ML091530274
| ML091530274 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/01/2009 |
| From: | Calvert Cliffs |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML091530275 | List: |
| References | |
| ISI-04-04 | |
| Download: ML091530274 (6) | |
Text
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 Calvert Cliffs Nuclear Power Plant, Inc.
June 1, 2009
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 NRC RAI No. 1:
Code Case N-716 relies on American Society of Mechanical Engineers (ASME) flooding analysis.
Enclosure (1) of the submittal dated December 29, 2008 states reviews were conducted by SAIC Inc. of the Calvert Cliffs Probabilistic Risk Assessment (PRA) flooding model as part of an update to Electric Power Research Institute guidance. This flooding model was determined to be of high quality. Describe what Standard or other documents SAIC used to review the Calvert Cliffs PRA flooding model.
CCNPP Response:
The SAIC review consisted of their review of a previous Calvert Cliffs conducted self-assessment and a review of our current documentation to determine which upgrades were needed for compliance with ASME RA-Sb-2005. The SAIC review was informal in the extent that a formal gap analysis was not prepared however, SAICs observations were presented to PRA supervision and subsequently added to the previous self-assessments.
Following their review, SAIC was contracted to upgrade Calvert Cliffs internal flood analyses to meet ASME RA-Sb-2005 requirements for internal floods. Calvert Cliffs guidance to them is to upgrade to a Category II flood model per the requirements for all Internal Flood High Level and Supporting Requirements for Internal Floods from ASME RA-Sb-2005. This effort is currently in progress.
Separately, Calvert Cliffs has verified our flood analyses against Regulatory Guide (RG) 1.200, Revision 1, Appendix A requirements. This subsequent review identified no significant additional items that need to be addressed. Additional discussion is included in responses 3, 5, and 6 of this document regarding the screening process, flood induced failures and plant specific experience which are noted in RG 1.200, Appendix A.
NRC RAI No. 2:
Please provide a description of outstanding issues proposed by SAICs evaluation of Calvert Cliffs PRA flooding model and describe how these issues have been resolved.
CCNPP Response:
The major outstanding issues identified by SAICs review include:
a) Documentation -- Documentation is out-of-date and does not have a road map to show how each ASME element is addressed.
SAIC is currently developing an internal floods notebook which will provide a centralized location for documentation of compliance to ASME RA-Sb-2005. Development of such an internal floods notebook meets the requirements of Table 4.5.7-2(f) of ASME RA-Sb-2005. Included in this effort are updating the documentation of all flood sources, flood screening, human actions for recovery, flood initiating event frequencies, and preliminary quantification.
b) Walkdowns -- New walkdowns should be conducted to ensure flood model reflects modifications installed in the plant after original walkdowns were performed.
Subsequent to SAICs review, walkdowns were re-performed by SAIC and site personnel to ensure the as-built plant is addressed in the flood model. This action addresses Supporting Requirements (SR)
IF-A4, IF-B3a, IF-C2, IF-C9, IF-E8 of ASME RA-Sb-2005.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 c) Flood Initiator Grouping -- Review and revise potentially overly conservative subsuming of flood initiators within the model.
Supporting Requirements IF-D3 notes Avoid subsuming scenarios into a group unless the impacts are comparable and it is demonstrated that such grouping does not impact accident sequences.
There are some floods in the original analysis which include multiple sources which may be overly conservative in the binning. One example includes the Truck Bay area on the 45 of the Auxiliary Building. This flood conservatively fails all of the potential flood sources in this area for this initiator. One of these sources is Component Cooling Water whose failure causes a reactor coolant pump seal challenge due to loss of cooling and is assumed for all floods in this area. This causes overly conservative results for this flood. However, this conservative binning of flood initiators does not result in challenging the 1E-6 criteria so revising the binning to reduce this conservatism is not necessary for this application.
d) Human Action Recoveries -- Update flood discovery and recovery probabilities to reflect latest practices and procedures.
Operator interviews have been performed to update flood discovery and recovery probabilities based on the latest plant practices and staffing. Based on these interviews, conservative screening values for recovery likelihoods were developed. See Response No. 5 for more discussion on human intervention.
e) Quantification -- Update flood basic events into the CAFTA fault trees.
