BSEP 09-0068, Submittal of Supporting Documentation for the July 28, 2009, Regulatory Conference Regarding Preliminary White Finding

Submittal of Supporting Documentation for the July 28, 2009, Regulatory Conference Regarding Preliminary White Finding

ML092120222
Person / Time
Site:	Brunswick
Issue date:	07/22/2009
From:	Waldrep B Carolina Power & Light Co, Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 09-0068, EA-09-121
Download: ML092120222 (29)

Text

~ Progress Energy Benjamin C.Waidrep ViceBrnsin NcPresident Brunswick Nuclear Plant Progress Energy Carolinas, Inc.

July 22, 2009 Serial: BSEP 09-0068 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Submittal of Supporting Documentation for the July 28, 2009, Regulatory Conference Regarding Preliminary White Finding

Reference:

Letter from Kriss M. Kennedy, Director, Division of Reactor Safety, U.S. NRC, to Benjamin C. Waldrep, Vice President, Carolina Power &

Light Company, Brunswick Steam Electric Plant, dated June 17, 2009 Ladies and Gentlemen:

On June 17, 2009, the NRC informed Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., of a preliminary white finding (i.e.,

EA-09-121), associated with the Brunswick Steam Electric Plant (BSEP), Units 1 and 2.

The finding documents a wiring error, identified on August 18, 2008, that impacted the ability of the Emergency Diesel Generators (EDGs) Nos. 2, 3, and 4 to perform their intended Alternative Safe Shutdown (ASSD) function. This condition did not affect the Technical Specification (TS) operability of the EDGs and they remained fully capable of performing their intended safety system functions. On June 18, 2009, CP&L requested a Regulatory Conference to discuss finding EA-09-121; which has been scheduled for July 28, 2009.

The purpose of this letter is to provide documentation supporting CP&L's conclusion that the proposed finding should be categorized as having very low safety significance (i.e.,

Green) versus low to moderate safety significance (i.e., White). This information is being submitted as requested by the June 17, 2009, referenced letter.

Enclosure 1 to this letter provides a summary of the Probabilistic Risk Assessment (PRA) approach used by CP&L to evaluate the impact, from a risk perspective, that the condition had on Units 1 and 2. Enclosure 1 also provides a detailed comparison of major differences between CP&L's and the NRC's evaluation methodologies, which lead to the differences in safety significance determined for the condition. Enclosure 2 provides P.O.Box 10429 Southport, NC 28461 ,4o0 T> 910.457.3698 / ,(Z

Document Control Desk BSEP 09-0068 / Page 2 supplemental details of differences between CP&L's and the NRC's evaluation methodologies.

No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Mr. Gene Atkinson, Supervisor - Licensing/Regulatory Programs, at (910) 457-2056.

Sincerely, Benjamin C. Wald ep MAT/mat

Enclosures:

1. Summary of Probabilistic Risk Assessment Evaluation

2. Supplemental Details of Major Differences between CP&L's Evaluation and the NRC's Evaluation

Document Control Desk BSEP 09-0068 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor McCree, Deputy Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Len Wert, Director, Division of Reactor Projects Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Kriss Kennedy, Director, Division of Reactor Safety Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Harold Christensen, Deputy Director, Division of Reactor Safety Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Ms. Rebecca Nease, Chief, Engineering Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Randy Musser, Chief, Reactor Projects Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931

Document Control Desk BSEP 09-0068 / Page 4 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. George MacDonald, Senior Reactor Analyst Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Reinaldo Rodriguez, Senior Reactor Inspector Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

BSEP 09-0068 Enclosure 1 Page 1 of 9 Summary of Probabilistic Risk Assessment Evaluation

Background

On August 18, 2008, during the performance of surveillance OPT-12.14.L, "Diesel Generator 4 Local Control Operability Test," Emergency Diesel Generator (EDG) 4 failed to start from the local control panel. The biennial (i.e., every 24 months) surveillance is a non-Technical Specification (TS) related test to demonstrate that control and indication for EDG 4, and the EDG 4 to Bus E4 output breaker, can be isolated from the Control Room and be controlled from the respective local control station, as required by the Alternative Safe Shutdown (ASSD) analysis.

Troubleshooting determined that the Lockout Control Relay (LOCR), installed in June 2007, was wired such that power was lost when the associated ASSD switch was in the LOCAL position, preventing EDG 4 from being reset so that it could be started locally. The plant modification was installed on all four EDGs, and it has been concluded that this condition impacted the ability of EDG Nos. 2, 3, and 4 to perform their intended ASSD function (i.e., local control of EDG 1 is not credited in the ASSD analysis). This condition did not affect the TS operability of the EDGs and they remained fully capable of performing their intended safety system functions.

A revised plant modification was implemented on all four EDGs that re-wired the LOCRs such that they do not lose power when the ASSD switches are operated. With completion of this activity on August 21, 2008, the affected EDGs were fully capable of performing their ASSD function. This event was reported in Licensee Event Report (LER) 1-2008-006, dated October 14, 2008.

Probabilistic Risk Assessment (PRA) Evaluation Summary Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., and the NRC's risk assessment focuses on fires in Alternate Safe Shutdown credited areas of which the Main Control Room and Cable Spread Room are the dominate scenarios. Fires in these areas could result in a Loss of Off-site power (LOOP) with auto-start of the EDGs and subsequent postulated implementation of the alternate safe shutdown procedures.

