ML20138F092

TER on IPE Back End Analysis

ML20138F092
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/30/1995
From:	Meyer J, Hanry Wagage SCIENTECH, INC.
To:	NRC
Shared Package
ML20138F088	List: ML20138F081 ML20138F092 ML20138G522 ML20138K043 ML20138K056 ML20138K062
References
CON-NRC-05-91-068, CON-NRC-5-91-68 SCI-NRC-233-94, NUDOCS 9607250053
Download: ML20138F092 (45)

Text

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! SCI-NRC-233-94 l

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Arkansas Nuclear One Unit 2 Technical Evaluation Repon on the Individual Plant Exam *mation Back-End Analysis H. A. Wagage ,

J. F. Meyer l

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l Prepared for the U.S. Nuclear Regulatory Commission Under Contract NRC-05-91-068-37 November 1995 SCIENTECH, Inc.

l 11140 Rockville Pike, Suite 500 Rockville, Maryland 20852 ENCLOSURE 4C I

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TABLE OF CONTENTS

PAEC J

E Executive S ummary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E- 1

, l E.1 Plant Characterization .................................E-1  !

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i E.2 Licensee IPE Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1 l l

E.3 Back-End Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-3 i E.4 Containment Performance Improvements (CPI) . . . . . . . . . . . . . . . . . . E-4 l E.5 Vulnerabilities and Plant Improvements . . . . . . . . . . . . . . . . . . . . . . . E-4 l

E.6 Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-5 4

1. INTRODUCTION ....................................... I 1.1 Review Process ............... ...................... 1 1.2 Plant Characterization .................................. 1 l

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2. TECHNICAL REVIEW ....................................

. 2.1 Licensee IPE Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1

2.1.1 Completeness and Methodology ........................ 3 2.1.2 Multi-Unit Effects and As-Built, As-Operated Status . . . . . . . . . . . . 4 2.1.3 Licensee Participation and Peer Review . . . . . . . . . . . . . . . . . . . . 4 2.2 Containment Analysis / Characterization . . . . . . . . . . . . . . . . ...... 5 2.2.1 Front-end Nack-end Dependencies . . . . . . . . . . . . . . . . . . . . . . 9 2.2.2 Event Tree Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 12 2.2.3 Failure Modes and Timing ..........................16 2.2.4 Isolation Failure . . . . . . . . . .......................18 19 2.2.5 System / Human Responses . . . . . . . . . . . . . . . . . . . . . . . . . . .

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TABLE OF CONTENTS (cont.)

Page 2.2.6 Radionuclide Release Characterization . . '. . . . . . . . . . . . . . . . . . 20 2.3 Accident Progression and Containment Performance Analysis . . . . . . . . . 21 2.3.1 Severe Accident Progiession . . . . . . . . . . . . . . . . . . . . . . . . . 21 2.3.2 Dominant Contributors: Consistency with IPE Insights . . . . . . . . . 23 2.3.3 Characterization of Containment Performance . . . . . . . . . . . . . . . 24 2.3.4 Impact on Equipment Behavior . . . . . . . . . . . . . . . . . . . . . . . . 25 2.3.5 Uncertainty and Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . 27 2.4 Reducing Probability of Core Damage or Fission Product Release . . . . . . . 29 2.4.1 Definition of Vulnerability . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 2.4.2 Plant Impmvements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 2.5 Responses to CPI Program Recommendations . . . . . . . . . . . . . . . . . . . 31 2.6 IPE Insights, Improvements and Commitments . . . . . . . . . . . . . . . . . 32

3. CONTRACTOR OBSERVATIONS AND CONCLUSIONS . . . . . . . . . . . . . . 33

4. REFEREN C ES . . . . ,. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4 Appendix 8.

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l E EXECUTIVE SUMhiARY This technical evaluation repon documents the results of the SCIENTECH review of the back-end portion of the Arkansas Nuclear One, Unit 2 (ANO-2) individual Plant  ;

Examination (IPE) submittal. ,

E.1 Plant Characterization The Arkansas Nuclear One, Unit 2 plant is a pressurized water reactor (PWR) with a two-loop by four-loop nuclear steam supply system, designed by Combustion Engineering (CE) and engineered and constructed by Bechtel. ANO-2 has a steel-lined, prestressed, ,

large, dry containment.

ANO-2 centainmen' and reactor coolant system (RCS) characteristics important to the results of the inck-end assessment include .

• The relatively large amount of zirconium in the core, typical of CE PWR designs, which could yield large amounts of hydrogen;

- De absence of vessel lower head instrument tube penetrations, which could lead to

.nore catastrophic vessel failure modes at vessel high p.essures;

• The presence of qualified sprays and fan coolers that reduces the probability of late containment failures; I

• De relatively low susceptibility of the reactor cavity to flooding, compared with other l PWR reactor cavities, and the configuration of the cavity in such a way that the effects l of high-pressure core ejection would be large; and l 1

• The strength and free volume characteristics of the large, dry containment, which make these types of PWR containments so robust.

E.2 Licensee IPE Process Entergy Operations, Inc., (Entergy) perfonned the IPE with suppon from Science Applications International Corporation (SAIC) and ERIN Engineering and Research, Inc.

(ERIN). The utility invested 8 person-years of staff time in the IPE effon, more than 50%

of the total engineering effon. De IPE team worked on site and panicipated in plant walk-throughs and inspections. An independent, in house Entergy team reviewed the results of the IPE with assistance from ERIN.

He IPE team performed modeling that was scoping in nature and relied on p:evious PRA results, mainly from the IDCOR PWR reference plant analyses and the NUREG-1150 Surry Plant study. He team perfonned a limited deterministic analysis and developed a l Arkansas Nuclear One, Unit 2 E-1 November 1995

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! containment event tree (CET), determining the top events from a set of fault trees, which made the CET straightforward to evaluate. However, the team made conservative assumptions, particularly in the phenomenological areas, that are inconsistent with analytical determinations reached in other PWR IPEs. For example, for the two plant damage states i that drive early failure (PDS IVKi and PDS IIKi), the probability of containment failure l was 100%. His certainty of failure could be overly conservative, and thereby skew the containment failure profile. Further, it may reflect an incomplete understanding of the j progression of severe accidents.

j With nine top events, the CET was sufficiently detailed for the purposes of the IPE, integrating the systemic with the phenomenological aspects of severe accident progression.

, De IPE team did not take human actions into account in the CET because no back-end j recovery actions or other actions were assumed after core damage. Entergy has no formali7M Emergency Operating Procedures (EOPs) for beyond-core-damage events.

l j nrough the CET top events, the ANO-2 IPE team was able to directly address j phenomenological issues concerning hydrogen burn, direct containment heating, steam i explosions, molten core-concrete interactions (CCI), and steam /noncondensable gas pressuruation.

To characterize and quantify accident sequences and containment responses, the ANO-2 IPE

, team used a combination of:

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• Existing reference plant analyses; i

• Scoping calculations to account for ANO-2-specific containment features and j responses; and

• Conservative adiabatic calculations of pressure loads from hydrogen bums.

l l The IPE team did not perform most of the EPRI-recommended MAAP sensitivity analyses.

However, their sensitivity analysis was instructive. Members of the IPE team did perform plant-specific structural analyses of the ANO-2 containment, which were scoping in nature and addressed:

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• Structural failure due to pressure-induced yielding of the building tendons; i

• Pressure-induced liner tearing; i

• Imal failures at penetrations due to high temperature and/or pressure conditions;

• Failure of containment isolation system to close; and I

• Containment bypass. i E-2 November 1995 Arkansas Nuclear One, Unit 2 f .

1 E.3 Back-End Analysis The ANO-2 IPE submittal reports a 12.2% conditional probability of early containment failure (not including a " bypass" failure of 1.1 %) which is a relatively large value for a PWR with a large, dry containment. He other values that shape the condidonal containment failure profile (late failure - including basement penetration -- of 13.9% and

" intact" of 72.8%) are comparable to values for other PWRs with large, dry containments.

Because Entergy calculated the frequency of core damage to be relatively low (3.7E-5/ year), it is their position that no vulnerabilities exist at ANO-2, despite the large, early containment failure value. The containment failure profile, especially the early failure component, was driven by three factors: (1) the makeup of the sequences that lead to core damage, (2) specific containment and RCS charactedstics, and (3) modeling assumptions. i Superimposed on to these three factors was the approach that Entergy took, namely the performance of a limited " scoping" assessment, based on extrapolations from previous probabilistic risk assessments (PRAs), specifically the IDCOR PWR reference plant analyses and the NUREG 1150 assessment of the Surry Plant.

The nonstation blackout (SBO) transients dominated the core damage contributors at 84.3%.

Unlike in other IPEs, SBO sequences accounted for only 3.5% of the ANO-2 core damage frequency (CDF). Hus, AC power would be available in most situations, which would permit the use of containment spray and fan coolers. (However, sprays and fan coolers i would fail in the recirculation mode for other reasons, e.g., loss of service water). It was predicted that many of the transients would experience core damage under high-vessel pressure.

Further, the contributors to containment bypass, i.e., the interfacing systems loss of coolant accident (ISLOCA) at 1.0% of the total CDF and the steam generator tube rupture (SGTR) at 0.15% of the total CDF together contributed to more than 87% of the "large" release, that is, the ponion of the early release that would have a high radionuclide content. The other contributor to early release -- early containment failure -- did not make a substantial contribution to large release because mitigative factors were present in the containment at or after containment failure. The ddver of an ISLOCA would be the failure of an RCP seal cooler tube.

The A'IWS sequences (at 3.0%) and the LOCAs (at 8.1%) were the remaining contributors to core damage. ,,

The limiting failure was determined to be liner tearmg at the personnel hatch penetration at a pressure of 141.3 psig.

ne ANO-2 IPE team used the same definition for early contairunent failure as in the NUREG-1150 study, namely, ". . . carly containment failures occur prior to or at about the same time (or shortly after) vessel failure . . . ." The IPE team defined large radionuclide releases as those with a source term equal to or greater than the PWR-4 source term of WASH-1400 (releases of 9% iodine and 4% cesium). The total conditional probability for large release was determined to be 6.9% of all severe accidents.