Updated flood basic event impacts are being developed for the CAFTA fault trees. Preliminary runs using the current CAFTA model will be performed to evaluate the accident sequences related to flood (this step meets the requirements of SR IF-E). Final quantification will be performed via the overall internal events upgrade which is currently in process.
The scope of the work in regards to the above issues is not expected to significantly increase the importance of any piping segment to the point where it would challenge the Code Case N-716 1E-06 threshold.
NRC RAI No. 3:
The supporting requirement (SR), IF-C3, in ASME PRA Standard RA-Sb-2005 identifies the failure mechanisms that shall be evaluated to determine the susceptibility of each safety-related structure, system, and component (SSC) in a flood area to flood-induced failures. Capability Category II identifies failure by submergence and spray as requiring detailed analysis. Capability Category III includes jet impingement, pipe whip, and humidity, condensation, and temperature concerns. Risk informed inservice inspection (RI-ISI) requires that all SSC failures induced by a pipe break be considered. Please demonstrate that all SSC failures that are induced by a pipe break are adequately addressed in your analysis.
CCNPP Response:
The Calvert flood model has a detailed analysis of submergence levels and propagation paths. Unless specifically evaluated, all components within a flood source room are assumed failed via spray or other impacts such as those noted in Category III for SR IF-C3. While this method does not completely address Category III of SR IF-C3, it is conservative when determining whether a flood could exceed a core damage impact of 1E-06 per Code Case N-716.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 Calvert Cliffs original flood analysis did not directly address the impacts of high energy piping as the only piping that met those criteria was main steam and main feedwater (MFW). These were considered bounded by the internal events initiators. The only room of potential concern where additional failures may have needed to be considered was the main steam isolation valve (MSIV) room. The internal events initiators do not consider the pipe whip and steam impingement failure mechanisms on SSCs within the room. To address this, we evaluated the high energy lines in the MSIV rooms.
The feedwater and main steam lines in the MSIV rooms are encapsulated at the areas of higher stress.
The encapsulation is a 1.125 inch thick steel sleeve surrounding the pipe. The encapsulation provides a controlled and balanced relief of the high energy fluid in the event of a break at one of the encapsulated portions, by venting the steam out both ends of the encapsulation. The encapsulation directs the steam so as not to impinge on other important components. This encapsulation also contains restraints inside the encapsulation that limit the movement of the pipe, should a break occur there. In addition, pipe whip restraints on the outside of the feedwater and main steam lines in the MSIV room restrict the movement of the pipe in the event of a break. Therefore, the pipe whip and jet impingement failure mechanisms are not considered to have a significant effect on other components in the MSIV room.
The impacts of humidity and temperature on components in the MSIV room were also considered for impacts in these initiators. The main components of concern are the auxiliary feedwater steam admission valves, steam admission bypass valves and the MFW isolation motor-operated valves. However these safety-related components are Quality Category 1M and are designed to operate in a high energy line break environment.
NRC RAI No. 4:
Were new examination locations identified, if so, were the new examination locations included in the change in risk estimate? Using an upper-bound estimate for new locations would be non-conservative.
Please demonstrate that this non-conservative approach, if corrected, would not exceed the delta risk guidelines.
CCNPP Response:
New examination locations were identified and were included in the change in risk estimate. The inspection selections for the original ASME Section XI program, the proposed Code Case N-716 program and the difference between those selections are contained in Tables 3.4-1 and 3.4-2 under the columns entitled SXI, RIS_B and Delta respectively. The risk impact analysis included changes made to the original ASME Section XI inspections as a result of implementing Code Case N-716 and the results are displayed in the Delta column of Tables 3.4-1 and 3.4-2 as either no change (represented by 0), an increase (represented by a positive number) or a decrease (represented by a negative number). A risk impact sensitivity calculation was also performed, with lower-bound conditional core damage probabilities used whenever new examination locations were identified (represented by a positive number in the delta column). The results of this sensitivity study meet Code Case N-716 acceptance criteria.
NRC RAI No. 5:
The SR, IF-C6, permits screening out of flood areas based on, in part, the success of human actions to isolate and terminate the flood. The endorsed RI-ISI methods require determination of the flood scenario with and without human intervention which corresponds to the Capability Category III, i.e. scenarios are not screened out based on human actions. Therefore a Category III analysis would be acceptable and is 3
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 preferable. To provide confidence that scenarios that might exceed the quantitative Core Damage Frequency and Large Early Release Frequency guideline are identified, please describe how credit is given to human actions.
