pre-existing open pathways, manual isolation, and x+;-:-:= failures. ;

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As indicated in Table 4.5-5, page 4-159, the containment isolation failure is defined as a ;

i line, hatch, or pe . union opening with a diameter greater than 2 inches. h IPE modeled l a containment isolation failure as a large (2 ft') failure in the drywell. !

1he IPE team calculated the DAEC containment isolation failure probability using a fault tree j for containment isolation failure coupled with data for containment isolation failure as

extrapolated by Pacific Northwest laboratory for the NRC. [3] The frequency of isolation failure is not reported in the submittal. Table 4.8-6 of the submittal lists the following with l regards to isolation failure

i l Containment isolation system is highly reliable, h operating experience of DAEC i and other BWRs indicates that containment isolation is reliable and that early release j due to containment isolation failure is a negligible contribution to risk.

l The DAEC IPE team appears to have assessed and identified contributors to containment i isolation failure.

i 2M Svevammuman Pamonse.

i l Section 4.6.2.5, page 4-195 of the submittal, notes that the DAEC IPE team considered proceduralized operator actions (i.e., directed by current EOPs).

l l Table 4.6-2, page 4-196, lists the following proceduralized operator actions considered in the ,

back-end analyses: i l

Isolation of primary containment pathway, given failure of automatic isolation 4

RPV depressurization i

Injection recovery

Off-site power recovery i

Emergency diesel generator recovery i

Combustible gas venting ,

Containment flooding

Containment venting

Manual alignment of alternate injection systems

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RPV venting.

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i Duane Amold Energy Center 7 August 1995 IPE Back-End Review

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j The IPE team analyzed operator involvement in the following CET top events:

Operator depressurizes RPV (IS)

Core melt progression anested in-vessel (RX)

Combustible gas venting initiated (GV)

Containment flooding initiated (FC)

Drywell steel shell intact (SI)

Venting initiated and successful (CV).

De DAEC IPE team appears to have taken significant credit for the back-end operator actions and the back-end msuhs were driven by these operator actions. For example, the operator action to flood the containment resulted in a relatively low conditional probability (given core damage) of drywell shell failure of 1 percent. Drywell venting, which was performed to release combustible gas, had a relatively high conditional probability of 29 Percent.

116 Radionuclide Release Characterization Section 4.7, pages 4-224 through 4-272 of the submittal, describes radionuclide release characterization for the DAEC plant. As noted on page 4-224:

De radionuclide release sequences determined from the CET evaluation that exceed the screening criteria frequency (i.e., reporting criteria) have been assessed to determine their radionuclide release magnitude.

De DAEC IPE team used the following to determine the radionuclide release magnitude (page 4-224):

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Existing Mark I radionuclide releases for a similar plant to charactenze some release sequences

Plant-specific DAEC calculations to both confirm the surrogate plant calculadons and to fillin missing sequence calculations.

heed on discussion on another page in the submittal this surrogate plant appears to be Peach Bottom.

De DAEC IPE defined release categories based on release timing and release severity. The three release timing categories used were as follows:

Early (E) Less than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months from accident initiation

Intermediate (I) Greater than or equal to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , but less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Iate (L) Greater than or equal to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Duane Arnold Energy Center 8 August 1995 IPE Back-End Review

4 i De five release severity categories used were as follows:

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High (H) Early fatalities Moderate (M) Near-term health effects Low (L) Iatent health effects Iow-I.aw (LL) Potential for latent health effects

Negligible (No Release) Iess than or equal to the cantninment design base leakage.

De IPE team did not perform consequence analyses to determine the release fractions for the above categories. Instead the team myiewed the msuhs of daenitari consequence analyses from previous IDCOR nudies, PRAs, and NRC studies and found the following Csl release fractions: High - greater than 10%, Moderate - 1 to 10%, Low - 0.1 to 1.0%, Iow-Low

- less than 0.1 %, and Negligible - much less than 0.1 %. ,

Table 4.6-5, page 4-202 of the submittal, lists the release frequencies for 13 accident classes under 13 release categories, defined as listed above. The DAEC used the MAAP computer code to calculate the source term.