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The IPE team presented radionuclide releases in terms of 51 release modes, which are also

{ the CET end states. The relationship of the 51 release modes to the plant damage states (PDSs) and the percentile contributions to the CDF from the release modes were provided in tables in the submittal. Five groupings of radionuclides were assessed (NG, I, Cs, Te,

! and Sr).

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j In many of the accident sequences, the ANO-2 containment cavity stayed dry. However, the IPE team did assume that a reactor cavity check valve had a high probability of sticking i open, thus allowing water to enter the cavity during several PDSs.

1 Back-end insights gained from this IPE were derived from considering those PDSs that led 1

to large releases of radionuclides: i iU

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• For the PDSs that did not bypass the containment, loss of service water was a significant contributor to both core damage and containment failure resulting from l containment spray and fan-cooler failures; l

• By stopping one train of containment sprays, or even one train of the HPSI, the time  ;

i to core damage could be extended, thus increasing the potential for recovery; i

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• Containtnent venting before core damage could reduce the probability of early failure;  !

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• Manual depressurization of the vessel could also reduce the probability of early failure;

• The most important " containment isolation failure" PDS is dominated by the failure of i

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the containment vent header isolation valves. Recovery after such a failure could be i achieved by closing a manual isolation valve; and  !

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• In SBO events, recovery of AC power, and thereby of sprays, could cause a deinerting i j of the containment, hydrogen burn, and containment failure.

E.4 Containment Performance Improvements (CPI) j

'Ihe containment performance improvement (CPI) issue regarding global and local hydrogen f bums and detonations was not addressed as such in the original submittal. [1] However, j the IPE team did perform patametric analyses of the effect of hydrogen burns on early and late containment failures. Entergy satisfactorily addressed local pocketing effects in their response to the staff's Request for AdditionalInformation (RAI). [5]

l E.5 Vulnerabilities and Plant Improvements J

In order to identify potential containment response concerns, the IPE team adopted back-1 end screening criteria from NUMARC 91-04. The team compared the IPE results to these i criteria and identified no back-end vulnerabilities for the ANO-2 plant. This finding was l based on the following:

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• 'nie largest CET end state group involving containment failure had a frequency of 4.8E-6 per reactor year, which was less than the 1.0E-5 per reactor year criterion;

• The dominant large release CET end state represented 3.8% of the total CDF, which was less than the 10% criterion; and -

• The largest CET end state group involving containment failure represented 13.0% of the total CDF, which was less than 20% of the total CDF.

Entergy discussed potential plant improvements in two categories: procedures and -

hardware. Most were front-end improvements. However, procedures were suggested for:

• Coping with failed or degmded service water; and i

• Flooding the fuel transfer tube to protect the seal from high-temperature-induced failure caused by HPME events. In response to the staff RAI, Entergy stated that '

funher investigation revealed that the fuel transfer-tube seal failure is no longer risk-significant.

Entergy provided a list of procedural and hardware improvements including their status in their response to the staff's RAI. [5] The only back-end hardware fix currently "under evaluation" is removal of the reactor vessel cavity check valve intemals so that more water can flood the reactor cavity. ,

E.6 Observations Although the IPE team used a scoping approach, the back-end ponion of the IPE appears to satisfy the intent of Generic letter (GL) 88-20. No accident progression analyses were perfonned (for example, using MAAP). Instead, Entergy used the results of existing reference plant analyses combined with separate-effect, scoping calculations. For example, the pressure rise from hydrogen burn was calculated separately and compared with NUREG-1150 Surry calculations. Radiological source terms were assessed in a similar manner. The individual calculations were entered into the quantification of the containment event tree.

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1. - INTRODUCTION

, 1.1 Review Process This technical evaluation report (TER) documents the results of the SCIENTECH review of

the back-end portion of the Arkansas Nuclear One, Unit 2-(ANO-2) Individual Plant j Examination (IPE) submittal. [1,5] nis technical evaluation report complies with the l 1 requirements for IPE back-end reviews of the U.S. Nuclear Regu!atory Commission (NRC) 3 in its contractor task orders, and adopts the NRC review objectives, which include the

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• To help NRC staff detennine if the IPE submittal provides the level of detail requested 3

in the " Submittal Guidance Document," NUREG-1335; i

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• To help NRC staff assess if the IPE submittal meets the intent of Generic j Letter (GL) 88-20; and ,

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• To complete the IPE Evaluation Data Summary Sheet.

i SCIENTECH sent the NRC a draft TER on the back-end portion of the ANO-2 IPE submittal in December 1994. Based in part on this draft submittal, the NRC staff submitted

a Request for Additional Information (RAI) to Entergy Operations, Inc. (EOI) on May 8,

- 1995. EOI responded to the RAI in a document dated October 5,1995. [5] nis final 3

TER is based on the original submittal and the response to the RAI.

Section 2 of the TER summarizes SCIENTECH's review and briefly describes the ANO-2 IPE submittal, as it pertains to the work requirements outlined in the contractor task order.

Each portion of Section 2 corresponds to a specific work requirement. Section 2 also l 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 i

conclusions. References are given in Section 4. He Appendix contains an IPE evaluation

and data summary sheet.

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1.2 Plant Characserization ne Arkansas Nuclear One, Unit 2 plant is a pressurized water reactor (PWR) with a two-loop by four-loop nuclear steam supply system, designed by Combustion Engineering '

(CE) and engineered and constructed by Bechtel. ANO-2 has a steel-lined, pre-stsessed, large, dry containment.

i ANO-2 containment and reactor coolant system (RCS) characteristics important to the i results of the back-end assessment include: i

• The relatively large amount of zirconium in the core, typical of CE PWR designs,

which potentially could yield large amounts of hydrogen; 1

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• The absence of vessel lower head instrument tube penetrations, which potentially could lead to more catastrophic vessel failure modes at vessel high pressures;

• The presence of qualified sprays and fan coolers that reduces the probability of late containment failures;

• The relatively low susceptibility of the reactor cavity to flooding, compared with other PWR reactor cavities, and the configuration of the cavity in such a way that the effects  ;

of high-pressure core ejection would be large; and

• 'Ihe strength and free volume characteristics of the large, dry containment, which make these types of PWR containments so robust.

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i 2. TECHNICAL REVIEW ,

j 2.1 Licensee IPE Process 2.,11 Comoleteness and Methodolony. .

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The ANO-2 IPE submittal contains a substantial amount of information in accordance with l 1 the guidance in GL 88-20, its supplements, and NUREG-1335 and appears to satisfy the ,

level of detail requested in NUREG-1335. l 4

l The methodology used to perform the IPE is described clearly in the submittal. The

approach taken, which is consistent with the basic tenets of GL 88-20, Appendix 1, also is

described along with the team's basic underlying assumptions. The important plant l l information and data are documented and the key IPE results and findings are well j i presented.

In performing the ANO-2 IPE, the IPE team used the "small event tree, large fault tree" j approach, which relies on the PRA technique called " fault tree linking." The team I performed a limited-scope Level 2 PRA in which " detailed plant specific phenomena

analysis and structural analysis were not performed, but ANO-2 design features were considered in the quantification of containment performance via limited engineering i 3 calculations or as a basis for scaling reference plant analysis." (Section 1.3, page 1.3-1) 4 L

The IPE team developed fault tree models of containment systems to ensure proper l interface between Level 1 and Level 2 analyses and to use plant-specific data for the l l  ;

4 containment performance analysis. These fault tree models were run to obtain the various combinations of containment system failures (sprays, fan coolers, isolation). The l i

Izvel 1/ Level 2 interface was maintained by combining the Level I cutsets with the Level 2 l l l

! cutsets for each accident sequence.

1 The combination of level 1 and Level 2 cutssets linked the core damage end state to a plant

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i j damage end state. The result was used in the containment event tree (CET) which modeled the containment response, severe accident phenomena, containment failure modes, and l fission product release information in an event tree format. The IPE team quantified the CET by first integrating the containment system failure cutsets, core damage cutsets, and l j  !

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related source term, and then calculating the magnitude, frequency, and contributing failures that characterize radionuclide release.

F De ANO 2 IPE team defined "carly" containment failure as that occurring before or at the l.

occurrence of vessel failure. (Section 4.6.3.4, page 4.6-7) nis definition is consistent with that used in NUREG-1150 for PWRs.  ;

3 j Re ANO-2 IPE team predicted that core damage would occur if the core heated up  !

significantly (e.g., beyond 2,200'F) and used a 24-hour mission time for the IPE analysis.

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1 2,12 Multi-Unit Effects and As-Built. As-Onerated Starus.

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The ANO-2 IPE team used the results of walkdowns, interviews, procedure and drawing ,

reviews, and the results of independent peer reviews that ANO-2 plant personnel and PRA 1 l

j contractors conducted to ensure that "the PRA models accurately reflect the plant design l and operation and are consistent with accepted PRA practices." (Section 1.4, page 1.4-1) 1' 2.,13 r ie*nw Panicination and Peer Review.

i To bring PRA technology and expertise permanently into the organization, Arkansas

Nuclear One (ANO) established a group in its Nuclear Engineering Design Depanment .

(NEDD) to be responsible for developing and conducting probabilistic risk assessments. l De ANO-2 IPE team consisted of members of the PRA group and individuals from the two l

cont:3ctor organizations, Science Applications International Corporation (SAIC) and ERIN l

Engineering and Research, Inc. Figure 5-1 of the submittal shows the ANO IPE l

organizational chan (fx both Units 1 and 2) at the time of the IPE submittal. (In March

! 1988, ANO initiated a Level 1 PRA and limited-scope Level 2 PRA for both ANO Units 1 and 2. He results of the ANO-2 IPE were submitted in August 1992. The eventual l

! membership of the ANO-2 IPE team was slightly different than originally planned, but the

submittal states that the personnel changes had no significant impact on the conduct of the examination.

