ML17265A551
| ML17265A551 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/25/1999 |
| From: | Vissing G NRC (Affiliation Not Assigned) |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TAC-M83624, NUDOCS 9903020221 | |
| Download: ML17265A551 (20) | |
Text
, Dr. Robert C. Mecredy ~
Vice President, Nuclear rations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649 Feb I y 25, 1999
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONON THE R.E. GINNA NUCLEAR POWER PLANT IPEEE SUBMITTAL(TAC NO. M83624)
Dear Dr. Mecredy:
Based on our ongoing review of the Ginna Individual Plant Examination of External Events (IPEEE) submittal, we have developed the enclosed request for additional information (RAI).
The RAls are related to the seismic and fire analyses in the IPEEE. There are no RAls related to the high winds, floods, and other external events.
The RAls were developed by our contractors, Brookhaven National Laboratories and Sandia National Laboratories, and were reviewed by the "Senior Review Board" (SRB). The SRB is comprised of RES and NRR staff and RES consultants (Sandia National Laboratories) with probabilistic risk assessment expertise for external events.
The NRC review of the internal flooding analysis for Ginna, that is now due to be submitted in February, is being reviewed separately from the IPEEE review. Therefore, any RAls associated with the internal flooding review willbe transmitted separately.
A draft of this RAI was e-mailed to your staff on February 12, 1999, for the purpose of making a determination as to when you would be able to respond to the RAI. This was discussed with your staff on February 22, 1999, and considering the upcoming refueling outage and other activities within your organization, it was agreed that you could respond to this request within 120 days of the date of this letter.
Sincerely, ORIGINAL SIGNED BY:
Guy S. Vissing, Senior Project Manager Project Directorate l-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
RAls ccw/encl: See next page DISTRIBUTION:
Docket File S. Little PUBLIC G. Vissing PDI-1 R/F OGC J. Zwolinski ACRs S. Bajwa A. Rubin r ~ 9."! g 4 lc)
A. Blough, Region I
mme~~>>
DOCUMENT NAME: G:iGINNAlRAI83624.WPD t
I p(v'l To receive attachment a co of this docunent indicate in the box:
"C" = Co without attachment enclosure "E" = Co with enclosure "H" = Mo co OFFICE PH:PDI-1 LA:PDI-1 D:PDI-1 HAHE Gvissingt lcc SLitt e Sga wa DATE 02 99 02 99 02 99 02 99 9'P03020221.
990225 PDR ADOCK OS000244 P
PDR OFFICIAL RECORD COPY
(,
e
,I
Dr. Robert C. Mecredy Rochester Gas and Electric Company R.E. Ginna Nuclear Power Plant CC:
Peter D. Drysdale, Sr. Resident Inspector R.E. Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Aliendale Road King of Prussia, PA 19406 Mr. F. WilliamValentino, President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Charles Donaldson, Esquire Assistant Attorney General New York Department of Lalw 120 Broadway New York, NY 10271 Nicholas S. Reynolds Winston 8 Strawn 1400 S Street N.W.
Washington, DC 20005-3502 Ms. Thelma Wideman, Director Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 West FaIls Road, Room 11 Rochester, NY 14620 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 Febr vary 25, 1999
~
~Pg RE00 0
Cy
~g 0
lg gO Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONON THE R.E. GINNA NUCLEAR POWER PLANT IPEEE SUBMITTAL(TAC NO. M83624)
Dear Dr. Mecredy:
Based on our ongoing review of the Ginna Individual Plant Examination of External Events (IPEEE) submittal, we have developed the enclosed requests for additional information (RAI).
The RAls are related to the seismic and fire analyses in the IPEEE. There are no RAls related to the high winds, floods, and other external events.
The RAls were developed by our
, contractors, Brookhaven National Laboratories and Sandia National Laboratories, and were reviewed by the "Senior Review Board" (SRB). The SRB is comprised of RES and NRR staff and RES consultants (Sandia National Laboratories) with probabilistic risk assessment expertise for external events.
The NRC review of the internal flooding analysis for Ginna, that is now due to be submitted in February, is being reviewed separately from the IPEEE review. Therefore, any RAls associated with the internal flooding review willbe transmitted separately.