As reported in the utility responses to NRC requests for additional information, the DAEC has a "large" release frequency of 1.3E-6 per year.

Although not explicitly stated in the submittal, the DAEC IPE team appears to have documented the radionuclide release categories appropriately for accident sequences excaarhng the Generic letter 88-20 screening criteria.

2.3 Accident Progression and Containment Performance Analysis LL1 Severe Accidant Prorression.

De DAEC used the BWR Mark I version of the MAAP computer code (version 3.0B, nevision 7.03) to calculate the containment survivability under the postulated severe accident conditions. Table 4.6-1, page 4-179, of the submittal lists a sample of four MAAP calculation cases. In Section 4.6.3.3, Figures 4.6-2 through -9, give the predicted drywell !

pressure and temperature transients. (Dese figures appear on pages 4-181 and 4-182,4-185 through 4-188, and 4-190 and 4-191, respectively.)

Table 4.2-2, page 4-64 of the submittal, shows that the DAEC IPE team did not assess the following sequences or phenomena by using MAAP calculations, but did assess them using probabilistic analyses, j l

Ex-vessel steam explosion )

Mark I shell failure

Direct impingement-induced failure

Dimet containment heating

Reactivity insertion during core melt progression.

Duane Arnold Energy Center 9 August 1995 l IPE Back-End Review l

., e i i In the sensitivity assessment for drywell shell failure, it was noted that such a failure,

' induced by molten debris, has the potential to be a beneficial effect, " creating a radionuclide release pathway through the reactor building with increased DF potential."

I (See page 4-385). 1 The DAEC IPE team appears to have used a reasonable process to understand and quantify severe accident progression and to have addressed the phenomenological uncertainties of !

J accident progression.

l W rh=ian=* Cantributors Can=letancy with IPE Taciehts.

j In Table 1, below, dominant contributors to DAEC containment failure are compared with l those contributors identified during individual plant ernminations performed at nimilar plants, and with the NUREG/CR-1150 PRA results obtained at Peach Bottom. No major differences 1

l exist among the various results.

i i Care should be taken in making such comparisons (as shown in Table 1 below) because the ,

! definitions of failure categories may vary from one IPE to another. Note, for example, the l j different definitions of "early" in Figure 4.7-5 on page 4-244 of the submittal. l i

! Table 1. Containment Failure as a Percentage of CDF: i

! DAEC Results Compared with other IPEs and with

{ Peach Bottom NUREG-1150 PRA Results Containment Fitzpatrick Oyster Browns Nine Mile Peach Duane j

Failure IPE Creek Ferry Point 1 Bottom / Arnold ;

IPE IPE IPE' NUREG- IPE' [2]

1150 CDF (per year) 1.9E-6 3.2E-6 4.8E-5 5.4E-6 4.5E-6 1.5E-5 Early Failure 60 16 46 25 56 41 ;

Bypass na 7 na 0.5 na 0.6 Late Failure 26 26 26 62 16 29 3 0 3 13 18 292 Intact w/ Vessel Breach Intact w/o 11 51 25 na 10 na Vessel Breach _

3Ca=am.==aar faGure time is daA=ad relative to accadent ==aaaaa 2in the DAEC response to the NRC RAI Figure A-6 abows a oce&tional probability of co=======* insect of 30%,

bisblearly releases of 95, and all onbar releases of 61 %. ne== the v.-? * " of aa===i=--* faGure modes abown in his amble were not emphcaly given, abey were comiputed from the release '.:;- --y values givenin Table A-2 of the same reference. [2]

ma Nat available Duane Arnold Energy Center 10 August 1995 IPE Back-End Review

1 2J.J Characterization of Containment Performance. l l

I

'Ihe DAEC IPE team categorized the containment challenges under accident conditions i accordmg to one of the following four regimes (Section 4.4.3, pages 4-94 and 4-97 of the J submittal): -

Pressure-induced containment challenge

Temperature-induced containment challenge

Combined pressure- and temperature-induced containment challenge

Dynamic loads.