CP&L has reviewed the NRC's PRA evaluation included in the referenced June 17, 2009, letter.

Using the methodologies described in the attached evaluation, CP&L has determined that the total Core Damage Frequency (CDF) associated with this condition is 1.05E-07 for both Units 1 and 2, which should be categorized as having very low safety significance (i.e., Green). This differs with the NRC's evaluation, which used NRC Inspection Manual Chapter IMC 0609, "Significance Determination Process," resulting in a CDF of 7.63E-06, which is categorized as having low to moderate safety significance (i.e., White).

The NRC's Phase II analysis uses the guidance found in the Fire Protection Significance Determination Process (SDP) in IMC 0609. Phase III of the SDP allows for further refinement of critical inputs that go beyond the IMC 0609 guidelines such as NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed

BSEP 09-0068 Enclosure 1 Page 2 of 9 Methodology," as used by CP&L. Significant differences in the CP&L and NRC calculated CDF are a result of the following major contributors:

• Reduction in Fire Ignition Frequencies based on NUREG/CR-6850 combined with NRC clarifying information (joint NRC/EPRI fire PRA Frequently Asked Question (FAQ) 08-0048).

• Reduction in the extent of fire induced damage by applying the following types of inputs or analysis.

o Cable trays that are greater than or equal to 17 feet above the Cable Spread Room floor are beyond the zone of influence for any credible fire.

o The cable trays above the fire sources contain IEEE-383 qualified cables, covered with a flame-retardant coating, and are contained in solid-bottom trays in key locations, significantly limiting fire propagation.

o Detailed fire modeling using Fire Dynamics Tools (FDTs).

o NRC clarifying FAQ 08-0050 provided updated curves for non-suppression probability concerning Fire Brigade response.

0 Probability of success for alternate safe shutdown methodologies o Operators will not automatically enter the ASSD procedures and leave the Control Room when there is a fire in the areas of concern. They will evaluate conditions, as trained, and enter ASSD procedures as conditions warrant.

o Even if ASSD-02, "Alternate Safe Shutdown Procedure for Control Building," is entered, there are alternate safe shutdown methodology options.

o Further refinement of ASSD baseline procedure risk.

Oualitative Risk Insi2hts The following items are qualitative insights which provide additional margin to the low safety significance of the event.

Based upon the analyzed fire sources and realistic impact combined with the passive fire features of the plant, it is highly unlikely that the ASSD procedures would be implemented. If operators do not enter the ASSD procedure, the EDG ASSD lockout is not a factor; there is no increase in plant risk.

The cable spread rooms have solid-bottom cable trays with IEEE-383 qualified cable that is also coated with a spray-on flame-retardant coating, making the expected cable performance significantly better than those that were tested in the Sandia Labs studies as discussed in Appendix Q of NUREG/CR-6850. This combination of IEEE-383 qualified cables that are sprayed with flame-retardant coating and solid bottom cable trays in key locations provides an additional level of defense-in-depth that is not typically found in the industry. The conclusion is that the credible fire scenarios will not propagate beyond the ignition source.

In order to quantify the Conditional Core Damage Probabilities (CCDPs) of the Cable Spread Room scenarios, the failure of the components were compared against the base case for the

BSEP 09-0068 Enclosure 1 Page 3 of 9 applicable situations. The Severe Accident Mitigation Alternative (SAMA) Diesel Generators were installed for approximately 9 of the 12 months that the condition existed. Cases were run both with and without available based on the time that these supplemental diesel generators were actually installed. Crediting the SAMA Diesel Generators would further reduce the risk associated with Cable Spread Room fires.

The ignition sources evaluated are assumed to cause the LOOP, or EDG auto start, if components and/or cables related to off-site power or the EDGs are within the zone of influence of the fire or the postulated fire growth. The components are assumed to be failed; although, this may not be the case for all fires.

Discussion of Maior Differences between CP&L and NRC Methodolouv Applying NRC reviewed or approved methodologies, now available to the industry, allows for a more accurate characterization of the scenario as compared to some of the conservative screening values used in IMC 0609 for Phase II of the SDP. The following discussion provides an overview of the basis for, and impact of, these refinements for a more accurate result. provides supplemental details of these refinement areas.

CP&L has determined that the total CDF associated with this condition is 1.05E-07 for both Units 1 and 2, which should be categorized as having very low safety significance (i.e., Green).

The NRC's evaluation determined that the total CDF associated with this condition is 7.63E-06, which is categorized as having low to moderate safety significance (i.e., White).

CP&L's evaluation of the three major contributors (i.e., fire ignition frequencies, fire induced damage, and ASSD implementation) identified the following eight primary refinement areas as those that constitute the majority of the difference in calculated CDF results between the CP&L and NRC assessments.

1. Fire Ignition Frequencies

2. MCC High Energy Arcing Fault

3. Source Applicability

4. Source Heat Release Rates (HRR)

5. Motor Control Center (MCC)

a. MCC Fire Growth Factor

b. MCC Configuration

6. Solid Bottom Trays

7. Non-Suppression Probability

8. ASSD Implementation

1. Fire Ignition Frequencies The NRC's Phase II analysis uses the fire ignition frequencies found in the Fire Protection Significance Determination Process in IMC 0609, Appendix F, Attachment 4. Phase III of the SDP process allows for further insights that go beyond the IMC 0609 guidelines.

