ML17158B947

Summary of 961217 Meeting W/Licensee to Discuss Containment Issues & IPE for SSES

ML17158B947
Person / Time
Site:	Susquehanna
Issue date:	02/12/1997
From:	Poslusny C NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9702190025
Download: ML17158B947 (56)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&4001 February 12, 1997 5o-Kv i'3'K LICENSEE: Pennsylvania Power and Light Company (PP&L)

FACILITY: Susquehanna Steam Electric Station (SSES),

Units 1 and 2

SUBJECT:

MEETING WITH PP&L STAFF REGARDING SSES, UNITS 1

AND 2, SELECTED CONTAINMENT ISSUES AND THE INDIVIDUALPLANT EXAMINATION (IPE)

On December 17,

1996, PP8L and NRC staff met to discuss containment issues and the IPE for SSES.

The first session on containment issues included those listed in Enclosure 1 and the session on the IPE included those listed in Enclosure 2.

CONTAINMENT ISSUES The second item dealt with the design of the standby gas treatment system (SGTS) and its ability to deal with a single failure.

PP8L has been studying potential single failures that could affect the SGTS operability and effectiveness and most recently completed a modification to a controller circuit affecting outside air damper operation.

However, a number of failure modes (8 or 9 being the most significant) are believed to be important enough to be considered for additional potential design changes.

These failure modes had not been considered in the original design basis which had been approved by the staff during the licensing review.

PP&L indicated as noted in Enclosure 4 that it would be considering potential plant modifications, procedural

changes, Final Safety Analysis Report (FSAR) changes, and even new offsite dose analyses.

The staff indicated that it would entertain a

submittal that would discuss the details of PP&L's findings and proposals relative to the newly identified failure modes-for the 'SGTS; IPE During this session, PP&L staff provided a brief overview of the development of its IPE and how it has been used in the design, maintenance, and operation of the SSES units (Enclosure 5).

The remainder of the meeting was dedicated to a discussion of review comments on the SSES IPE by the Office of Research (RES).

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97OZXS0025 9702am PDR "ADQCK 05000387 PDR PP8L discussed the status of two items that were being tracked by the SSES Condition Reporting system.

.The first dealt with the design basis. secondary containment bypass leakage (SCBL) for'the units.

Based on the identification of additional bypass

pathways, the licensee has modified the original leakage rate of 5 standard cubic feet per hour (SCFH) to a higher rate of 9 SCFH under the provisions of 10 CFR 50.59.

As indicated in Enclosure 3,

PP&L has identified a number of additional pathways and plans to increase the SCBL to 28 SCFH with a minimal increase to the offsite dose value.

PP8L staff stated that this change would be submitted to the staff for review and approval.

The staff agreed that this would be a reasonable approach and that review by the Containment Systems and Radiation Protection staff would be necessary for Commission approval.

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As noted in Enclosure 6,

RES staff has some concerns about the IPE submittal in that some clarification is required and more significantly, that additional work is needed in the IPE to account for the contribution of human error and common cause failures in the specific modeling.

PP&L agreed to consider the staff comments summarized in the enclosure and committed to providing submittals to address and respond to the concerns.

It was also suggested that follow-up conference calls or a meeting would be held as necessary to ensure mutual understanding of the PP&L product and NRC requirements is maintained.

/s/

Chester

Poslusny, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos.

50-387/388

Enclosures:

1.

Meeting Attendees List for December 17,

1996, Containment Issues Session 2.

Meeting Attendees, List for December 17,

1996, IPE Session 3.

Containment Issues Slides (PP&L) 4.

Standby Gas Treatment System Design Issues Slides (PP&L) 5.

Susquehanna IPE Slides (PP&L),

6.

Areas of Concern for Susquehanna IPE (NRC) cc w/encls:

See next page DISTRIBUTION:

HARD COPY *w/Enclosures 1 and 2 "

• • w'/Enclosure's 1 - 6

• • Docket File

• • PUBLIC

• • PDI-2 Reading

• OGC

• ACRS

• • WPasciak, RGN-I

• • CPoslusny E-MAIL w/Enclosures 1 and 2

FHiraglia/AThadani (A) (FJM/ACT)

RZimmerman (RPZ)

SVarga (SAV)

JZwolinski (JAZ)

JStolz (JFS)

MO'Brien (HBO)

EJordan JKR HDrouin JSchiffgens WDean (WHD)

MCunningham (HAC3)

ELois (EXL1)

JLan'e (JCL1)

JKudrick (JAK1)

HSnodderly (HRS1)

WLon WOL

• Previously Concurred OFFICE PDI-2 P

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RES:DST*

D -2 PD NAME CPoslusny:cdw HO ie JKudrick ELois JS o.