'Ibe CETs are grouped into three categories: CErl--containment initially intact; CET2--containment initially failed or seriously challenged before core melt; and CET3---containment bypassed. Of the three CETs, CETI had the most top events with a total number of 19 as follows (CET2 and CET3 had 14 and 5 top events, respectively):

Core damage entry state, IA

Containment isolated, IS

Operator depressurized the RPV, OP

Core melt arrested in-vessel, RX

Combustible gas venting initiated, GV

Containment intact, CZ

Mark I shell failure precluded, SI

Injection established to RPV or drywell, TD e Containment flooding initiated, FC

Containment intact during flood or RPV breach, CX

Flood completed, FC

Contalunent heat remuval initiated, HR

Ventir.g imitated and successful, CV

Suppression pool not bypassed, SP

No large containment failure, NC

Coolant inventory makeup, MU

Drywell intact, DI i

Wetwell airspace breach, WW l

Release mitigated in reactor building, RB. )

The quantification of the Level H IPE model merged all of the deterministic thermal-hydraulic calculations, the postulated containment failure modes, the assessment of the nnntainment ultimate strength, the assessment of mitigation, and the probabilistic assessment of the likelihood of each.

I Duane Amold Energy Center 11 August 1995 IPE Back-End Review

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

e .O The CET functional top event, " Containment Remains Intact (CZ)," was used to analyze the a sics phenomena landing to early containment failure. The IPE team concluded that m-eful prevention of early containment failure required the following (Table 4.5-5, i p. 4-162):

No direct containment hanting (DCH was precluded if the RPV was e de,ressu,aed, i

j

No in-vessel steam explosion (in-vessel steam explosions were precluded if either the RPV was at high pressure (> 100 psig) or the core did not fragment

into fine particles before dropping on to the bottom head)

No ex-vessel stamm explosion 1

No failure of vapor suppression (i.e., the suppression pool was not bypassed and j vapor suppression success was guaranteed by having no more than I drywell to j wetwell vacuum breaker failed open) i

No high pressure spikes sufficient to cause containment failure at the time of j vessel melt-through (i.e., extreme pressure spikes were precluded if the RPV q bottom head penetration failed locally; or the RPV remained at low pressure)

I l .

No hydrogen deflagration or detonation (i.e., if the containment remainad inert

! or effective combustible gas vent was operated successfully then hydrogen i detonation or deflagrations was guaranteed not to occur)

No RPV blowdown from high pressure with the suppression pool t.mperature

above 24(T F 1

1

No recriticality due to an unusual core configuration that may be achieved

! during the melt progression.

j If the above failure modes could not be prevented, containment failure was assumed to ,

occur. The failure location was assumed to be in the drywell head region and was classified as a large failure. The probabilistic assessment used probability values based on industry or ,

l NRC studies. If a substantial energetic event occurred probabili=tiem11y, it was assumed to l

fail the containment.

t The DAEC IPE team used the MAAP computer code to calculate the containment

performance in CET sequence evaluation. Table 4.6-1, page 4-179, lists a sample of four MAAP calculation cases for four accident sequences representing four accident classes.

l Figures 4.6-2 through -9 give the predicted drywell pressure and temperature transients. The

. following is a summary of resuks:

4 i,

i Duane Arnold Energy Center 12 August 1995 IPE Back-End Review

k Class IA: Inss ofMakeup at High RPV Pressure with the Containment Initially Intact. ne

- timing of key events for this sequence was as follows

i Core uncovered 0.69 hours7.986111e-4 days 0.0192 hours 1.140873e-4 weeks 2.62545e-5 months l Initiation of core damage 1.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months L Initiatian of core melt 1.28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months

!. RPV failure /bmach 3.38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months .

l 1 In this sequence the containmant. penninad intact and negligible releases occuned. A drywell

! pressure spike of 67 psig and a temperature spike of 580' F occurred at vessel breach, which

! did not cause containment failure. Following these spikes containment pressure and j temperature increased slowly up to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months into the accident and then started to level off j and no cantninmant challenge occuned.