CCNPP Response:
The original flood analysis performs a screening review that generally meets Category III criteria with the following exceptions where operator actions are credited.
- Fire Protection Floods in the -10/-15 hallway of the Auxiliary Building (Rooms 100 and 103)
- Fire Protection Floods in Emergency Core Cooling System (ECCS) Pump Rooms 11 or 12 (21 and 22 for Unit 2)
- Component Cooling Water Floods in Room 100, -10 Hallway of the Auxiliary Bldg Although these do not meet Category III criteria for IF-C6, they are not risk significant. The basis for this is documented below:
a) Fire protection floods (-10/-15 Hallway) or other areas that do not directly impact mitigating components in the Auxiliary Building where flow is unimpeded to the Auxiliary Building lower levels. Design (800 gpm) fire protection flood flow rates will cause a forced shutdown due to loss of the charging pumps at approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after flood initiation. Assuming this was to occur, the conditional core damage probability would be less than 7E-06. Alarms in the Control Room indicating that the fire protection pumps are running will provide clear indication of a potential flood or fire. The updated likelihood of failure for identification and recovery from this flood is 7E-03, based on operator interviews. Thus, the overall core damage frequency impact is approximately 5E-08 if the flood was assumed to occur once annually. The frequency, if calculated for the flood, would be orders of magnitude lower than one for a large flow rate. Thus, it does not challenge the Code Case N-716 criteria of 1E-06 core damage frequency per year.
b) Fire protection floods in ECCS Pump Rooms - A fire protection flood in these rooms would fail the ECCS pumps in the room. Assuming design fire protection flow rates, the analysis notes that it would take over 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before a plant trip or forced shutdown would occur. The failure likelihood of identification and recovery is less than 1E-03. The conditional core damage probability if this flood was not recovered prior to a loss of the charging system is less than 1E-05. Thus, even if one flood is assumed to occur annually the core damage frequency impact would be 1E-08. Even if no recoveries are ever assumed, the large flow pipe break frequency would have to be as high as 1E-01 to exceed the Code Case N-716 criteria. Given the experience of Calvert Cliffs (> 65 years on line considering both Units) that this flood has not occurred, the frequency is well below this value. As a result this flood scenario does not challenge the Code Case N-716 threshold.
c) Component cooling water floods in Room 100, -10 Hallway of the Auxiliary Building - After the initial component cooling inventory is exhausted the makeup is limited to 160 gpm from demineralized water. The event would be assumed to cause a plant trip on loss of component cooling.
It would then require an additional 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before the failure of additional systems (charging system) would occur. For non-large loss-of-coolant accident events, loss of charging will not have a significant impact that late in the event. The charging systems primary functions modeled in the PRA are for once-through core cooling and anticipated transient without scram mitigation. Charging is not required for successful once-through core cooling or for anticipated transient without scram mitigation at the 5.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> point. Loss of charging in this flood scenario does not appear to have any impact greater than a loss of component cooling and is appropriately binned. Conditional core 4
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST ISI-04-04 damage probability for a loss of component cooling is approximately 1E-05. As a result this flood scenario does not challenge the Code Case N-716 threshold.
NRC RAI No. 6:
The SR, IF-D5a, addresses the development of flood initiating (pipe rupture) frequencies for use during the scenario development. RI-ISI is premised on inspecting locations with the highest risk, driven mostly by failure frequency. The plant specific information collected and used should include experience related to degradation mechanisms that could indicate increased likelihood of pipe failure at particular locations. Please describe how plant specific operating experience was used to identify experience related to degradation mechanisms and how this experience was incorporated into the development of pipe failure frequencies.
CCNPP Response:
A review of plant failure mechanisms was performed for this Code Case N-716 submittal. This included a review of condition reports in the sites Corrective Action Program and a review of the flow assisted corrosion program. No plant specific issues were noted within either program that would indicate there are piping locations which should use values above generic frequencies. The original and in-process analyses use generic pipe break frequencies. The review determined there were no cases where Calvert Cliffs would be an outlier and would require the use of more conservative frequencies. Given the review results, there is no apparent gap for meeting Category II/III for SR IF-D5a.
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