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l ANO personnel were involved in the development, quantification, refmement, and l interpretation of the model used to conduct the Level 1 and limited-scope PRAs. At the outset, SAIC delineated the tasks and managed the overall project plan and individual work j task development or guidelines involved in the conduct of the IPE. ANO personnel assisted

in the necessary technology transfer. His organizational approach was successful, in that ANO personnel were performing most of the work before half of the effort was completed.

l At the end of the project, ANO had total responsibility for the IPE and contractor suppon l

j was needed only on a supplemental basis or for purposes of review. As a consequence of performing the IPE, ANO has within its nuclear engineering design department the benefit l

j of a PRA group whose members include two safety analysis supervisors and 10 engineers, i

all of whom are knowledgeable about some elements of PRA and severe accident issues.

ANO and SAIC jointly performed the back-end analysis: SAIC provided the methodology, 1 ANO provided plant-specific,information, and together they performed the analysis at the l

ANO engineering facility. .

j In addition to the independent reviews that the ANO and SAIC analysts conducted during j

the performance of the IPE, ANO separately and independently reviewed the ANO-2 IPE to I ensure accuracy and " fulfillment of ANO objectives and provide traceability of all l f

assessment data to reliable or controlled design documentation and information sources to i

facilitate future updating efforts." (Section 5.2, page 5-5) The ANO Independent Review Team consisted of ANO Operations, Design Engineering, Training, and Licensing l I employees and ERIN Engineering and Research, Inc. Before staning the review, ERIN j

PRA expens trained the members of the Independent Review Team in two sessions. The first session was to familiarize the team members with the IPE objectives, the basic November 1995 Arkansas Nuclear One, Unit 2 4 l

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___s UNREVIEWED DRAFT first session was to familiarue the team members with the IPE objectives, the basic  !

technical elements of the PRA, the expected role of each participant, and the project '

schedule. The second session was to train the reviewers in key aspects of PRA technology and modeling techniques.

'Ihe independent review covered the following seven aspects of the PRA, of which the first  :

and the last affected the back-end analysis: l c

• Overall PRA methodology;  ;

• Initiating events and accident sequences; j

• System models; j

• Data analysis;

• Human reliability and recovery analysis; i

• Model quantification; and  !

• Containment analysis and release characterization.

The objective of the review of the overall PRA methodology was to provide the ANO PRA team with feedback on the adequacy and accuracy of the PRA. Although this review was performed on the ANO-1 PRA, which was initiated first, the team members considered both ANO units in reviewing project plans and procedures. After the initial review of the ,

PRA work packages and overall methodology, it was concluded that the IPE project was j meeting IPE and ANO objectives and no major changes were necessary. After the final  ;

review, the need was recognized for a dedicated group of ANO individuals to maintain j familiarity with the PRA models and their use. j i

Review of the containment analysis and release characterization involved study of the documentation associated with the draft submittal report and plant-specific calculations.  !

This review was performed in a 2-day meeting with " ERIN Review Team, key ANO l Review Team members, specific design engineering individuals knowledgeable of I containment features, and SAIC and ANO level 2 PRA experts." (Section 5.3.7, page 5-9)

The review team found that the methodology used in the back end analysis was consistent with the NRC guidance and recommended that additional description of the plant-specific analysis be provided.

It appears that the ANO-2 IPE submittal received adequate peer review.

2.2 Containment Analysis / Characterization The ANO-2 plant is located adjacent to the Arkansas Nuclear One Unit 1 (ANO-1) plant on

, a peninsula in Dardanelle Reservoir on the Arkansas River in Pope County and about 2 miles southeast of London, Arkansas. ANO-2 has a capacity of 2815 MWt core thermal power output, which corresponds to a net electrical power output of 912 MWe. ANO-2 is a pressurized water reactor with a Combustion Engineering designed nuclear steam supply system and a Bechtel designed large, dry containment system.

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In Table 1 of this report, key plant and containment design characteristics of ANO 2 are 5

compared with those of the Zion and Surry plants. The containment robustness to decay

heat measure shows that ANO-2 is slightly more robust than Surry, but less robust than '

Zion. The containment robustness to zircaloy (oxidation and hydrogen burn) measure shows that the ANO-2 containment would perform significantly less robustly than Zion 1 would and also less robustly than Surry. However, the IPE analysts concluded that the

ANO-2 containment would perform similarly to Surry after the impact of zircaloy oxidation

'and hydrogen burn was taken into account, as described below under " Reactor Core and Reactor Coolant System." Without apparent further justification, the IPE team chose the  ;

Surry plant as the primary reference plant for ANO-2 severe accident response analysis (Section 4.1, page 4.1-1).

q J- Specific plant features and their impact on the containment response are summarized as ,

e follows: .

1 R**ctor Core and Remetor Coolant System. "A comparison of the ANO-2, Zion, and Surry l  ;

j core thermal power, core mass, core zircaloy mass, containment volume, and containment '

! design pressure indicates that containment pressure loads resulting from zircaloy oxidation i and hydrogen burning will not be significantly more severe for ANO-2 than those at the i Surry reference plant. His judgment is reasoned as follows: (1) the ratio of core thermal j power to core mass is about equal for ANO-2 and Surry and (2) the ratio of zircaloy mass j to containment volume is higher for ANO-2 than for Surry; however, the ANO-2 containment design pressure is also higher than that of Surry." (Section 4.1.1, page 4.1-1) e j Secondary Coolant System. In many of the severe accidents simulated during the ANO-2 IPE, the steam generator inventory was considered a passive heat sink for core decay heat,

' even when feedwater pumps were not available. He water inventory in the steam generators was 276,000 lb; the ratio of steam generator water inventory to thermal power i was 98, which was close enough to be comparable to the 110 ratio at Zion. The impact of j initial steam generator water inventory was relevant only to accident sequences in which the emergency feedwater was lost, and the reactor coolant system remained at a high pressure, l  !

i e.g., during a station blackout accident in which the time to core uncovery might be i

delayed until steam generator boils dried.

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Table 1: Comparison of ANO-2 Key Plant and Containment Design  !

Characteristics with Those of the Zion and Surry Plants 1

Charactenstic Zion Surry ANO-2

, Manufacturer Westinghouse Westinghouse Combustion )

i Engineering s

Year of Commercial Operation 1972 1973 1980 i l

Thermal Power, MWt 3,236 2,441 2,815 RCS Water Volume, m3 368 261 324 1 Contamment Fme Volume, m' 81,000 49,000 49,800 3  !

i (ft ) (2.86E6) (1.73E6) (1.78E6)

Mass of Fuel, kg 98,250 79,600 83,500 i

, Mass of Zircaloy, kg 20,230 16,463 23,700 )

i l Design Pressure, psig 47 45 54 l

, Operating Pressure, psia 14.7 10 14.7 l

Mean Failure Pressure, psig 134 126 141 Concrete Type Basaltic Limestone / Basaltic common sand Floor Area (Cavity and ICIR), m' 37.1 57.6 13.4

{ ,

Contamment Robustness to Decay 110 78 84 Heat Measure' Containment Robustness to 537 375 296 Zircaloy Measurt2

[ Cont. Free Vol., m'] x [Mean Failure Pree., psig) i ' Coat. Robust. to Decay Heat Meas. -

[ Mass of Fuel, kg)

[ Cont. Free Vol., m') x [Mean Failure Pres., peig) j 8

Cont. Robust. to Zircaloy Meas. =

[ Mass of Zarcaloy, kg) i I

Arkansas Nuclear One, Unit 2 7 November 1995

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Emereency Core Cooline Systems. De emergency core cooling systems (ECCSs) j include the high pressuit safety injection (HPSI) system, the low pressure safety injection (LPSI) system, and safety injection tanks (SITS). It is assumed that the 4 l'

HPSI system at ANO-2 would fail, if fan cooler and spray systems were not available J to reject heat to the environment. De ANO-2 SIT discharge setpoint was about the same RCS pressure value as at the Surry and Zion plants. The SIT water inventory l

j was highet at ANO-2 than it was at Zion and Surry (5,652 ft' at ANO-2 compared

' with 3,400 ft' at Zion and 2,850 at Surry). Hus, for small and medium break

~

LOCAs, it is expected that the SITS could delay core damage significantly. In high '

pressure sequences that preclude SIT injection until vessel failure, it is expected that.

more water would enter the ANO-2 cavity than at Zion and Surry. j 1 i i

Containment Saferuards Fantures. He ANO-2 containment safeguard features consists of the containment spray system (CSS), the containment cooling systems (CCSs), and the containment isolation system (CIS). During severe accidents the CSS i

and CCSs mitigate containment pressure challenges and the CSS scrubs fission product vapors and aerosols from the containment. The CIS ensures that the l containment is leak-tight, preventing release of fission products to the environment. l j

Section 2.1.2.4 of this report describes the CIS in more detail.

l 1

The ANO-2 CSS is set to actuate at a containment pressure of 23.3 psia, which is  !

{ slightly lower than the actuation valc. .et at the PWRs examined during the NUREG- ,

1150 study. (Section 4.1.4, page 4.1-2) Because of this lower setpoint, it was assumed that, if available, the containment sprays would actuate (independent of the l

' operation of CCSs) before vessel failure, ne ANO-2 containment sprays provide -

only a limited amount of water to the reactor cavity. The IPE team assumed that the

$ containment sprays would continue to operate after containment failure and that

containment failure would contribute to containment spray failure. ,

I For the sequences that involve early containment failure, the IPE team assumed that

, the containment fan coolers would fail. The IPE team assumed that the containme i i

would not overpressurize if a single train of the containment spray or containment fan cooling continued to operate after core damage and vessel failure. '

i

Containment and Auvilimrv Buildines. He total volume of the ANO-2 cavity, '

excluding the region displaced by the reactor vessel, below the bottom of the hot leg is 10,675 ft8 The volume of the access tunnel is 19 ft'.

ne debris dispersal paths that are possible between the cavity and the conta'mment i include the access tunnel, the annular region between the reactor vessel and the inside of the cavity wall, and along the hot legs and cold legs. The path through the access i tunnel is confined to 10 ft2 , which is further confined by the 10-inch cavity check valve, 2BS-46, on the cavity access door. The 2 annular path between the vessel and the cavity wall has a cross-section area of 134 ft . This is the dominant path for

i 1

't g November 1995 I Arkansas Nuclear One, Unit 2 i

l 8 1

e

l LWREVIECED DRAFT missile shield, and the debris that missed the missile shield would be discharged into the upper containment and participate in direct containment heating (DCH).