A draff of this RAI was e-mailed to your staff on February 12, 1999, for the purpose of making a determination as to when you would be able to respond to the RAI. This was discussed with your staff on February 22, 1999, and considering the upcoming refueling outage and other activities within your organization, it was agreed that you could respond to this request within 120 days of the date of this letter.
Sincerely, Guy S. Vissing, Senior Project Manager Project Directorate I-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
RAls cc w/encl: See next page
~ ~
I
RE UEST FOR ADDITIONALINFORMATIONON THE R.E. GINNA NUCLEAR POWER PLANT IPEEE SUBMITTAL A.
Seismic The Ginna IPEEE submittal (Section 2 System Analysis) states that "Rather than performing a system analysis to define the scope of the review, this assessment includes all safety related components in the plant." Success path logical diagrams (SPLDs) and success paths as described in EPRI NP041 are not developed in the Ginna IPEEE.
Of the equipment included for assessment, fiftytwo (52) items were identified in the.
IPEEE process as "SMA Mechanical and Electrical Equipment Outliers."
However, no further actions are planned for 38 of the 52 outliers that are not related to A-46 closeout.
These 38 equipment outliers include those related to the safety injection (SI) system, the residual heat removal (RHR) system, and the component cooling water (CCW) system.
The plant's ability to safely shutdown under small loss-of-coolant accident (LOCA) conditions may be questionable without these systems.
Although the inclusion of all safety-related components seems to provide a bigger equipment list, it cannot provide the valuable insights a structured evaluation using success paths can provide.
For example, the impact of the failure of the above systems cannot be assessed without the help of defined success paths.
As stated in EPRI NP<041, a primary purpose of a seismic margin review is to identify any "weak links" which might limitthe plant's shutdown capability. The approach used in a margin review is to demonstrate that a reliable operational sequence exists to shutdown and maintain the plant in a safe shutdown condition,for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. To achieve this purpose, the establishment of a preferred and an alternative success path, with one capable of handling a small LOCA, are recommended in the Electric Power Research Institute (EPRI) margin method.
According to NUREG-1407 (Section 3.2.5), the NRC panel that reviewed the EPRI methodology recommended that "a reasonably complete set of potential success paths be set down initially, rather than a very small number, since limiting the number of success paths too quickly can prevent the identification of some plant-level high confidence of low probability of failure (HCLPF) insights, and can mask plant differences regarding defense-in-depth."
The development of success paths for plant evaluation is therefore an important part of the margin method. According to EPRI NP-6041, the selection of the success paths is a joint responsibility of the plant operators, system engineers, and seismic capability engineers.
The Ginna USI A-46 report, that was submitted as part of the IPEEE submittal, shows a single success path.
It includes the use of the chemical and volume control system (CVCS) for reactor coolant system (RCS) inventory control and secondary auxiliary feedwater (SAFW) for decay heat removal. Since this path may not be able to handle a small LOCA condition, another success path for a small LOCA condition needs to be established.
Enclosure
~ ~
.(a)
Please followthe method described in EPRI NP-6041 to develop a success path that can handle a small LOCA. Please identify the front line and support systems required forthe four safety functions identified ln the EPRI report (l.e., reactor reactivity control, RCS pressure control, RCS inventory control, and decay heat removal). Please Identifyand discuss any "weak links"in the success path (e.g., the impact ofthe Mechanical and Electrical Equipment Outliers, shown in Table 3 ofthe IPEEE submittal, on the success paths)., Please describe the analysis performed in the Systematic Evaluation Program (SEP) forthe systems that are weak links in the success path.
(b)
.Please address the non-seismic failure and human action issues associated with the success paths as requested in EPRI NP-6041 and Section 3.2.5.8 ofNUREG-1407.
For the success path selected for USI A-46, the auxiliary feedwater (AFW) source seems to be from the service water (SW) system.
It is not clear whether there are
'ondensate storage tanks associated with the AFW or the SAFW system and whether these tanks are included in the equipment list for seismic assessment.
Please describe the seismic capabilities ofthese tanks ifthey are Included
, in the list; or, ifthey are notincluded in the equipment list, describe the equipment, procedures, and operator actions required to assure proper isolation ofthese systems such that feedwateris not lost through failed tanks.
The Ginna IPEEE submittal described the seismic evaluation of the Ginna IPEEE program as being consistent with a reduced-scope evaluation via the Seismic Margins Methodology, in that ail safety-related equipment was evaluated using the Gilbert SSE floor response spectra which were generated based on the 0.2g Regulatory Guide (RG) 1.60 spectrum ground input.