BSEP 09-0068 Enclosure 1 Page 4 of 9 The following two recent fire PRA refinement documents show that the existing IMC 0609 fire ignition frequencies are a conservative understanding of the postulated fire scenarios.

" EPRI document 1016735, "Fire PRA Methods Enhancements: Additions, Clarifications, and Refinements to EPRI 1019189," Table B-5, and

• Joint NRC/EPRI fire PRA FAQ 08-0048.

Applying the FAQ 08-0048 fire ignition frequencies reduce the NRC calculated CDF by approximately 34.7%.

For supplemental details see Enclosure 2, Item 6.

2. MCC High Energy Arcing Fault (HEAF)

Approved FAQ 06-0017, Revision 2, states:

... MCCs with molded-case circuit breakers should not be counted as HEAF sources unless it is associated with switchgear that is used to directly operate equipment such as load centers.

The MCCs in the Cable Spread Rooms clearly match this condition and should not be counted for HEAF.

Application of this refinement reduces the NRC calculated CDF by approximately 4.4%.

For supplemental details see Enclosure 2, Item 10.

3. Source Applicability Based on the actual plant physical configuration, two vertical sections in each Cable Spread Room for the Common D(C), and 2L(1L) Substations have no ignition source within the cabinets and thus are not considered as fire sources. These vertical sections are included as "General Electrical" sections of the substations. These cabinets fit into two categories as follows.

" Cabinets containing incoming power cables and manual disconnect switches. These disconnect switches are similar to a bus bar in a static position.

" Cabinets containing only incoming power cables to transformers. This type of electrical cabinet is a wireway or an enclosed cable tray for routing purposes and is not counted as electrical cabinet ignition sources.

Application of this refinement reduces the NRC calculated CDF by approximately 27.4%.

For supplemental details see Enclosure 2, Items 10 and 12.

BSEP 09-0068 Enclosure 1 Page 5 of 9

4. Source Heat Release Rates (HRR)

A specific assessment was made to one of the vertical sections within the substations. These cabinets are identified as "General Electrical" cabinets and are actually labeled "Control Power and Neutral Resistor" cabinets. They are present in all four Substations. Each cabinet has one bundle of cables which runs vertically on one side within the cabinet. There are two small 120 V breakers, a meter in the upper face of the door, and a coil in the bottom of the cabinet. Per guidance in Appendix G of NUREG/CR-6850 this vertical cabinet section can be modeled with a 98% HRR of 211 kW, and a 75% HRR of 69 kW.

The dry-type transformers in the Cable Spread Rooms are considered, by NUREG/CR-6850, as electrical motors for the purposes of deriving HRRs. This is discussed in Table 6-1, Table 8-1, and Table 11-1 of NUREG/CR-6850. Table E-l and Table G-1 assigns a HRR for motors (i.e., the 9 8 th percentile) as 69 kW and the 75% HRR as 32kw for electric motors.

The NRC analysis uses a HRR of dry transformers in the Cable Spread Room of 70 kW and 200 kW.

Per Section Q.2.2 ofNUREG/CR-6850, the propagation of lower HRR fires (i.e., 69 kW for dry-type transformers) to solid-bottom cable trays is prevented.

Application of this refinement reduces the NRC calculated CDF by approximately 48.8%.

For supplemental details see Enclosure 2, Items 9, 10, 11, and 12.

5. Motor Control Center (MCC)

(a) MCC Fire Growth Factor The MCCs were constructed without any ventilation openings, and the individual doors of the cubicles are attached with hex head mechanical fasteners. In addition, the electrical connections were made via conduits. This configuration is consistent with well sealed electrical cabinets as described in NUREG/CR-6850, and could be excluded. However, for this analysis, CP&L conservatively included these MCCs in the evaluation. The fraction of fires propagating beyond the MCC cubicles is estimated to be 0.05 (i.e., 5.0%), based on the following.

• Not all fires originating in the MCC cubicle will be energetic enough to blow open the rigidly secured door. It is conservatively assumed that 10% of the fires are capable of this.

• Based on detailed fire modeling, only the upper 1/2 of the MCC sections will be capable'of producing a zone of influence that can impact any of the overhead cable trays. The bottom half of each vertical section is excluded, due to the source to target distances.

BSEP 09-0068 Enclosure 1 Page 6 of 9 Therefore, considering only the top 1/2 of each vertical section, and excluding the bottom half, would yield 0.05 (i.e., 0.10 x 0.5), or a 5.0% MCC fire growth factor.

(b) MCC Configuration - Removal of Two Vertical Sections The NRC evaluation assumes all MCCs have eight vertical sections. Unit 1 MCC 1CB has only six vertical sections. This conservatism can be removed for the one MCC in Unit 1.

Application of refinement (a) reduces the NRC calculated CDF by approximately 13.9%. No credit was taken for refinement (b).

For supplemental details see Enclosure 2, Items 7 and 8.

J

6. Solid Bottom Trays Additional credit for solid bottom trays to both delay fire growth and prevent fire damage for low HRR fires as discussed in NUREG/CR-6850 Attachment Q has been factored into CP&L's final analysis.

Application of this refinement reduces the NRC calculated CDF by approximately 36.7%.

For supplemental details see Enclosure 2, Item 11.

