ML20207L547
| ML20207L547 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 02/26/1999 |
| From: | Croteau R NRC (Affiliation Not Assigned) |
| To: | Maret G VERMONT YANKEE NUCLEAR POWER CORP. |
| References | |
| TAC-M83689, NUDOCS 9903180178 | |
| Download: ML20207L547 (10) | |
Text
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- i. .
.. February 26, 1999 [
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Mr. Gregory A. Maret !
Director of Operations
- Vermont Yankee Nuclear Power Corporation ,
185 Old Ferry Road !
Brattleboro,VT 05301 l
i
SUBJECT:
' REQUEST FOR ADDITIONAL INFORMATION ON VERMONT YANKEE I NUCLEAR POWER STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL (TAC NO. M83689)
Dear Mr. Maret:
I i
' Based on our ongoing review of your Vermont Yankee IPEEE submittal dated June 30,1998, l we have developed the attached requests for additional information (RAls). The RAls are !
related to the fire, seismic, intemal flooding, and high wind, flood, and other extemal events !
areas discussed in your IPEEE submittal. The RAI on fire was developed by our contitsetor l Sandia National Laboratories; the RAI in the seismic area was developed by Brookhaven ;
National Laboratory; and the RAls in the internal flooding and the HFO area wee developed by i NRC's Office of Research.
We request that you respond or provide a schedule for responding to the enclosed RAI within !
60 days.
Sincerely, !'
/s/
Richard P. Croteau, Project Manager Project Directorate 1-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271-
Enclosure:
Request For AdditionalInformation cc w/ encl: See next page DISTRIBUTION Docket File 7 PUBLIC E. Adensam [, l PDl-2 Rdg. J. Zwolinski C. Anderson, RI l j A. Rubin T. Clark 1
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DOCUMENT NAME: G:\CROTEAU\RAl83689.WPD !
To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy !
with ettachment/ enclosure "N" = No copy I OFFICE PM:PDI-2 E, LA:PDI-2 lM D:Pl3P2 ; l -- l l~ l l NAME- RCroteauP TClarid/& EAdWam ,
DATE- 02/s./99 02/5499 Mk99 0y ;
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006H001 O %
% February 26, 1999 4 :
Mr. Gregory A. Maret Director of Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road Brattleboro, VT 05301
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON VERMONT YANKEE NUCLEAR POWER STATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) SUBMITTAL (TAC NO. M83689)
Dear Mr. Maret:
i Based on our ongoing review of your Vermont Yankee IPEEE submittal dated June 30,1998, !
we have developed the attached requests for additionalinformation (RAls). The RAls are '
related to the fire, seismic, internal flooding, and high wind, flood, and other external events i areas discussed in your IPEEE submittal. The RAI on fire was developed by our contractor Sandia Nstional Laboratories; the RAI in the seismic area was developed by Brookhaven :
National Laboratory; and the RAls in the internal flooding and the HFO area were developed by i NRC's Office of Research.
We request that you respond or provide a schedule for responding to the enclosed RAI within I 60 days.
Sincerely l
bI Richard P. Croteau, Project Manager Project Directorate I-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosure:
Request For AdditionalInformation cc w/ encl: See next page
l 1
3 G. Maret Vermont Yankee Nuclear Power Station cc:
Regional Administrator, Region i U. S. Nuclear Regulatory Commission Mr. Rayn .2.d N. McCandless 475 Allendale Road Vermont Division of Occupational King of Prussia, PA 19406 and Radiological Health Administration Building l Mr. David R. Lewis Montpelier, VT 05602 1 Shaw, Pittman, Potts & Trowbridge )
2300 N Street, N.W. Mr. Gautam ben 1 Washington, DC 20037-1128 Licensing Manager Vermont Yankee Nuclear Power Mr. Richard P. Sedano, Commissioner Corporation Vermont Department of Public Service 185 Old Ferry Road 120 State Street,3rd Floor Brattleboro, VT 05301 Montpelier,VT 05602 Resident inspector Public Service Board Vermont Yankee Nuclear Power Station State of Vermont U. S. Nuclear Regulatory Commission 120 State Street P.O. Cox 176 Montpelier, VT 05602 Vernon, VT C5354 Chairman, Board of Selectmen Mr. Peter LaPorte, Director Town of Vernon ATTN: James Muckerheide P.O. Box 116 Massachusetts Emergency Management Vernon, VT 05354-0116 Agency 400 Worcester Rd.-
Mr. Richard E. McCullough P.O. Box 1496 i Operating Experience Coordinator Framingham, MA 01701-0317 I Vermont Yankee Nuclear Power Station P.O. Box 157 Jonathan M. Block, Esq.
