ML14178A927
| ML14178A927 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 04/17/1996 |
| From: | Mozafari B NRC (Affiliation Not Assigned) |
| To: | Hinnant C CAROLINA POWER & LIGHT CO. |
| References | |
| GL-88-20, NUDOCS 9604260103 | |
| Download: ML14178A927 (8) | |
Text
April 17, 1996 9
.Mr. C. S. Hinnant, Vice President Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550 As a result of our ongoing review of your submittal dated June 30, 1995, its associated documentation in accordance with Generic Letter (GL) 88-20, Supplement 4, "Individual Plant Examination if External Events (IPEEE) for Severe Accident Vulnerabilities, 10 CFR 50.54(f)," for the H. B. Robinson Steam Electric Plant, Unit No. 2, we have determined a need for additional information. The Enclosure contains the detailed request for additional information (RAI) developed with the assistance of our contractor, Energy Research, Inc., and reviewed by a senior review board (SRB) comprised of staff from the Offices of Nuclear Regulatory Research (RES) and Nuclear Reactor Regulation (NRR), as well as RES consultants (Sandia National Laboratory) with probabilistic risk assessment expertise for external events. The RAI is related to the external event analyses in the IPEEE, including the seismic analysis, the fire analysis, and the analyses on effects of high winds, floods, and others.
We request that you provide written responses to the RAI within 60 days of the receipt of this letter.
Sincerely, (Original Signed By)
Brenda Mozafari, Project Manager Project Directorate II-1 Division of Reactor Projects -
I/Il Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosure:
RAI cc w/enclosure: See next page Distribution RD k-etTFiler MCunningham PUBLIC RHernan PDII-1 RF TKing SVarga MWHodges JZwolinski OGC FILENAME -
G:Robinson\\ROB83668.RAI ACRS OFFICE LA:PDII-P :PDII-1 D:PDII-1 NAME EDunnington ozafari Elmbro DATE 04//7/96 04//1/96 04/ /1/96 COPY es No Yes/No Yes/N 4a
__II OFFICIAL RECORD COPY~
9604260103 960417 PDR ADOCK 05000261 C
I LE CENTR COP p
__PDR
Mr. C. S. Hinnant H. B. Robinson Steam Electric Carolina Power & Light Company Plant, Unit No. 2 cc:
Mr. William D. Johnson Mr. Dayne H. Brown, Director Vice President and-Senior Counsel Department of Environmental, Carolina Power & Light Company Health and Natural Resources Post Office Box 1551 Division of Radiation Protection Raleigh, North Carolina 27602 Post Offilce Box 27687 Raleigh,North CarolCa 27611-7687 Ms. Karen E. Long Assistant Attorney General Mr. Robert P. Gruber State of North Carolina Executive Director Post Office Box 629 Public Staff -
NCUC Raleigh, North Carolina.27602 Post Office Box 29520 Raleigh,North Carolina 27626-0520 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mr. Max Batavia, Chief H. B. Robinson Steam Electric Plant South Carolina Department of Health 2112 Old Camden Road Bureau of Radiological Health Hartsville, South Carolina 29550 and Environmental Control 2600 Bull Street Regional Administrator, Region II Columbia, South Carolina 29201 U.S. Nuclear Regulatory Commission 101 Marietta St., N.W., Ste. 2900 Mr. J. Cowan Atlanta, Georgia 30323 Manager Nuclear Services and Environmental Mr. Dale E. Young Support Department Plant General Manager Carolina Power & Light Company Carolina Power & Light Company Post Office Box 1551 -
Mail 0HS7 H. B. Robinson Steam Electric Plant Raleigh, North Carolina 27602 3581 West Entrance Road Hartsville, South Carolina 29550 Mr. Milton Shymlock U. S. Nuclear Regulatory Commission Public Service Commission 101 Marietta Street, N.W. Suite 2900 State of South Carolina Atlanta, Ga. 3023-0199 Post Office Drawer 11649 Columbia, South Carolina 29211 Mr. R. M. Krich Manager -
Regulatory Affairs Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550
Enclosure REQUEST FOR ADDITIONAL INFORMATION REGARDING THE INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS FOR THE H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 A.2 Seismic
- 1.
NUREG-1407 requests that a screening criteria be applied to non-seismic failures and human actions. Please provide the results of your related screening analysis and a list of operator actions that are required to ensure integrity of the chosen success paths.
Include in this list any operator actions that may be required to recover from relay chatter.
(Note the submittal's relay evaluation credited operator recovery actions for 14 essential relays.)