The ANO 2 basemat is about 9 feet thick and is made of a basaltic type of concrete.

The cavity floor area is 144 2ft , which is smaller by a factor of four than the cavity floor area at Surry, which also is made of basaltic concrete.

i 12J Front-end Back-end Deoerdencies.

The ANO-2 IPE team developed a bridge tree, which was used to sort front-end sequences into plant damage states (PDSs). PDSs were used as input to the containment event trees, which were developed and used to analyze the containment performance for the front end accident sequences to determine radionuclide releases to the eiwironment. The bridge tree combines the front-end accident sequences with key core and containment system factors that affect the magnitude and timing of severe accident fission product releases. Figure 4.3-2, page 4.3-22 of the submittal, shows the bridge tree entry state; this is reproduced as Figure 1 of this report. Figures 4.3- '

3A through 4.3-3F, pages 4.3 23 through 4.3-28 of the submittal, show the bridge tree transfer trees; Figure 2 of this report reproduces one transfer tree.

In developing the ANO-2 bridge tree, the team considered the following attributes (Section 4.3.2, pages 4.3-1 thmugh 4.3-3):

• Time of core damare initiation. Very early (0 to 30 minutes), Early (30 minutes to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), or Late (peater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />);

l I

!

• RCS oressure durine core meltine. High (more than 2500 psia), Moderate (between 200 psia and 2500 psia), and Low (lower than 200 psia);

• Containment status and oressure at vessel failure. Containment integrity status at l

l vessel failure: Isolated, Unisolated, Eailed due to overpressurization before core l

damage, and Bypassed by RCS leakage. Containment pressure at vessel failure:

Low or High;

• Aynilability of containment mitientine systems (heat removal / fission product removal) at vessel failure. Containment spray system operation: Containment sprays available in both the injection and recirculation modes, and actuated before vessel failure; containment sprays available in both the injection and recirculation modes, and not actuated before vessel failure (CS/0);

i November 1995 Arkansas Nuclear One, Unit 2 9

UNREVIEWED DRAFT l Ig  :

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November 1995 Arkansas Nuclear One, Unit 2 10

l UNREVIEWED DRAIT i

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Arkansas Nuclear One, Unit 2 11 November 1995 l l 4

]

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

LNuMEDED DRArr l

containment sprays available in the injection mode, fails in the recirculation l

mode, and actuated before vessel failure (CS/R); containment sprays available in i the injection mode, fails in the recirculation mode, and not actuated before vessel failure (CS/OR); and containment sprays not available (CS). Containment fan cooler operation: fan coolers available and fan coolers not available (FC);

• M:ence of water in the mactor cavity before vessel failure. Water is available l

in the reactor cavity and Dry reactor cavity; 1

• RCS ratantian ennahilitv. Whether steam generators are in the leakage path; j whether steam generators boil dry; fission product _ residence time in the RCS (e.g., it would be short in a large break LOCA, thus initial retention would not

be significant); and

! e Nance of eention bincknut (SBO) conditions. Whether an accident involves the loss of all offsite and onsite AC power.

Of the 145 possible ANO-2 bridge tree PDSs, the IPE team found only 56 nonzero sequences, which were further collapsed to a total of 32. De final PDSs selected for the back-end analyses are listed in Table 4.3-3, page 4.3-16 of the submittal, along with the front-end sequences involved in each PDS. De dominant PDS was "IIIAi,"

which represented 62.8 % of the tetal CDF. This PDS was characterized by transient-initiated events, with loss of secondary heat sink, and failure of once-through cooling in the injection mode. Containment isolation was successful and both the containment fan cooling and spray systems were available, ne next most dominant PDS was "IVKi," which represented 5.1% of the total CDP. His PDS also was characterized by transient-initiated events, with loss of secondary beat sink and failure of once-through cooling in the injection mode. Although containment isolation was successful, however, containment cooling was not available.

The IPE team appears to have adequately considered front-end back-end (pendencies by using the ANO-2 bridge tree.

W Event Tree Develonment.

For the back-end analysistthe ANO-2 IPE team used an event tree / fault tree approach, similar to that used in the front-end analysis. The CETs that were developed included "the important phenomenological and system-related events identified in the NUREG-1150 reference PWR accident progression analysis and previous PRAs, including the Zion, Surry, and Oconee PRAs." (Section 4.5.1, page 4.5-1) ne team developed two CETs, i.e., the " Generic" and the " Bypass." De .

Generic CET modeled accidents in which fission products were released initially to i the containment building; the Bypass CET modeled accidents in which fission products were released directly to the environment, thus bypassing possible mitigation j

through initial release in the containment building. De Bypass CET modeled  !

accident scenarios involving interfacing systems loss of coolant accidents (ISLOCAs) and steam generator tube rupture (SGTR) sequences. The Generic CET modeled the

)

November 1995 i Arkansas Nuclear One, Unit 2 12

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remaining accident sequences. Figure 3, reproduced from Figure 4.5-2, page 4.5-16 of the submittal, shows the structure of the Generic CET, which was similar to that of  ;

4 the Bypass CET. The only difference is that "DP" was the second top event in the l Generic CET while "SRV" (secondary side n: lief valves reclose) was the second top ,

event in the Bypass CET. "Ihe Generic CET wasisted of the following nine top events (Table 4.5-1, page 4.5-12 of the submiray: {

PDS Plant damage state-Entry state to the CRT and defined core melt progression boundary conditions of core damage; j

DP RCS depressurized before vessel failure-Defined by the entry state, operator action, or phenomena following core damage; .

REC Coolant recovered in-vessel-Defined operator recovery action, or a passive actuation, should the conditions that preclude initial operation be removed in the DP event; VF No vessel failure-Implied arrest of core melt progrenion, terminating in- ,

vessel, and the subsequent formation of coolable debris geometry; CFE No early containment failure-Implied challenges to containment integrity '

insufficient to fail containment before or at vessel failure; DC Coolable debris formed ex-vessel-Implied formation of coolable debris  ;

geometry ex-vessel, precluding significant core-concrete interaction; i

CFL No late containment failure-Implied mitigation of long-term containment challenges their incapacity to fait containment pressure boundary; i

FPR Fission product removal-Characterized potential fission product release magnitudes; considered as the mitigation of releases from fuel in- or ex-vessel and in-containment removal processes; and ,

CFM Containment failure modes-Characterized implications of. containment failure .

on the magnitudes and duration of fission product release to the environment.

I

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November 1995 Arkansas Nuclear One, Unit 2 13 f

UNREVIEWED DRAFT R

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8 UNREVIEwED DRAFT 1 l

The first four CET top events, i.e., "PDS," "DP" (SRV for the Bypass CET),

" REC," and "VF," were used to analyze events occurring before the vessel failure. 1 The next three top events, i.e., "CFE," "DC," and "CFL," were used to analyze )

events occurring after vessel failure and before containment release. The last two  ;

events, "FPR" and "CFM," addressed containment release mechanisms. De i following are some of the important assumptions used in developing the CETs j (section 4.5.2, pages 4.5-3 and 4.5-4): j

• During a sequence in which RCS inventory was not recovered in vessel, it was assumed that the vessel bottom head would fail. He possibility that external  ;

i cooling of the vessel could avert vessel failure was ignored because (1) insufficient water exists in the cavity, (2) the reflective insulation on the outside of the reactor vessel could impede efficient cooling, and (3) tenlimd regions of high heat flux (resulting from nonuniformity of the core material and crust formation) would result in vessel wall thinning and creep rupture;

• Sequences with high RCS pressure would not permit the recovery of coolant in-vessel, except for station blackout (SBO) events in which offsite power was ,

recovered after vessel failure; and l 1

• The timing of both ECF and LCF are relative to the time of vessel failure.

Early containment failure was assumed to occur as a consequence of the following (Section 4.5.3.4, page 4.5-7):

• Pressure spikes resulting from RCS blowdown and/or hydrogen burning during _

core damage or at vessel failure;

• Energetic fuel-coolant interactions within the vessel at core slump or in the reactor cavity;

• Pressure loads caused by high-pressure melt ejection (HPME) loads at vessel breach at high RCS pressure (HPME included direct containment heating and attendant hydrogen burning); and

• Vessel acting as a " rocket" generating a missile that fails the containment.

Late containment failure was assumed to occur as a consequence of the following (Section 4.5.3.6, page 4.5-8):

• Containment pressurization generated during long-term debris-coolant interactions in the cavity without decay heat removal capability;

• Containment pressurization from noncondensable gas generated during CCI of noncoolable debris beds; Arkansas Nuclear One, Unit 2 15 November 1995

^

UNaEVIEWED DaAFT i t

• Containment pressurization from combustion of hydrogen and other combustible gases produced during CCI;

+

• Basemat melt-through resulting from CCI; and a

• Overtemperature failure of containment seal materials.

, The IPE team appears to have adequately addressed the important issues of ANO-2 l containment performance using CETs.

l 5

W Failure Modes and Timine. ,

I 'Ihe ANO-2 IPE team considered the following containment failure modes (Section 4.4, page 4.4-1):

i.

• Containment structural failure resulting from pressure-induced yielding of the j building tendons;

• Pressure induced containment liner tearing near penetration or major discontinuities; l

e ,

1

• Containment penetration failures (electrical; mechanical; and hrge penetrations, i j such as equipment hatch, and personnel and emergency esez; , diocks) resulting ,

l from high-pressure / temperature conditions;

• Failure of containment isolation system to close normally-open containment

. penetrations; and ,

i

• Containment bypass events, i

Containment Structural Failure. The IPE team calculated the best estimate value of containment failure pressure by assuming that the containment would fail when hoop tendons underwent a strain of 1%. This assumption was based on NUREGICP-0033 i

in which it was noted that a prestressed concrete containment is structurally weakest under internal pressure loading in hoops far from penetrations. [2] The team calculated a failure pressure of 151 psig, which was 2.8 times the containment design pressure. .