7. Non-Suppression Probability (NSP)

Fire modeling was conducted to determine the potential for fire induced damage to critical circuits that would lead to operators having to implement the ASSD procedure.

NUREG/CR-6850 fire modeling guidance was used in conjunction with the FDTs.

The guidance in NUREG/CR-6850 for determining NSP has been updated in FAQ 08-0050, which is an NRC interim position for calculating NSP. This removes the conservatism associated with Fire Brigade response times which further reduces the likelihood of ASSD entry.

Special attention was placed on the Cable Spread Rooms for Unit 1 and Unit 2. Consistent with the plant licensing basis, the cables are IEEE-383 qualified and those in cable trays have spray-on Flame-Master 71A or Flame-Master 77 flame-retardant coatings applied to them.

Additionally, there are solid bottom trays above fixed sources in key locations.

NUREG/CR-6850 states that damage to coated, non-qualified cables will occur in 6 to 10 minutes in the first tray, and 11 minutes in the second and third tray. Ignition of coated, non-qualified cables will not occur for at least 13 minutes in the first tray, with no ignition in the second or third tray. Based on field walk-downs and research of design documents, all of the cables, except for fiber optic, in the Cable Spread Rooms are IEEE-383 qualified cables with flame-retardant coating.

BSEP 09-0068 Enclosure 1 Page 7 of 9 Further research, sponsored by the NRC and performed by Sandia National Labs in 1978, concluded that IEEE-383 qualified cable that is coated with a flame-retardant coating had significant ignition delays when subjected to severe fire scenarios (i.e., diesel fuel pool fire) and did not propagate vertically to the next cable tray. The same research indicated that for IEEE-383 qualified cable located in a solid-bottom cable tray, no fire developed after six cycles of the ignition source. The Cable Spread Rooms have solid-bottom cable trays with IEEE-383 qualified cable that is also coated with a spray-on flame-retardant coating, making the expected cable performance significantly better than those that were tested in the Sandia Labs studies.

The research and test data support CP&L's analytical assumption that vertical fire propagation is halted for coated IEEE-383 qualified cable located in a solid-bottom cable tray when subjected to a severe fire scenario. These are significant contributors when calculating the NSP because the test data (i.e., Appendix Q of NUREG/CR-6850) suggests that the cable tray configuration and characteristics are not conducive to vertical flame spread from the originating ignition source. Further analysis demonstrates that the initiating ignition sources alone do not produce conditions that would damage component circuitry needed for a safe reactor shutdown.

The fire analysis demonstrates that the plausible fire scenarios do not involve fire sizes that are capable of vertical propagation given the presence of IEEE-383 qualified cable, spray-on flame-retardant coating, and solid-bottom cable trays near the ignition sources. Based on existing cable routing information, key safe shutdown components are not impacted and entry into the ASSD procedure is unlikely.

Application of this refinement reduces the NRC calculated CDF by approximately 83.2%.

For supplemental details see Enclosure 2, Items 1, 2, and 3.

8. ASSD Implementation In order to address the unlikely event that the ASSD procedures are entered, a decision tree was developed to determine the change in risk rather than assuming the generic industry value 0.1 failure probability.

A significant factor to determine the risk involved with this analysis is the decision of the operators to implement the ASSD procedures. To support the assessment, the target raceways in the zone of influence were examined to identify the set of equipment that might be expected to fail as the fire progresses. No key safe shutdown components were identified in the zone of influence.

To assess the significance of the adverse EDG condition identified, the focus of the CP&L analysis is to determine the set of circumstances that would cause Control Room Operators to implement the ASSD procedure. For the EDG modification to have an impact, a LOOP condition must be in effect, followed by entry into the ASSD procedure. If operators do not enter the ASSD procedure, the EDG ASSD lockout is not a factor and, therefore, cannot

BSEP 09-0068 Enclosure 1 Page 8 of 9 adversely impact the calculated plant risk. The fire modeling and risk insight that are included in the CP&LASSD event tree provide an assessment of the likelihood of Control Room Operators entering the ASSD procedure. This analysis was based on interviews with licensed reactor operators and the conclusions were validated by performing simulator scenarios based on the potential impact of the evaluated fires.

The potential equipment losses associated with this condition are not sufficient to result in the implementation of the ASSD procedures. Additionally, this conclusion is based upon the analyzed fire sources and realistic impact.

Application of this refinement reduces the NRC calculated CDF by approximately 81.9%.

For supplemental details see Enclosure 2, Item 5.

Conclusion Based on the above discussions, CP&L has concluded that a more accurate characterization of the preliminary white finding, using NRC reviewed or approved methodologies, now available to the industry, is appropriate. The following table summarizes the impact of revising the more significant areas of conservatism.

Fire Ignition Frequencies XU-2 1.58E-03 XU-2 3.2E-03 34.7%

XU-5 1.58E-03 XU-5 3.2E-03 Breakers 2.88E-05 Breakers 5.5E-05 HEAF 3.32E-06 HEAF 1.6E-06 Transformers 8.91E-05 Transformers 1.1E-04 MCC High Energy Screened Out 1.0 4.4%

Arcing Fault Source Applicability Screened Out 1.0 27.4%

Source Heat Release Transformer 32-70kW Transformer 70-200kW 48.8%

Rates Gen Electrical Cab 70- Gen Elec Cab 200-200kW 650kW

BSEP 09-0068 Enclosure 1 Page 9 of 9 Using the methodologies described in the above evaluation, CP&L has determined that the total CDF associated with this condition is 1.05E-07 for both Units 1 and 2, which should be categorized as having very low safety significance (i.e., Green). This differs with the NRC's evaluation, which determined that the worst case total CDF associated with this condition is 7.63E-06, which is categorized as having low to moderate safety significance (i.e., White).

