ML20138K043
| ML20138K043 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/05/1997 |
| From: | Kalman G NRC (Affiliation Not Assigned) |
| To: | Hutchinson C ENTERGY OPERATIONS, INC. |
| Shared Package | |
| ML20138F088 | List: |
| References | |
| GL-88-20, TAC-M74376, TAC-M74377, NUDOCS 9705120077 | |
| Download: ML20138K043 (11) | |
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- % UNITED STATES
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NUCLEAR REGULATORY COMMISSION l
WASHINGTON. D.C. 20006 4 001
%...../ May 5,1997 Mr. C. Randy Hutchinson
. Vice President, Operations ANO '
Entergy Operations, Inc. j 1448 S. R. 333 '
Russellville, AR 72801 1
SUBJECT:
STAFF EVALUATION REPORT (SER) AND TECHNICAL EVALUATION REPORT (TER FOR ARKANSAS NUCLEAR ONE, UNITS 1 AND 2, INDIVIDUAL Pl. ANT EXAMINATION (IPE) SUBMITTALS - INTERNAL EVENTS (TAC NOS. M74376, ,
M74377)
Dear Mr. Hutchinson:
On April 29, 1993, and August 28, 1992, Entergy Operations, Inc. (E01) submitted its responses to the Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-20, " Individual Plant Examination for Severe Accident vulnerabilities," and its associated supplements, for Arkansas Nuclear One, Unit I and Unit 2 (AN0-l&2), respectively. On December 1, and May 8, 1995, the NRC sent requests for additional information (RAls) regarding those submittals to E01. E01 responded to these RAls by. letters dated May 9, 1996,
- and October 5, 1995.
The staff's evaluations of the ANO-l&2 IPEs are documented in the enclosed '
reports. Enclosure 1 is the Staff Evaluation Report and Enclosure 2 is the contractor Technical Evaluation Report (TER) for the front-end, back-end and human reliability (HRA) analyses for ANO-1. Enclosure 3 is the Staff 1
Evaluation Report and Enclosure 4, Appendices A, B, and C, are the contractor Technical Evaluation Reports for the front-end, back-end and human reliability analyses, respectively, for ANO-2. Based on these reviews, the staff
, concludes that: (1) the AN0 IPE submittals for internal events, including internal flooding, are complete with respect to the information requested by GL 88-20 (and associated guidance in NUREG-1335) and (2) the IPE results are reasonable, given the ANO design, operation, and history. The staff concludes that the ANO IPE process was capable of identifying the most likely severe accidents and severe accident vulnerabilities, and that the ANO-l&2 IPEs met the intent of GL 88-20. It should be noted that this review was not intended to validate the accuracy of the IPE findings or quantification estimates.
Therefore, the attached SERs do not constitute NRC approval or endorsement of any IPE material other than meeting the intent of GL 88-20. ,
i The ANO-1 IPE estimated a total core damage frequency (CDF) of 4.n. 05 per !
reactor year, excluding internal flooding events and Anticipated Transients Without Scram (ATWS). This CDF compares reasonably with other Babcock and Wilcox (B&W) type plants. However, the staff identified weaknesses in the front-end and HRA portion of the IPE. The staff also noted that the licensee also relied on the acceptable reference-plant method for the back-end and containment analyses, but did not use plant specific data. Use of this IPE to support future risk-based regulatory applications will require additional treatment in these areas. !
9705120077 970505
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l Mr. C. Randy Hutchinson l The ANO-1 IPE used the criteria in Nuclear Management and Resources Council :
(NUMARC) 91-04, " Severe Accident Issue Closure Guidelines," to screen for J
plant-specific vulnerabilities. On the basis of these criteria, no vulnerabilities were identified; however, several potential improvements to i
plant equipment and procedures were recommended. Also, while no back-end vulnerabilities were identified, back-end issues regarding containment liner failure and containment isolation failure were identified for further investigation. In response to an RAI regarding the status of these issues, the licensee stated that the containment isolation failure had been addressed.
