ML16060A361
| ML16060A361 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/24/2016 |
| From: | George Gellrich Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16060A366 | List: |
| References | |
| TAC MF2993, TAC MF2994 | |
| Download: ML16060A361 (85) | |
Text
SECURITY-RELATED INFORMATION - WITHHOLD UNDER 10 CFR 2.390 George Gelirich Site Vice President Exeton Generation° Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657 410 495 5200 Office 717 497 3463 Mobile www.exeloncorp.com george.gellrich@exeloncorp~com 10 CFR 50.90 February 24, 2016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318
Subject:
Request for Additional Information Regarding the National Fire Protection Association Standard 805 License Amendment Recluest
References:
- 1.
Letter from G. H. Gelrinch (CCNPP) to Document Control Desk (NRC), dated September 24, 2013, License Amendment Request re: Transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection
- 2.
Letter from N. S. Morgan (NRR) to G. H. Gellinch (Exelon), dated January 12, 2015, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Request for Additional Information Regarding the National Fire Protection Association Standard 805 License Amendment Request (TAC Nos. MF2993 and MF2994)
- 3.
Letter from G. H. Gelrinch (Exelon) to Document Control Desk (NRC), dated April 15, 2015, Request for Additional Information Regarding the National Fire Protection Association Standard 805 License Amendment Request In Reference 1, Calvert Cliffs Nuclear Power Plant, LLC submitted a license amendment request to transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection.
In Reference 2 the Nuclear Regulatory Commission staff requested additional information regarding this amendment request. The response to PRA RAI 03 and PRA RAI 19 provided in Reference 3 noted that the final risk quantification and related actions would be provided to the NRC after acceptance of the approach. This letter provides that information.
Attachment (1) and the Enclosure provide the response to the request for additional information. contains page replacements for various pages in Section 4.0 of the license package. These page replacements address the change in fleet and station procedures due to the merger with Exelon Generation Corporation. Please note that procedures that may be listed elsewhere in the original license amendment package have been or will be transitioned to Exelon procedures with similar elements, as appropriate.
Additionally, Enclosure 1 contains A)*
Upon removal of Attachment C, 5, and W pages in Enclosure 1, this submittal is not restricted
Document Control Desk February 24, 2016 Page 2 replacements for Attachments C, G, 5, and W of the license amendment package and supersedes the previously provided Attachments.
Attachments C, S, and W in Enclosure 1 contain security-related information and are requested to be withheld from public disclosure under 10 CFR 2.390.
This additional information does not change the No Significant Hazards Determination provided in Reference 1. No regulatory commitments are contained in this letter.
Should you have questions regarding this matter, please contact Mr. Larry D. Smith at (410) 495-5219.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 24, 2016.
Respectfully, George H. GellInch Site Vice President GHG/PSF/bjm
Attachment:
(1)
Request for Additional Information Regarding the National Fire Protection Association Standard 8905 License Amendment Request
Enclosure:
1 Section 4.7 pages and Attachments C, G, 5, W cc:
NRC Project Manager, Calvert Cliffs NRC Regional Administrator, Region I NRC Resident Inspector, Calvert Cliffs S. Gray, MD-DNR
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Calvert Cliffs Nuclear Power Plant February 24, 2016
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST By letter dated September 24, 2013 Calvert Cliffs Nuclear Power Plant, LLC, submitted a license amendment request (LAR) for Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (CCNPP), to transition its fire protection licensing basis from Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48(b) to 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," "2001 Edition. The licensee submitted request for additional information (RAI) responses. Based on its review of the RAI responses, the U.S. Nuclear Regulatory Commission (NRC) staff requests the foliowing additional information to complete its safety evaluation of the LAR:
Probabilistic Risk Assessment (PRA) RA1 03. Integrated Analysis:
Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 fudther states that the change in public health risk arising from transition from the current fire protection program to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF and LERF, identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis, and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff's review of the information in the LAR has identified additional information that is required to fully characterize the risk estimates.
The PRA methods currently under review in the LAR include:
PRA RAI 01.a regarding loss of MCR HVAC PRA RAI 01.b regarding division of PAUs into "sub-PA Us" PRA RAI 01.c regarding treatment of propagation in the MCA PRA RAI 01. d regarding unreliability and unavailability of active barriers PRA RAI 01.e regarding adverse operator actions PRA RAI 01.f regarding 120V panel breaker coordination issues PRA RAI 02.a regarding alignment of OC diesel generator PRA RAI 02.b regarding HRA methods, including sequential timing of operator actions PRA RAI 04 regarding placement of transient fires PRA RAI 05 regarding transient influence factors PRA RAI 06 regarding reduced transient HRR PRA RAI 07 regarding self-ignited cable fires and those caused by welding and cutting PRA RAI 08 regarding treatment of junction boxes PRA RAI 09 regarding treatment of sensitive electronics PRA RAI 10 regarding circuit failure probabilities PRA RAI 11 regarding counting and treatment of Bin 15 electrical cabinets PRA RAI 12 regarding treatment of HEAF PRA RAI 13 regarding MCR modeling PRA RAI 14 regarding credit for MCR abandonment actions PRA RAI 15 regarding MCR abandonment on loss of control PRA RAI 16 regarding application of the state-of-knowledge correlation (SOKC)
PRA RAI 18 regarding ACDF, ALERF and additional risk of RAs PRA RAI123 regarding other deviations from acceptable methods Provide the following:
a)
Results of an aggregate analysis that provide the integrated impact on the fire risk (i.e.,
the total transition CDF, LERF, ACDF, ALERF, and additional risk of RAs) of replacing specific methods identified above with alternative methods that are acceptable to the NRC.
In this aggregate analysis, for those cases where the individual issues have a synergistic impact on the
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done. It should be noted that this list may change depending on NRC's review of the responses to other RAIs in this document.
b)
For each method (i.e., each bullet) above, explain how the issue will be addressed in 1) the final aggregate analysis results provided in support of the LAR, and 2) the PRA that wili be used at the beginning of the self-approval of post-transition changes. In addition, provide a process to ensure that ali changes will be made, that a focused-scope peer review will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings will be resolved before self-approval of post-transition changes.
c)
In the response, explain how RG 1.205 risk acceptance guidelines are satisfied for the aggregate analysis. Additionally, discuss the likelihood that the risk increase in any individual fire area would exceed the acceptance guidelines, and if so, why exceeding the guidelines should be acceptable. If applicable, include a description of any new modifications or operator actions being credited to reduce delta risk as well as a discussion of the associated impacts to the fire protection program.
d)
If any unacceptable methods identified above will be retained in the PRA and will be used to estimate the change in risk of post-transition changes to support self-approval, explain how the quantification results for each future change will account for the use of these methods.
3a. See Enclosure 1, specifically Attachment W for the updated results addressing the resolution of all RAIs.
3b. See the table below for documentation of how the changes in methodology applied to any of the RAIs will be addressed in the final aggregate analysis and in the PRA that will be used at the beginning of the self-approval of post-transition changes.
None of the responses have resulted in changes in methodology from that used at the time of the peer review that would constitute a PRA upgrade as defined in the PRA standard.
Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
sumtalomandying thes PRAIta Number Submittal sbitl n
ntePAta Dae will be used at the beginning of the self-approval of post-transition changes)
E PE-OI A thermal inSulation materials, radiation shielding materials, veritiiatior"
/11/2015 No c*hange :to PRA methods :'..
duct materials,,and soundproofing mater'ials shall be nonconrbustible or linjited co nbustibli~e.*
i* ;
2
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe sseRsubmitta submittal, and in the PRA that Dae will be used at the beginning of Date the self-approval of post-transition changes)
Clarify in the compliance bases whether thermal insulation materials, FE0B radiation shielding materials, ventilation duct materials, and 31/05 N
hnet R
ehd FE0B soundproofing materials that are either permanently or temporarily 31/05 N
hnet R
ehd installed in the plant are noncombustible or limited combustible.
If installed materials are not noncombustible or limited "combustible, FPE-01C describe how these materials are accounted for and managed in the fire 3/11/2015 No change to PRA methods
~protection program.
Revise the compliance statement for Section 3.3.4 of NFPA 805 using one or more of the compliance strategies described in NEI 04-02 FPE-01.01A Appendix B, such as evaluating the condition in an existing engineering 7/6/2015 No change to PRA methods equivalency evaluation or submitting a performance based evaluation approval request in accordance with 10 CFR 50.48(c)(2)(vii).
Provide additional informaf*i0on characterizing the installed con'ditions FP-1.1B that do not meet the NFPA 805, Section 3.3.4, requirement (i.e., types, 7/6/2015 No change to PRA methods,
- FPE-1.01.B quantity, permanent or temporary. installation, locations,, installation,
.detJails, etc.).*
Describe the administrative controls and the criteria for evaluating the FPE-01.01 C acceptability of future uses of materials that do not meet the 7/6/2015 No change to PRA methods requirements of NFPA 805, Section 3.3.4.
Fire.Brigade Training - In Attachment A, the licensee stated that it
~~~~~complies and references P~robedure SA-1 -105, Fire Brigade Training,
- ° Section 4.4.A.1, which includes the NFPA 805, Section 3.4.1 (c)
FPE-02 requirement as a responsibility for the shift manager to assure the fire 2/205 NchnetPRmtod FPE-02
~brigade members have the requisite training and knowledge. Provide 2/205 NchnetPRmtod.
additional detailregarding the trainin'g that is provided tothe fire brigade leader and members that addresses their ability to assess the effects of fire and fire suppressants on NSPC.
3
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that Dae will be used at the beginning of Date the self-approval of post-transition changes)
Table S-2 Item 18 - In the compliance bases in Attachment A for NFPA 805, Sections 3.10.1 (2) and 3.10.3, the licensee refers to a required action in Attachment 5, Table S-2, Item 18 of the LAR. Attachment 5, FPE-03 Table S-2 does not include an Item 18; however, Attachment S, Table 2/205 NchnetPRmtod FPE-03 S-2, Item 17 appears to address these elements. Confirm that2/205 NchnetPAmtod Attachment S, Table S-2, Item 17 is the correct reference for the implementation item or provide the correct implementation item for the Halon system actions identified in the LAR.