Based on the analysis documented in this submittal, CP&L contends that the finding is most appropriately classified as having very low safety significance (i.e., Green).

BSEP 09-0068 Enclosure 2 Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 1 of 15 Item 1: Cable Spreaid Room Tlargets~

NRC Evaluation Targets in the Cable Spread Room (CSR) are those cables just beyond the first three cable trays which, when damaged, result in implementation of Brunswick Steam Electric Plant (BSEP) procedure OASSD-02, "Control Building."

CP&L Evaluation It is unlikely that the cables within the zone of influence, if damaged would result in OASSD-02 entry.

References

• BNP-PSA-073, "Risk Evaluation for SDP Associated with EDG Local Transfer Switch Modification"

• NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology," Final Report.

" NUREG/CR-0381, "A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings Tests" Discussion Targets should be those cables that are critical to equipment required for safe shutdown, which at CP&L, exist in trays higher than 17' off of the CSR floor or are shielded. None of these critical control cables are in the first three cable trays, which results in at least 25 minutes for Non-Suppression Probability (i.e., reference BNP-PSA-073). Thus, damaged targets that would result in implementation of OASSD-02 are in much higher cable trays and are most likely never to be damaged. The Loss of Off-site Power (LOOP) is not enough to require implementation of the Alternative Safe Shutdown (ASSD) procedure, as discussed below.

Attachments 16 and 18 to calculation BNP-PSA-073, provides discussion of the detailed fire modeling, Operations interviews, and cable identification process that was evaluated to determine that the LOOP would not be a deciding factor to enter the ASSD procedures. The LOOP is a practiced scenario in the simulator and is a frequently trained topic. Brunswick also has previous operating experience in actual LOOP events, and based on that performance, it is reasonable to assume that this condition alone will clearly not result in Main Control Room (MCR) evacuation.

Operations expressed in multiple interviews, that in addition to the LOOP, there must be complicating conditions and loss of control functions of ASSD equipment to prompt entry into OASSD-02. 6 of BNP-PSA-073 states, in part: In general, there was no concern for the loads associated with the 2L substation. The condition was assessed and it was determined that there would be adequate power available, injection sources, and heat removal capabilities and procedural guidance for each system to mitigate the event. Although there was more concern due to batteries and DC battery chargers, the groups concluded that there is procedural guidance to shift DC control power for diesels in the Diesel Building and may require manual actions to obtain power from the other unit, but clear procedural guidance is available. It was noted that this condition may hinder the application of the Severe Accident Mitigation Alternative (SAMA) Diesel Generators (DGs), but consideration would be given to their use once conditions in the room were acceptable and if the need for additional DC power necessitated this action. The consensus from those interviewed was that there is adequate procedural guidance to

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 2 of 15 1 "beSpread R~oomII Tar1ets <

mitigate the failures, thus no entry into 0ASSD-02 would be made from the posed fire (in Unit 2 CSR 2L Substation) unless habitability of the MCR came into question or safe shutdown was not assured. Otherwise, the applicable Abnormal Operating Procedures (AOPs) and Emergency Operating Procedures (EOPs) would provide adequate guidance to combat the posed scenario. The same conclusions were reached for each of the applicable fire sources in the Cable Spread Rooms. 8 of BNP-PSA-073 states, in part: Several fire scenarios were considered for the eight ignition sources in CSR Unit 1 and Unit 2. The Motor Control Center (MCC) fire scenarios involve a fault that blows open a well-sealed MCC cubicle and results in a 200 kW fire. The unit substation breaker cubicles are considered to be 200 kW. The targets of concern are not in the lower three cable trays, which provide ample time for suppression. Both the 200 kW and 702 kW originating fire scenarios (not considering vertical flame propagation), do not have the ability to damage any cables beyond 17' (such as the diesel generator control cables) because of the target distance relative to the source fire (i.e., reference BNP-PSA-073).

The second scenario that involves the potential for vertical flame spread to the cable trays above the ignition sources is unlikely given the flame-retardant coated, IEEE-383 qualifiedcable, and the presence of solid-bottom trays. The noted reference (NUREG/CR-0381) suggests that vertical flame propagation for the configuration that is present in the CSRs will not occur in the time frame associated with the NUREG/CR-6850 heat release rate profile.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 3 of 15 Itm2: C-A& Spr~ead Ro on-Suppression ProhabiitN NRC Evaluation Probability for Non-Suppression in the CSR was based on time to damage those cables beyond the first three cable trays.

CP&L Evaluation Conservatively assumes 25 minutes for time to suppress the fire, based on BNP-PSA-073, Attachment 18.

References

• BNP-PSA-073, Attachment 18

" NRC/EPRI fire PRA FAQ 08-0050 Discussion See Item 1 discussion above on timing of Cable Spread Room Non-Suppression Probability.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 4 of 15 hn3:Fire Goýh'ýoen NRC Evaluation Fire growth modeling in the CSR is modeled based on IMC-0609, and does not give credit as allowed per NUREG/CR-6850.