Coneninment Y iner Tearine. Following the methods outlined in three reports prepared by the Electric Power Research Institute (EPRI) on calculating the onset of liner tearing and subsequent leakage, the IPE team calculated the containment liner failure pressures given in Table 4.4-2, page 4.4-16 of the submittal. These are reproduced in Table 2 of this report, where it is shown that liner tearing is most likely to occur at the personnel hatch penetration at a pressure of 141.3 psig.

I I

Arkansas Nuclear One, Unit 2 16 November 1995

l (WaEVIEWED DaAFT l

~

i

i Table 2. Estimates of ANO-2 Containment Liner Failure Pressure i

4 '

Discontinuity location Leakage Pressure P (psig) P/P%

4 Steamline Penetration 141.5 2.62 j Equipment Hatch 141.7 2.62 Wall-Skin Juncture 146.1 2.71 Springline 149.3 2.76 {

Personnel Hatch 141.3 2.62 P% = 54 psig 1

Contmirtment Penetration Failures. The IPE team reviewed ANO-2 penetration types  !

and penetration failure analyses from available repons and PRAs for the j following likely failure locations:

• Electrical penetration assemblies;

• Mechanical penetrations, such as purge valves, fuel transfer tube, and fluid l'ines;  :

i and

• Large penetrations, such as the equipment hatch, and personnel and emergency  ;

escape airlocks, I

As noted in Section 4.4.3, page 4.4-7 of the submittal, " qualitative and scoping rather than quantitative and detailed analyses were performed for these penetrations.

Conclusions were drawn via engineering judgment and on insights drawn from the Containment Performance Working Group Repon." [3]

Electrical penetration assemblies (EPAs) permit circuit conductors to pass through the containment wall while maintaining containment integrity. ANO-2 has 38 EPAs, all of which consist of three basic components: a welded flange, a header plate, and electrical feedthrough modules. Amphenol supplied all of the ANO-2 EPAs, except for the eight feedthrough modules provided by Conax. The IPE team found that "high temperature degradation and failure of the Amphenol EPA module u-cup seals cannot be ruled out during extremely severe accidents. This temperature-induced i

containment integrity failure mode requires additional investigation." (page 4.4-9)

The team concluded that the Conax modules would not be expected to fail during a severe accident. l The IPE team funher concluded that purge valves, purge lines, and their respective seals would not be likely to fail during severe accident conditions. In severe accidents involving HPME, the fuel transfer tube seals might fail as a result of l

exposure to high temperature. This seal failure would be a mechanism of late l

J November 1995 j Arkansas Nuclear One, Unit 2 17 i

t

4 d

i tWaEVIECTED Da AFT i

i containment failure. However, it was not modeled in the ANO-2 CET. De team  !

i recommends further investigation of fuel transfer tube seal failure.

De ANO-2 equipment hatch, and personnel and emergency escape airlocks all i employ seals made of an EPDM compound. He team concluded that the seals were ,

not likely to fail under severe accident conditions.

! Containment Isolation Failures. Our review of ANO-2 containment isolation failure is i described in Section 2.1.2.4.

4 Containment Bypass Events. As evaluated in the Level 1 portion of the study, the

IPE team identified two containment bypass accidents that could have an important '

bearing on ANO-2: the steam generator tube rupture event and the interfacing 2

systems LOCA. Of these, the ISLOCA was found to be the more likely bypass ,

mechanism. In characterizing bypass releases, the IPE team assumed that they would '

escape directly from the break to the environment without being mitigated by any water pool. It was found that an ISLOCA could occur in the auxiliary building penetration room and the RCS water would be ejected to this room by the CCW pipe break drains to lower levels of the building before core damage. Containment bypass contributed to 1.1 % of the total CDF.

The IPE team appem, to have adequately considered the ANO-2 containment failure modes.

21s Isolation Failure.  ?

ANO-2 containment isolation was actuated at a containment pressure of 18.4 psia at which pressure all containment penetrations closed, except those of the engineered safety features and the recirculation coolant pump cooling lines. To determine the likelihood of failure to isolate, the IPE team used an isolation system fault tree.

(Section 4.1.5, page 4.1-3) A description of the containment isolation system is provided in Section A2.1, pages A-97 through A-108 of the submittal.

De IPE team concluded that containment isolation failure was most likely to occur at the containment vent header isolation valves 2CV-2400-2 and 2CV-2401-1, which were assumed to be normayy open during power operation.

Containment isolation failure contributed to 2.3% of the total CDF, which is relatively high compared with that for similar plants (see Table 3 of this report). The reason for the high value, in part, is that Entergy def'med the isolation failure as follows:

Centainment isolation failure was modeled as the failure of any size penetration connecting the atmosphere, or a system, within containment with the atmosphere, or a system, outside of containment. A minimum Arkansas Nuclear One, Unit 2 18 November 1995 f

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UNREVIEWED DRAFT l

l opening size criterion was not imposed. His modeling not only tends to  ;

l '

conservatively overestimate the probability of containment failure but also 2

tends to lead to conservatively large fission product releases for very small j openings. [5]  !

Many IPE submittals assume no loss of containment isolation if openings are all l less than 2 inches in diameter. ,

(

Table 3. Containment Failure as a Percentage of Total CDF: I ANO-2 IPE Results Compared with the Zion NUREG-1150 PRA Results and with the Results of Other IPEs l

CDF Early Late Bypass Isolation Intact j Study (per rx yr) Failure Failure Failure Zion /NUREG-1150 6.2E-5 1.5 25 0.5 na 73 Palo Verde IPE 9.0E-5 10 14 4 0* 72 Palisade:; IPE 5.2E-5 32.5 15.1 5.5 0.4 46.3 i San Onofre IPE 3.0E-5 0 . 9.4 6.7 0.07 83.8 "

Millstone 2 IPE 3.4E-5 9.5 32.4 2.2 0.02 55.9 ANO-2 IPE 3.7E-5 9.9 13.9 1.1 2.3"* 72.8 j na . not available

• Probability is less than 0.001, conditional on core melt

" includes MCCI basemat penetration failures

• • • the submittal reports both isolation failure and early containment failure as early containment failure of 12.2%

Mi System / Human Responses.

Although the potential for human response to severe accidents (e.g., for setting the stage for accident management) was recognized in the submittal, operator recovery actions were not formally included in the back-end analysis. On page 4.6-3 of the submittal, the following statement is made, " Generally, no credit was taken for operator recovery beyond core damage." Specifically discussed were the possibilities of:

• Introducing a containinent failure mode by recovering containment sprays after the recovery of AC power during an SBO sequence. De previously inerted '

containment would become deinerted with the subsequent potential for large hydrogen burns (page 4.3-4); and

• Having operator-initiated depressurization of the RCS as part of the CET top event, "DP" (page 4.5-5).

1 The scant attention that human recovery actions receives in the ANO-2 back-end analysis is typical of IPEs for PWRs. This is in contrast to the major consideration that human action receives in many of the IPEs for boiling water reactors (BWRs). A November 1995 Arkansas Nuclear One, Unit 2 19 ,

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UNaEVIEWED DRAFT 3

4 l major reason for this is that the BWR Owners Group Emergency Operating

. Procedures (EOPs) for BWRs (Rev. 4) go well beyond the point of core damage while the EOPs for PWRs generally do not. ,

, 216 Radionuclide Release Characterization.

! The IPE team presented radionuclide releases in terms of 51 release modes, which are also the CET end states. Each release mode is defined in Table 4.7-1, page 4.7-10.

The relationship of the 51 release modes (or CET end states) to the PDSs and the i percentile contributions to the CDF from the release modes appear in Table 4.6-2 of i the submittal. De team assessed five groupings of radionuclides (NG, I, Cs, Te, and Sr) and detennined the fractions that each group constituted of the initial core inventory of the 51 release modes. (See Table 4.7-4 of the submittal.) Thus, unlike many of the IPE submittals, it is a straightforward matter to determine the releases for ,

any of the plant damage states.

De methods used to detennine the release fractions were " based on simple hand calculations of release and removal terms using insights from reference plant analysis of fission product releases." Although not unique to the ANO-2 IPE, this approach is ,

in contrast to the more standard one of using the MAAP code. Such hand calculations can be very instmetive, allowing analysts to gain an understanding of release and retention mechanisms.

The IPE team determined release and removal mechanisms from values established in ,

NUREG/CR- 4881 and NUREG/CR-4551. The release and removal mechanisms l included: (1) in-vessel release, (2) ex-vessel release, (3) delayed release (revolatilization), (4) in-vessel retention, (5) scrubbing of fission products by sprays, and scrubbing of fission products by water in flooded reactor cavity; (6) scrubbing of ,

fission products by steam generator water (the quantitative analysis assumes no credit j for this scrubbing mechanism), and (7) scrubbing of fission products by flooding the auxiliary building at the ISLOCA location (the quantitative analysis assumes no credit for this scrubbing mechanism). For any of the 51 release modes, the total release, could be determined by summing up the releases (and subtracting the eventual removal of some fission products) for four types of release: in-vessel, high-pressure 1 melt injection, core-concrete interaction, and late (iodine and cesium). l

~

The results, that is, the " Summary of the Fission Product Releases" in Table 4.7-3, appear to be reasonable and the discussion of the elements of the source term assessment is sound. However, it is difficult to audit these results because Table 4.7-2, which is supposed to provide values sufficient to calculate releases, appears to be incomplete. (For example, in calculating the CCI release for the E4 release mode, the value is not given for the FCCID (fraction of the core inventory that is released during CCI in dry cavity cases).

Arkansas Nuclear One, Unit 2 20 November 1995

, l

i tJNaEVIEWED DRAFT i

l 2.3 Accident Progression and Containment Performance Analysis 4 i l'

4 Lij, Severe Accident Prorression.

The IPE team used the following to determine the accident progression at ANO-2 (Section 4.2, page 4.2-1):

i

• Data from the evaluations of the containment response to accident phenomena of l other plants that have undergone analysis; i

!