Govemor Hunt Road Main Street Vernon,VT 05354 P. O. Box 566 l Putney, VT 05346-0566 G. Dana Bisbee, Esq.
Deputy Attorney G3neral Mr. Michael J. Daley j 33 Capitol Street Trustee and Legislative Representative i Concord, NH 03301-6937 New England Coalition on Nuclear i Pollution, Inc. j Chief, Safety Unit Box 545 .
)
Office of the Attomey General Brattleboro,VT 05301 One Ashburton Place,19th Floor i Boston, MA 02108 I Ms. Deborah 3. Katz !
Box 83
- Shelburne Falls, MA 01370 l
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- Request for Additional Infonnation (RAI) on Vermont Yankee Individual Plant Examination of External Events (IPEEE) Submittal '
FIRE
- 1. From the submittal the bases for crediting fire detection and suppression systems are not clear. In particular, it appears that carbon dioxide (cardox) systems have been credited in several of the risk daminant areas. However, cardox systems are especially vulnerable to inconsistent performance if not installed and maintained in accordance with i accepted national codes and standards. Generic system reliability values were apparently applied to these systems. These values assume fully code compliant systems.
]
Please confirm whether or not the fire detection and suppression systems including the cardox suppression systems at Vermont Yankee have been installed and maintained in accordance with accepted national standards. If any of the installed systems do not meet these standards, assess the impact on core damage frequency (CDF) if the non-compliant system is not credited.
- 2. In the assessment of compartment-specific fire frequencies, both fire severity and fire data correction factors were used to modify the fire ignition frequencies for many postulated fire sources. Further, the assessmer.t has also applied fire frequency weighting factors for some fire sources that are not consistent with the Electric Power Research Institute (EPRI) fire- i induced vulnerability evaluation (FIVE) methodology.
The first point of concem is that the " correction factors" that have been independently applied may have inappropriately reduced compartment fire frequencies. In particular, the licensee has applied a correction factor based on the fraction of fires in the data base that occurred while a plant was "at power." The fire frequencies cited in the FIVE study include consideration of this fact, and to apply an additional correction is contrary te the FIVE methodology. Typically, fire events are considered relevant if they could have occurred at power even if the event actually occurred during some other mode of operation. This would include, for example, events occurring during pre-operational testing, during start up operations, and during certain shutdown events where the shutdown mode of operation is not identified as a specific contributing factor in the event. Events that are typically excluded l would Nelude events such as fires caused by welding or cutting in areas where such l activities are not permitted during power operation, or events attributed to construction !
activities that are no longer present. Again, these factors have already been incorporated I into the generic fire frequencies cited in procedural guidance documents such as the EPRI FIVE methodology.
The second point of concem relates to the licensee's application of the FIVE methodology with respect to ignition source weighting factors. This impacts the switchgear rooms, battery rooms and intake / discharge structures. The licensee applied the analysis of " Method B",
which is dividing the number of ignition sources in the fire compartment by the number in the :
selected location, rather than the FIVE approach of " Method A," which does not use an ignition source weighting factor. This may have artificially reduced fire frequencies. In particular, it is important that the overall plant wide frequency of fire events be preserved, i 1
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and it is not clear that the licensee IPEEE analysis has done so. Since the licensee used the FIVE methodology as its basis for its IPEEE fire analysis, the licensee has not provided adequate justification for its deviations from the FIVE's prescription of methods of applying weighting factors.
The third point of concem relates to the application of fire severity factors. The licensee cites the EPRI Fire Probabilistic Risk. Assessment (PRA) Implementation Guide as the source of these values. The severity factors were used to adjust the basic ignition frequencies of the associated components for those areas surviving screening. These adjusted ignition frequencies were apparently used in scenarios where fire suppression was credited. Since the success of fire suppression would reduce the potential for a large fire, there appears to be a significant possibility that the use of a fire severity factor, when fire suppression is modeled, double counts for suppretsion efforts.
Considering the points above, please provide the following:
(a) For the compartments that were quantitatively screened, please reassess the screening analysis using defensible fire ignition frequencies. For each compartment, identify the fire sources, the baseline fire frequencies, and any correction or severity factors used to develop the final compartment frequency.
Provide justification for any modifying factors used including consideration of the concerns discussed above. Discuss any changes to the original screening results.