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As noted in Enclosure 6,

RES staff has some concerns about the IPE submittal in that some clarification is required and more significantly, that additional work is needed in the IPE to account for the contribution of human error and common cause failures in the specific modeling.

PP&L agreed to consider the staff comments summarized in the enclosure and committed to providing submittals to address and respond to the concerns.

It was also suggested that follow-up conference calls or a meeting would be held as necessary to ensure mutual understanding of the PP&L product and NRC requirements is maintained.

Docket Nos. 50-387/388 i~(,. D;.,/

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t I a wv4 Chester

Poslusny, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.

Meeting Attendees List for December 17,

1996, Containment Issues Session 2.

Meeting Attendees List for December 17," 1996, IPE Session 3.

Containment Issues Slides (PP&L) 4.

Standby Gas Treatment System Design Issues Slides (PP&L) 5.

Susquehanna IPE Slides (PP&L) 6.

Areas of Concern for Susquehanna IPE (NRC) cc w/encls:

See next page

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Mr. Robert G.

Byram Pennsylvania Power 8 Light Company Susquehanna Steam Electric Station, Units 1 5 2 CC:

Jay Silberg, Esq.

Shaw, Pittman, Potts 5 Trowbridge 2300 N Street N.W.

Washington.

D.C.

20037 Bryan A. Snapp, Esq.

Assistant Corporate Counsel Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown. Pennsylvania 18101 Mr. J.

M. Kenny Licensing Group Super visor Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. K. Jenison Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.

Box 35 Berwick, Pennsylvania 18603-0035 Mr. William P. Dornsife, Director Bureau of Radiation Protection Pennsylvania Department of Environmental Resources P. 0.

Box 8469 Harrisburg, Pennsylvania 17105-8469 Mr. Jesse C. Tilton, III Allegheny Elec. Cooperative.

Inc.

212 Locust Street P.O.

Box 1266 Harrisbur g, Pennsylvania 17108-1266 Mr. Robert G.

Byram Senior Vice President-Nuclear Pennsylvania Power 5 Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101.

Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406

. Mr. George Kuczynski Plant Manager Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick. Pennsylvania 18603 Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1.

Box 1797 Berwick. Pennsylvania 18603 George T. Jones Vice President-Nuclear Operations Pennsylvania Power and Light Company 2 North Ninth Street Allentown. Pennsylvania 18101 Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803 Chairman Board of Supervisors 738 East Third Street

Berwick, PA 18603

Containment Issues Meeting December 17, 1996 Meeting Attendees game C. Poslusny J. Kudrick M. Snodderly W. Long J.

Kenny A. Roscioli Or anization NRC NRC NRC NRC PP&L PP&L Enclosure 1

PP&L IPE Meeting December 17, 1996 Meeting Attendees game Mark Cunningham Mary Drouin C.

C. Lin John Lehner John Wreathall Jim Kenny Glenn Miller Greg Butler Michael W. Simpson Tony Roscioli Casimir Kykielks Erasmia Lois Jordan Musichi Chet Poslusny John C.

Lane Or anization USNRC/RES/DST/PRAB USNRC/RES/DST/PRAB BNL BNL TWWG PP&L PP&L PP&L PP&L PP&L PP&L NRC BNL NRC NRC/RES Enclosure 2

Enclosure 3

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~ SSES DesignlBackground

~ Issues

~ Strategy For Closure

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I j4 Reactor essel To Reacto Vessel F011A MO Fee dwater System F011A MO F010A Primary Containment Reactor Bldg F0108

<<I(IIIlIIII.:I~>l; Primary Containment Reactor Bldg 107A 1078 MO MO RCIC F082A F039A F0398 RWCU Hpc)

F0828 MO 149F013 155F006 F032A F0328 "ggI I(f~cfg+'1Jj Reactor Bldg k>s~ aa '

Turbine Bldg Reactor Bldg Turbine Bldg Keepfill C RFP 8 RFP I

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ri ina na sis ssum tions

~ 5 SCEH Bypass Leakage

~ Eeedwater Water Seal

~ 5 GPM Water Leakage

~ X/Q Differences Between ESAR and NRC

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Current Containment Leak Paths 8 Limits Secondary Containment 1.0 La (Analysis)