• Scoping calculations to account for the ANO containment response to mass and '

energy additions to the containment during core damage, debris dryout, and core-concrete interactions; and i

• Conservative adiabatic calculation of pressure loads as the result of hydrogen j bums.  ;

4 By referring to the results of other plant analyses, the team made use of information on event timing, pressure rise in the containment as a result of vessel failure, and consideration of uncertainties in severe accident phenomena such as higt pressure melt ejection loads associated with vessel failure. Sources of information included the following: (Section 4.2.2, page 4.2-3)

• Iarge, dry PWR plant analyses of selected accident conditions, such as the IDCOR Task 23.1 technical reports on the Zion Integrated Containment Analyses l using MAAP; l I

• I.arge, dry and subatmospheric PWR plant analyses reported in the BMI-2104 and NUREGICR-4624 studies of Zion and Surry using the Source Term Code Package;

• NUREG/CR-4551 assessment of containment loads due to HPME at vessel failure;

• NUREG 1150 consideration of uncertain issues and EPRI parametric analysis (EPRI NP-6111 and EPRI NP-7192) on modeling parameters using MAAP; and

• Selected MAAP plant-specific calculations to confirm assumptions relative to the more severe accident conditions.

The team used scoping analyses to determine event sequence timings of the ANO-2 plant compared with the timings determined in another plant severe accident analysis.

For this purpose, the IPE team developed a Comparative Core / Containment Response '

Scoping (CCRS) model. The CCRS model simplified the conservatior, equations of mass and energy using the containment as the control volume to estimate containment response under severe accident conditions. 'Ihe CCRS integrated sources of mass and '

energy, including heat sinks, to determine containment response. 'Ihis model was Arkansas Nuclear One, Unit 2 21 November 1995 t

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

l. UNREVIEWED DRAFT used to estimate the containment response of the various core-damage sequences at

, key times to support CET quantification for the IPE back-end analysis. To be able to better review the CCRS model, however, the licensee needs to describe it in more detail and show some results obtained.

S ne IPE team developed a spreadsheet to calculate the rise in containment pressure

resulting from a global hydrogen burn. His calculation was based on the following criteria

• Hydrogen bum is possible when its concentration is above 4%;

• Hydrogen burn is not possible when oxygen concentration is below 5 %;

i l

• Hydrogen bum is not possible when the containment is steam-inerted to a steam concentration of above 55%;

i

• Hydrogen concentration limits of hydrogen bum are: diffusion / slow bum,4%;

deflagration, 8 %; and detonation,14 %;

• Total hydrogen buming is assumed if the deflagration (or detonation) limit exists in the containment; and l

• Bum efficiency is assumed to vary linearly between the diffusion and i deflagration limits of hydrogen concentration.

The licensee needs to show the results obtained from this spreadsheet in order for it to j be reviewed adequately. l The IPE team assessed the possibility of in-vessel recovery using the CET top event, l "VF-Vessel failure occurs," which was quantified using a fault tree shown in ,

Figure 4.6-7, page 4.6-32 of the submittal. Two mechanisms of in-vessel recovery l were considered: lower head cooling via ex-vessel heat removal and cooling debris ,

in-vessel. Assuming that the ANO-2 vessel could not be submerged deep enough to  !

prevent upper vessel walls from becoming excessively hot, the team took no credit for j ex-vessel cooling. In-vessel cooling requires (1) the formation of a coolable debris bed in-vessel and (2) the recovery of coolant injection. The IPE team assumed that debris is not coolable whei about 25 % of the fuel in the core exceeds the core material eutectic temperature of 4130*F. Based on the expert opinion used in the Surry NUREG-ll50 study, the team assumed a 0.9 conditional probability that a coolable debris bed would form at ANO-2. Recovery of coolant injection was assessed with a separate CET top event, " REC-Coolant not recovered in-vessel." In-vessel recovery was not considered possible for the three PDSs (IIIAi, IV Ki, and HKi) that constituted 85% of the total core damage sequences, and for which the only CET quantification results are presented in the submittal, because no coolant injection was recovered.

Arkansas Nuclear One, Unit 2 22 November 1995 t

l 1

' 1 UNaEVIEWED DaAFT l

f Figure 4.6-14, page 4.6-63 of the submittal, shows an event tne used to calculate the -l containment pressure rise at vessel failure, depending on the particular containment, j vessel, and accident conditions given, including fraction of melt ejected from the l

vessel, which is designated as "high," " medium," or " low." It is not clear from the i

submittal what fraction of melt was assumed would be ejected from the vessel at the

~

time of its failure. l J

i W Dominant Contributors: Contietancy with IPE Insiehts.

The dominant contributors to the ANO-2 containment failure profile were high RCS ,

pressure and nonSBO transients with or without loss of service water. (Other ,

contributors included IDCAs and SBO transients.) If service water is lost, these transients dominate early containment failure. If service water is available throughout  :

an accident, and active containment heat is therefore removable, these transients result l

! in no containment failures.

e

! Two other important factors contribute to early containment failures at ANO-2:

i l'

• De reactor cavity and general containment layout are conducive to HPME/DCH containment failures; the vessel has no lower head penetrations, thus conducive  !

I to vessel failures with larger openings and rapid blowdowns; and the core l

l contains relatively large amounts of zirconium with the associated potential for i large amounts of hydrogen generation; and i t

• The relatively conservative assumption that, if there is a high-pressure vessel ,

i failure into a containment with no active heat removal, the containment will fail I i 100% of the time.

This combination of sequences, facility characteristics, and conservative assumptions i

! is unique to the ANO-2 IPE submittal and needs to be kept in mind when ::omparing the ANO-2 containment failure' characteristics with those of other, similar PWRs. .

Table 3 in this TER provides such a comparison. (All of the plants compared have l CE nuclear steam supply systems (NSSSs) with the exception of Zion, which is a ,

Westinghouse PWR.) De "early failure" values for ANO-2, Millstone-2, and Palo '

Verde are very similar. His is probably fortuitous because the elements that make up this failure mode are all different.

As shown in Table 3, San Onofre stands out from the rest of the plants because the ,

IPE team there postulated no early containment failures. Like ANU-2, the San l Onofre reactor vessel does not have bottom-head penetrations. De San Onofre IPE team assumed that the vessel or reactor cooling system piping would fail as a result of i creep mpture and depressurize the vessel and would make HPME sequences unlikely. '

As Table 3 also shows, the Palisades plant has a unique design feature-a cavity tunnel through which debris relocates early into the auxiliary building-which affected early containment failures.

Arkansas Nuclear One, Unit 2 23 November 1995

f LNaEVIEwED DRAFT  !

i

i e

! Except for San Onofre (where basemat penetation was a nonfailure), ANO-2 has the l i lowest " late failure." He reason is that active containment cooling (quahfied fan coolers and sprays) is available at ANO-2 for a large percentage of the time because

! SBO is a low probability sequence rvative to the other plants.. De probability of l

_ bypass failures at ANO-2 was predicted to be relatively low, in large measure because l l the probability of SGTR as an initiator was low compared with the probability identified during other plant IPEs. Finally, comparisons were not helpful with regard to " isolation failure" (highest for ANO-2) because the calculations for this type of l' failure at ANO-2 were plant-design-specific and containment-failure-dermition-specific i (i.e., what is the containment hole size below which isolation failures are ignored?).

LJ.J Charactarhation of Contain nent Performaner, j The IPE team quantified the CET end state frequency based on an EPRI generic framework for IPE back-end analysis, which involved scoping. [4] In order to i quantify the CET top events, the team developed fault trees, the basic events of which  :

were quantified using screening values based on estimates at other, similar plants, '

plant-specific infomtation, simplified calculation models, and comparative evaluations.

The basic event probability values used were relative ones that provide insights into containment performance under postulated severe accident conditions, and are  !

" intended to rank the importance of key issues on the overall risk, not to produce  :

absolute probability values." (Section 4.6.2, page 4.6-2) The I.evel 2 basic events )

quantification process involved recognition of the following general classifications:

• System-related events that are defined by the PDS or human recovery issues (Category 1);

• Phenomena-related events that are subject to large uncertainties, which can be characterized via other plant probability estimates (Category 2); and

• Phenomena-related events that are subject to uncertainties, but influenced by plant-specific features (Category 3).

The basic events in Category I were quantified in one of several ways. Basic events that were PDS-dependent were used with PDS dapaadant flags, or were renamed as PDS-specific basic events. Generally, each of these events was set either to zero or one (or true or false) for a given PDS but, ideally, was set to an intermediate value.

Generally, no credit was taken for operator recovery actions beyond the occurrence of core damage. De only recovery action modeled was the recovery of offsite power, which is not actually an operator action. De basic events in Category 2 were quantified based mostly on the point estimates for the Surry plant as reported in NUREGICR-4551. He basic events in Category 3 were quantified by biasing reference basic event probabilities using plant-specific design considerations or Arkansas Nuclear One, Unit 2 24 November 1995 t

l UNaEVIEWED DaAFT ,

i qualitative assessments of plant specific probabilities. The quantitative probability j ranges, adopted from the Surry plant PRA as reported in NUREGICR-4551, are given j in Table 4.6-1, page 4.6-20 of the submittal, which is reproduced as Table 4 of this report.

Table 4. ANO-2 CET Quantification Probability Ranges Description Probability Range Point Estunate  !

Certain P=1 1 i

Highly Likely 1 > P > = 0.995 0.9925  !

Very Likely 0.995 > P > = 0.9725 O.95 ,

l Likely 0.95 > P > = 0.70 0.825 Indeterminate 0.70 > P > = 0.30 0.5 Unlikely 0.30 > P > = 0.05 0.175 Very Unlikely 0.05 > P > = 0.0275 0.005 Highly Unlikely 0.005 > P > 0 0.0025 Impossible P=0 0 i Figures 4.6-16 through 4.6-18, pages 4.6-65 thmugh 4.6-67 of the submittal, show quantified CETs for the three dominant PDSs, i.e., IIIAi, IVKi, and IIKi, ,

respectively. Figure 4 in this report is a reproduction of Figure 4.6-16 in the submittal.  ;

The IPE team appears to have adequately chameterized the ANO-2 containment performance using CETs.