The licensee also stated that they had not initiated further study related to containment liner failure, but would monitor industry progress in this area.
The ANO-2 IPE estimated a total CDF of 3.4E-05 per reactor year, excluding internal flooding events. This CDF compares reasonably with other Combustion Engineering (CE) type plants. However, the staff identified weaknesses in the front-end and HRA portions of the IPE. Use of this IPE to support future risk-based regulatory applications will require additional treatment in these
, areas.
$ The ANO-2 IPE used the criteria in Nuclear Management and Resources Council (NUMARC) 91-04, " Severe Accident Issue Closure Guidelines," to screen for ,
plant-specific vulnerabilities. On the basis of these criteria, no '
vulnerabilities were identified; however, potential improvements to plant i' equipment and procedures were recommended.
E01 has indicated that it plans to maintain a "living" probabilistic risk I analysis (PRA) for ANO and will continue to consider plant modifications or 1 changes, as appropriate, based on revised risk insights.
l In accordance with GL 88-20, E01 proposed to resolve Unresolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements," for ANO. Based on the !
IPE process used to search for decay heat removal (DHR) vulnerabilities, and ;
our review of ANO plant-specific features, the staff finds that E0!'s DHR I evaluation is consistent with the intent of the USI A-45 resolution, and is therefore acceptable. The ANO-1 IPE also concluded that no vulnerabilities existed with respect to GSIs 23, 105, and 121. According to the guidance provided in GL 88-20, the staff concludes that the licensee has resolved GSIs 105 and 121. Regarding GSI-23, the Commission intends to issue a GL on this issue at a future date; therefore, the staff cannot conclude that GSI-23 has been resolved. -
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l Mr. C. Randy Hutchinson l l
This completes the staff review of TAC Nos. M74376 and M74377. If you have any questions, please call me at (301) 415-1308.
Sincerely, *
..- e n" ~
' Ge rge Kalm , Senior Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation r
l Docket Nos. 50-313 and 50-368 l
Enclosures:
- 1. Arkansas Nuclear One, Unit 1, Nuclear Power Plant Individual Plant Examination Staff Evaluatian Report
- 2. Arkansas Nuclear One, Unit 1, Nuclear Power Plant Individual Plant Examination Technic Evaluation Report
- 3. Arkansas Nuclear One, U , Nuclear Power Plant Individual Plant Examination Staf. ..aluation Report 4A. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual Plant Examination Technical Evaluation Report (Front-End ;
Analysis) ,
- 48. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual '
Plant Examination Technical Evaluation Report (Human Reliability Analysis) 4C. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual Plant Examination Technical Evaluation Report (Back-End Analysis) cc w/encls: See next page s
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May 5, 1997 Mr. C. Randy Hutchinson i i
This completes the staff review of TAC Nos. M74376 and M74377. If you have '
s any questions, please call me at (301) 415-1308.
Sincerely, Original signed by ;
4 George Kalman, Senior Project Manager Project Directorate IV-1 1 Division of Reactor Projects III/IV i
- Office of Nuclear Reactor Regulation <
1 1
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Docket Nos. 50-313 and 50-368 i !
Enclosures:
- 1. Arkansas Nuclear One, Unit 1, Nuclear Power Plant Individual l Plant Examination Staff Evaluation Report
! 2. Arkansas Nuclear One, Unit 1, Nuclear Power Plant Individual j Plant Examination Technical Evaluation Report
- 3. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual
. Plant Examination Staff Evaluation Report 4A. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual Plant Examination Technical Evaluation Report (Front-End Analysis)
- 48. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual 1 Plant Examination Technical Evaluation Report (Human
- Reliability Analysis) 4C. Arkansas Nuclear One, Unit 2, Nuclear Power Plant Individual Plant Examination Technical Evaluation Report (Back-End l
- Analysis) cc w/encls
- See next page l DISTRIBUTION:
Docket File PUBLIC- PD4-1 r/f GKalman CHawes OGC AC".S PGwynn, RIV JRoe EXdensam (EGAl) ,
Document Name: AR74376.LTR I
0FC PM/PD4-1 LA/PD4-1 l NAME GKaNn/vw CHawe(/MJ DATE df/9D 5 /s /97 COPY YES/N0 YES/N0 0FFICIAL RECORD COPY l
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Mr. C. Randy Hutchinson Entergy Operations, Inc. Arkansas Nuclear One, Units 1 & 2 i
i cc:
Executive Vice President Vice President, Operations Support !