SGTS Part of 3.3 gpm Part of 3.3 8 5.0 gpm SLC RWCURpv Air SCBL 0.6 La (LLRT) 0.75 La (ILRT)

MSIV MSL 9 scfh (air) total 300 scfh total to condenser System Boundary Leakage 5.0 gpm (water)

ESF Systems Drywell Suppression Pool Part of CRD 50 gpm Valve Hydrotest Water Seal 3.3 gpm (water)

Part of

'P Cleanu~3 3 gpm

~ Feedwater Leakage Is Now Considered SCBL

~ With New Secondary Containment Bypass Leakage Pathways and Primary Containment Water Leakage Pathways Identified, Difficult to Meet Leakage Assumptions in Offsite Dose Analysis

~ SGTS/Recirc System Single Failures

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~ Increase SCBL to 9 SCFH Using Scrubbing

~ Increase SCBL to 28 SCFH

~ Increase Water Leakage to 20 GPM

~ Rely on Addition of Water to Supp. Pool per EOPs

~ Submittal Required

Final Containment Leak Paths 8 Limits Secondary Containment 1.0 La (Analysis)

SGTS SCBL 28 scfh (air) total SLC RWCU Rpy Air 0.6 La (LLRT) 0.75 La (ILRT)

MSIV'SL HPCI/RCIC Stm Ln Drain 300-scfh total to condense System Boundary Leakage 20 gpm (water)

ESF Systems Drywell Suppression Pool CRD Part of 20 gpm water leakage SP Cleanu

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~ SSES Design

~ Issues

~ Possible Solutions

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~ Common SGTS and Rx Bldg Recirc

~ Reg Guide 1.52 Compliance Discussed in ESAR

~ Designed With Redundant Components to Assure System Operation During DBA

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RECIRC. AND SGTS SIMPLIFIED DIAGRAM Zone l Zone ill sEcoNDARY coNTAINMENT Rec/regulation ISOLATION DAMPERS Plenum RECIRC FANS Zone ill Zone.ll Zone ll Zone l Vent Stack SGTS FILTER TRAIN

.Outside Air Cooling Air Damper Recirculation Fan Dampers SECONDARY CONTAINMENT ISOLATION DAMPERS SGTS EXHAUST FAN SOTS M

PDDC trolled PDD C ntrolled SGTS FILTER TRAIN Outside Air Makeup Dampers M

Outside Air SGTS EXHAUST FAN Cooling Air Damper

lg~~t SSUCS

~ System Alignment After Test

~ Failure Modes Not Previously Considered in FSAR FMEA

~ New Worst Case Single Failure Possible

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~ Plant Modifications

~ Procedure Changes

~ FSAR Change Explicit Discussion of Failures Considered

~ New Offsite Dose Analysis

, New NRCNQs Revised. Source Term

us ue anna PP&L-Susquehanna 1 &2 Enclosure 5

verview

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Background

PPkL Approach SSES IPE Volume 6 Safety Improvements NRC Issues PPEcL - Susquehanna 1 &2

~

11/88

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10/89 PPXL Response to GL 88-20 Identifies "Other Systematic Examination Method"

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1987-1990 PPAL Meets with NRC to Discuss PP8zL Approach and Modeling NRC Issues GL 88-20

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12/91

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1/93 NRC Responds to PPXL IPE Plan PPkL IPE Submitted to NRC IPE Volume 6 Submitted PPEcL - Susquehanna

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~ Realistic Probabilities Assigned

~ Defense in Depth

~ Identify Plant and Procedure Vulnerabilities

~ ModifyPlant/Procedures/Training Programs to Increase Plant Safety PP&L - Susquehanna 1 &; 2

~ Identifies NRC Concerns in 4 Areas

~ Provides Analyses/Sensitivity Studies to Address the Issues PAL - Susquehanna 1 4, 2

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8 Plant Modifications to Reduce Equipment Vulnerabilities

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8 Procedural Changes to Improve Operator Response and to Reduce the Probability for Operator Error

~ Extensive Operator Training on EOP Scenarios

~ Maintenance Configurations Evaluated with Respect to Impact on Risk PP&L - Susquehanna I 4 2

NR Issues PP&L willUse NRC Comments to Improve Our Understanding ofRisk and to Assure Uncertainties have been Identified

~

PP&L willAddress AllNRC Comments PP&L-Susquehanna I &2

AREAS OF CONCERN FOR THE SUSQUEHANNA INDIVIDUALPLANT EXAMINATION DISCUSSION WITH PENNSYLVANIAPOWER 4 LIGHT CO December 17, 1996 Erasmia Lois John Lane Probabilistic Risk Assessment Branch Office of Nuclear Regulatory Research fnclosure 6

As BuildAs Operated

~

Itis not clear to what extend plant Mtalkdowns ~ere performed.