'=i.. ~Attachmrfent'L Approval Request 1 - Clarify if Attach'men~tL, Appr~oval:
'.:i i=
FP-0 Request 1, is also applicable to NFPA 805, Section 3.11.3(2), and2/205 NchnetPRmtod FpE'04 revise Approval Request 1 as necessary to accomnmodate the additional 2//05 N
hne oPAmtol Provide further details that describe the extent of use of extension cords that are located above the suspended ceilings, such as number, length, 4/13/2015 No change to PRA methods FPE-05A size, use (e.g., types of electrical loads), and if the extension cords are for permanent or temporary use.
...Describe the adminis~trative controls that ar:e (or will be) in place to
.. *=
- FPE-05B maintain the technical bases for the request (e.g:, prevent/limit future 31/05 N
hnet R
ehd FP-5 placemnent of ignition sources and combustiblerrmaterials~periodic
°
/1205 NchnetPRmtod "surveillance above the ceiliri~g etc_.).
Clarify if the Nuclear Safety Capability Assessment (NSCA) credited cables that are routed in metal conduit above the suspended ceiling FPE-05C(i) need to be free from fire damage in order to support a nuclear safety 2/9/2015 No change to PRA methods function or fire risk evaluation (FRE) for a fire in the fire areas described in this request.
, ~The NSPC discussion implies fire damage will not occur because, in, p iart, thie cables are protected* in metal conduit Or in' metal covered trays.
i*:"'.....
Metal conduit and metal trays are not generally sufficient to protect FPE-05C(ii) cables from exposure fire damage. P~rovide additional discussion and/or "2/9/2015 No change to PRA methods details that provide assuranc*e that NSCA credited cables are not susceptible to damage from extension cords or other Potential fire hazards in the area above the ceiling.
- i 4
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
The licensee appears to conclude that because defense-in-depth (DID)
Echelon 1 is satisfied, that Echelons 2 and 3 are also satisfied. The FPE-05D NRC staff notes that DID is based on a balance of the three echelons.
4/13/2015 No change to PRA methods Provide additional details related to how Echelons 2 (fire detection and suppression) and 3 (safe shutdown) of the DID concept are maintained.
In the bases for the request, the licensee stated that this request is applicable to any fire area-containing a sprinkler system, as identified in P,RE-06A,
Attachment Ci Table C-2. Discuss the bases for limitin'g this hot work" 2/9/201 5',
No change to PRA methods :
~~procedure req'uest to on~ly fire areas that contairn required fire sprinkler systems identified in Attachnment C, Table C-2.
Describe the hot work administrative controls for the fire areas that contain a suppression system that is not identified as a required FPE-06B suppression system in Attachment C, Table C-2, and whether the 2/9/2015 No change to PRA methods administrative controls are different than those for fire areas with required tire suppression systems.
~ ~~~In, the bases for the requesti, the licensee stated that permanent
,:o...
' =
combustibles located within 35 feet of the work area that cannot be '"
removed must be covered with the appropriate style of blanket. Clarify if =
,FPE-06C
- the "appropriate style of blanket" is a listed or approved welding curtain, 2/9/2015
,No change to PRA methods
, i
- welding blanket, welding pad, o~r equivalent, as required by NFPA 51B,
" 'standard for Fire Prevention, During welding, cutting, :and Othe~r"Hot
" ~Work."
Describe any additional actions/controls to be used when hot work is FPE-06D performed in fire areas/zones where one or more sprinkler systems are 2/205 NchnetPRmtod impaired above and beyond those taken for any other hot work activity 2/205 NchnetPRmtod conducted when sprinklers are in service.
5
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NCRAI IseRsos accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Section 4.1.3 :and Attachment' I, Table 1.1, "Definitiion of Power *Block,"'
'L ii i *!*
of the LAR state that b~uildings that are required for n~uclear plant
° *,= i, i,,i*i
.iii,
!i
.;i:I operations (ie., required to meet t~he n~clear safety or' radioactive.
j;elease (RAD) perfo~rmance criteria identified in Sections *1.5..1 and,,:
reviewed thie plant for compliance with the RAD performance criteria,..
. ~and the res~ilts are documented in Attachment E, Which includes the followinig comnpartments as screened in for RAP revie*W, but are not.
de.....
scribed as part o*f th~e power block inAtcnet
,TbeI i
, *,o Interim Resin* Stor~age Facility (Lake Davies) "
- /:
.~ Material,Ri-0ce~sng Facility'
-t, i;*
FP..
~i: : Offi*e 'andc Training Facility
-.*-* 1:"
11 P-7
.*Orgngal Steam:Generator Storage Facility 3/11/2015.
No* ch to"PRA methods
[,":
- i"*
":*" Pre-Assembiy Facility"(Upper° Laydownh Area);
°'t i
° ~Sewage Treatment Plant.i
"¢t:*','", 't~'::*
Describ~e the, basis for excluiding these structures from -the power blocki,
,[!;
.',,:,tL
. "°: -ft h~~a~sedl on the criter'ia stated in se(*tion 4.1.3, "..th0'se~tha'tcontain
- *
- , e q u i p m e n t r e q u ir l, d t o m e e t t h e n u J c l e a r s a f e t y a n d R A D c r i t e r 'i a -..
., " : t ',,,,
! '., * ' ' :,i i,...
and.(consequently, exclusionfr;
- the, PA,80.5, Chapter 3 elements[
!"':I:
FP 8 Provide the basis for excluding the 45'-0" elevation of the North Service 2/9/2015 No change to PRA methods FPE-08 Building from the power block.
-'when there are less tha*n 5 fire briga~de members dhsite.: :."
Electrical Raceway Fire Barrier System / NFPA 805 Chapter 4 - Clarify FPE-1 0
if there were any cable resolutions in the NSCA that credit an ERFBS to 2/205 NchnetPAmtod FPE-10 protect the affected cables to meet NFPA 805, Chapter 4. If yes, then 2/205 NchnetPRmtod clarify if the ERFBS are in compliance with NFPA 805, Section 3.11.5.
6 6
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RA!
ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
The approval request states that the storaeliodo afrea cotin thseareaus is extpescand, qatties oficomb utbe tothe g
ncuig u d
limited nmeofixdgition rakstourage, and that roims ond thoedquralcntitoes of sotworag antrenin FPE-11 adnomnistibelyesmterablishe Dsrby the fiypes rot iedgntion enieeshoughctes in
/321 ocanetR ehd oA96r neartibefecdi storage airooms, a' nd10 andh exposuretfire hzreltatie'81/05 NchnetPAmeod contiutiponaat to mutihe lacd-ingstfrahe nond ptrentalyintedwo tchned stoohrecmusil materials.i hs om Theappova rquet sate tat he ikliod*
of at fies intheerasi FPE-1 lB Dicmusstibe maerales.
Desribteedn the typstalto of fieagiin sourcestin, 81c05 N
hnet R
ehd sorkearethfencedion sytmtopragie aeasy warnnd thxosure fire hazrd thatened FPE-1 could propageateas toumn thefecdi storage anda potntialyignithe theu stlored muanteials.
o-rae od n te obsils.Dsrb
/321 o hnet R
ehd the apprtoval frequs stotetionavalal in the aevnofa ofir ithe storageros tharewt spipeourmticdfe supi-dpressiongispoovidedation h ardeasingbovers).
7
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Element 1 of defense-in-depth, as described in NFPA 805, Section 1.2, 1
~~is associated with fire prevention, which includes controlling the.
elements Of fuels (i.e., Comb-ustibles) and~ ignition tlhat are necessary for FP-1E fire tO occur. In Approval ReqUest 7, the discussion of Element 1 8/13/2015 No change to PRA methods FPE-11E (Echelon 1) of defense-in-depth only addresses ignition sources.
., ~~Provide additional discussion of how the storage and use of untreated
~wood in Rooms 1101 and 11:09 meets or has Compensated for Element 1 of defense-in-depth relative to control Of combustibles.
For use of non-metallic raceways (conduits), the approval request states that the use of non-metallic conduit is required by CCNPP drawings/specifications for concrete-embedded and underground installations where metal raceways do not meet design requirements.
The approval request further states that new applications of non-metallic conduit are approved and evaluated in accordance with design procedures which include a review of fire protection design FPE-12a(i) requirements. Describe the acceptance criteria that is included in the 8/13/2015 No change to PRA methods current design procedures that allow the installation of non-metallic conduit at CCNPP, and clarify whether the current design procedure includes (or will include) criteria that involve (A) satisfying the nuclear safety and radiological release performance goals, performance objectives and performance criteria, (B) maintaining safety margins, and (C) maintaining fire protection defense-in-depth. Also, identify the implementation item to revise the procedure(s), if needed.
L,
'~~For use of non-metallic race ways (conduits), the licensee _stated during
'~~~
aJune 4i,2015, public meeting that Apprioval Request 8 w Il be r-evised
°"
FPE-12A(i~i) and resubmitted to remove the use of non-metallic raceways (conduit) 8/13/2015 No Change to PRA methods in applications that are neither embedded in concrete nor buried
~
~~underground.. Confirm that this is correct.
/:
For use of electrical metallic tubing (EMT), provide a technical FPE-12B(i) justification which relies on neither an unendorsed edition of NFPA 805 81/05 N
hnet R
ehd (i.e., the 2015 edition), nor another unendorsed NFPA code (i.e., NFPA 81/05 N
hnet R
ehd 70).
~~~~For use of electrical metal'lic tubing (EMT), confirm that the EMT is not....
"FPE-12B(ii) installed in any location subject to physical damage, and describe the "81/05 N
hnet R
ehd criteria which will be used to ensure that.future EMT installations will not 81/05 N
hnet R
ehd be in locations subject to physical damage.
8
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NCRAI IseRsos accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
For use of electrical metallic tubing (EMT), provide additional detail regarding the extent of installation of EMT at CCNPP (e.g., a few FPE-12B(iii) specific fire areas or throughout the plant). Confirm whether EMT is 8/13/2015 No change to PRA methods used for any power, control, or instrument cables associated with NSCA components.
For use of electrical metallic'tubing (EMT), the-licensee stated that EMT is non-combustible and will not contribute to the fire load., and states that neither non-EMT nor EMT metallic conduits are c~reditedJ in the""*
FPE-12B(iv)
NF:PA 805 analyses to prevent or delay damage due to the fire.