For each human action, indicate the time the operator action is required and its location. Indicate also the human error probabilities which account for seismic affects on operator actions. Also, please provide a list of the random failures and their failure rates having the most significant potential to compromise integrity of the success paths. Indicate the screening criteria applied to rates of random failures and operator errors, and report the results of your screening evaluation.
Please also discuss how plant emergency operating procedures have been modified to ensure availability of success paths. If no changes to the emergency operating procedures have been made as a result of the seismic internal plant examination for external events (IPEEE), please justify why not.
- 2.
The evaluation of soil liquefaction described in the submittal assumes an earthquake magnitude of M 5.5. Liquefaction-induced soil strains, and other manifestations of accumulated damage in soils, are sensitive to magnitude selection (which implicitly determines the level of strong motion duration). The magnitude used for the soils evaluation in the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBR2) seismic IPEEE has not been determined to be consistent with the review level earthquake (RLE).
(The selected magnitude is based on the mean magnitude associated with the 10,000-yr uniform hazard spectrum (UHS) for the Savannah River site; hence, not only is it not consistent with the RLE spectrum, but it is also not even based on the seismic hazard at the HBR2 site.)
Please select a magnitude consistent with the RLE for HBR2, for use in the soils evaluation, and justify your magnitude selection.
Please discuss the impacts of such magnitude selection on your assessment of liquefaction and of other soil failures, describing the details of your evaluation procedures. (Alternative to selecting a single, fixed RLE-consistent magnitude, analyses of the sensitivity of soil responses to various chosen magnitude levels of interest [e.g., MW 5.5, 6.0, and 6.5] would suffice.)
The submittal has noted that isolated lenses of soil liquefaction may occur at the site. Please identify the expected locations and sizes of
-2 such liquefaction lenses.
Please provide soil profiles for those locations where liquefaction is most likely to occur and corrected blow count data for these profiles. For a magnitude selection more consistent with an RLE motion, such isolated lenses may become significant zones of liquefaction. Hence, describe the extent (size and location) of liquefaction zones (together with the extent of any lateral spreading), for an RLE-consistent magnitude selection, for zones that may affect plant structures (including performance of foundation pile systems), components, and embankments (including Lake Robinson Dam).
- 3.
The submittal has not discussed the potential failures of buried piping, of below-grade trenches containing vital piping and equipment, nor of other important components that are buried, embedded, or founded in soil.
Effects of differential soil displacements/settlements are important considerations for these items. Potential failures of the fuel oil transfer system and of the service water system are just a couple of critical safety concerns. The discussion in Section 5.3.4.2 of the IPEEE submittal reveals that fuel oil transfer lines are not inherently robust; the submittal has also noted past concerns pertaining to buried service water piping. Please discuss the potential for failure of buried components that are important to success paths.
Justify whether or not the high confidence in low probability failure (HCLPF) capacities of such buried components meet the RLE.
- 4.
NUREG-1407 requests that state-of-the-art approaches to soils evaluation be implemented. Use of the simplified Newmark sliding-block analysis alone, for embankment deformation/stability assessment, is considered to be inconsistent with this request, unless the assessment can be shown to be bounding. Because the dynamic factor of safety of Lake Robinson Dam, for SSE input (0.2g), was earlier assessed as being only 1.02, and because sliding instability was predicted (in the submittal) to occur for the RLE, seismic-induced deformations and potential instability of the dam are important concerns. Hence, please justify, through either a state-of-the-art assessment or a bounding analysis, whether or not the HCLPF capacities of Lake Robinson Dam and of other important embankments (to resist dynamic instability/deformation failure) meet the RLE.
Also, please address and discuss the potential for seismic-caused movements of submerged slopes, or dispersal of lake sediments, which may interfere with service water intake or discharge. Justify whether or not such concerns exist for RLE motions.
- 5.
The discussion in Section 8 of the seismic IPEEE report by your consultant, EQE Engineering Consultants, pertaining to seismic-induced floods, does not appear to address failures of non-safety system equipment list (SSEL) tanks nor failures of piping other than fire-water piping. Please report the findings of your walkdown and resulting evaluation pertaining to seismic-induced floods that may be caused by non-SSEL tank failures and non-fire-water piping failures.
-3
- 6.
Unless components of the containment heat removal system are known to have high seismic capacity, NUREG-1407 requests that the seismic walkdown include an examination of such components and their anchorages.
Excluding fan coolers from the seismic IPEEE, on the basis of MAAP calculations, does not meet this request, and is inconsistent with the principal intent of the IPEEE in finding potential vulnerabilities from plant walkdowns. Please report the findings of your walkdown pertaining to containment heat removal systems, including the fan cooler system.