I i Class ID: Loss ofAdequate RPVMakeup at High RPV Pressure. %e key events in this l postulated sequence involved the plant response when the RPV had been secesdully depressurized, but no injection was available to the RPV. The timing of key events for this l sequence was as follows:

3

{ RPV depressurization 0.59 hours6.828704e-4 days 0.0164 hours 9.755291e-5 weeks 2.24495e-5 months

Core uncoveied 0.62 hours7.175926e-4 days 0.0172 hours 1.025132e-4 weeks 2.3591e-5 months j Initiation of core damage 0.64 hours7.407407e-4 days 0.0178 hours 1.058201e-4 weeks 2.4352e-5 months

! Imtiation of core melt 1.39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months

! RPV failure / breach, 1.91 hours0.00105 days 0.0253 hours 1.50463e-4 weeks 3.46255e-5 months Containment failure ~27 hours Radionuclide release 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months .

Cantninmant failure occurred at the drywell, which was of large size, ne radionuclide release magnitude was moderate. A pressure spike occurred at vessel breach, which was less than that for a high-pressure blowdown. Over the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months containment pressure and temperature rose until the primary cantainment failed 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months into the accident.

Class H: Loss ofAdequate Containment Heat Remonrl. nis accident involved core damage l only after containment failure occurred. Containment heat removal was postulated to fail but l the capability of injecting coolant into the RPV remainad available until containment failure i occurred. De timing of key events for this sequence was as follows:

1 Core uncovered 0.02 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Initiation of core melt 26.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months RPV failure / breach 28.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months Containment failure 25.0 hours0 days 0 hours 0 weeks 0 months Radionuclide release 28.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . '

Containment failure occurred at the drywell which was of large size. De radionuclide release magnitude was High. De containment failed at relatively low containment ,

temperature 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months after scram and loss of containment beat removal. I I

Duane Arnold Energy Center 13 August 1995 l IPE Back-End Review

Class IV: A1WS Induced Containment Failure Followed by Core Damage. In this accident sequence containment failure was induced by a rapid increase in containment pressure, which !

preceded core damage. The timing of key events for this sequence was as follows:

Core uncovered 0.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

Initiation of core damage 0.99 hours0.00115 days 0.0275 hours 1.636905e-4 weeks 3.76695e-5 months Initiation of core melt 1.39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months RPV failure / breach, 4.20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months Containment failure 1.0 hours0 days 0 hours 0 weeks 0 months Radionuclide release 1.0 hours0 days 0 hours 0 weeks 0 months . i r'antninment failure occurred at the drywell head, which was of large size. The mdionuclide release ==gniawia was moderate. The containment failed "early," i.e., within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and core damage followed soon afterwarti.

214 Imnace on h..:...=. nahnvior.

Section 4.6.2.3, pages 4-192 through 4-194, describes equipment survivability in severe accident environments. Before taking credit for equipment operation under severe accident conditions, the IPE team assessed the capability of the equipment "to perform the function for a specific period of time considering exposure to temperature, pressure, aerosol loading, ;

radiation, and moisture." The IPE team reviewed the results of research studies and i equipment survivability tests of the following components:

Cables

Electrical penetration assemblies

Electrical connections

Solenoid valves

Motor-operated valves

Motor-driven pumps

Motor control centers.

The DAEC IPE team appears to have considered the impact of severe accidents on equipment behavior.