8/13/2015 No change to PRA methods Confirm that fire damage and circuit failure assumptions for circuits in non-;EMT metallic conduit, and EMT :condUit, are9 the same, or describe...
the differences.*,
SSA-01A Clarify if any rising stem valves involved in an RA are subjected to fire 2/9/2015 No change to PRA methods damage.
If any of the valves in the fire area of Concern being repositioned by an RA are risinig stem valves, then clarify if an engineering evaluation, was
~~~performed to evaluate the exposure fire damage to manual valves and i SSA-01 B piping to determine if the exposure to fire would adversely impact their 2/9/2015 No change to PRA methods ability to perform their pressure boundary or safe shutdown function. If,:,"
,~:*,
r*;
evaluation.
Clarify if cables that supply loads not required to meet the NSPC off of the nuclear safety buses are classified as "required" cables. If non-SSA-02A nuclear safety cables are not included, then provide the justification for 3/11/2015 No change to PRA methods not considering the failure of non-nuclear safety cables in meeting the breaker coordination criteria for protection.
The alignment basis states that plant modifications have been identified
~~to 'achieve selective coordination of breakers/fuses and identified as
...SSA-02B; implementation items in Attaclhment S, T'ableS:-2. iden~tify the specific 3/11/2015, No change to PRA methods modifications that are required to achieve the selective coordination of breakers/fuses.
9
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
For multi-conductor cables routed in dedicated conduit, provide a description if intra-cable hot shorts (wire-to-wire shorts) are considered SSA-03 as a potential impact of fire damage on required position of the NSCA 3/11/2015 No change to PRA methods equipment (i.e., the function of the initial position of circuit contacts are not affected by intra-cable hot shorts).
In Section.4.2.1.2, subsection "'Methods to Maintain 'Safe and stable' and Extend Hot Standby ConditionS," of the LAR, local manualiactions S
are described to align variOus systems and functions. In Item No. 8, the licensee stated that should alternating current (AC) charging sources be lost, local manual operator action may be required, and that station SSA-04A batteries are capable of providing a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of 125 V direct 3/1 1/2015 No change to PRA methods current power to their respective loads during a station blackout Without i~~~~A Acharging soUrces. The licensee further stated that this time':-o S
allowance credits securing 1 iNV1Tll1 in the cable Spreading room o,
- i'""
(CSR) within 45 minutes. Clarify if this local manual action is credited as an RA in any fire area.
In Section 4.2.1.2, subsection "Assessment of Risk," the licensee stated that the ERO provides sufficient resources for assessment of fire damage and completion of repairs to equipment necessary to maintain hot standby for an extended period, transition to cold shutdown, or SA4B return to power operations as dictated by the plant fire event. Describe 3/11/2015 No change to PRA methods SSA-04B if any repair activities are necessary to maintain hot standby for an extended period (safe and stable conditions), including a detailed description of the specific repairs that would be needed, the success path(s) being restored, and the time frame required to complete the repair.
Clarify if th~e control room remains the command and control location for SSA-05A a fire in Fire Areas 16 and 17, and if so, discuss how the RAs at the 3/11/2015 No change to PRA methods PCS are evaluated for compliance with NFPA 805, Section 4.2.4.
10
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NCRAI IseRsos accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes) in Fire Areas 16 and 17, there are RAs at the PCS that are not associated with a variance from deterministic requirement (VFDR).
- For Fire Area 16, the RAs are:
61CHECKRXSD1 ; 161CONSERVE1; 1 61SECHTR1 1_13; 1 61ADV1 C43; 1611 C43CONTROL; and 1 61RCSTEMPI.
SSA-05B 3/11/2015 No change to PRA methods
- For Fire Area 17, the RAs are:
1 71CHECKRXSD2;.1 71CONSERVE2; 1 7ISECHTR21_23; 1 71ADV2C43; 17i2C43CONTROL; and 17iRCSTEMP2.
Clarify the purpose of performing these RAs, and whether the actions are required to meet the NSPC required by NFPA 805, Section 1.5.1.
In Attachment G, Table 0-1 of the LAR, disposition of VFDR 16-19-1 credits PAs :at the PCS to energize pressurizer backup heater banks 11
,i:SSA-05C, and 13; however, another non-VFDR related RA (16ISECHR!11J3) is 3/11/2015 No change to PRA mrethods credited to secure the pressu~rizer backup heater baniks 11 anid"13.'
Discu~ss how the contradiCting RAs are evaluated in the NSCA.
in LAP Attachment G, Table G-1, RAs are credited to disposition VFDRs 16-27-1 and 17-25-2 to control atmospheric dump valve (ADV) hand valves to support control of the ADVs at the PCS locations 1 C43 SSA-05D and 2043, respectively. However, the RAs (1 61ADV1C043 and 3/11/2015 No change to PRA methods 171ADV2C43) to control the ADVs at the PCS location do not have a VFDR associated with them. Discuss the method for crediting RAs to support the VFDR disposition without crediting the RA at the PCS.
In Attachment C, Table 0-1, Fire Area 34 is identified as transitio)ning i',
deterministically in Unit 2 (Section 4.2:3,2.of NFPA 805) with no 'VFDRS'
' ~~identified. However, in Attachiment W, Table W-7', Fire Area 34 is..
SSA-O6A identified as transitioning using performance-based methods (Section
-4/13/2015
,No change to PPRA methods
~4.2.4.2 of NFPA 805), VFDRs are identified, RAs are credited, and the
~~~~~risk of the PA was calculate'dClarify the correct nuclear safety
~~~~Compliance strategy for Fire Area 34.
11
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
In Attachment C, Table C-1, the following fire areas are identified as transitioning deterministically with no VFDRs identified. However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7 (for Unit
- 2) identify that these fire areas have VFDRs identified. Further, an FRE was performed that calculated a delta core damage frequency (CDF) and delta large early release frequency (LERF) value as follows:
SSA-06B Unit 1: 2, 8,13, 18, 18A, 22, 23, 25, 26, 27, 28, 31, 38, 40, and 2CNMT 4/13/2015 No change to PRA methods Unit 2: 3, 4,6, 14, 15, 19, 19A, 21, 30, 33, 39, and 1CNMT For each of these fire areas, clarify the correct nuclear safety compliance strategy, and justify the bases for performing an FRE that is not discussed in the NSCA in LAR Attachment C, Table C-i, and the bases for crediting RAs that are not included in LAR Attachment G, Table G-1.
In Atahmn
,, Tal C-i, the folng fieareas are identifedaas
.":.° Stransitioning using performance-based methods (FRE).to meet the.
.' 'NSPC, and no RAs were credited (either for risk or DID).,However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-.7 (for Unit
calculated:*
SS-6 4
- /13/20)15 N0 change to PRA methb*ds
~ ~Unit 1:1i2', 14,'-15, 19A, 21, 30, 32, 33, 35, 36:, 39, 1 CNMT, and lIS Unit S
2:12,13, 18A, 20, 26, 27, 28, 32, 34, 35, 36, 40, 2CNMT and Is For each of *these fire areas, clarify the correct nuclear safety compliance strategy for these fire areas and tlhe bases for crediting RAs
"°...'*
,...... **..that are not :included in Attachmnent6, Table G-l.*
12 12
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI RAI ssueResp nse subm ittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
In Attachment C, Table C-1, the following fire areas are identified as transitioning using deterministic methods to meet the NSPC, and no RAs were credited (either for risk or DID). However, Attachment W, Table W-6 (for Unit 1) and Attachment W, W-7 (for Unit 2) identifies these fire areas as crediting RAs and the risk of the RA was calculated:
SSA-06D Unit 1: 13, 18, 18A, 22, 23, 25, 26, 27, 28, and 2CNMT Unit 2: 14, 15, 4/13/2015 No change to PRA methods 19, 19A, 21,30, 33, 39 and 1 CNMT For each of these fire areas, clarify the correct nuclear safety compliance strategy and justify the bases for not including these RAs in Attachment G, Table G-l, if these RAs are actually credited in the NSCA.
~Attachment s, Table s-'2,Item7, inv01*veS modifying control circuits for"
°
/
.i:
. :':--:*-""i.
thef Pressuriizer Power Operated Relief Valves (PORVs), 1 (2)ERV4O2 i.'
i
.'J*:
'i; i.
~andl1(2)ERV404, todprevent the PORVs from spuriously oPening.
SSA-07A*
However, VFDRs 16-46-1, 24-26-1, 16-47-1, 24-27-1, 17-41-2, 24-63-2, 4/13/2015 No change to PRA methods
~~~17-42-2, and 24-64-2 involve fire damage to cables which could result o..
°
......in spurious opening of the Pressurizer PORV, and t.he VFDR* *.**i
- o. -* "'-
' dispositionscredits an.RA forDID.
=
i
,i-Attachment S, Table S-2, item 8, involves modifying the control circuits for the auxiliary feed water (AFW) steam admission valves 1 (2)CV4070 and 1 (2)CV4071 to ensure adequate separation such that one set of valves will be available during a fire in either the CSR or switchgear SSA-07B rooms. However, VFDRs (16-22-1, 17-16-2, 16-26-1, and 17-26-2) 4/13/2015 No change to PRA methods involve fire damage to cables that could cause the loss of control and/or spurious operation of 1 (2)CV4070 and 1 (2)4071, and the VFDR dispositions credit an RA either to reduce risk (VFDRs 16-22-1 and 17-16-2) or for DID (VFDRs 16-26-1 and 17-26-2).
13
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Attachment S, Table S-2, Item 11, involves modifying control circuits for the Main Steam Isolation Valves (MSIVs), 1 (2)CV40430P and 1 (2)CV40480P, to ensure at least one solenoid dump valve can be SSAO7C. energized to close the MSIVs. However, VFDRs 16-31-1, 16-32-1, 17-SS-07 23-:2, and 17-24-2 involve fire damage to cables that co'uld cause a loss 4/13/2015 No change to PRA methods of control and/or spurious operation of the associated MSIV, and the VFDR dispositions credit an RA for DID (VFDRs 16-31-1, 16-32-1, 17-23-2, and 17-24-2).
Describe if the fire zone numbers listed in Attachment C, Table C-2, are the same as the room numbers listed in the fire area summary in SSA-08A Attachment C, Table C-i. Describe if the room numbers in Attachment 2/9/2015 No change to PRA methods C correspond with the room numbers cited in the previous licensing actions in Attachment K.