(b) For the compartments which survive the revised screening analysis from (a) above, please analyze or re-analyze the detailed fire scenarios. For the source / target sets analyzed in each compartment, discuss the partitioning used to develop the scenario fire event frequency. As part of the re-analysis, identify any cases where both a severity factor and independent credit for fire detection and suppression were applied in the original analysis. For those cases, either eliminate the severity J factor or eliminate the detection suppression credit. Based on the abcyo approach, l please recalculate the CDF contribution for each compartment and provide the l results.
(c) Given the above results, identify those compartments / scenarios that dominate the fire CDF. Re-assess the implications of the analysis with regard to the identification of plant vulnerabilities, modifications, and improvements.
- 3. The heat loss factor (HLF) is defined as the fraction of energy released by a fire that is transferred to the enclosure boundaries. This is a key parameter in the prediction of component damage, as it determines the amount of heat available to the hot gas layer. A I larger HLF means that a larger amount of heat (due to a more severe fire, a longer buming time, or both) is needed to cause a given temperature rise. It can be seen that if the va!ue assumed for the HLF is unrealistically high, fire scenarios can be improperly screened out.
Figure R.1 provides a representative example of how hot gas layer temperature predictions can change assuming different HLFs. Note that: 1) the curves are computed for a 1000 kW fire in a 10m x Sm x 4m compartment with a forced ventilation rate of 1130 cfm; 2) the FlVE-recommended damage temperat'.sre for qualified cable is 700*F for qualified cable and 450*F for unqualified cable; and,3) the Society for Fire Protection Engineers (SFPE) curve !
in the figure is generated from a correlation provided in the SFPE Hanc; book [R.2]. i 2-I
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t Based on evidence provided by a 1982 paper by Cooper et al. [R.3], the EPRI Fire PRA (
Implementation Guide recommends a HLF of 0.94 for fires with durations greater than five minutes and 0.85 for " exposure fires away from a wall and quickly developing hot gas layers." However, l as a general statem3nt, this appears to be a misinterpretation of the results. l 1
Time-Temperature Curves I i
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-*- H.F = 0.85 1 E 800 - -+-itF = 0.94
-x- H.F = 0.99 500- /
400 300 -
200 -
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- m. f.3 Figure R.1 Sensitivity of the hot gas layer temperature predictions to the assumed heat loss factor.
Reference R.3, which documents the results of multi-compartment fire experiments, states that the higher HLFs are associated with the movement of the hot gas layer from the burning compartment to adjacent, cooler compartments. Earlier in the experiments, where the hot gas layer is limited to the buming compartment, Reference R.3 reports much lower HLFs (on the order of 0.51 to 0.74).
These lower HLFs ars more appropriate when analyzing a single compartment fire.
In summary, (a) hot gas layer predictions are very sensitive to the assumed value of the HLF; and (b) large HLFs cannot be justified for single-room scenarios based on the information referenced in the EPRI Fire PRA Implementation Guide.
For each scenario where the hot gas layer temperature was calculated, please specify the heat loss factor value used in the analysis. In light of the preceding discussion, please either: a) justify the value used and discuss its effect on the identification of fire vulnerabilities, or b) repeat the analysis using a more justifiable value and provide the resulting change in scenario contribution to CDF.
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- 4. Th'ere are a number of points related to the analysis of the Reactor Building Cable !
Separation Zone (CSZ) that require clarification. First, the treatment of manual suppression i appears optimistic. A 90% manual suppression reliability appears to have been assumed !
(a manual non-suppression probability of 0.1 is cited). A fire brigade initial response time .
of ten minutes was estimated for this zone; however, the initial brigade response time should l l not be equated to the fire suppression time. Suppression time must include the time needed l
! to detect the fire, verify the fire, assemble and equip the fire fighting team, assess the fire l situation, and actively suppress the fire. Further, manual suppression reliability should
) consider the time to fire damaga as compared to the time required for suppression. In the j study a damage time of 12.5 minutes is cited for damage to cable tray Division S1 and Division S2. Given an initial brigade response time of 10 minutes (after detection), it would appear highly optimistic to assume a 90% manual suppression reliability before damage occurs. Secondly, the bases of the fire ignition frequencies, including any severity or l
correction factors, used for this zone were not described. This aspect of the analysis may also be impacted by the concems discussed in RAI number 2 above. Third, the dependence between failure of the automatic suppression system and failure of manual fire fighting support systems (e.g., hose stream stand pipes) was not discussed. For example,if the fire water supply to the sprinkler system fails, then it would appear likely that the water supply to manual hose streams might also fail. Fourth, the compartment is cited as being covered by a partial pre-action sprinkler system. It is not clear what coverage this system provides, how this system was credited in the analysis, or how its effectiveness against the postulated fire scenarios was assessed. Finally, it is not clear what systems are potentially impacted by fires in this area.