211 Unc,*=inty and Sancitivitv An=1yses.

Section 4.2.3.1, page 4-67, of the submittal notes that the DAEC IPE team treated phenomenological uncertainties using MAAP sensitivity studies and insights from other unwiian. Section 4.9, pages 4-326 through -425, describes the DAEC sensitivity study of which the key results are summarized in Table 4.9-25, pages 4-417 through -425. As noted on page 4-337, the DAEC addressed sensitivities to the folicaing:

Core-melt progression

In-vessel hydrogen generation

's RPV pressure at vessel failure Duane Arnold Energy Canter 14 August 1995 IPE Back-End Review

=

f

Late Cs! revapuriunion from the RPV Debris spread in containment l

j

Amount of debris retained in RPV

Ex-vessel debris coolability

!

Shell failure

!

Containment failure location 4

Cantainment failure area

Reactor building effectiveness.

1 2.4 Reducing Probability of Core Damage or F1ssion Product Release 4

W Depaitian of Vidnerability.

4 j 'Ibe DABC IPE team defined vulnerability based on the answers to the following questions,

! which were refened to as " criteria" (Section 3.4.2, page 3-444):

Are there any new or unusual means by which core damage or containment l

failure occur as compared with those identified in other PRAs?

Do the results suggest that the DAEC core dmnage frequency would not be able l l to meet the NRC's safety goal for core damage?

i

Are there any single failures of components that lead directly to a core damage l l state (not including the common cause failure of multiple components of similar I 3

^

types)?

I Based on the answers, Iowa Electric found no vulnerabilities at DAEC.

t W Plant Imorovements 1

'Ihe back-end insights gained through the conduct of this IPE, as well as suggestions for possible improvements / strategies (Section 6.2.6, pages 6-9 and 6-10) were the following:

j

Termination (in accordance with the procedures) of RPV injection when the l

l containment pressure exceeds a set limit can lead to core damage and a

< subsequent containment challenge. 'Ibe prudency of terminating water injection to the containment under any circumstances for which core degradation may be aggravated should be evaluated.

! . As an ex-vessel recovery action, the use of containment sprays and drywell i sprays in lieu of low-pressure coolant injection appears to be most useful m l pse to degraded core conditions. Prioritization of injection systems could be included in future accident management development.

l i

Initiation of drywell sprays before RPV breach would preclude debris attack and failure of the drywell shell for some of the accident sequences that would allow i

15 August 1995 j

Duane Arnold Energy Center IPE Back-End Review l

< a

! or call for spray initiation before vessel breach. (Class IIIC, Class V, and Class ID sequence types would not allow or call for spray initiation before vessel

breach.) Consideration of changes to BOPS allowing the use of drywell spray initiation as well as removing any ambiguity regarding the diversion of injection j sources away from the RPV when adequate core cooling is not assured (i.e., low

. reactor water level) could be included as part of future accident management development.

Drywell sprays offer an additional alternative to control of the drywell i

temperature to avoid premature containment failure. Palavation of the i' restrictions on the use of the drywell sprays in the drywell spray initiation curve of the EOPs may be a possible future accident management item to develop. l i

i

!'

EPG duections with regard to containment flooding sequences can result in the j i highest consequences at the earliest time. Future accident management strategies

should provide guidance to the operator on protecting containment and cooling debris using methods that do not require venting of the RPV and that avoid using the drywell vent unless no other altemative exists, j 2.5 Responses to CPI Program Racammendations !

i Generic Ietter No. 88-20, Supplement No.1, reiterates the following recommendations made during the Containment Performance Improvement Program (CPI) pertaining to the Mark I containments: !

Create alternate water supply for drywell spray / vessel injection

Enhance reactor pressure vessel depressurization system reliability

Implement emergency procedures and training.

Supplement No. I also notes that the above improvements should be considered in addition to improvements that stem from the evaluation and implementation of the hardened vent.

DAEC plans to install a hardened piped vent and the responses to other recommendations are as follows (pages Q#17-1 through -13 of the submittal [2]):

Several alternate injection sources that can provide external water sources to the RPV or drywell sprays were included in the DAEC Emergency Operating Procedures. Alternate water injection was included in a number of accident sequences as a potential method of preventing core damage, preserving containment integrity, or flooding containment. If alternate water injection could be made perfect (failure probability of 0) for all sequences, the base CDF would be reduced by 1.8E-8 per year and the high/early release would decrease by a factor of 5.