ProVide a.description of the water curtain arrangement, including the.:
sprinkler systems that supply the required sprinkler heads using the SSA-08B current terminology for rooms, fire areas, and/or fire zones such that the 2/9/2015, No change to PRA methods staff can fully understand the installation and how the installation is represented in the VariouS tables in the submittal and the previous licensing actions.
Clarify the basis for discussing the fire suppression effects for a fire SSA-09A modeling performance-based approach when the fire areas used a risk 3/11/2015 No change to PRA methods evaluation performance-based approach.
Provide additional discussion for those fire areas where VFDRs are SSA-09B identified, but the suppression effects discussion states there is no 3/11!/2015 No change to PRA methods NSCA equipment in the fire area.
14
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
In Section 4.5.2.2, the licensee stated that there are no VFDRs that involved performance-based evaluations related to wrapped or embedded cables. However, in Attachment C, Table C-1, Fire Areas 18, 19, 35, 36, and TB/NSB/ACA are performance-based tire areas and credit EEEE, "ECP-1 3-000359-"Generic Letter (GL) 86-10 Evaluation of SSA-1 0
Embedded Conduit in the Turbine Building and Barrier Thickness ot the 31/05 N
hnet R
ehd SSA-10 Floor/Ceiling Barrier between AB-4/AB-5 and 517/51 8," which justifies 31/05 N
hnet R
ehd the acceptability of conduits embedded in the Turbine Building floor slab (elevation 27'), the floor/ceiling slab between stairwells AB-4 and AB-5, and the horizontal cable chases (Rooms 517 and 518). Clarify if the disposition of the VFDRs in Fire Areas 18, 19, 35, 36, and TB/NSB/ACA credit the embedment as evaluated in the EEEE.
Describe the extent that Marinite boards are credited for Chapter 4 separation ("S") and for risk significance ("R) in the Unit 1 and Unit 2 PRA method is changed SSA-11A ContainmentS. In addition, describe the design and plant configuration 4/13/2015 Consistent with the referenced RAI of the Marinite boards and the nuclear safety functions that the passive response fire protection features are protecting.
Provide previous NRR staff approval (if any) for the use of Marinite boards in containment to demonstrate meeting the requirements of PRA method is changed SSA-11B Appendix R, Section lll.G.2, which can be credited to meet the 41/05 cnitn ihterfrne A
SSA-1 1 B requirements of NFPA 805, Section 4.2.3.4, or evaluate acceptability 41/05 cnitn ihterfrne A
using a performance-based analysis approach in accordance with response NFPA 805, Section 4.2.4 NFPA 805 Section 4.1 states that once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design and qualification shall meet the applicable requirement of Chapter 3. Based on discussions SSA-1 1.01 with the licensee during a public meeting on June 4, 2015, the NRC 8/13/2015 No change to PRA methods staff understands that the Marinite boards have been installed and are being credited in a similar manner to metal cable tray top and bottom covers. Describe the Marinite board performance assumptions, as credited in the performance-based analysis for the Marinite board.
15
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NRC RAIaccompanying this RA!
RAI Issue
Response
sbitl n
ntePAta Number Submittal sbitl n
ntePAta Dae will be used at the beginning of the self-approval of post-transition changes)
In Attachment C, Table C-2, the licensee makes reference to "Unit 1 Containment (App-R Purposes Only)" and "Unit 2 Containment (App-R Purposes Only)," for fire protection systems and features. The fire protection systems and features are identified as required for'"S" (Required for Chapter 4 Separation Criteria), "R" (Required for Risk SSA-12 Significance), and/or "0" (maintain an adequate balance of DID in a 2/9/2015 No change to PRA methods change evaluation or FRE).
Clarify the meaning of "Appendix-R Purposes Only" and if these fire protection systems and features are credited with respect to compliance with NFPA 805, Chapter 4.
Section 4.3.2 and Attachment D state that incorporation of the recommendations from the "KS F (key safety function] pinch point" evaluations into appropriate plant procedures prior to implementation willibe done to ensure the reqluirements of NFPA 805 are met. Identify o
SSA-13A and describe the changes to outage management procedures, risk 2/9/2015 No change to PRA methods management tools, and any other document resulting from incorporation of KSFs identified as part of NFPA 805 transition. Include changes to any administrative procedures such as",'Control of.
Combustibles."
For those components that had not previously been analyzed in support of the at-power analysis or whose functional requirements may have been different for the non-power analysis, cable selection was performed in accordance with approved project procedures. Provide a list of the additional components and a list of those at-power2/205 NchnetPRmtod SSA-13B components that have a different functional requirement for NPO.2/205 NchnetPR mtod Describe the difference between the at-power safe shutdown function and the NPO function. Include with this list a general description by system indicating why components would be selected for NPO and not be included in the at-power analysis.
16
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( an final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Section 4.3.1 and Attachments D and H state that the NPO analysis S'
was performed in accordance with FAQ 07-0040, "Non-Power.
Operations Clarifications (ADAMS.Access~ior No.,ML082200528).
- i However, :the LAR did not provide the results of the KSF pinch, point...
SSA-1 3(
analysis. Provide a list of KSF pinch points by fire area that were 2/9/2015 No change to PRA methods identified in the NPO fire area reviews using FAQ 07-0040, including a summary level identification of unavailable paths in each fire area.
Describe how these locations Will be identified to the plant staff for I ~implementation.
During NPO modes, spurious actuation of valves can have a significant impact on the ability to maintain decay heat removal and inventory control. Provide a description of any actions being credited to minimize SSA-13D the impact of fire-induced spurious actuations on power operated valves 3/11/2015 No change to PRA methods (e.g., air-operated valves and motor-operated valves) during NPO (e.g.,
pre-fire rack-out, actuation of or pinning of valves, and isolation of air supplies).
During normal outa.ge evolu~tions, certain NPO credited equipmnent will.
- SSA-1i3E have to be removed from serviCe. Describe the types of compensatory 3/11/2015 No Change to PRA methods' actions that will be used during such equipment down-time.
The description of the NPO review for the LAR does not identify locations where KSFs are achieved via RAs or for which instrumentation not already included in the at-power analysis is needed SSA-13F to support RAs required to maintain safe and stable conditions. identify 2/9/2015 No change to PRA methods those RAs and instrumentation relied upon in NPO and describe how RA feasibility is evaluated, Include in the description whether these variables have been or will be factored into operator procedures supporting these actions.
"* Describe Whether these cross-connecting RAs require staff from both SSA-1i4A
.units. If so, describe how the'feasibility *analYsis reflects the unit* 1 and
.2/9/20)15 No chag oPAmt~d
.Unit 2 staffing, communication, and operation~al in~terface.
hange to RA.method Describe the operational impacts (by fire), if any, on the unaffected unit SSA-14B created by cross-tying these systems. Describe whether Technical2/205 NchnetPRmtod SSA-14B Specification 3.0.3 is entered once the cross-tie with the opposite unit 2/205 NchnetPRmtod has been completed for fire safe shutdown.
17
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Calculation CA07971 - Provide justification that 50 psig of instrument air SSA-15A pressure Will not prevent instrument air operated valves from changing 4/13/2015 No change to PRA methods position.
Calculation CA07971 - Provide justification for limiting the size of the line to 1' soldered joints being susceptible to separation during a fire.
SSA-15B Describe the soldered joints used in the plant instrument air system. For 2/9/2015 No change to PRA methods any soldered joints larger than 1 ", describe how they were treated in the NSCA and Fire Probabilistic Risk Assessment (PRA).
Calculation CA07971 - For Several fire areas in Attachment C (such as Fire Areas 18;,19, 20, 21, and 22), the licensee stated in the method of accomplishment for the vital auxiliaries performance goal that instrument air may be recoverable from the opposite unit plant air system. However, the VFDJRs associated with the fire areas (such as VFDRs 18-16-2, 19-01-'1, 20-02-1, 21-02-1, and 22-05-2) state that
- SSA-15C plant air from the opposite unit cannot be used because of failure of 4/13/2015 No change to PRA methods 1 CV2061 or 2CV2061, and the VFDR disposition credits an RA that involves aligning backup nitrogen to the affected unit control valves.
Clarify the discrepancy between the method described in the subject
~fire areas for achieving the performance' goal, theVFDRs th'at state this method is not available, and the RAs cited in LAR Attachment G for resolution of the VFDRs.
Calculation CA07971 - in Attachment C, the discussion of fire suppression effects on the NSPC for Fire Areas 39 and 40 addresses the impact of suppression damage to redundant instrument air compressors and the saltwater air system, and states that the AFW air accumulators can be charged from the nitrogen system with an RA.
However, the disposition of VFDRs 39-01-1 and 40-01-2, which address fire damage to the respective unit's instrument air system, stated that SSA-15D the VFDR has been evaluated with no further action required. In 4/13/2015 No change to PRA methods addition, the RA to align the nitrogen system to the AFW air accumulators is not discussed in LAR Attachment G for these fire areas. Clarify the apparent discrepancy between the effects of fire damage and suppression damage on the instrument air system and salt water air compressors (SWAC) with regard to the need for an RA. If an RA is necessary to mitigate the suppression effects on the instrument air compressors and SWAC, then describe the feasibility and additional risk of the RA.
18
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NRC RAIaccompanying this RAI RAI Issue
Response
sbitl n
ntePAta submttaradSintheiRAtha NubeSbmtale will be used at the beginning of Date the self-approval of post-transition changes)
In Attachment C, the licensee stated in the summary of vital auxiliaries
~~for Fire Area 178 that the control room and CSR heating, ventilating,
' ~~and air conditi6ning'(HVAO) is not available without an RA and referenced VFDR 178-01-0. However, the disposition discussion for SSA-16 VFDR 178-01-0 states that no further actions are required based on the 2/9/2015 No change to PRA methods performance-based analysis forthe VFDR, and noRAs required for risk Ilo or DID we're identified in Attachment G. Clarify the bases for thel discrepancy between the description of the vital auxiliaries' discussion
., and the VFDR disposition.....
Attachment G / Portable Fans for Switchgear Room Cooling - Describe the location of the portable generators and the location of NSCA structures, systems, and components (SSCs), if any, in the vicinity of SSA-17A these location(s). In your description, include a summary of the 4/13/2015 No change to PRA methods procedure guidance for the use of portable gas generators and how the RA aligns with each of the feasibility criteria of FAQ 07-0030 (i.e.,
training, procedures, drills, etc.).