For the Reactor Building CSZ, please discuss in more detail the fire protection features of this area including a description of the fire detection system, the fixed fire suppression system, and the systems needed to support manual fire fighting activities. Describe how the effectiveness of the fixed suppression system was assessed and credited in the analysis.
Discuss the time required for fire detection and verification, initial fire brigade response, assembling and equipping the brigade, fire situation assessment, and actual fire suppression for the postulated fires in this zone. Identify the plant systems and equipment located in the .
area and discuss their importance to plant safety. Describe the fire scenarios postulated for l this area, including the bases for the assumed fire ignition frequencies associated with each ;
fire scenario and consideration of the concerns identified in RAI number 2 above. Also j discuss the conditional core damage probability (CCDP) for loss of all equipment in the room ;
for each of the postulated fire scenarios. Discuss the estimated time (s) to critical component
- i. damage that have been assumed in the analysis and their bases. Provide an explicit justification for the assumed manual suppression reliability estimate used in the analysis or repeat the analysis for the CSZ using more realistic manual non-suppression probabilities.
Assess the potential for dependence between failure of automatic suppression and failure of manual fire fighting support systems. Given the above factors, reassess the CDF contribution of this compartment.
- 5. The detailed analysis of Reactor Building compartment RB3 resulted in a CDF of 3.1E-6/yr.
l- The quantitative screening analysis for this compartment shows a CDF of 1.5E-4/yr based l on an ignition frequency of 1.7E-3/yr. This implies a CCDP of about 0.1, presumably I assuming loss of all equipment in the area. However, it does not appear that this area is j protected by automatic suppression, and the analysis states that manual suppression was
- ' not credited in the analysis of this compartment. Hence, it is unclear how the screening CDF j of 1.5E-4/yr was reduced to the final estimated CDF of 3.1E-6/yr.
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l Please provide a detailed discussion of how this compartment was analyzed, starting with the ignition frequency and concluding with the CDF. Explain the factors (ignition frequencies, severity factors, partitioning factors, weighting factors, etc.) that were used to estimate fire ignition f requencies and include in this discussion consideration of the concems l l raised in RAI number 2 above. Describe the plant equipment and systems that are located l
in this room. Describe the fire source / target sets considered in the detailed quantification {
l scenarios and the basis for their selection. Include a discussion of the CCDP for each fire scenario analyzed. l
! l l RAI References l
1 l R.1 J. Lambright, et al., "A Review of Fire PRA Requantification Studies Reported in l l
NSAC/181," prepared for the iInited States Nuclear Regulatory Commission, April !
1994. !
l R.2 P. J. DiNenno, et al, eds., "SFPE Handbook of Fire Protection Engineering," 2nd '
i Edition, National Fire Protection Ascociation, p. 3-140,1995.
R.3 L. Y. Cooper, M. Harkleroad, J. Quintiere, W. Rinkinen, "An Experimental Study of i Upper Hot Layer Stratification in Full-Scah Multiroom Fire Scenarios," ASME Journal of Heat Transfer, 104,741-749, November 1982. )
i SEISMIC
- 1. With the exception of floods due to actuation or failure of the fire suppression systems, seismic induced internal floods are not discussed in the submittal. This is not consistent with i' NUREG-1407. It is stated in NUREG 1407 that "The scope of the evaluation of seismically induced floods, in addition to that of the external sources of water (e.g., tanks, upstream dams), should include the evaluation of some internal flooding consistent with the discussion in Appendix 1 of EPRI NP-6041." According to EPRI NP-6041,"the effects of possible ruptured vessels or piping systems that could flood or cascade onto essential equipment should be considered." The consideration is not limited to the flooding due to actuation and failure of the !
fire suppression syums. Please describe how seismically induced intemal flood sources were identified and evaluated in the IPEEE. {
- 2. Please provide the results of the evaluation of the seismic improvement opportunities described in Section 7.2.2 of the submittal. Please provide a schedule for those items selected for implementation.