11.4 Impact on Equipment Behavior.

The original submittal [1] did not include an evaluation of the effects of the severe '

accident environmental conditions on the performance of key containment safety system equipment neededto mitigate severe accidents (e.g., containment fan coolers and containment sprays). 'In response to the staff's RAI, Entergy stated the following: )

Containment conditions and equipment vulnerability given harsh conditions was evaluated in NUREG 4551, and the same basis applied to the ANO-2  :

study. Spray survivability was divided into two categories: sprays survive given containment fails, and sprays survive given no conainment failure. It was assumed that the containment sprays at ANO-2 would be as robust as  ;

the sprays at Surry, so the same failure probability used for sprays after containment failure was used here (0.1). Whether the sprays continue to ,

l operate while the containment is intact was based on a 24 hr mission time.

f Arkansas Nuclear One, Unit 2 25 November 1995 l t

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UNREVIEWED DRAFT j Since no information is available on equipment survivability given the conditions that  ;

i are to be expected in the containment, it was conservatively assumed that the failure  :

rate would be an order of magnitude higher than the nominal failure rate (2.03E-2).

The same probability was assumed for fan cooler failures.

LL1 Uncereminty and Sentitivity Analyses. .

The IPE team performed sensitivity analyses conceming the uncertainties associated l with nine issues of accident progression (See Sections 4.8.1 through 4.8.9 of the

submittal for details and Table 4.8 2 and Figure 4.8-1, pages 4.8-10 and 4.8-11, i respectively, for summaries of the results.) The nine issues are summarized as l follows:

L

• In-vauel coolability G==n* 11. The baseline value used for the conditional probability of an uncoolable debris geometry forming in-vessel was 0.1. By I 1

changing this probability value to 0 and 1, the early containment failure I

i frequency became 1.07 and 0.99, respectively, of the baseline value. The team

! considered this low sensitivity an artifact of the relatively low assumed i probability for successful depressurization or successful coolant injection I recovery following a core damage event.

l l

$

• Induced mpture of the primarv system assue 2). In the baseline analysis, the l team assumed a 0.72 conditional probability of hot leg failure in cases where the l RCS pressure remained at SRV set point; in other cases, the team assumed values that reflected the unlikelihood of hot leg failure occurring. Assuming that jl the hot legs would remain intact increased the nominal early containment failure 2 (ECF) frequency by 1%, from 4.48E-6 to 4.54E-6 per reactor year. The same assumption decreased the total containment failure (TCF) frequency by 31 %,

from 9.69E-6 to 6.64E-6 per reactor year. An increase in the ECF resulted

[ from an increase in the probability of HPME failures. The ECF was relatively l

' insensitive to the hot legs remaining intact because "IVKi," the dominant PDS

! contributor to ECF, involved very high containment pressure before vessel failure. This guaranteed containment failure at vessel failure, independent of depressurization of the primary system. The decrease in TCF was possible l because, for the sequences that survived HPME containment failures, less debris j remained in the cavity and the debris was more likely to be coolable.

1 1he assumption that the hot legs would fail decreased the nominal ECF l

frequency by 7%, from 4.48E-6 to 4.18E-6 per reactor year, and increased TCF l

i by 139%, from 9.69E-6 to 2.32E-5 per reactor year. The decrease in ECF 4 resulted from a decrease in HPME failures. The TCP frequency was highly i sensitive to hot leg failure because an increase in the amount of core debris remaining in the cavity would render the debris uncoolable, thus causing i

4 containment failures from CC1s.

a Arkansas Nuclear One, Unit 2 27 November 1995

l

! 1 o .- -

b i- UNREVIEwED DRAFT

• DCH. HPME. and early hydrogen bum loads assue 31. 'Ihe IPE team assessed l

in combined form the impact of DCH, HPME, and early hydrogen bum on early

' containment failure because the effects were difficult to assess individually.

Containment pressure rise at vessel failure was compared with the assumed i containment fragility curve to determine the probability of ECF. In the baseline 4 analysis it was assumed that 90% of the vessel failures would create a "large hole" in the vessel, causing a high containment pressure rise at vessel failure. ,

The other sequences were assumed to cause a "small hole" at vessel failure. l

'Ihe baseline calculation assumed that, for an HPME event,25 % of the initial debris would participate in DCH, and 100% of that participating debris would

' i transfer its energy to the atmosphere. l l

- In the baseline calculation a variable probability was used based on ECF l

resulting from DCH, HPME, and early hydrogen bum, which would range between 0.0001 and 0.415, depending on the PDS. j i

e Hydronen lenition assue 4). Issue 4 addressed only the sensitivity of hydrogen l burn on late containment failure (LCF) because that on ECF was addressed in l Issue 3. The overall effect of hydrogen bum on late containment failure was found to be negligible. A main reason for this finding is the conservative modeling assumption that, even if hydrogen bum did not fail the containment, it would most likely fail in the long term owing to one or more of the following: l steam generation, pressurization from noncondensable gas generation, or basemat I melt-through. The only mechanism that the team identified to prevent long-term l containment failure was to operate containment sprays and fan coolers. It was not expected that a small hydrogen bum would threaten the integrity of the containment when containment cooling was available.

• Conininment failure nuessure Gssue 5). Using 5% and 95 % confidence limits of containment failure pressure for sensitivity analysis, the IPE team found that the total containment failure frequency was not sensitive to containment failure pressure (only a 3% change was observed).

• Hieh-temccrature effects on r.Aueine failure or*==nte assue 6). In the baseline calculation, a 0.0275. conditional probability was assumed that the dry containment would fail because of a drop in containment pressure capacity caused by high temperatuit. Assuming that the containment temperature under sevent accident conditions would always be high enough to fail the containment, the LCF frequency would increase by 140%. Excluding the high temperature as a contributor to containment failure would decrease the IfF frequency by 4%.

November 1995 Arkansas Nuclear One, Unit 2 28 f

1 1

J UNREVIEWED DRAFT

• Ex-vessel debris coolability Gssue 7). Assuming that the debris was coolable if I water was available, the LCF frequency decreased by 18% of the baseline value.

i Assuming that the debris bed was not coolable, the LCF frequency increased j by 39%.

• Cavity check valve failure effects on CCI assue Bh The IPE team examined the effect of the cavity check valve, "2BS-46," on the occurrence of basemat '

penetration resulting from CCI in the cavity. Assuming that the check valve was i removed, the probability of basemat penetration decreased by 5 % of the nominal value of 1.57E-6 per reactor year. Assuming that the check valve never failed,

! thus always preventing water from entering the cavity, the probability of basemat

penetration increased by 160%.

!

• Containment isolatinn (T==* 9). Section 4.8.9, page 4.8-8, notes that "If containment isolation failures are set to zero, the early containment failure frequency decreases by 19%." This conclusion was a direct observation from

the back-end results, and it is not clear what the analysts did or what they learned from this sensitivity study.

. 2.4 Reducing Probability of Core Damage or Fission Product Release ,

L4J. Definition of Vulnerability.

In order to identify potential containment response concems, the IPE team adopted l-i back-end screening criteria from NUMARC 91-04, which are listed in Table 3.7-4, >

j page 3.7-25 of the submittal; this table is reproduced as Table 5 in this repon. The l

team compared the IPE results with these criteria and identified no back-end vulnerabilities for the ANO-2 plant. This finding was based on the following (Section 4.9.7, page 4.9-6):

1 j

• The largest CET end state group involving containment failure had a fiequency

) of 4,8E-6 per reactor year, which was less than the 1.0E-5 per reactor year i

criterion '

• The dominant large release CET end state represented 3.8% of the total CDF, i which was less than the 10% criterion; and ,

• The largest CET end state group involving containment failure represented i 13.0% of the total CDF, which was less than 20% of the total CDF.

1 The ANO-2 IPE team postulated large (early, high) releases to occur at a frequency of 1.4E-6 per teactor year. Ihis is higher than IE-6 per reactor year, which was the

screening criterion used in several IPEs.

1  :

Although members of the IPE team found no ANO-2 back-end vulnerabilities to exist, l they recommended that several issues relating to ANO-2 containment perfonnance be j

investigated funher. These include the following:

Arkansas Nuclear One, Unit 2 November 1995 f 29 1 ,

t

4 UNREVIEWED DRAFT I I

i i~ ,

1 Table 3.7 4 '

ANO 2 IPE Containment Response Screening Review Mean Core Damage Frequency (CDF)

%:ennal AcnonsTo Be Taken i

With ComainmentFailm/ Bypass Per CET Endstate Group fDer nactor WET) Find a coat effective plant  ;

1.

Greater than IE-5 administradvs. procedural or i

) hardware modification with emphasis l or on eliminating or reducing the  ;

likelihood of the source of the accident 7 l arge nicase greater than 1% of total CDF sequenceinidator.

or

2. If unable to satisfy above esponse, find cost effective treatmentin EOPs or any nicase greater than 2% of total CDF other plant procedure with emphasis on prevention of core damage.

3.If unable to sadsfy above responses, ,

ensure SAMG isin place with emphasis J on prevention / mitigation of core damage or vesselfailure, and  ;

contamment failure. i 1E 5 to IE 6

1. Find a cost effective neatment in EOPs or other plant procedure or gimg hardware change with emphasis on or prevention of core damage.

large nicase 3% to 1% or total CDF 2. If unable to satisfy above response, s ensure SAMG isin place with emphasi or on prevention / mitigation of core damage or vesselfailure,and any release 5% to 2% of total CDF containment failure.

Ensure SAMG isin place with 1E 6 to IE 7 emphasis on prevention / mitigation of core damage or vessel failure,and containment failure.

l No specific action recuired.

Less than !E 7 l

l November 1995 Arkansas Nuclear One, Unit 2 30 j

tWREVIERED DRAFT i

4 2

i

• Potential for a high temperature-induced failure of the o-ring seals of the fuel transfer tube, which could result in late containment failures;

\

• High-tempenture degradation and failure of the Amphenol electrical penetration l module u-cup seals during severe accidents; j 4

)

• Containment isolation failures resulting from the failures of containment vent

• beader isolation valves, 2CV-2400-2 and 2CV-2401-1;

• Whether removal of the internals of the cavity check valve,2BS-46, would increaw the likelihood of water entering the cavity area; and 1

• Uncertainty of HPME and DCH effects on containment pressurization.