CP&L Evaluation CP&L applies the values contained in NUREG/CR-6850 for fire growth modeling. Additionally, the credit associated with passive fire protection features, as shown in the Table below (i.e., reference Appendix Q, Table Q-1), has been used in this analysis.

References

" NUREG/CR- 6850, Appendix Q.

• NUREG/CR-0381.

" Inspection Manual 0609, "Significance Determination Process" Discussion Tray~espone G~atingTime to giLIt[01n (m111intes) Time to Dmg riuej NULJREG/CR-6850~ NR \-U~IGC-6850. NRC~

Lower Tray Flame-Master 71A 13 No credit for flame 10 No credit for flame Response Flame-Master 77 13 coating. 6 coating.

Upper Tray Flame-Master 71A No ignition. No credit for flame 11 No credit for flame Response Flame-Master 77 No ignition. coating. 11 coating.

Table data is excerpted from NUREG/CR-6850, Appendix Q.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 5 of 15 Itemn 3: Fire Growth M1odeling~

The noted references provide guidance for the credit of passive fire protection features. An excerpt from Appendix Q of NUREG/CR-6850 is provided below:

Sandia Laboratories (SL) performed tests to evaluate the effects of cable coatings in:

" reducing the flammability of cable material,

• preventing or delaying the spread of fire, and

• preventing fire-induced cable failures.

Thirty-three full-scale tests were performed using trays loaded with either qualified or non-qualified cables. The cable tray configurations included both a single cable tray and a two-tray stack. Exposure fires included either a gas burner or a diesel fuel pool fire. Flammability was evaluated by measuring time to ignition, time to maximum heat release rate, and cumulative heat release at various times after initiation of the exposure fire. The diesel fuel pool fire was selected because the exposure fire was more intense, and therefore, conservative.

The Cable Spread Room (CSR) flame-retardant cable coatings are the Flame-Master 71A and Flame-Master 77. NUREG/CR-6850 states that damage to coated, non-qualified cables will occur in six to 10 minutes in the first tray, and 11 minutes in the second and third tray. Ignition of coated, non-qualified cables will not occur for at least 13 minutes in the first tray, with no ignition in the second or third tray. Based on field walk-downs and research of design documents, all of the cables in the cable spread rooms are qualified cables with fire retardant coating, and are expected to respond similarly when exposed to a large diesel fire (i.e., at a time much greater than 13 minutes).

Further research sponsored by the NRC and performed by Sandia Labs in 1978 suggests that IEEE-383 qualified cable that is coated with a flame-retardant coating had significant ignition delays when subject to severe fire scenarios (diesel pool fire) and did not propagate vertically to the next cable tray. The same research indicated that for IEEE-383 qualified cable located in a solid bottom cable tray, no fire developed after six cycles of the ignition source. The cable spread room has solid bottom cable trays with IEEE-383 qualified cable that is also coated with a spray-on fire retardant coating making the expected cable performance significantly better than those that were tested in the Sandia Labs studies.

Based on the available literature and test data, it is justified to assume that vertical fire propagation is halted for coated IEEE-383 qualified cable located in a solid bottom cable tray when subjected to a severe fire scenario.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 6 of 15 Item 4. SAtMA(Severe-Accident Mtlitigation Alternative) DliselGnerators NRC Evaluation The SAMA DGs are not credited for any CSR scenarios due to their need to be connected to either MCC 2CA or 2CB.

CP&L Evaluation Credit for the SAMA DGs should be allowed for CSR scenarios.

References

. BNP-PSA-073, Attachment 16 and 17.

Discussion From interviews with licensed Reactor Operators (Attachment 16), it is reasonable to assume that even with a fire in the CSR, if a lack of power to the plant was a major concern, the operators would attempt to use the SAMA DGs in order to combat this condition. Even if the fire was from one of the MCCs, the operators would attempt to attach the SAMA diesels to the opposite train, and some credit should be granted for this probability.

Provided that Reactor Core Isolation Cooling (RCIC) is operable initially, the MAAP4 runs indicate that core damage would be postponed for greater than four hours, which is ample time to extinguish the fire and electrically connect the SAMA DGs (Attachment 17).

The interviews showed that operators would put a priority on regaining the electrical system as soon as safely possible. Only one of the MCCs at a time (in the CSR) would be damaged from the posed fire, and it would still be possible to attach one of the SAMA DGs to the undamaged MCC to power one DC bus.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 7 of 15 Intom proere iAnSDMa" r R.o Panel .':U,;

NRC Evaluation Fires in the Unit 2 MCR panel 2-XU-2 that meet the fire modeling assumption no. 2; will damage enough ASSD equipment to meet the entry criteria into procedure 0ASSD-02.

CP&L Evaluation Fire damage to the 2-XU-2 panel does not automatically prompt entry into procedure OASSD-02.

References

• BNP-PSA-073, Attachment 16.