Explain the importance of the integrity of fan coolers to availability of component cooling water and to the prevention of core damage.
- 7.
It is not clear that the scope of the relay evaluation has addressed all components needed for successful containment performance. The submittal does not provide a list indicating which relays are essential to successful containment performance, including successful containment isolation functions.
Please provide such a list and discuss the findings pertaining to evaluation of these relays.
- 8.
The IPEEE program, as well as the findings in this review, must rely heavily on the judgment of the seismic review team (SRT), which has the responsibility for making screening decisions. The submittal states (Section 5.7.2 of Appendix A) that "The reactor internals HCLPF capacities are probably larger than 0.3g based on Electric Power Research Institute (EPRI) NP-6041, Appendix A and limited available information. Further research is not expected to identify any significant vulnerabilities." Again, Table 5-3 of IPEEE Appendix A states that the HCLPF capacity of reactor internals is "probably larger than the RLE."
These statements imply that reactor internals could not be screened out by the SRT. [According to seismic margins analysis (SMA) guidance, to screen an item, the SRT should be confident that the HCLPF capacity of the item exceeds the chosen seismic margins earthquake (SME).] Hence, please clarify: Based on its combined experience and judgment, and based on available information, does the SRT find that reactor internals can indeed be screened out? Please provide your justification for making a definitive decision. If reactor internals can not be screened out, please report the results of your relevant HCLPF calculations.
- 9.
Please provide fragility calculations, completed screening evaluation work sheets (SEWSs), walkdown notes/checklists and photographs for the following components:
- Motor-Operated Valves RHR-750 and RHR-751
- Diesel Fuel Oil Storage Tank, refueling water storage tank (RWST),
condensate storage tank (CST)
Service Water Pumps 125 VDC MCCs A & B
-4 A.3 Fire
- 1.
For inter-zone fire propagation, it is argued that the fire barriers are adequate and that fire compartment interaction analysis (FCIA) screening criteria are, therefore, satisfied. This line of thinking may be acceptable if there are no active fire barriers (e.g., dampers, louvers, or normally open doors).
It should be noted that the failure rate of such devices can be as high as 0.2 per demand. Please identify adjacent fire compartments linked with active fire barriers that contain cables and equipment from multiple safety trains. Assuming a failure rate of 0.2 per demand, please provide a discussion on inter-zone fire propagation and the risk significance of this phenomenon.
- 2.
FCIA should consider fire brigade accessing the fire area through adjacent fire zones that contain cables and equipment from an opposite safety train. Please provide fire scenarios that involve this situation, and describe how they have been considered in the IPEEE submittal.
- 3.
From the submittal, it is clear that consideration has been given to the possibility of interfacing LOCAs caused by hot-short failures in control cables. Hot shorts can also affect normal system operation by moving valves from their normal operating positions. This possibility is especially important when there are cross-ties between the redundant trains of a system. Please provide a discussion regarding the inclusion of the possibility of valves moving from their safe position as a result of hot shorts.
- 4.
Although there is only one nuclear unit at HBR2, from the IPEEE it can be deduced that there are shared elements (e.g., compartments and systems) between the-nuclear and the non-nuclear units.
Please provide a brief description of the shared areas, and especially of all shared systems and a discussion of how these instances have been modeled in the fire analysis.
- 5.
From the information provided in the submittal, it is difficult to deduce the system failures and plant damage states leading to core damage for various fire scenarios. Please provide a listing of dominant core damage sequences in terms of system-train failures, and other pertinent information, for the most significant fire scenarios
- 6.
The probability of operator failure in achieving hot shutdown, after evacuation of the control room, has not been presented in the IPEEE submittal.
However, from the discussions provided in Section 4.6.1, it appears that a probability value of 0.01 has been used. This value is significantly smaller that the HEP typically used for analysis of other power plants,even for those having a centralized alternate shutdown panel.
Please explain and justify the derivation and source of the 0.01 HEP considering the fact that the alternate shutdown system is a collection of panels dispersed throughout the plant.
-5
- 7.
The submittal's discussion regarding human error probabilities (HEPs) mentions that the HEPs used in the IPE model have been modified to take into account fire conditions. No quantitative examples have been provided that display how the HEPs have been modified and how dependencies among different HEPs have been modeled. Please provide examples, using dominant fire scenarios, to show how the HEPs have been modified and how dependencies have been included in the calculations of fire core damage frequency.