~~Attachment. G / Fuel for Portable Generators - Describe the type of fuel and quantity associated with the portable generators and the availability SSA-17B
'and the location(s) of sufficient fuel sources to.support maintaining safe:
4/13/2015...
No change to PRiA meth~ods
~and stable conditions for the time period reqtiir~d.
Attachment G / Hazard from Refueling Portable Generators - Provide SSA-17C justification that refueling the generators does not present a fire 4/13/2015 No change to PRA methods exposure hazard to NSCA SSCs.
A ttachm en lt G / T m po rary P o w er C a b lei D 'esc h
i n o
- I
/ee rib e t e inqsta llat o of
- ,;i:
SSA-1 7D temporary power cables, connections to distribution panels, and any 4/13/2015 No change to PRA methods S,
disruptions to fire area boundaries.,
iI Attachment G / Analyzed Ventilation Path Describe the method (e.g.,
SSA-17E the analyzed ventilation path configuration) of providing temporary 4/13/2015 No change to PRA methods cooling when portable fans are used for these RAs.
19
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att, W update RA!
ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
Acceptability of the PRA Approach, Methods and Data - Identify" whether any fire modeling tools and methods have been used in the
.PRA method is changed
...FM-01A development of the [AR that are not disCussed in Attachment J. In 4/13/2015 '
Consistent witih'the refer~enced RAI addition, identify any fire modeling tools and methods that* are discussed in' Attachment J th'at were not used in the fire modeling response
~analyseS performed at the :plant.
Acceptability of the PRA Approach, Methods and Data - It is discussed in the detailed fire modeling analysis that, "the FDTs are not setup for PRA method is changed FM-01 B secondary ignition or for the effects of suppression systems on a fire 4/13/2015 consistent with the referenced RAI scenario." This implies that secondary combustibles were not considered for any fire modeling analysis at the plant, except those response using FDS.
~~AcceptabilitY of the PRA Approach, Methods and Data -! Provide..
justification for ignoring the effects of flame spread and :fire propagation PRA method is Changed FM-OlC:
ifn second~arycombustibles (for exampl~e, cable trays) and the *.
4/13/20'15 consistent With the referenced RAi corresponding heat release rate (HRR) on the calculated ZOI and HGL.
response S* temperature.
Acceptability of the PRA Approach, Methods and Data - Provide PRA method is changed EM-O1D information on how non-cable intervening combustibles were identified 4/13/2015 consistent with the referenced RAI and accounted for in the fire modeling analyses.
response i*I '
- ".',Acceptability of the PRA Approach, Methodsa~nd Data - Typically,
,I....
during maintenance or measurement activities in the plant, electrical F-1E cabinet doors are opened forea certain period oftime. Explain what 4/13/2015 :.No change to PRA methods FM-01E.
admhinistrative Controls are in: place~to mninimize the likelihood of fires...!:*
' i "ihivo1ving such a cabinet, arid describe how cabinets with tempor'ary' open doors were treated in the fire modeling analyses.
20
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Number Submittal submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
Acceptability of the PRA Approach, Methods and Data - Describe the criteria that were used to decide whether a cable tray in the vicinity of an electrical cabinet will ignite following a high energy arcing fault PRA method is changed FM-01F (HEAF) event in the cabinet. Explain how the ignited area was 4/13/2015 consistent with the referenced RAI determined and subsequent fire propagation was calculated, If response applicable, describe the effect of tray covers and fire-resistant wraps on HEAF-induced cable tray ignition and subsequent fire propagation.
~Acceptability of the PRA AppracMethods and Data'- ExplaindoeR mto shne FM-01H w~ifall~ r fn ornter eff
'uet fire arcuneas aondletin th atir modein used05 cosstn it herinece A
FM-al cah'culationsi Zor rovide jutific.pation if w
thee rrfoet asuwrenpticons ineed res32p5ochngetoPAmhd txp'ris of o6the oor fteMRwr sue oblsdo pn antHR ao wll nimes brwee ofsme transioent combsteibles tine aDire s
areab Acce1i iptabilty of thes a
RAsApprio ach, Mheaulthods abndonData t xinmoePAmehdsschne FMescriHewall andcrnrefetsoa werkae paccuted foawrepcfe in the fieDoeln 4/13/2015 cochnsisen wth PAmtheoeeecdsRA calcultfions, ord provide jutiftehication sisfo the e efecsuwere notucnsdred.
resons arse'oaFs.
i ~c bnomn aclain t~ ~sta h
....i...i.....
height.. of.......
the....
M..R..used... in..
the....
cac lto si ahr=~ -7f te....di.n.ons.pe
.fie in..the..D..inpu fil
.s were.......
21
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RA!
provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
The MCR abandonment calculations for a specified ignition source appear to include EDS runs for 10 HRR bins. Appendix-E of NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear PRA method is changed FM-011(iv)
Power Facilities: Volume 1 : Summary and Overview," September 2005 4/13/2015 consistent with the referenced RAI (ADAMS Accession No. ML052580075) uses a 15-bin discretization.
response Explain why only 10 bins were used, and describe how the 1 0-bin discretization was established.
r ii i
i~i..:rodeledlin FDS, and confirm.that oatmn th hod......is....
changed.........
Explin ow he aeand location s if thetia caberetr area tandsten PRAmehdicand FM-Oll(vi) fack weedtrieaddmntaetatteasmdaesad 41/05 consisten~t with the referenced RAI pleatinsl area e consistrent forh pnondthtyns of ieanition sousrces.v estProidaeso tehnia 'jsiiatinonme ~not cim nsidrnieseaiswt h response-isnitonda somurc ista les orte in a c nornmer.
c oa ns:;+*J: +!.!
Explain how the aRea anfeeatono electrical cabinets anedete ranined nt PRA method is changed FM-011(vii) frswherher dtherminued, andr emons tratentwt that thpes ocassumeds areasen and 4/13/2015 consistent with the referenced RAI etieMCRateso the abandomnttme.response i~i.........................
.. P..
A m etho.is.cange FM++:+:+++
Oil+:+ ++*
+ ++v ii)
++++*+++
Provide
++++*+
justifcatio for+:+:+!:!::+++: '+:++++:++++++++++::::+*
+:;+*++:+
- not:;
coni der+"++
i ng:
+
scenar+++
i os?*:;+:++;++
that involve++++++:++'*:++ ::":+*:::++++++:
- ++':*:+++I+
i+i+i*+++,+
- +*+ +.+ ++ *:++ ::
+thatpmsecon atoi ryv+
- +combusibale+siia+n+(i++°s i+n+*+i++"+
++ =+'+
idacn thcbiets?
++:o abad+nen
"+
':+
trcal'cuations.+[*'
4/1P/
++:+:*:++205 conistent+PO.g w~
- ++:+;+++
+:c++*+ith the-" referenced+[?:,
RA+!*:I:++#
Exli o h
Rsfo lcrclcbnt were:".:"
determ:inedan PRA..
method
- -is:.:
changed*;
,.:::*::*+::::i:
FM-0,1:.'?:
(viii):+:
'+::::
wheter he vlue are: consisten with the type?
s) of, cabiets:reset in 4/13201 conisen with:: th refeence RA: I 22
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI RAmbe IseRSponseta submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Provide the technical basis for the material properties that were specified in FDS for the cables inside the cabinets in the MCR. Provide FM-01 1(x) confirmation that the assumed soot yield and heat of combustion values 4/13/2015 No change to PRA methods (the latter either explicitly or implicitly through the specified fuel composition) lead to conservative estimates of the soot generation rate.
S
- Describe the transient fire growth rate(s) used in the control room
'=
IPRA method is Changed Provide the technical basis for the material properties that were specified in FDS for the transient combustibles in the MCR. Provide PRA method is changed FM-011(xii) confirmation that the assumed soot yield and heat of combustion values 4/13/2015 consistent with the referenced RAI (the latter either explicitly or implicitly through the specified fuel response composition) lead to conservative estimates of the soot generation rate.
- / ~~Descrie thehabitability conditions tihat were used todetermine the",=
- ="i;
.i
.:time to :MOR abandonment. FDS 'devices' (temp~erature and optical::',
i::
,i:.density) wer'e placed at a height of 6 feet and at four different locations,:
- F,
ll*,x... :;,;in the MOR: Describe the basis for choosing these locations andd
,i:
i":
PRA,meth~od is changed :'*
- M-0 (ii),:'
- demonstrate that the~se locations are either: representative of wlh&re
- 4/13/2015 consistent with the referenc'ed RAI*:
- operators iare expected tobe, or lead to conserva~tive abandonmfent.
- .
- iresponse:
- = "= : -::
so, provide te~chnical justification for using temp'erature sensor* as a'
- i
.:=:*,
Variations in the input parameters such as ambient temperature, soot yield of the fuel, fire base height, etc., affect the output of FDS calculations. The abandonment analyses for the MCR were performed PRA method is changed FM-011(xiv) using a single set of input parameters for each scenario. Demonstrate 4/13/2015 consistent with the referenced RAI that the FDS calculations obtained using this set of input parameters provide conservative or bounding results. Alternatively, demonstrate response that the abandonment times for a given scenario are not sensitive to variations within the uncertainty of the input parameters.
23
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Art. W update RAI ssueRespnse accompanying this RA!
RAI ssueRespnse submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
EMO F.0I(xv)
Explain how the results of the MCR abandonment time calculations 41/13/2015 No change to PRA methods were used in the FPF3A.
Specifically regarding the MCA - Describe the criteria that were used to PRA method is changed FM-01 J(i) screen multi-compartment scenarios based on the size of the exposing 4/13/2015 consistent with the referenced RAI and exposed compartments.
response Explain how the methods described in Chapter 2 of NUREG-1 805, "Fire Dynamics Tools (FDTs): Qu~antitative Fire Hazard Analysis Methods for the U.S. Nuclear RegUlatory Commission Fire Protection Inspection PAmto scagd FM-01J(i'i)
Program, "December 2004 (ADAMS Accession No. ML043290075) 4/13/201.
onsistentho with theaefrned.
RA were used in the calculations to screen an ignition source based on insufficient HRR to generate a HGL condition in the exposing response compartment. In addition, clarify which FDTs were Used for the HGL calculations.