& Chief Operating Officer Entergy Operations, Inc. l Entergy Operations, Inc. P. O. Box 31995 i P. O. Box 31995 Jackson, MS 39286-1995 '
Jackson, MS 39286-1995 Wise, Carter, Child & Caraway
. Director, Division of Radiation P. O. Box 651 Control and Emergency Management Jackson, MS 39205 Arkansas Department 'of Health :
4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 l
Winston & Strawn 1400 L Street, N.W. i Washington, DC 20005-3502 i
Manager, Rockville Nuclear Licensing Framatone Technologies !
1700 Rockville Pike, Suite 525 1 Rockville, MD 20852 l Senior Resident Inspector I U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801 e
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. . , . 3 J. INTRODUCTION On April 29,1993, the Entergy Corporation submitted the Arkansas Nuclear One-Unit 1 (AMO-1) Nuclear Power Plant Individual Plant Evaluation (IPE) submittal in response to Generic Letter 88-20 and associated supplements. On December 1,1995, the ,
staff sent questions to the licensee requesting additional information. He licensee '
responded in a letter dated May 9,1996. 1 A " Step 1" review of the ANO-1 IPE submittal was performed and involved the efforts of Brookhaven National Laboratory in the front-end, back-end, and human reliability analysis (HRA). The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities. Therefore, the review considered: (1) the completeness of the information, and (2) the reasonableness of the results given the ANO-1 design, operation, 1 and history. A more detailed review, a " Step 2" review, was not performed for this IPE submittal. A summary of staff findings is provided below. Details of the contractor's findings are in the attached technical evaluation report (Appendix A) of this staff evaluation j report (SER). '
l In accordance with GL 88-20, ANO-1 proposed to resolve Unresolved Safety Issue (USI) ;
A-45, " Shutdown Decay Heat Removal (DHR) Requirements." The licensee also proposed, j and the staff agreed to consider USI A-17, " Systems Interactions in Nuclear Power Plants," ,
for resolution with the submission of the internal flood portion of the IPE submittal. l l
In addition, the following Generic Safety Issues (GSIs) were included by the licensee in the IPE submittal for resolution:
GSI-23, " Reactor Cec!st Pump (RCP) Seal Failures,"
GSI-105, " Interfacing Systems Loss-of-Coolant Accidents (ISLOCAs) in Pressurized Water Reactors (PWRs),"
l GSI-121, " Hydrogen Control for Large Dry Containments."
l Section II of this SER discusses resolution of these USIs and GSIs. i 1
ne licensee intends to maintain a "living" probabilistic risk assessment.
I II. EVALUATION - l ANO-1 is a PWR with a large dry containment. The ANO-1 IPE has estimated a core damage frequency (CDF) of 4.7E-05 per reactor-year from internally initiated events, not 1 including an additional contribution from internal floods of IE-06. The ANO-1 CDF compares reasonably with that of other Babcox and Wilcox PWR nuclear plants. Transients contribute 66 percent (slightly over half of this due to loss of offsite power (LOOP)/ station -
blackout (SBO)) and loss of coolant accidents (LOCAs), primarily small LOCAs, contribute l 34 percent. Interfacing systems LOCAs and steam generator tube rupture (SGTR) each contribute less than 0.5 percent. A scoping study was conducted for anticipated transients without scram (ATWS) and it was found to contribute two percent to the CDF. Like ENCLOSURE 1 f
internal flood, ATWS results were not included in the reported CDF and, consequently, both accident types do not appear in importance rankings conducted for this IPE and will not appear in future rankings conducted for other regulatory applications, unless specifically incorporated by the licensee.