Duane Arnold Energy Center 16 August 1995 IPE Back-End Review

. s i

l

h safety relief valves were found to reliable for reactor pressure vessel depressunzation. Alternate hardware modifications to further improve the

reliability were not found to be justifiable.

!

DAEC had incorporated the BWROG Rev. 4 EPGs into EOPs. Iowa Electric j had participated in the BWROG Emergency Procedures Committee during i development of Rev. 4 and has been involved sinx then in examining potential l j changes. l 1 1 i 2.6 IPE Insights, Improvements, and Commitments l

I Section 4.8 of the submittal describes the IPE insights formulated based on specific DAEC l calculations and probabilistic modeling of accident progression. The following is a summag I as listed in Table 4.8-6:

! I j

Containment isolation system is highly reliable. The operating experience of DAEC and other BWRs indicates that containment isolation is reliable and that early release due to containment isolation failure is a negligible contribution to risk.

DAEC EOPs specify depressurization for most situations required. DAEC EOPs specify halting depressurization at 200 psig when turbine-driven systems are available but low-pressure injections systems are not.

h DAEC EOPs are directed to the restoration of adequate core cooling even during degraded core states.

Based on MAAP calculations, the containment or drywell spray may be used in lieu of low-pressure coolant injection in response to degraded core conditions. ,

i

Drywell shell failure due to debris attack can be prevented if drywell sprays are initiated before reactor pressure vessel breach and the drywell floor is filled with I water to quench the debris.

Iowa Electric plans to install a hardened pipe vent. No other commitments were made for plant or procedure improvement.

l Duane Arnold Energy Center 17 August 1995 IPE Back-End Review

i j 3. CONTRACTOR OBSERVATIONS AND CONCLUSIONS i

l As discussed in Section 2 of this TER, the DAEC IPE submittal contains a large amount of

! back-end information, which contributes to the resolution of severe accident vulnerability j issues at the DAEC plant.

l 'Ihc key points of the SCIENTEf.H technical evaluation of the DAEC IPE back-end l submittal are mmmarized as follows:

i

In our opinion, the back-end portion of this IPE submittal provides a thomugh, i detailed, and well-written narrative on all aspects of severe accidents, l containment (and reactor) building respense, and anticipated radiological releases j to the environment. The results are displayed in a variety of ways and in a

! graphical format that is easy to understand (see, for example, the series of

figures on pages 4-218 through 4-223 and on pages 7-10 through 7-23).

i

The quantitative results appear consistent with severe accident phenomenology and past analyses of similar BWRs using Mark I containments. However, it was i difficult to track the quantitative analysis through (from PDS to release category) j because CET top event split fraction values were not listed for the important i sequences. The quantification specifics were not provided in the original submittal. However, the quantification pmcess and values were provided for an important top event in the DAEC response to the staff's RAI. This indicates i that the DAEC did employ an appropriate quantification methodology.

i

The sensitivity analyses performed were extensive and informative (see pages i 4-326 through 4-425), but they relied heavily on the MAAP code, with little

indication of the validity of the code over the ranges of all the sensitivity j parameters. The IPE used EPRI guidance for sensitivity studies.

f

The IPE team appears to have responded adequately to the recommendations of

. the Containment Performance Improvement Program.

i

Although it is not explicitly stated in the submittal, the DAEC IPE team appears l

to have documented the radionuclide release categories appropriately for accident j sequences exceedmg the Generic Ixtrer 88-20 screening criteria.

i

!

The DAEC IPE team appears to have taken significant credit for the back-end

! operator actions and the back-end results were driven by these operator actions.