In the MCA scenario analysis, explain the technical basis of modeling the ZOI as a vertical cylinder with the radius equal to 0.2 times the PRA method is changed FM-01J(iii) ceiling height in scenarios where the fire occurs near the opening 4/13/2015 consistent with the referenced RAI between the two compartments and damages items on both sides response within its ZOI.
- ~~Some, of the FDT calculations make the following assumption: "It is
,*°,
assumed that the forced ventilation of air flow rate is distributed among PRA method is changed FM-01J(iv) the interconnected compartments, especially corridors, based on the
- 4/13/2015 consistent with the~referenced RAI.
volume of the compartments."` Provide technical juStification for this response
~~~~assumption.
The screening process based on the ZOI specifies that if there are cable trays, conduits, or targets on the exposed side of the barrier PRA method is changed FM-01J(v) within the ZOI, which may not be the same as those inside the exposing 4/13/2015 consistent with the referenced RAI compartment, the scenario should be analyzed further. Provide details response about this additional analysis.
"Sp~ecifically regardingr the use Of
_FDS in the CSR (physical,
_ analysis..
units (PAUs) 306 and 302) calculations - It is stated that engineering PRA *method is chage FM-01K~i) judgment is used'to assess that the delay in smoke detector activation,'
41/05:i hne MO
~)
which is associated with cross-train logic that is not possible to 41/05 consistent with the referenced RAI incorporate in FDS, would be in the range of 2 to 1 0 seconds. Provide response technical justification for this estimate.
24
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubminsa submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
The FDS "devices" (temperature and heat flux) were placed at different PRA method is changed FM-0l1K(ii) locations around the switchgear rooms. Describe the basis for choosing 4/13/2015 consistent with the referenced RAI these locations.
response The analysis highlights the location of possible electrical cabinet fires PRA method is changed
.i FM-01 K(iii) that were considered. Provide technical justification for selecting these
.i.4/13/2015 bonsistent with the referenced RAI specific fire locations or-demonstrate that thes* locations lead to.
response
~~~bounding or conservative estimates.
epos A number of transient fires were postulated in the CSRs, but the documentation indicates that the walkdown identified no transient FM-0l1K(iv) combustibles and there were no storage areas for more permanent 4/13/2015 No change to PRA methods combustibles in the fire areas. Provide justification for selection of the transient fire areas and indicate if this selection is dependent on any administrative controls of transient combustibles in the CSRs.
i:.=
TheHRR Used for the cabin'et fires indicates that the Cabinet doors'
- .I'
'i FM-0l1K(v) were assumed to be closed. Provide justification for this assumption 4/13/2015 No change to PRA methods" (e.g., on the basis of the actual plant configuration{ or operational
~~condition).
o As stated in FM RAI 1.b, it is expected that secondary combustibles (ignition, flame spread, and cable tray fire propagation) would be part of PRA method is changed FM-0l1K(vi) the FDS analysis for the CSRs. Clarify how secondary combustibles 4/13/2015 consistent with the referenced RAI were considered in the FDS analysis of the CSRs, and if they were not response considered, provide justification for their omission.
The main horseshoe and back panel cabinet configur'ations consist oif open cabinets with a steel mesh open top with the open sides faCing PRA method is changed FMa ~)
each other across anarrow aisle. The FDS analysis utilizes an HRR 3/11/2015 cOnsistent with the referenced RAI
' FM0!
L(i) case from Ap~pendlix G of NUREG-CR 6850, which assUmes closed cabin~ets. Prov*ide justification for not using an HRR case applicable to resPonse open cabinets or update the analysis with the appropriate HRR.".
During the discussion about the open cabinets, it was also discussed that the current analysis does not consider the potential for fire spread PRA method is changed FM-al1L(ii) across the aisle (i.e., within the horseshoe) from the front to back or 4/13/2015 consistent with the referenced RAI vice versa. Provide justification for not considering this potential fire response spread or update the analysis to include this scenario.
25
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
During the walkdown of the MCR, several combustible items, which could be considered transient fire sources, were observed that could
, potentially have an HRR of greater than 317 kW. Examples include the,
PRArmethod is changed
- ,=
- ,,,* ~selected in the' EDS analysiSis conservative and bounding.
Describe how the installed cabling in the power block was characterized, specifically with regard to the critical damage threshold FM-02A temperatures and critical heat fluxes for thermoset and thermoplastic 41/05 N
hnet R
ehd cables as described in NUREG/CR-6850. If thermoplastic cables are 41/05 N
hnet R
ehd present, explain how raceways with a mixture of thermoset and thermoplastic cables were treated in terms of damage thresholds.
" ",,'* ~Explain how th~ damage thresholds frnoni~*-cable comnponents *(iie.,,:
~~pumps, valves, electrical cabinets, etc.) were determined. Identify any FM-02B nonmcable components that were assigned damage thresholds different 3/11/2015 No change to PRA methods
.. ~~from those for thermoset and thermoplastic cables, a~nd provide a
- i i
- i*,
. ~tec*hnical justification for these damage thresholdS.:
- ,i**
,,,,o.
Explain how exposed temperature-sensitive equipment was treated, PRA method is changed FM-02C and provide a technical justification for the damage criteria that were 4/13/2015 consistent with the referenced RAI used.
response For any tool or method identified in the response to FM RAI l.a above, provide the V&V basis if not already explicitly provided in the LAR (forPR mehdicand FM0 example, in Attachment J).. Provide technical details to demonstrate that 3/11/20A cossenth with theaefrned RA F-03 these models were applied within the validlated range of input..
/05 cnitn ihterfrne A
S parameters, or justify the application of the model outside the validated response range in the V&V basis documents..
Identify uses, if any, of FDS and the FDTs outside the limits of PRA method is changed FM-04 applicability of the model, and for those cases, explain how the use of 4/13/2015 consistent with the referenced RAI FDS and the FDTs was justified, response 26
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that Dae will be used at the beginning of Date the self-approval of post-transition changes)
.Regarding qualifications of users of,en gineering analyses, and E
- 'M-0SA*
":numerical m~d'els (i~e., fire rnodelih'g techniquJes) -* Describe8 the:',*
2/9/2015..
' :* No chg to"Amehd
- ° :
'reqUirements to qualify persosnnel for perfmrring fire modeling calcuiati6'ns: in, the NE=pA 805 tr~ansition?,.
Describe the process for ensuring that fire modeling personnel have the FM-05B appropriate qualifications not only before the transition, but also during 3/11/2015 No change to PRA methods and following the transition.
': ' " '.D s~ib how the qualifications of personnel performing fire modeling
,,'.,,i,...
FM-O5.01 caclain met rwlfnee th requiement of..
N A 8's.to 7/2015l Ndochange to PRAmethods 2.7.3*.4 d uring the,development of the application, before transition,,',,,
- °
<°:
during the transition per'iod, and after tr-ahsition:
i".'I When fire modeling is performed in support of the FPRA, describe how FM-05C proper communication between the fire modeling and FPRA personnel 3/11/2015 No change to PRA methods is ensured.
Regarding the' uncertainty analysis.for fire modeling - Describe how: the' PRA methodis cS'hanged
- '" :"i, F,:
M-06A, ;....
uncertainty associated with the fire model inp*utiparameters was-4./13/2015
., consistent with thereferenced jA.,
PRA method is changed FM-06B Describe how the "model" and "completeness" uncertainties were 4/13/2015 consistent with the referenced RAI accounted for in the fire modeling analyses.
response induced loss of MCR HVAC, Which the 'peer review suggests has a *
, ':conditional Core'dam~age prob*ability (CCDP) of 1.0, by incre'asing the:'
P*RA*0A, likelihood of functional failures in :lieu of assuming their occurrence.
2/9/,20115 m,,N:canet R
ethods
- i
- ,.
OlA Jus~tify the func*tional failures moddeled by th* FP"RA to address this lossI No Iane,*,r-
degradation"of equ~ipment dlue to eleVat*ed tempe'ratures cau~sed by loss:
{
- ,"i'":
- of HVAC as an inCrease in equipment failure rates, and provide a
- technical basis for doing so.!,:,::
' :.... i' 27
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR At., W update RAI ssueRespnse accompanying this RAI RAI ssueRespnse submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
F&O FSS-A5 This F&O states that some PAUs are further divided into "sub-PAUs" and appears to indicate that there is no explicit process for evaluating the fire spread across sub-PAU boundaries, which, as the peer review noted, are not defined by physical barriers. The disposition to this F&O, however, does not discuss such a process, and by PRA method is changed PRA-01 B referencing a sensitivity analysis limited to a number of "representative" 4/13/2015 consistent with the referenced RAI PAUs, suggests that this apparent deviation from acceptable methods response has not been fully addressed for all PAUs for which sub-PAUs have been defined. Explain how the fire effects across non-physical sub-PAU boundaries are identified and evaluated. Discuss how this approach is consistent with or conservatively bounds acceptable methods.
F&O FSS-G4 The disposition to this FOindicates tha the MCA,
,didl not postulate a propagation scenario if doi~ng so would rquir'e "
failure of a penetration seal. The licensee's analysis (CO-FSS-08) -
~~suggests that a similar-approach may have also been followed for other.*..
biarrier type~s (e.g., Walls). As a result, idehtify each'barrier type for S
which propagation Scenarios were not postulated, and provide PRA method'is changed PRA-01 C quantitative justification (e.g., an evaluation demonstrating that MCA 4/13/2015 consistent with the referenced RAI
- : :scenarios involving barrier failure are low, risk, even considering the risk" response
~~~associated with the multi-compartment fire) for not addressing ProPagation. As an alternative; provide updated risk results as part of the integrated analysis requested in PRA RI0,s'umming thegener~c barrier failure probabilities for" each type of barrier present between
° communicating compartments, consistent with NUREG/CR-6850.
F&O GSS-G5 The disposition to this F&O indicates that unreliability values were applied to all normally open, self-closing dampers and doors; however, the disposition neither provides a basis PRA-01 D for the values applied nor mentions active elements discussed 2/9/2015
- No change to PRA methods elsewhere (e.g., water curtains in F&O PP-85-01). Summarize the types of active fire barrier elements credited in the FPRA, and provide quantitative justification for their unreliability and unavailability.