He important system / equipment contributors to the estimated CDF that appear in the top sequences are: AC power, service water, low pressure injection (LPI), emergency feedwater (EFW), ventilation systems, main feedwater (MFW), and high pressure injection (HPI).
He licensee's front end analysis appears to have ex=inad the significant initiating events and dominant accident sequences. A total of 19 initiating events were identified plus three ATWS initiators and six ISLOCA initiators. De IPE developed seven functional event trees. ANO-1 data are generally in agreement with NUREG-1150 and NUREG/ R-4550.
However, some plant-specific values which may affect CDF, such as, turbine driven EFW pump failure to start and/or run and the emergency diesel generator failure to start, appear .
to be significantly lower.
Weaknesses were identified, however, in some front-end data:
- 1. Although the IPE did not take credit for recovery or repair of the diesel generators in station blackout scenarios, it did take credit for recovery of offsite power, as is typical.
However, the power non-rec.overy factors appear to be optimistic in comparison to NSAC-147 data. For instance, non-recovery of offsite power is about four times less likely in the IPE than in NSAC-147. In addition, the LOOP frequency appears lower than would be expected based on the plant location (ANO-1 did experience a LOOP in 1978, but this was outside the plant-specific data window). As reported, only 15 percent of the LOOP frequency is due to weather causes.
- 2. Common cause failures were not modeled between turbine driven EFW and motor driven EFW pumps. Also, many components were not modeled in the common cause analysis, such as, circuit breakers (other than reactor trip breakers), electrical switchgear, air operated valves, air compressors, inverterr., relays, switches, and transmitters.
Based on the licensee's IPE process used to search for DHR vulnerabilities, and review of the ANO-1 plant-specific features, the staff finds the l' ensee's DHR evaluation consistent with the intent of the USI A-45, Decay Heat Removal deliability, and is, therefore, acceptable. Funthermore, the licensee did not identify any vulnerabilities with respect to l GSIs 23,105, and 121. According to GL 88-20, if a licensee concludes "that no vulnerability exists at its plant that is topically associated with any USI or GSI, the staff will consider the USI or GSI resolved for a plant upon review and acceptance of the results of the IPE." De staff concludes, therefore, that the licensee has resolved USIs A-17, GSI-105, and 121. Regarding GSI-23, the Commission has decided not to take additional rulemaking action at this time and plans to issue a generic letter on this issue at a future date; therefore, the staff cannot conclude that GSI-23 has been resolved.
He licensee performed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events. De !
licensee identified the following operator actions as important in the estimate of the CDF (in descending importance): j i
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- 1. Failure to manually open CV-1405/06 upon failure to deliver flow from the sump. !
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- 2. Failure to trip the RCPs within 30 minutes ofloss of intermediate cooling water seal ;
cooling or HPI seal injection. ;
- 3. Failure to prevent steam generator overfill due to excessive main feedwater flow.
- 4. Failure to start and align operating service water pump including available power source l l '
given total loss of service water; failure to manually open service water cooling jacket upon t valve failure signal.
- 5. Failure to attempt HPI cooling, i.e., feed and bleed cooling.
4 He HRA methodology employed by the licensee addressed both pre-initiator actions
! (performed during maintenance, test, and surveillance), including miscalibration and i restoration faults, and post-initiator actions (performed as part of the response to an i accident), including response type (proceduralized) and recovery-type (non-proceduralized)
{ actions. However, several weaknesses exist, as indicated below. While the weaknesses are l not severe enough to conclude that the submittal failed to meet the intent of Generic Letter
! 88-20, they do suggest that the licensee may not have learned as much about the role of j humans during accidents as they could have.