• BNP-PSA-073, CP&L ASSD Event Tree Discussion The loss of off-site power scenario is practiced frequently on the simulator as a training topic for continuing operator training,,and there are several methods to combat this scenario other than MCR evacuation. If habitability is a concern, or if all methods of regaining power to the blacked out unit are exhausted, then entry into OASSD-02 would be the correct procedure to enter. Entry into the procedure is based on the loss of control, and not the loss of power as described in assumption no. 2 of NRC Inspection Report 2009009. 6 discusses in detail the fact that the licensed operators would not enter the OASSD-02 procedure based on the LOOP and loss of the applicable Units EDG controls from the respective XU-2 panel. All SROs are trained on evacuation of the Control Room and implementation of the ASSD procedure. Each interviewed operator expressed that maintaining control of the plant was of paramount importance, and some provided examples of equipment malfunctions that would prompt them to evacuate the Control Room for the remote shutdown panels (i.e., SRV's cycling sporadically, MOV's randomly repositioning, pumps stopping and starting for no apparent reason).

During interviews with the Reactor Operators, most expressed the position that the LOOP was not, by itself, criteria to abandon the MCR; that this event is trained on extensively and that entry into OASSD-02 will not be required until all efforts to cross-tie the electrical plant have been exhausted.

This conclusion was based on the extent of training and working knowledge of the Station Blackout procedure. The operators discussed the available options of cross-tying the electrical plant from the opposite Units control board, as well as locally at the breakers.

A crew of licensed operators were exposed to the XU-2 MCR fire scenario on the simulator and concluded that entry into OASSD-02 would not be made based solely on the presence of fire (provided that satisfactory environmental conditions were not in jeopardy). The operating crew concluded that they would remain in the MCR until it is determined that the cross-tie of the electrical plant to the blacked-out unit or safe shut down is not possible from the MCR.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 8 of 15 Itemn 6: Fire lignition Frequenicies NRC Evaluation Fire Ignition Frequencies were taken from Fire Protection Significance Determination Process 0609, Appendix F, Attachment 4. XU-2/XU-5 Fire Ignition Frequency is 3.20E-03.

CP&L Evaluation XU-2/XU-5 Fire Ignition Frequency is 1.58E-03, 33% reduction of the NUREG/CR-6850 values.

References

• EPRI 1016735, "Fire PRA Methods Enhancements: Additions, Clarifications, and Refinements to EPRI 1019189," Table B-5.

• Joint NRC/EPRI Fire Probabilistic Risk Assessment (PRA), FAQ 08-0048.

" NUREG/CR-6850 Discussion This data was developed by a joint effort of EPRI and NRC research.

From Table B-5 of EPRI 1016735, Bin no. 4, Fire Ignition Frequencies = 4.8E-03*0.33 = 1.58E-03.

From Bin no. 15.1, Electrical.Cabinets, Non-High Energy Arcing Fault (HEAF), have been reduced from 1.5E-3 to 1.06E-3 (0.71). This factor was multiplied by the applicable Fire Ignition Frequencies.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 9 of 15 litin7: F~ire Growth Damnage NRC Evaluation Assumes that all MCC fires grow and are capable of causing fire damage.

CP&L Evaluation Only a fraction (i.e., 0.05, or 5.0%) of MCC breakers is capable of causing damage to target cables.

References

• BNP-PSA-073

• NUREG/CR-6850 Discussion The MCCs were constructed without any ventilation openings, and the individual doors of the cubicles are attached with screws. In addition, the electrical connections were made via conduits. This configuration is consistent with well sealed electrical cabinets as described in NIUREG/CR-6850, and could be excluded. However, for this analysis, CP&L conservatively decided these MCCs will be included in the evaluation. The fraction of fires propagating beyond the MCC cubicles is based on the fact that not all fires originating in the MCC cubicle will be energetic enough to blow open the rigidly secured door. It is conservatively assumed that 10 percent of the fires are capable of this. An additional best estimate factor, based on detailed fire modeling, is that only those fires in the upper 1/2 of the MCC will be energetic enough to be capable of dislodging the door and allowing the fire to propagate. The bottom half of each vertical section is excluded, due to the source to target distances. Therefore, conservatively assuming only the top half of each vertical section and excluding the bottom half, would yield 0.05 (i.e., 0.10 x 0.50, or 5.0%).

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 10 of 15 Itemn 8: MCC I1(B VetclScin NRC Evaluation MCCs in the CSR each have 8 vertical sections.

CP&L Evaluation Based on visual inspection, Unit 1 MCC 1CB has only 6 vertical sections.

References

• Plant physical configuration Discussion Only six vertical cabinets exist in one of the two MCCs in the Unit 1 CSR. Eight vertical cabinets exist in the MCCs in the Unit 2 CSR.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 11 of 15 Item 9: Hleat Release Rate K NRC Evaluation The Heat Release Rate (HRR) of dry transformers in the CSR is 70 kW and 200 kW.

CP&L Evaluation Dry transformers are treated as motors per NUREG/CR-6850, and are assigned the HRR of electrical motors, which is 69kW (9 8th percentile).

References

• NUREG/CR-6850, Table 11-1, Table E-I, and Table G-1.

Discussion Per NUREG/CR-6850, Table 11-1, dry transformers as ignition sources should use HRR Probability Distribution to calculate the Severity Factor using that of electric motors. Table E-I of NUREG/CR-6850 (Case 7), states that the HRR = 69 kW (Reference Table G-1) for electric motors. Thus, all dry transformers are 69kW, and 200kW does not apply. The 69kW HRR targets are shielded by solid bottom cable trays and do not contribute to fire growth.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 12 of 15 NRC Evaluation Per Fire Protection Significance Determination Process IMC 0609, Appendix F:

The distribution of General Electrical cabinet fires are:

200 kW and 650kW.