~~~F&O HRA-"B2-01
- Provide justification for the assumption: that moideling PRA-01 E(i)
"adverse" actions as successful is conservative. Note that guidance in 29/05 NchnetPRmtod NUREG-1 921 offers considerations for evaluating fault clearing2/205 NchnetPRmtod strategies in the FPRA human reliability analysis (HRA)..
28
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI RAI ssueRespnse submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
F&O HRA-B2 Clarify how "adverse" actions are addressed by the PRA-O01E(ii)
FPRA HRA dependency analysis, given that these actions are modeled 2/9/2015 No change to PRA methods by failing associated equipment directly within the PRA logic model.
- ' "F&O HRA-B21 - Explain the statemen in Attachment Gthat' [none of the re~covery actions were found to have. an adverse impact on the FPRA." in doing so, clarify how "adverse" risk impact was defined. Note PRA-01 E(iii) tactin in the03 saetht
"[i]f activities (recovery actions or other
- 2/9/2015 No change.to PRA methods advernse riskth post-tir'e operational guidance) are determ~ined to have an avJsrikimpact, they should be resolved during NFPA 805 implementation via an alternate strategy that eliminates the need for th-e action* in the NSCA.".
F&O CS-B1 The licensee's analysis (Appendix F of ECP 000321, "Common Power Supply and Common Enclosure Study")
identifies several MCC 208/120 Volts alternating current load breakers that were not coordinated with their respective feed breakers. The disposition to this F&O indicates that these 1 20V panel breaker2/205 NchnetPRmtod PRA-01 F coordination issues are to be addressed by plant modification; however, 29/05 NchnetPRmtod Attachment S does not appear to contain such a modification, Identify the Attachment S modification(s) being credited to resolve the 120V panel breaker coordination issues identified in the disposition to this F&O.
Inf~ernal EvnsF&O 4 This.F&On0 atsta h
lignmerit,.
strategy assuJm~ed by the PRA for the 00 diesel generator (DG) is not FPRA, modeling of th OC Diesel S
appropriately justified and may be non-conservative. While the Genrato aodligmnt strathegyC Diese PA0A disposition to this F&O Clarifies how alignment of the 00 DO is modeled 2/9/2015 supported by committed operator PA0A in the PRA, a justification for this treatment is not provided. Provide a itriw n
iuao
" i:*
technical and/or procedural basis for the alignment strategy assumed in observations.inevws dsmuao,.,,
~~~the PRA for the' OC DO, anid indicate Whether any OPerator interviews were conducted to support the analysis.
Internal Events F&O 6 Explain how the HRA methods used by the FPRA for developing HEP and joint HEP values are consistent with or PR-2~)
conservatively bound NRC-accepted guidance in NUREG/CR-6850 or See PRA RAIs 02.b.i.01 and NUREG-1921. Alternatively, provide updated risk results as part of the 2/9/2015 02bi.1 aggregate change-in-risk analysis requested in PRAN RAI 03 applying HEP and joint HEP values developed using NRC-accepted guidance.
29
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Provide jus~tification for the conclusion that the HEPs* based on a "Tdetay Adequate justification or an PRA-of zero that might not be appropriate" have a negligible impact on the; PRA,
transition risk results (iLe., Core Damage Frequency (ODF), Large Early
- 7/6/2015 :
adequate Tdelay has been provided 02B(i).01 a
,Release Frequency (LERF), delta CDF and delta ILERF) based on the follH
, curnl orfrmry collective contribution to risk and not on the fraction of HEPs affected.d.
with a Tdelay of zero.,
Provide justification for the conclusion that the HEPs based on a "Tdelay Adequate justification or an PRA-of zero that might not be appropriate" have a negligible impact on theadqteTtyhsbenpodd 02~)0b post-transition self-approval risk results (i.e., CDF, LERF, delta ODE 7/6/2015 adequate THErhs, berentl prfovidedly andi).lb delta LERE) based on the collective contribution to risk and not on forh alEs curentlyf oer formry the fraction of HEPs affected.
If it is not possible to demonstrate that the HEPs based on a "Tdelay of Adqut jutfcto ora P' RA-
- zero that might ndt be appropriate' have a negligible im'pact on both th~e Aeut utfctora i, 0B~i)01 c
transition and post-transition self approval risk results, then provide 7/6/2015 fradlequatea~
EsTdetacrety has b0eenoproreYprvided updated riSk 'results as part~of the Change-in-risk analysis, reque'sted in with a T~eiayOf zero'......
- PRA RAI 03, using apPropriate timing.
Internal Events F&O 6 NUREG-1 921 indicates and NUREG-1 792 (Table 2-1) states that joint HEP values should not be below 1.0E-05.
Confirm that each joint HEP value used in the FPRA below 1.0E-05 See PA0Bi)0aadPA PRA-02B(ii) includes its own justification that demonstrates the inapplicability of the 2/9/2015 PA0Bi)OaadPA NUREG-1792 lower value guideline. Provide an estimate of the number 02B(ii).01 b of these joint HEPs below 1.0E-05 and at least two different types of justification.
30
ATTACHM ENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR At., W update RAI Issue
Response
accompanying this RAI Numbe Submttal submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Justifications for each joint HEP value less than 10-5 have been developed.
The response PRA RAI 02.b.ii.01 stated that justification will be rodeajsiiainfor each HEP value less than 10s that will be available when the response to PR-rovaidedcnimta a justificatio n o ahsc E
au a
en RAI PRA 03 is submitted. The PRA reaind, onfrm ha a ustfictio fo eah uchHERvale hs ben 7/6/2015 final number of joint HEPs less 02B(ii).01a developed, provide further examples of justifications developed, andthtaelstan1-adsvrl provide the number of such HER values that will be retained,.xml utfctosta a
requested in 02.b.ii.01 but not provided in the July 6, 2015, response is provided as an augmented response to PRA RAI 02.b.ii.01 or as part of the response to PRA RAI 03.
For all other joint HEPs, apply a lower bound value of 10- in the FPRAPRmehdicand PRA-that will be used to support post-transition evaluations, and provide 7/6/2015 consistent with the referenced RAI 02B(ii).01 b updated risk results as part of the aggregate change-in-risk analysis rsos requested in PRA RAI 03.
rsos PRA Integrated Analysis - Results of an aggregate analysis that provide the integrated impact on the fire risk (i.e., the total transition CDF, LERF, ACDF, ALERF, and additional risk of RAs) of replacing specific methods identified above with alternative methods that are acceptable to the NRC. In this aggregate analysis, for those cases where the issued with An integrated analysis has been PRA-03A individual issues have a synergistic impact on the results, a this performed incorporating all of the simultaneous analysis must be performed. For those cases where no submittal cages in the referenced RAI synergy exists, a one-at-a-time analysis may be done. For those cases responses that have a negligible impact, a qualitative evaluation may be done. It should be noted that this list may change depending on NRC's review of the responses to other RAIs in this document.
31
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results-NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI RAI ssueRespnse submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
PRA Integrated Analysis - For each method (i.e., each bullet) above, An integrated analysis has been explain how the issue will be addressed in 1) the final aggregate performed incorporating all of the analysis results provided in support of the LAR, and 2) the PRA that will isewth change nterfrne A
PRA-03B be used at the beginning of the self-approval of post-transition changes, this responses. The impact of these In addition, provide a process to ensure that all changes will be made, submittal changes is detailed in this table.
that a focused-scope peer review will be performed on changes that are NO methodology changes that PRA upgrades as defined in the PRA standard, and that any tindings would require focussed scope peer will be resolved before self-approval of post-transition changes.
review have been identified.
. " See RAI submittal Enclosure 1 for udtdLRAtcmnPIA Integrated Analysis - In.the response, explain how RG 1.205 risk demntating LRAtahmet Wl G125 S
acceptance guidelines are satisfied for the demoneaayss rstratingtathatgualllin 1.205
-Additionally, discuss the likelihlood thtte agginregate an anysirs cetnegieie r
indvidal ireare wold xced tat he iskinceas inanyissued with met. None of the fire areas result PRA.03C whi ecedi ngua tire grawuidelineed the acceptance guidelines, and if so, this in increases that exceed the RG" wh xedn h~udlnsShould be acceptable. If applicable, sumta 125gidlns Adtoa include a'description of any new modifications or operator actions being modifiitalcationsid inesupporttiofa the credited to reduce delta risk as well as a discussion of the associated Attachm eniicton W rsulpots arte spcfe impacts to the fire protection program.
- in the changes to Att. S (Enclosure PRA Integrated Analysis - If any unacceptable methods identified above PRA methods have been changed will be retained in the PRA and will be used to estimate the change in issued with cesonsisent wit therefrcedpRabl PRA-03D risk of post-transition changes to support self-approval, explain how the this rsoss ouacpal quantification results for each future change will account for the use of submittal methods are used in the analysis these methods.
supporting the final RAI 03 quantification.
Transient Fire Placement at Pinch Points -Clarify how *pinch points"....
PRA methods have been changed SPRA-04A.
'were identified and modeled for general transient fires andI transient 4/13/2015
'consistent with the referenced RAI tires due to hot work.
responses Describe how general transient fires and transient fires due to hot work PRA methods have been changed PRA-04B are distributed within the PAUs at Calvert Cliffs. In particular, identify 41/05 cnitn ihterfrne A
the criteria used to determine where such ignition sources are placed 41/05 cnitn ihterfrne A
within the PAUs.
responses 32
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Transient Influencing Factors / FAQ 12-0064 - Appendix H of the LAR does not indicate that FAQ 12-0064, "Hot Work/Transient Fire Frequency Influence Factors," dated January 17, 2013 (ADAMS Accession No. ML12346A488), was used in preparation of the FPRA.
According to this FAQ, transient influence factor may not be assigned a ranking value of 0, unless associated ac~tivities and/or entrance during power operation are precluded by design and/or operation. The PRA method is changed PRA-05 licensee's analysis (Table 0-2 of CO-iGN-001 ) indicates, however, that 4/13/2015 consistent with the referenced RAI a large number of PAUs are assigned ranking values of 0 for one or more of the transient influence factors. As a result, clarify whether response ranking values assigned to transient influencing faCtors were developed consistent with the guidance in NUREG/CR-6850 and FAQ 12-0064, in particular Section 6.5.7.2, and if not, provide justification. If justification cannot be provided, then provide treatment of transient influence factors
~consistent with NRC guidance in the integrated analysis provided in response to PRA RAI 03.