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j HRA weaknesses noted include- i
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- 1. Most plant-specific performance shaping factors appcar to have been left at their default j values, thereby eliminating an opportunity to " customize" the analysis. By leaving these L factors at their default values, the licensee is basically assuming that ANO-1 is an " average"
- plant. The resulting analysis is, therefore, generic instead of plant-specific. De licensee did indicate that the burden factor incmporated in the analysis does consider the difficulty
- of the task. However, there appears tc he no specific guidance as to which actions should be assigned to include burden.
l 2. Two human error probabilities (HEPs) in SGTR sequences appear exceptionally low. ;
, Dey include operator action to maintain reactor coolant system (RCS) pressure below a I
specified main steam safety valve set point and the action to cooldown the RCS and isolate j the break with the DHR system. The staff notes that the CDF estimates for SGTR
- sequences at ANO-1 were four to five times lower than those found at similar plants. I
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) 3. Here appear to liave been no simulator exercises conducted to evaluate operator .
! response nor walkdowns conducted to determine response times for operator actions outside !
! the control room. Dese appear to have been based primarily (if not solely) on interviews !
! with operators. Walkdowns, appear to have been conducted, however, for the flooding analysis and to resolve specific front-end modeling issues.
j De licensee evaluated and quantified the results of the severe accident progression through i the use of a containment event trees (CETs) and considered uncertainties in containment
< response through the use of sensitivity analyses. A generic and bypass CET were i developed to assess accident scenarias. As allowed by Generic Letter 88-20, the licensee
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- i i used a limited scope " reference plant" approach. Most quantifications of the CETs were ,
i based on point estimates taken from the Surry NUREG-1)50 analysis, although a few plant !
- specific MAAP runs were carried out to confirm assumrnons made in scoping calculations.
i The licensee's back-end analysis appeared to have considered important severe accident :
, phenomena. However, because of the simplified scoping analysis used and the lack of l
- detailed plant specific calculations, the back-end analysis is likely to be oflimited use for :
) future applications beyond fulfilling the intent of Generic Letter 88-20. !
J l According to the licensee, the ANO-1 conditional containmant failure probabilities are as ;
j follows: early containment failure (defined as failure prior to or approximately coincident .;
j with reactor vessel failure) is six percent with liner failure due to direct impingement of
- molten core material (dry cavity, high pressure melt ejection case) and impulse loads from !
ex-vessel steam explosion (wet cavity case) the primary contributors; late containment .
failures is 12 percent with containment overpressure (from noncondensible gas and/or !
l hydrogen burn) being the primary contributor. Bypass and isolation failure together add to j one percent with SGTR the primary bypass contributor. (Induced SGTR is not considered
- likely due to hot leg failure.) According to the licensee, the containment remains intact 81 ;
i percent of the time, i i
j De largest radiological releases resulted from containment bypass events and accidents ,
- involving early containment failure along with loss of containment sprays. De licensee's ;
l response to containment performance improvement program recommendations is consistent
- with the intent of Generic Letter 88-20 and associated Supplement 3.
- r l According to the licensee, some insights and unique plant safety features identified at l
- ANO-1 are
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! 1. He ANO-1 safety batteries have a two hour life upon loss of battery charging (see I l following discussion of new non-safety battery and alternate AC power source).
- 2. ANO-1 shares the control room, %e emergency cooling pond, and start-up transformer
l #2 with ANO-2.
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! to either or both steam generators. One train contains a motor driven pump and the other i
! contains a turbine driven pump. Dere are three water sources for the EFW; the condensate ,
j storage tank (CST), a back-up condensate storage tank (interconnected with Unit 2), and the .
j service water system.
- 4. Manual transfer is required to switch from injection to recirculation mode of core !'
! cooling. During high pressure recirculation following a small break LOCA, HPI pump I suction is aligned to the reactor building sumps in a " piggy back" fashion via the low j pressure recirculation pump.