- 8.
The submittal does not provide any quantitative example for the evaluation of fire detection and suppression effectiveness, for the cases where a comparison is made between the time for fire propagation and the time for fire detection and suppression. Please provide a quantitative example that clearly displays how fire suppression system effectiveness is modeled.
- 9.
In Table 4.6-5, the third column provides notations that are not explained. Please provide an explanation of the notations under the column heading of "Transient / LOCA."
- 10.
Please provide the correct conditional core damage frequency for A20-12 in Table 4.6-5.
- 11.
On page 4-10, in Section 4.1.3.1.2 of the submittal report, one of the two screening questions is stated as "there are Appendix R equipment in the compartment."
Does this mean that only Appendix R equipment and cables are addressed? Has the effect of non-Appendix R cables and equipment been modeled?
Please provide information regarding cables and equipment that have been considered in addition to those associated with Appendix R.
- 12.
Transient fires are not included in the cable spreading room fire analysis (as evidenced in Table 4.6-2e). It is NRC staff's position that administrative controls are insufficient basis for eliminating transient combustible fire from consideration. Therefore, please provide a discussion on the impact on CDF of including transient combustible fires for all those fire areas for which transient fires had not been considered.
- 13.
On page 4-10 in Section 4.1.2.5 of the submittal, it is stated that the coated non-qualified cables are assumed to be equivalent to IEEE-383 qualified cables. Please provide the technical basis or applicable test data supporting this assumption and that coated cables will not support self-ignited fire. In addition, provide an assessment of the impact of self-ignited cable fire on core damage frequency.
- 14.
On page 4-19, in Section 4.3.2 of the submittal, it is stated that "(ii) the fire compartment is assumed to be unventilated....".
Please provide information regarding how this was applied in the COMPBRN runs.
It should be noted that COMPBRN incorporates a ventilation controlled fire model that in the case of no ventilation flow will predict pre mature fire extinguishment.
-6
- 15.
On page 4-28, it is stated that manual fire suppression is considered as equal to fire brigade response time. It should be noted that fire brigade response time is based on arrival time during the drills. This does not take into account the time that it takes to extinguish the fire after arrival.
Please provide an assessment of the impact of added extinguishment time on core damage frequency.
- 16.
Possibility of equipment failure from suppression system activation has been addressed but no details are provided. Please provide a discussion of whether the possibility of diesel generator failure from CO2 actuation exists. This may occur from CO discharge into a diesel room
- that takes suction from inside the room itself. Also, potential for CO damage to cables and cable trays located in a cable vault may exist. I inadvertent CO2 actuation occurs, cable trays can potentially structurally fail from an over dump of CO2. Please provide either the justification that the above failure mechanisms cannot occur at HBR2 or the CDF analysis from inadvertent CO2 actuation.
- 17.
In Section 4.6.3 of the submittal it is stated that Nuclear Safety Analysis Center (NSAC) 181 is used. NRC staff has not approved the use of the various models presented in NSAC 181 for use in the preparation of IPEEE submittals. For example, the failure probability for manual suppression of control room fires may yield an optimistic value. Please provide a discussion of the specific data and quantification models adopted from NSAC 181.
In addition, provide an assessment on the impact on CDF if models and data other than those from NSAC 181 are employed.
A.4 High Winds, Floods, and Other Accidents (HFOs)
- 1.
Please discuss the effects of wind-induced loss of offsite power (excluding effects of wind-induced missiles, but including other potential wind effects) on the conditional probability of core damage.
(Address any hindrances to recovery actions due to the extreme winds.)
Explain how the conditional probability of core damage, given wind induced loss of offsite power, differs from the corresponding probability developed in the IPE.
- 2.
Please provide figures which indicate (in plan and elevation) the configuration of service water intake and discharge facilities and explain the potential impacts of floods on these facilities. Also, describe the potential effects of flooding on operation of service water pumps, diesel fuel oil transfer pumps, deep well pumps, and any other important, potentially exposed components.
- 3.
Please report and justify the conditional probability of core damage due to loss of service water, fire water, and other effects, resulting from failure of Lake Robinson Dam (and the consequential loss of reservoir impoundment).
Explain how the conditional probability of core damage, given failure of Lake Robinson Dam, differs from the IPE finding for conditional probability of core damage given loss of service water.
- 4.
Please describe the procedures/methods followed in your HFO walkdown.
Please indicate how many walkdowns were performed, and describe (for each walkdown) the number and expertise of walkdown participants.
Please identify all areas/components/systems included in the walkdowns.