In cases where an item of safety-related equipment was found to be rigid, the calculated factor of safety from its anchorage evaluation was reduced by 1.5 (equivalent to increasing the zero period acceleration (ZPA) from 0.2g to 0.3g). Furthermore, the submittal compared the 0.2g RG 1.60 spectrum with the 0.3g NUREG/CR-0098 spectrum for rock (the IPEEE seismic input designated for Ginna) and characterized the two spectra as being similar and differing by less than 20% for frequencies below 10 Hz. However, for frequencies higher than 10 Hz, the comparison shows that the 0.2g RG 1.60 spectrum decreases more rapidly relative to the 0.3g NUREG/CR-0098 spectrum with a difference between the two as much as 50 To assist our evaluation, please provide the followinginformation:
(a)
A list ofall equipment which was evaluated forseismic adequacy using the ZPA ofthe floorresponse spectra (i.e., assumed to be rigid) and reduced by 1.5 in the calculated factor ofsafety forIPEEE screening.
(b)
A detailed quantitative discussion on how the seismic capacities (in terms ofHCLPF) for the safety-related equipment items with frequencies between 10 Hz and the ZPA were determined.
( ~
B.
Fire It is unclear how room-to-room fire scenarios were treated in the Ginna IPEEE analysis.
Section 3.1 of the submittal indicates that fire areas were screened individually according to the FIVE (Reference 1) criteria (i.e, the area contains no Appendix R equipment, and a fire in the area would not cause a demand for safe shutdown).
This section also indicates that fire compartments were further screened ifthey had no credible potential for fire spreading to other fire compartments.
While the submittal indicates that the qualitative screening conformed with Phase I of the FIVE methodology, fire propagation potential between fire compartments is reviewed in the FIVE methodology during the Fire Compartment Interaction Analysis (FCIA)~ The submittal does not indicate that the FIVE FCIA criteria were used to determine iffire propagation between fire zones was possible.
The submittal does provide qualitative criteria for grouping plant locations, but it is
. unclear that this approach has adequately treated room-to-room fire scenarios.
One of the cited criteria indicates that plant locations were grouped together when a physical barrier (not necessarily fire-rated) separates the subject locations from the rest of the plant and there is a significant time delay for fire propagation from the subject locations to other adjacent locations.
In addition, the submittal indicates that one consideration for determining the importance of a location was whether it contains a sufficient amount of combustible material that, ifignited, could potentially propagate to adjacent zones.
The basis for making these judgments is not provided.
Further, it is not clear that the analysis has adequately considered the potential that active fire barrier elements (e.g.,
normally open fire doors, ventilation dampers, etc) might fail to activate, or that passive fire barriers (e.g., various fire barriers both rated and unrated and barrier penetration seals) might be challenged by local concentrations of flammable materials.
Finally, it is not clear that the analysis has considered the potential for the spread of smoke and heat from one compartment to another in addition to the consideration of actual fire spread.
Please clarify the bases used to assess the potential for cross-zone spreading of fire, heat,'nd smoke.
Please provide an analysis forall fire areas ofthe effect on fire-induced core damage frequency (CDF) that includes consideration ofthe failure potential ofactive barrier components such as doors and dampers.
Please provide an analysis ofthe potential for cross-zone fire propagation forhigh hazard areas such as the turbine building, diesel generator room, switchgear rooms, and lube oilstorage areas that includes consideration ofthe potential to challenge passive fire barrier elements.
The Ginna fire IPEEE submittal indicates that cable wrap was credited in the quantitative assessments, but the treatment that was given is not clear and may have-led to "double counting" of suppression effectiveness.
It is also not clear ifthis approach was used in the screening analyses as well. Section 3.3.4 (Assumption 2) indicates that "a probability of 0.15 was assigned to the failure of cable wrap to account for the probability that a fire is not suppressed within the one hour time frame associated with the fire rating of the cable wrap." This description implies that the
V
'odeling of the cable wrap failure implicitlycredits fire suppression in the quantitative screening of fire zones that contain the wrap.
In the detailed fire PRA evaluations, an additional independerit credit for fire suppression efforts would result in double counting suppression efforts.