- 1. The Time Reliability Correlation method from NUREG/CR 1278 was used for estimating human error probabilities (HEPs) for internal flooding events. The submittal states that "for the relatively routine operator actions needed for the isolation of a pipe," failure to diagnose a l flooding event was judged to be small, and optimum stress levels were judged to be I reasonable; therefore, nominal HEP values were used. The submittal also states that because
- - "the typical action (s) necessary to mitigate a flooding event are not complex," the HEP values l
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associated with action execution (" manipulative error") were judged to be small (on the order of 1E-4/yr) and hence adequately accounted in the HEP values related to diagnostic error; as a result only HEPs for failing to diagnose an event were derived.
However, the submittal neither provided a clear description of the human actions involved in the accident sequences of each flooding initiating event nor demonstrated how these assumptions were applied in the quantification of each particular action under different accident conditions. The use of these assumptions is of particular interest for the initiator "RBTRF2: un isolable service water retum line break in the reactor building" the mitigation of which depends on several human actions some of which do not appear to be " routine operator actions." Please address the following:
(a) Provide a detailed description of the accident sequences (and pertinent failure probabilities) associated with the initiator RBTRF2, depicting how this event leads to core damage.
(b) For each human action credited in RBTRF2, please describe how associated HEPs were derived on the basis of these assumptions. That is, for each particular action Ansn each particular accident sequenco in which the action is modeled, explain: (i) why the probability to fail to recognize the need for the action is small, (ii) why stress would be optimum, (iii) how much time is available (from the moment it is recognized the action is needed), (iv) how much time is needed to accomplish the action, (v) are there procedures available, (vi) would it be a " routine action" under the specific accident conditions.
(c) Describe how the times needed to perform an action were estimated. For example, were they estimated by walkthroughs?
(d) Describe how the dependencies among human actions in this particular initiator were treated.
- 2. (a) The CDF from intemal flooding of 9E-6/yr takes credit for several improvements l most of which are stated in the submittal as being "under evaluation." (This CDF is 200% higher than the total CDF of 4.3E-6/yr from all internal events calculated in the IPE. If the internal flooding CDF is included, the total internal event CDF becomes about 1.3E-5/yr, and the flooding contributes about 70% of the total intemal event CDF.) Furthermore, the actual internal flooding CDF may be even ,
higher if improvements credited will not be implemented. The submittal does not i provide a CDF estimate without crediting the improvements which have not been '
implemented. Please provide: (a) a- schedule for those items selected for implementation, and (b) the CDF estimate without crediting any of the j improvements (described in Section 7.2.3 of the submittal) that are not planned to i be implemented.
(b) The internal flooding associated with "RBTRF2: un-isolable service water retum line break in the reactor building," yielded a CDF of about 6E-6, which is about 70% of the total internal flooding CDF. Section 7.2.3 of the submittal states that it has t een l proposed to evaluate hardware and procedural modifications (not credited);
however, it is not explained what exactly these improvements are. Since this is the most risk significant initiator, it is reasonable to expect a more focused emphasis
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in terms of improvements in this particular area. Please explain if any specific
' improvements have been identified and have been, or are planned to be, implemented.
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- 3. It is stated in the submittal that " Based on the comprehensive internal flooding evaluation performed, internal flooding events are judged to have insignificant influence on the
- reliability of containment performance as analyzed in the IPE."
A value of a large release frequency (LERF) less than 1E-6/yr was used in the IPE to screen for vulnerabilities. The IPE concluded that no vulnerability with respect to containment performance exists although a LERF of 9.7E-7/yr was estimated, which is very close to the IPE's criterion for vulnerability.
The internal flooding CDF is a factor of 2 (and the total intemal event CDF a factor of 3) higher than the IPE's CDF The submittal does not include a discussion of accident progression and containment performance due to flooding. For example, it is not clear if all sequences associated with RBTRF2 are long-term sequences. Also, it appears that sequences associated with other initiators have the potential for LERF, especially those leading to loss of switchgear or to the loss of the diesels. The low CDF of these initiators can be due to credit taken for improvements "under evaluation." The subr-ittal did not provide an assessment of containment performance for Vermont Yankee as-built as-operated. Therefore, it is unclear how it can be concluded that internal flooding has an
" insignificant influence on the reliability of containment performance as analyzed in the IPE." Please explain how you have reached this conclusion.
HIGH WIND, FLOOD, AND OTHER EXTERNAL EVENTS 4
- 1. Regarding gas pipeline rupture accidents, the IPEEE submittal did net state whether there are any gas pipelines near the plant site or provide any information relating to gas pipeline l rupture accidents. Please provide an assessment of gas pipeline rupture accidents at Vermont Yankee.
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