3 For example, the operator action to flood the containment resulted in a relatively low conditional probability (given core riamage) of drywell shell failure of

1 percent. Drywell venting which was performed to release combustible gas j from the containment had a relatively high conditional probability of 29%. A l part of the sequences involving drywell vertng resulted in early containment

failure (a value not reported in the submittal).

I 4

Duane Arnold Energy Center 18 August 1995 l IPE Back-End Review 4

. s-i I

4. REFERENCES

1. Iowa Electric Ught and Power Company, "Duane Arnold Energy Center Individual 3 Plant Examination Report," November 1992.

4 2. Iowa Electric Light and Power Company, "DAEC Response to Request for Additional Information on IPE," June 1995.

3. P. J. Pelto, K. R. Ames, and R. H. Gallucci, " Reliability Analysis of Containment l Isolation Systems," Pacific Northwest laboratory, NUREGICR-3539, April 1984. !

i i

4 I

i i

1 4

l i

l l

i 1

i I

4 Duane Arnold Energy Center 19 August 1995 IPE Back End Review I

1

1 APPENDIX i l !

IPE EVALUATION AND DATA

SUMMARY

SHEET l BWR Back-end Facts i !

Plant Name l 4

j Duane Arnold Energy Center i

Containment Type Mark I Unique Containment Features None found Unique Vessel Features None found Number of Plant Damage States i

13 l Ultimate Containment Failure Pressure 140 psig ;

Additional Radionuclide Transport And Retention Structures l Suppression pool retention is credited, Extensive analyses of the reactor building retention capabilities were performed. However, as stated on page 4-70 of the submittal:

I.ittle credit is allowed for the reactor building DF. The . . . DF is limited to no more than a factor of 10. . .

Note also the subsections, entitled " Reactor Building Effectiveness," beginning on page 4-152, and " Reactor Building Modeling Assumptions, beginning on page 4-385.

Conditional Probability hat De Containment Is Not Isolated Not available Duane Arnold Energy Center A-1 August 1995 IPE Back-End Review

4 l

i APPENDIX IPE EVALUATION AND DATA

SUMMARY

SHEET i

(continued)

Important Insights, Including Unique Safety Featums

De BOPS that now dictate termination of the RPV injection when containment pressure rises above a set limit may need to be modified.

l

Containment and drywell sprays may be used in lieu of low-pressure coolant

! injection into the RPV.

Current Emergency Procedure Guidelines on containment floodmg could result l in high consequences.

The likelihood that there will be a failure to depressurize is increased by the l policy ofimmediately locking the automatic depressurization system, which puts j

the burden of depressurization on the operator.

l Implemented Plant Improvements i

l None i

a j

i i

Duane Arnold Energy Center A-2 August 1995 IPE Back-End Review

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

3 J

l APPENDIX IPE EVALUATION AND DATA

SUMMARY

SHEET

{ (continued) l i C-Matrix * -

4 1

Accident CDF . Early Late Intact j Class (per year)

!- IA 4.14E-7 0.58 ~ 0.42 i.

j IB 1.92E-6 0.33 0.28 0.39

! IC 1.49E-7 0.48 0.52

ID 4.80E-7 0.45 0.08 0.47 a I

IE 1.01E-6 0.60 0.40 l l IIL 2.64E-7 1.00 !

4 .

IIT 1.64E-6 1.00 !

l IIIB 1.24E-8 0.01 0.99 IIIC 2.62E-8 0.06 0.01 0.93

IIID 1.35E-7 1.00

i. IVA 1.68E-6 1.00

! IVL 8.48E-8 1.00 t

l V 0

}

Notes:

i

1. For a comparable matrix of accident classes (PDSs versus Radionuclide j Releases), see Table 4.6-5, page 4-202 of the submittal.

! 2. These results were obtained from the original submittal. [1] An updated j

matrix may be computed using the results provided on page Q#20-6 of the j i

utility responses to NRC requests for additional information [2] which reports the Revision 3 results of the Probabilistic Safety Analysis.

i 4

i '

I Duane Arnold Energy Center A-3 August 1995 l IPE Back-End Review 1

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