The distribution of MCC fires are:

1. HEAF,

2. 70kW, and

3. 200kW.

MCCs contribute to CDF via the HEAF mechanism.

CP&L Evaluation Per NTUREG/CR-6850, Appendix G:

The distribution of General Electrical cabinet fires are:

69 kW and 211kW The distribution of MCC fires are:

1. 69kW, and

2. 211 kW.

HEAFs are not postulated for MCC scenarios.

References

• NUREG/CR-6850, Appendix M

• Physical configuration and inspection of cabinets

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 13 of 15 Item 10 Substation/General Electrica[jCabinets and MICC F~ire Distribution Discussion The importance of best estimate heat release rates determines the fire growth, fire suppression probability, size of the fire zone of influence and damage to cables in zone of influence.

The general electrical cabinets located in substation have two fire scenarios associated with them that consist of a 211 kW fire (98th percentile) and a 69 kW fire (75th percentile). The applicable heat release rates were determined by inspection of the inside of the general electrical cabinets.

MCC's are not considered HEAF sources.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 14 of 15

],tent It:, L ovN, H a ele aise R ateI(H R 6 n rb t onto C ore D m ge lcuencNw ( i .... ... ..*... ..

NRC Evaluation Transformers and MCC breakers of 70kW HRR contribute to CDF.

CP&L Evaluation Low HRR (i.e., 69 kW) transformers and MCC breakers are not capable of contributing to CDF.

References NUREG/CR-6850, Appendix for Chapter I1, Passive Fire Protective Features, Section Q.2.2, Cable Tray Barriers and Fire Stops.

Discussion Cable tray fire barrier tests were also performed by Sandia Labs. Thirteen tests were conducted in a manner identical to that used in the single tray and two-tray gas burner cable coating tests. The same cable types and the same gas burner exposure fire sources were used. Six potential fire barrier systems were tested:

1. Ceramic wool blanket wrap,

2. Solid tray bottom covers,

3. Solid tray top cover with no vents,

4. Solid tray bottom cover with vented top cover,

5. One-inch insulating barrier between cable trays, and

6. Fire stops.

The barrier test findings are as follows. Propagation of the fire to the second tray was prevented in each case. That is, each barrier prevented ignition of a cable tray when exposed to a cable tray fire in a lower tray. Barriers seem to substantially delay cable damage for qualified cable.

However, the barriers did not delay cable damage for non-qualified cable. For application to the Fire PRA, the barrier test findings are considered.

most appropriate to exposure fires with smaller heat release rates and to cable trays in a stack threatened by fires in lower trays. In these cases, each barrier prevents cable tray ignition until well after the fire brigade reaches the scene (i.e., greater than 20 minutes), and damage in qualified cable with solid tray bottom covers, until well after the fire brigade reaches the scene.

Supplemental Details of Differences between CP&L's and the NRC's Evaluation Methodologies Enclosure 2 Page 15 of 15 Item 12: Substation Ignition Sources ,

NRC Evaluation NRC considers all vertical sections of the Substations as electrical cabinet ignition sources. The General Electrical Cabinet of the Substations is a 650 kW source.

CP&L Evaluation There are two vertical sections in the CSR for the Common C/D, and IL/2L Substations that have no ignition sources within the cabinets. There are two vertical sections in the CSR for the Common C/D and 1L/2L Substations that have ignition sources of 200 kW vice 650 kW.

References

• BNP-PSA-073.

• NUREG/CR-6850, Appendix G.

• Brunswick Plant Drawing F-03022, "480 Volt One Line Lighting Distribution"

• Brunswick Plant Drawing F-30002, "Auxiliary One Line Diagram, 4160 Volt System, Switchgear 1B, IC, ID & Common A"

• Brunswick Plant Drawing F-03002, "Auxiliary One Line Diagram, 4160 Volt System, Switchgear 2B, 2C, 2D & Common B"

• Brunswick Plant Drawing F-30005, "480 Volt System, Unit Substation 1E, IF, E5, E6 and Common C Auxiliary One Line Diagram"

" Brunswick Plant Drawing F-03005, "480 Volt System, Unit Substation 2E, 2F, E7, E8 and Common D Auxiliary One Line Diagram" Discussion There are two vertical sections in the CSR substations referred to as "general electrical" sections of the substations in the NRC's analysis. Two of the Substation IL and 2L vertical cabinets contain incoming power cables and manual disconnect switches. These disconnect switches, identified as Load Interrupter switches in F-03022, are similar to a bus bar in a static position.

Two of the Substation Common C and Common D vertical cabinets contain only incoming power cables to transformers. This electrical cabinet is similar to a wire-way, or an enclosed cable tray, for routing purposes as shown in References F-30002, F-30005, F-03002 & F-03005 (i.e., Common C and Common D Substations, respectively).

CP&L further asserts that these vertical sections located in each of the four Substations should be appropriately classified as a 211 kW source. A specific assessment was made to one of the vertical sections within the Substations. These cabinets are identified as general electrical cabinets in the NRC report. These cabinets are actually ground fault meter cabinets and are present in all four Substations. Each cabinet has one bundle of cables which runs vertically on one side within the cabinet. Per guidance in Appendix G of NUREG/CR-6850, this vertical cabinet section can be modeled with a 98% HRR of 21 1kW, and a 75% HRR of 69kW.