9 i

112 Plant Improvements.  ;

In Section 6.2 of the IPE submittal, nine potential plant improvements are discussed, l l j

five of them involving procedures and four hardware. Most of the improvements would be front-end in nature, however, two relate directly to back-end issues, one

{

involving procedure and one hardware, Among the improvements proposed was that l the fuel-transfer tube be flooded in order to protect the tube seals from the high-l temperature failure that might result from HPME events. More recent analysis l

i shows that this flooding is not necessary. [5]

1 Another modification, which the submittal, states would reduce both CDF and early

[ containment failure, would be to improve the " loss of service waur procedure."

Service water is important for maintaining the containment cooling function, both I

from sprays and fan coolers.

The team further proposes that the internals be removed from the reactor vessel cavity check valve 2BS-46. *ntis would allow the reactor cavity to flood, with the associated benefits. As the submittal notes, one concern that such flooding raises is the enhanced potential for pressurized thermal shock.

Currently, Entergy is evaluating all of the potential improvements described in this section of the submittal. A summary of the status of the improvements is provided in the Entergy response to the staff's RAI. [5]

2.5 Responses to CPI Program Recommendations One of the Containment Performance Improvement (CPI) Program recommendations that pertains to PWRs with large, dry containments is that utilities evaluate their l

containment and equipment vulnerabilities to hydrogen combustion (local and global) l as part of their IPEs and that they identify the need for improvements in PWR procedures and equipment.

November 1995 Arkansu Nuclear One, Unit 2 31 t

LINREVIEQED DRAFT The CPI issue of global and local hydrogen burns and detonations was not in the original submittal. [1] However, Entergy did perform parametric analyses of the effect of hydrogen burns on early and late containment failures. The original submittal notes that the CPI item "related to localized hydrogen accumulation which was identified in Supplement 3 to GL 88-20 was evaluated and found not to be a significant contributor to failure for the ANO-2 containment design." (Section 1.4.2, page 1). More detailed information on local effects was provided by Entergy in their response to the staff's RAI. [5] The assessment is responsive to the CPI recommendations.

T 2.6 IPE Insights, Impreements and Come Jtments Back-end insights gained from this IPE were derived hom considering those PDSs that led to large releases of radionuclides:

• For the PDSs that did not bypass the containment, loss of service water was a significant contributor to both core damage and containment failure resulting f from containment spray and fan-cooler failures;

• By stopping one train of containment sprays, or even one train of the HPSI, the time to core damage could be extended, thus increasing the potential for recovery; l

• Containment venting before core damage could reduce the probability of early failure;

• Manual depressurization of the vessel could also reduce the probability of early failure;

• The most important " containment isolation failure" PDS is dominated by the ,

l E

failure of the containment vent header isolation valves. Recovery after such a failure could be achieved by closing a manual isolation valve; and

• In SBO events, recovery of AC power, and thereby of sprays, could cause a f deinerting of the containment, hydrogen burn, and containment failure.

Entergy discussed potential plant improvements in two categories: procedures and hardware. Most were front-end improvements. However, procedures were suggested for:

• Coping with failed or degraded service water; and

• Flooding the fuel transfer tube to protect the seal from high-temperature-induced failure caused by HPhG events.

The only back-end hardware fix considered was the removal of the reactor vessel cavity check valve intemals, so that more water could flood the reactor cavity.

November 1995 Arkansas Nuclear One, Unit 2 32

___J

UNREVIEWED DRAFT

3. CONTRACTOR OBSERVATIONS AND CONCLUSIONS Although the Entergy IPE team used a scoping approach in the back-end analysis of ANO-2, the back-end portion of the IPE appears to meet the intent of GL-88-20. In general, the qualitative portions of the assessment show an understanding of accident progression, phenomenology, conservatisms, containment response, and radiological source terms. However, no accident progression analyses were performed (for example, using MAAP). Instead Entergy used the results of other, similar plant analyses, combined with separate-effect, scoping calculations. For example, the pressure rise from hydrogen bum was calculated separately and compared with NUREG-ll50 Surry calculations. Radiological source terms were assessed in a similar manner. The individual calculations were input into the quantification of the containment event tree.

The containment performance improvement issue regarding global and local hydrogen bums and detonations was not addressed as such. However, the IPE team did perform parametric analyses of the effect of hydmgen bums on early and late containment failures. Although Entergy states that it analyzed local pocketing effixts, we were unable to find the results of such analysis in the submittal. Entergy did satisfactorily address these effects in response to the staff's RAI. [5]

Arkansas Nuclear One, Unit 2 33 November 1995 i

UNREVIEWED DRArr

4. REFERENCES

1. Entergy Operations, Inc. "ANO Unit 2 PRA Summary Report for Severe Accident Vulnerabilities." August 1992.

2. Shunmugavel P. and T. E. Anderson. " Internal Pressum Capacity of Prestmssed Concrete Containments for Nuclear Power Plants." Proceedings of the Workshop on Containment Integrity, SAND 82-1659, NUREG/CP-0033.

Albuquerque. October 1982.

3. U.S. Nuclear Regulatory Commission. " Containment Perfonnance Working Group Report," NUREG-1037. May 1985.

4. Z. T. Mendoza, et al. " Generic Framework for IPE Back-End (Ievel-2)

Analysis," Nuclear Safety Analysis Center for Electric Power Research Institute, NSAC/159. October 1991,.

5. Entergy Operations, Inc. " Individual Plant Examination Request for Additional Information," ANO Unit 2. October 5,1995.

l November 1995 Arkansas Nuclear One, Unit 2 34 f

UNREVIEWED DRAFT 7 l

APPENDIX IPE EVALUATION AND DATA SUhD,fARY SHEET l 1

i i

PWR Back-End Facts Plant Name Arkansas Nuclear One Unit 2 I

Containment Type 1 i

IJtrge, dry i

Unique Containment Features  !

A small cavity that does not mitigate debris dispersal during high-pressure melt l l

ejection and thus direct containment loads; l a

' A 10-inch swing check valve, 2BS-46, attached to the door of the cavity access

manway, which prevents water from entering the cavity from the containment floor unless the valve is failed open; A fuel transfer tube in which the o-ring seals may undergo a high-temperature-induced failure that could result in late containment failures; and Amphenol electrical penetration module u-cup seals that may degrade under j high temperature during severe accider.ts.

Unique Vessel Features Vessel does not have bottom head penetrations Important Insights, Including Unique Safety Features I Listed under Unique Containment Features i Number of Key Plant Damage States 17 Ultimate Containment Failure Pressure 141 psig j

1 November 1995 Arkansas Nuclear One, Unit 2 A-1 i ,

4 l Additional Radionuclide Transport and Retention Structures 1~

1 None credited i

Conditional Probability that the Containment Is Not Isolated  !

! 0.023 I l Implemented Plant Improvements l Entergy discussed potential plant improvements in two categories: procedures and j

. hardware. Most were front end improvements. However, procedures wele suggested for:

. I o Coping with failed or degraded service water; and  !

l e Flooding the fuel transfer tube to protect the seal from l high-temperature-induced failure caused by HPME events. In response to l the staff RAI, they stated that further investigation revealed that the fuel

{

transfer-tube seal failure is no longer risk-significant.  !

Entergy provided a list of procedural and hardware improvements including their status

]

in their response to the staff's RAI. [5] The only back end hardware fix currently "under ,

i evaluation" is removal of the reactor vessel cavity check valve internals so that more water can flood the reactor cavity.

l C-Matrix l i PDS CDF* (per Early Late Bypass intact rx yr) f J

IAi 9. " "-07 1 IlGi L.wi-08 1 IIKi i

1.21E 06 1 l

IIIAi 2.32E-05 0.006 0.138 0.856 j IllCu 7.88E-07 1 IIIEi 2.20E-07 1 IllFi 1.02E-06 0.015 0.985 IIIGi 1.22E47 1 IIIKi 1.23E-07 . 1 IVKi 1.89E-06 1 VAi 3.45E 07 1 VIGi 1.47E47 1 3.27E-07 1 CBISLOCA CBSGTR 9.53E48 1 SBOi 3.89E47 0.378 0.622 SBOu 4.31E-08 1 l

• The total CDF of the PDSs given in the table is 3.10E-5 per reactor year, which is 84% of the total CDF postulated for ANO-2 plant,3.71E-5 per reactor year.

A-2 November 1995 Arkansas Nuclear One. Unit 2 l

T I

Additional Radionuclide Transport and Retention Structures i

None credited Conditional Probability that the Containment Is Not Isolated 0.023 Implemented Plant Improvements Entergy discussed potential plant improvements in two categories: procedures and -

hardware. Most were front end improvements. However, procedures were suggested -

for:

e Coping with failed or degraded service water; and e Flooding the fuel transfer tube to protect the seal from high-temperature-induced failure caused by HPME events. In response to the staff RAI, they stated that further investigation revealed that the fuel transfer-tube seal failure is no longer risk-significant.

Entergy provided a list of procedural and hardware improvements including their status  ;

in their response to the staff's RAI. [5] The only back end hardware fix currently "under  ;

evaluation

• is removal of the reactor vessel cavity check valve internals so that more j water can flood the reactor cavity.

C-Matrix PDS CDF* (per Early Late Bypass intact rx yr)

IAi 9.65E-07 1 IIGi 3.40E-08 1 IIKi 1.21E 06 1 l IllAi 2.32E-05 0.006 0.138 0.856 IIICu 7.88E-07 1 IIIEi 2.20E47 1 IllFi 1.02E46 0.015 0.985 IIIGi 1.22E-07 1 IIIKi 1.23E-07 . 1 IVKi 1.89E-06 1 VAi 3.45E 07 1 VIGi 1.47E-07 1 i

1 CBISLOCA 3.27E-07 1

CBSGTR 9.53E 08 SBOi 3.89E-07 0.378 0.622 SBOu 4.31E-08 1

• The total CDF of the PDSs given in the table is 3.10E 5 per reactor year, which is 84% of the total CDF postulated for ANO-2 plant,3.71E-5 per reactor year.

A-2 November 1995 ]

Arkansas Nuclear One. Unit 2 .