Please Indicate Iffire suppression was credited in fire scenarios where cable wrap was Independently credited as protecting critical cables. Ifthere are any such scenarios, reevaluate the core damage frequency either (1) assuming that the 0.15 barrier faIlure probability fullycredits suppression, or (2) assuming an independent suppression credit and that the cable wrap fails with a probability of 1.0 forall fires lasting greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The 19 fire zones remaining after the qualitative and quantitative screening phases of the Ginna fire assessment were subjected to further detailed evaluation including the analysis of fire propagation and suppression.
Actual fire modeling using the FIVE methodology or other techniques was not performed.
Instead, probabilities for fire propagation were assigned based primarily upon physical separation of equipment.
However, it is not clear upon what basis these judgments were made.
For example, in the analysis of the AuxiliaryBuilding Operating Level (ABO, Page 3-14)
"a 0.01 probability was assigned that a fire occurring in the vicinitywould disable both CCW pumps, and a 0.99 probability was assigned that one CCW pump, in addition to one AC power electrical division, would be disabled."
It is not clear that this assumption is well founded. The submittal states the two CCW pumps are located within 9 feet of each other, and that there are cables in conduits in the area.
Presumably, loss of the cables may lead to loss of the second pump.
Further, the fact that the fire source appears to be the CCW pumps themselves, there would be a significant potential for large fires to occur. Given a large fire, 9 feet of spatial separation would likely not prevent thermal damage to the second pump, its power cables, or its control cables.
This is one speciTic example where the assumed damage probabilities may be optimistic.
For other fire areas, from the description of fire scenarios presented in the submittal, it appears that fires that are not suppressed were assumed to damage all the equipment in a fire zone.
Ifthe fire was suppressed, some level of damage was assumed to occur, and it appears that in many cases suppressed fires were assumed to damage just one electrical division. The submittal states that this is conservative since for many fire scenarios only a portion of components relying on the electrical division would be disabled.
In general, this approach is acceptable ifthe critical set of components and cables are relatively far apart, and therefore, itwilltake a long time for a fire to damage them. On the other hand, ifthe key cables and components are close together, critical damage may occur before successful suppression.
Please provide a general description ofhow the fire damage assumed for each of the fire scenarios considered in the detailed analyses was determined.
Include a description ofthe criteria used to determine the radius ofthe damage caused by suppressed fires and the tIming ofcomponent damage.
Also indicate to what extent the actual location ofcritical cables and components was verified and considered in the damage assessment.
For suppressed fires, indicate ifany time was assumed forthe suppression ofthe fire and ifthis tIme impacted the assumed damage.
Transient combustible fires were not analyzed separately in the Ginna fire assessment.
The submittal states that during the development of the fire frequencies, transient combustibles were grouped with the type of component that was primarily damaged. by or exposed to the fire. Thus, the submittal states that the impact and consequence's of transient combustible fires are accounted for in the modeled component fires, and no separate evaluation of transient fires was necessary.
Based on the limited descriptIon in the submittal, itis unclear ifthe methodology accounts for.transient fires at all critIcal locations In the plant. Specifically, itis unclear ifa portion ofthe frequency oftransIent fires was accounted forin the evaluation ofcable fires. Please provide a more detailed description ofhow the transient fire frequency was included in the analysis Including a descrIption of how the frequency was partitioned and the types ofcomponents assumed damaged or exposed to the transient fires. Ifcables were not in the list of components damaged or exposed to the transient fires, provide a separate assessment oftransient-induced fire scenarios involving cables in the unscreened fire zones containing cables.
5.
Two fire zones (IBN-1 and IBS-1) are identified in the submittal which are not listed as being either qualitatively or quantitatively screened.
Since the results of a detailed fire PRA evaluation for these fire zones are also not given in the submittal, the importance of these two fire zones is unknown.
Please indicate ifthese two fire zones were screened or subjected to a detailed fire scenario evaluation.
Provide descriptions forany fire scenarios modeled for these fire zones and list the estimated core damage frequencies.
NUREG-1407 (Reference 2), Section 4.2 and Appendix C, and GL 88-20, Supplement 4 (Reference 3), request that documentation be submitted with the IPEEE submittal with regard to the FRSS (Reference 4) issues, including the basis and assumptions used to address these issues, and a discussion of the findings and conclusions.