- 5. De unit has feed and bleed capability. The HPI shutoff head is greater than the safety
{ relief valve setpoint pressure. As a result, the HPI pump can provide adequate core cooling i injection without requiring RCS depressurization. (A successful bleed path can be j established by opening the PORV or one of the two SRVs.)
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i 6. De RCP seals at ANO-1 are Byron-Jackson N-9000 type. Each of these seals consists l of a series of three mechanical seals. The licensee claims that these seals are designed and j tested to minimize leakage following the loss of RCP cooling. Hey are cooled by both the j i
HPI seal injection flow and by the seal return coolers. Consequently, the assumed robust j behavior eliminates the probability of an RCP seal LOCA in a loss of offsite power event
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because the RCPs are tripped due to loss of power. . In other accidents, the operators have .j j 30 minutes to trip the RCPs to avoid seal damage. l i
- 7. Potential for containment failure exists if a pressure plate in containment fails, providing a path from the reactor building floor and cavity region to the outer reactor building wall j through the incore instrument tunnel.
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- 8. De containment safeguards system includet the reactor building spray system and the !
fan coolers, but only the fan coolers (or the LPI heat exchangers) may be used as a DHR mechanism because the sprays are not connected to heat exchangers. l t l De licensee defined a vulnerability as " sequence groups with a valid mean core damage l frequency greater than IE-4 per reactor year or containment event tree endstate groups with containment failure / bypass greater than 1E-5 per reador year." Based on this defm' ition, the licensee did not identify any vulnerabilities. J Plant improvements, however, were identified and evaluated. Rose implemented are listed ,
below:
Hardware:
- 1. Install new alternate AC power source, i.e., a diesel generator, in response to the SBO rule; not credited in the reported CDF.
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- 2. Remove non-safety loads from safety battery through the use of a new non-safety l' battery, thereby, increasing the safety battery life to four hours.
- 3. Remove internals to manual valve FW-1016 to reduce likelihood of flowpath blockage l for EFW (scheduled for October 1996 outage). !
Procedural: *
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- 1. Proceduralize the' process to recover the LPI failure combination of one LPI suction line i and one LPI unavailable by allowing flow from the available suction line to the operable i LPI pump. ;
- 2. Alter the "diaty liquid waste and drain processing" procedure to direct closure of the ;
normally open LPI/DHR and reactor building spray pump drain room isolation valves. ;
This action ensures that the LPI pumps will not be affected by ISLOCA discharges into the ;
auxiliary building.
- 3. Add verification for closure of SV-7454 to SBO procedure. )
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III. CONCLUSION :
Based on the above findings, the staff notes that: (1) the licensee's IPE is complete with regard to the information requested by Generic Letter 88-20 (and associated guidance in NUREG-1335), and (2) the IPE results are reasonable given the ANO-1 design, operation, )
and history. As a result, the staff concludes that the licensee's IPE process is capable of identifying the I most likely severe accidents and severe accident vulnerabilities, and therefore, that the ,
ANO-1 IPE has met the intent of Generic Letter 88-20. j it should be noted that the staff's review primarily focused on the licensee's ability to )
examine ANO-1 for severe accident vulnerabilities. Although certain aspects of the IPE '
were explored in more detail than others, the review is not intended to validate the accuracy !
of the licensee's detailed findings (or quantification estimates) that stemmed from the examination. Derefore, this SER does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of Generic Letter 88-20. De staff has identified weaknesses in the frontand and HRA portion of the IPE: and we also note that it relied on the (acceptable) reference plant method for back-end j containment analysis, which used non-plant-specific data; consequently, we believe that :
application of the IPE in support of risk-based regulatory applications, beyond those
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associated with Generic Letter 88-20, will require additional treatment in these areas.
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APPENDIX A ,
TECHNICAL EVALUATION REPORT 1
ENCLOSURE 2
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