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Itis not clear ifunit interactions were examined.

Itis not clear ifimprovements credited in the analysis have been implemented.

I Three hardware improvements "not yet known to be implemented" that have an important effect on the CDF are the following:

ii)

, iii).

Addition of the wetwell vent to be actuated in non-ATWS events, when there is no core damage ADS SRV control during isolation events LOCA load shed and high drywell pressure isolation.

It is not clear if all procedural improvements discussed in the IPE have been implemented.

Itis not clear what is the'impact on the CDFPom improvements credited and not implemented.

Itis not clear ifcertain improvements mere modeled (e.g., raising the suppression pool level to extend time to HCTL).

Human Reliability analysis Pre-initiator human erro s The objective of Generic Letter 88-20 was for the licensees to examine plant history and practices for identifying vulnerabilities and q quantifying events that. have occurred or have the potential to occur.

Pre-initiators have the potential to impact the results of an IPE in terms of:

determining the most important contributors to core damage developing a more clear understanding of the contribution of human error to plant risk identifying areas of improvement.

For example Pre-initiators can lead to common-cause failures of important equipment; such failures will not be identified without a careful examination of plant practices.

~

The SSES IPE assumed that equipment failures include the contribution of human errors performed during normal operotions.

It was not demonstrated that a systematic examination ofplant procedures and pr.actices was performed.

Broadly, our review examines ifthe licensee s IPK process included the following:

Maintenance, test and calibration procedures for the systems and components were reviewed by the systems analyst.

Discussions were held with appropriate plant personnel (e.g., maintenance, training, operations) on the interpretation and implementation of the plant's test, maintenance and calibration procedures to identify and understand the specific actions and the specific components manipulated when performing the maintenance, test or calibration task.

A lower adjustment of generic probabilities (to represent the specific plant) was appropriately justified by examination'of procedures, interviews with training, operations and various crews, physical observations of components, walkthroughs of procedures, and evaluation of administrative controls such as tagging or independent written verification.

Any applied recovery factors were appropriately justified; that is, the recovery action.

The effects of dependencies on pre-initiator human events were appropriately addressed by considering the following:

Plant conditions (e.g., poor lighting)

Human engineering (e.g., labels, accessibility etc.)

Performance by same crew, same tin e Adequacy of training Adequacy of procedures Interviews with training, operations

'nd various crews C

We were not able to verify that the SSES IPE team addressed these issues.

Post-initiator human errors

~

Limitedpost-initiator event analysis A limited post-initiator event analysis (i.e.

analysis of errors associated with human actions performed in response to the EOPs and those performed to 'recover a specific failure or fault) may have a significant impact of both the CDF and the most important sequences.

Several types of errors may occur in performing these actions:

failure to recognize the ne ed for an action failure to select the appropriate procedure failure to appropriately execute a procedure failure to appropriately perform the action

'failure to appropriately recover failed actions or equipment.

Success of an action depends on many factors:

time available vs time needed complexity of the action plant-specific factors (e.g., training) the influence of the accident on human performance the dependencies among human actions.

5

In the SSES IPE:

Errors associated with the selection and application ofprocedures were not modeled; i.e.,

the SSES excluded a whole class of errors, a significant weakness of the IPE.

Errors associated with action execution was modeled very minimally:

Only 3 types of error probabilities were used: 0, 1, equipment failures.

No substantive basis for the assumption that under moderate time pressure, the reliability of the execution actions will be similar to equipment reliability.

No explanation (or basis) for the times assumed to be available or need to perform an action.

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~

No evidence that the dependencies among human actions were considered.

no discussion of the influence of the accident on human performance.

no discussion of the influence of previous human failures.

~

Use of ext'remely optimistic hu~.tan error data for the ATWS modeL Failure probabilities for actions related to ATWS were assigned very low values 0.0, to 1E-4.