NUREG-1407 also requests that evaluation results and potential improvements be specifically highlighted. Control system interactions involving a combination of fire-induced failures and high probability random equipment failures were identified in the FRSS as potential contributors to fire risk.
The issue of control systems interactions is associated primarily with the potential that a fire in the plant (e.g., the MCR) might lead to potential control systems vulnerabilities.
Given a fire in the plant, the likely sources of control systems interactions are between the control room, the remote shutdown panel, and shutdown systems.
Specific areas that have been identified as requiring attention in the resolution of this issue include:
(a)
Electrical independence of the remote shutdown control systems:
The primary concern of control systems interactions occurs at plants that do not provide
independent remote shutdown control systems.
The electrical independence of the remote shutdown panel and the evaluation of the level of indication and control of remote shutdown control and monitoring circuits need to be assessed.
(b)
Loss of control equipment or power before transfer: The potential for loss of control power for certain control circuits as a result of hot shorts and/or blown fuses before transferring control from the MCR to remote shutdown locations needs to be assessed.
(c)
(d)
Spurious actuation of components leading to component damage, loss-of-coolant accident (LOCA), or interfacing systems LOCA: The spurious actuation of one or more safety-related to safe-shutdown-related components as a result of fire-induced cable faults, hot shorts, or component failures leading to
'omponent damage, LOCA, or interfacing systems LOCA, prior to taking control from the remote shutdown panel, needs to be assessed.
This assessment also needs to include the spurious starting and running of pumps as well as the spurious repositioning of valves.
It does appear that the assessment has included this aspect of the concern.
Total loss of system function: The potential for total loss of system function as a result of fire-induced redundant component failures or electrical distribution system (power source) failure needs to be addressed.
Please describe your remote shutdown capability, including the nature and location ofthe shutdown station(s), as well as the types ofcontrol actions which can be taken from the remote panel(s).
Describe how your procedures provide fortransfer ofcontrol to the remote shutdown station(s).
Provide an evaluation of whether loss ofcontrol power could occur prior to transferring control to the remote shutdown location and identify the risk contribution ofthese types of failures (ifthese failures are screened, please provide the basis forthe screening).
The submittal indicates that the "automatic fire detection and suppression systems at Ginna were assumed to be installed per design specifications, following the National Fire Protection Association (NFPA) and NRC guidelines." The submittal also states that fire protection systems were assumed to be maintained regularly and that generic failure rates were used in the analysis.
It is not clear that these assumptions were verified.
Please verifythat the automatic fire suppression systems at Ginna are, in fact, designed and maintained according to NFPA standards.
The Ginna fire IPEEE submittal identiTies one plant improvement planned for implementation and five additional plant modiTications that were being considered.
It is not clear ifthese improvements and modiTication were credited in the analysis, and whether or not the changes have been, or willbe, implemented.
V~
f Please provide the current status ofthese planned and proposed plant modifications and indicate whether or not the changes have been credited in the analysis.
Both manual actuation of automatic fire suppression systems and manual fire suppression were modeled for selected fire scenarios in the Ginna fire assessment.
In some scenarios', failure of automatic fire suppression, failure to manually initiate automatic suppression systems, and failure to manually suppress the fire were modeled.
The submittal does not address the potential for dependent failure of both automatic and manual suppression systems (e.g., common mode failures related to a common water source).
Please indicate ifdependent failures between the automatic and manual suppression systems were considered in the assignment ofthe suppression probabilities. Also indicateif dependencies between the failure ofpersonnel to manually initiate an automatic suppression system and failure ofpersonnel to manually suppress a fire were considered.
Indicate ifdependent failures that would cause the failure ofan automatic suppression system to actuate and also prevent manually initiating the system were considered in the analysis.
References 1.
EPRI, "Fire-Induced Vulnerability Evaluation (FIVE)," EPRI TR-100370, April 1992 2.
J. Chen, et al., "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," NUREG-1407, United States Nuclear Regulatory Commission, June 1991.
3.
"Independent Plant Examination for External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f)," Generic Letter 88-20, Supplement No. 4, United States Nuclear Regulatory Commission, June 1991.
4.
J. Lambright, et al., "Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, Including Previously Unaddressed Issues," NUREG/CR-5508, prepared for the United States Nuclear Regulatory Commission, January 1989.
C al
> '4+
Z)g