No basis was provided for these failure probabilities.

Failure probabilities for many actions that apparently were modeled were not provided.

~

Use ofsome extremely pessimistic values.

Manual scram and contain:nent venting were assigned probabilities of 1.0.

~

Excessive credit for repair/recovery Equi'pment repair in a PRA is typically limited to the recovery of offsite power for which there is adequate experience in nuclear power plants as well as established procedures

. and training.

There is no adequate data for repair of other equipment under accident conditions.

The success of equipment repair depends on many important plant-specific factors such as the type of failure, time needed for diagnosis, time needed for repair (which may range from a very few I:ours to several days), crew competing tasks under different accident conditions, and crew availability.

These factors do not appear to have been taken into consideration in the SSES IPE.

No apparent connection between several procedural improvements and IPE analysis results

~

The risk achievement importance measures for some actions is in excess of 10,000 C

Such values may be the result of:

'logic models in which failures of operators lead directly to core'damage very low probabilities used for these actions.

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Front end Analysis

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Inadequate common cause failure analysis.

The objective of CCF analysis is examine the plant for identifying and quantifying CCF events that have occurred or have the otential to o cur.

It was asserted in the IPE that no plant specific data show existence of CCFs for EDG, batteries and other DC power components.

This does not imply such failures are not possible.

Very few components are considered for common cause failures in the IPE.

Limited CCF treatment of injection systems (e.g.. no CCF for valves in HPCI or RCIC,

, no pumps in the condensate or RHR)

Limited CCF treatment of support systems (e.g.,

no CCF for ESW, or DGs, or batteries).

It seems to be implied that:

No failure possible due to maintenance, design, wearout Susquehanna staff can predict and fix a common cause failure with. 100% probability These assumptions do not appear to be reasonable

~

Lowfailure data.

Failure data of some important systems (e.g., HPCI, RCIC, and Fire Pump) are a lower

. by a factor of 5-to-1000 compared to generic or NUREG/CR-4550 data.

Some failure types are omitted, e.g., ESW failure to start or EDG failure to run is not included.

Blocks of totally differ components (e.g.,

a block comprised turbine, pumps, valves power supply, and flow controller).

Blocks of totally differ components have the same failure rates (e.g., blocks of valves have the same failure rates as the above block).

No basis provided for either the blocking of components or the data used.

~

Plant-specific assumptionslcalc. ilations No justification provided for several underlying hypotheses used.

Examples ATWS calculations predi".t in most cases core damage but no core melt.

Calculations of loss of H%'AC in the control building appears to neglect the potential new failures that may came to;ilay at elevated temperatures.

Procedure to increase the suppression pool inventory to prolong the time until HCTL are based on in-house calcul"Lions but are counter to BWROG EPGs.

11

Back end AnaLysis

~

Incomplete'examination ofenvironmental considerations The IPE considers harsh environments in the reactor building post containment failure would disable many systems located there, including HVACand the electrical switchgear.

However it is assumed that core damage can be prevented by utilizing the diesel driven fire pumps (which would not be affected) and opening the SRVs with the aid of the mobile diesel generator to supply the dc power (due to the reactor building damage to SRV power supplies).

This assumption appears to neglect possible consequential damage to, for example, the HPCI, RCIC, and/or core injection lines due to containment failure and a hot suppression pool.

Also, it does not explain why these conditions will allow the operators to align the required equipment locally.

12

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Lack of consideration of uncertainties in back-end accident progression:

no sensitivity studies performed (for example, to understand uncertainties associated with in-vessel and ex-vessel recovery) some vulnerabilities (due to uncertainties in code prediction, procedure descriptions, and operator actions) may not be identified from the PP8zL process

~

Incomplete treatment ofphenomena Major containment phenomena such as high-pressure melt ejection/direct heating and steam explosions were precluded in the Susquehanna IPH in the event trees and fault trees and consequently their impact on results is unknown.

Incomplete discussion of ex-vessel debris coolability, e.g., documentation of the geometric details of the cavity configuration to justify assumptions of eoolable debris bed.

~

Source term calculations Source term calculations may be biased by the low CDF so that only one. sequence meets the IPE screening criteria.

May not be representative ifhigher CDF had resulted.

13

~

Structural Analysis Containment failure pressure was chosen as a step function at 140 psig, not a distribution of failure pressures as requested in Generic Letter 88-20.

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