PRA-06A Reduced Transient Heat Release Rates - Identify all PAUs for which a 4/13/2015 See PRA RAI 06.01 reduction in the HRR below 317 kW for transient fires is credited.
For each location where a reduced HRR is credited, describe the PRA-06B administrative controls that justify the reduced HRR, including how 4/13/2015 See PRA RAI 06.01.
location-specific attributes and considerations are addressed.
Provide the results of a review of records related to violations of PA0C transient combustible and hot work controls, including how this review 4/13/2015 See PRA RAI 06.01 PRA-06C informs the development of administrative controls credited, in part, to justify an HRR lower than 317 kW.
Clarify how the administrative controls currently in place for switchgear room PAUs 311, 317, 407 and 430 can be used, in conjunction with PRA-06.01 specific attributes and considerations applicable to these locations, to 8/13/2015 No change to PRA mnethods support a justification for selection of a screening HRR that is lower than the 317 kW.
33
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAl Issue
Response
accompanying this RAI Numbe Submttal submittal, and in the PRA that Dae will be used at the beginning of the self-approval of post-transition changes)
Self-Ignited and Caused by Welding and Cutting / FAQ 13-0005 (Hot Work) - Appendix H of the LAR does not indicate that FAQ 13-0005,
'Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting," dated June 26, 2013 (ADAMS Accession No. ML133228260),
PRA-07 was used in preparation of the FPRA. Explain whether the treatment of 3/11/2015 See PRA-07.01 self-ignited fires and fires caused by welding and cutting is consistent with FAQ 13-0005, and if not, provide justification. If justification cannot be provided, then provide treatment of self-ignited fires and fires caused by welding and cutting consistent with NRC guidance in the integrated analysis provided in response to PRA RAI 03.
- ~uu "Mod;ify the "FPRA's trea~tment of selfu.--ign**ited cable fires and cable fires ':
due to welding and Cutting to be consistent with accepted methods (i.e.,
PRA method is changed P*RA-07.01 NUREG/CR-6850 and FAQ 13-0005), and provide updated risk results 7/6/2015,:'consistent with the referenced RAI as ~a~rt of the aggr~egate Change-in-;risk analysi§ requ~sted in PRA RAI response.... :!,
- 03.
Junction Boxes (FAQ 13-0006) - Appendix H of the LAR does not indicate that FAQ 13-0006, "Modeling Junction Box Scenarios in an Fire PRA," dated May 6, 2013 (ADAMS Accession No. ML13149A527), was PRA-08 used in preparation of the FPRA. Explain whether the treatment of3/125 SePR-80 junction box fires is consistent with FAQ 13-0006, and if not, provide 31/25 SePA-.0 justification. If justification cannot be provided, then provide treatment of junction box fires consistent with NRC guidance in the integrated analysis provided in response to PRA RAI 03.
- i",
,Modify the FPRA's treatment of Junction box~fires to be consiste~nt with: '
PRA mhethod is chianged PA001 accepted methods (i~e., NUREG/CR-6850 and FAQ 13-0006), and 7/6/2015
'consistent with the referenced RAI PA0.1 provide updated risk results as part of the aggregate change-.in-risk response
~ ~analysis requested in PRA RAI 03. :
.=,
34
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att, W update RAI Issue
Response
accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Sensitive Electronics (FAQ 13-0004) - The NRC staff could not identify in the LAR or licensee's analysis a description of how potential fire damage to sensitive electronics was modeled. Though the treatment of sensitive electronics may be consistent with recent guidance on the modeling of sensitive electronics, Appendix H of the LAR does not cite FAQ 13-0004, "Clarifications Regarding Treatment of Sensitive Electronics," dated December 3, 2013 (ADAMS Accession No. MLPRmehdicand PRA-09 13322A085), as one of the FAQ guidance documents used to support 4/PRA1mnssentho wish chaefrnged RA the FPRA. Describe the treatment of sensitive electronics for the FPRA 41/05 cnitn ihterfrne A
and explain whether it is consistent with the guidance in FAQ 13-0004, response including the caveats about configurations that can invalidate the approach (i.e., sensitive electronic mounted on the surface of cabinets and the presence of louvers or vents), If the approach is not consistent with FAQ 13-0004, justify the approach, or replace the current approach with an acceptable approach in the integrated analysis performed in response to PRA RAI 03.
Conditional Probabilities of SPUrious Operations - Provide an
" ~assessment of the assumptions used in the Calvert Cliffs FPRA relative.
to the updated guidance in NUREG/CR-71 50, Volume 2, specificallyPRmehdscand PRA-1 0 addressing each of the above items. If the FPRA assumptions are notPRmehdicand bounded by the new guidance, provide a justification for each 4/13/2015 consistent with the referenced RAI difference, or provide updated risk results as part of the aggregate response change-in-risk analysis requested in PRA RAI 03, utilizing the guidance in NUREG/CR-71 50.
Counting and Treatment of Bin 15 Electrical Cabinets - Per Section 6.5.6 of NUREG/CR-6850, fires originating from within "well-sealed electrical cabinets that have robustly secured doors (and/or access panels) and that house only circuits below 440V" do not meet the PRA-11lA definition of potentially challenging fires and, therefore, should be PRA method is changed excluded from the counting process for Bin 15. By counting these 4/13/2015 consistent with the referenced RAI cabinets as ignition sources within Bin 15, the frequencies applied to response other cabinets are inappropriately reduced. Clarify that this guidance is being applied, If not, then address the impact as part of the integrated analysis performed in response to PRA RAI 03.
35
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR At., W update RAI ssueRespnse accompanying this RAI RAI ssueRespnse su bmittal, and in the PRA that N u m breubitt a
be used at the beginning of Date the self-approval of post-transition changes)
Clarify 'it the criteria used to evaluate whether electrical cabinets belOw:
Nocag t:R'etosu PA11B 440V are 'well sealed' are Consistent with guidance in Cha~pter 8 of 3/11/2015e t
PRA isupa edtconsisten t
wit PRA11B Supplement 1 Of NUREG/CR-6850. If not, then address the impact as,,
1125 RA resupdae onseistneeded th part of the integrated analysis perfor~med in response to PRA RAI 03.
All cabinets having circuits of 440V or greater should be counted for purposes of Bin 15 frequency apportionment based on the guidance in No change to PRA methods, but PRA-11C Section 6.5.6 of NUREG/CR-6850. Clarify that this guidance is being 3/11/2015 PRA is updated consistent with applied. If not, then address the impact as part of the integrated RAI response as needed analysis performed in response to PRA RAI 03.
For thOse cabinets thdat house cirCUits of 440V dr greater, prop~agation of "i.
fire outside the ignition source should be evaluated based on guidance in Chapter 6 of NUREG/CR-6850, which states that "an arcing faultNocngtoPAmtdsbu co0uld compromise pa.nel integrity (a~n arcing fault Could bum through the
- N hnet R
el~s u
PRA-1 1D panel sides, but this should not be confused with the high energy arcing 3/11/2015 PRA is.updated consistent With fault type fires)." Desc~ribe how fire propgagation outside of cabinets
- ePRA
-1respons as.neede
-also greater-than 440V is evaluated (including those that are considered PA1...
1
~~~"well-sealed"). If propagation is not evaluated, then address the impact as part of the integrated analysis' performed in response to PRA RAI 03.
Confirm whether fire propagation outside of well-sealed and robustly secured cabinets that are not MCCs, but do house circuits greater than 440V, is evaluated consistent with guidance in NUREG/CR-6850. If it is PRA method is changed PRA-1 1D.01 not, then provide updated risk results as part of the aggregate change-7/6/2015 consistent with the referenced RAI in-risk analysis as requested in PRA RAI 03. The updated risk results response - also, see PRA-11 D for evaluating propagation of these cabinets should be consistent with NRG-accepted guidance.
36
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Dae will be used at the beginning of Date the self-approval of post-transition changes)
, ",,71*
Hi~gh Energy Arcin'g F~aults The NRb Staff could hoti identiify in thle 'L*R
~.::'
,V:Z
"- or licensee's analysis a description of how HEAF were modeled. The.
- ,*:i':;"
licensee's analysis (e.lg., Appendix B to CO-FQ-001) appears to indicat
....i
" ':'"':i?that HEAF igniti6n sourceS, ar~e combined t'ith other ignition soUr~ces i *,
- ji' (e.g., transients) to formn fire scenarios. Per Appendix P of NUREG/OR-i
'i: i*:!*,
6850, however;,,HEAF events and other types of fires have different
!.,:.,,il i::,o
'. i,, *!,.:i:
n #on-sUp)pre~ssion *probabilitycurves. in addition, thie NRC Staff's
...interpretation of the N U-REG/CR-6850 guidance is that the grOwth of a
°.
PR'- 2" fire *subsequenit to a HEAF event, unlike othai-types of frems,"
,i:
PRA mdethod-is changedl
}
i!
PRA-2 "
instantaneously starts at a non-zero HRR because of the intensity of the 4/13/2015' 'consistent with the referenced RAI'
,::...initial heat release from theHEAF. Asoa result, provide a detailed
- {..
re,*.on e
- *'/
.justification 6f~the&FPRA S treatm'ent of HEAF~evenits and the ens~uing,
! i
{:
" " i.,°!
- ~fire that includes a discussion of conservatisms and non-b:onservatisms
- .',,*i: *,
pproach with an acceptable appro0ach in the integrated analysis S
performed in response to PR A' RA! 03: Ndte theft the response Should !',-i..,
- *,
- 1 address the tre-atment of all HEAF Scenarios', including in the' HGL MOR Modeling - The licensee's analysis (Section 11.1 of CO-FSS-007) appears to assume that all of the wiring inside MCR control panels is qualified, even though unqualified wiring is known to be present as well.
PRA-13 Describe how the presence of both qualified and unqualified wiring is 4/13/2015 See PRA-13.01 incorporated into the NUREG/CR-6850 Appendix L evaluation.
Alternatively, provide treatment of qualification that is consistent with or bounds the actual MCR configuration in the integrated analysis provided in response to PRA RAI 03.
i*0~Th epps oPRA RAl,13 (in t~he letterdated April113, 2015),,;:.
[ersps
.part of the respons'e to PRA RAI 03 to address the NRC staff's Robservation thegarding the p5resence of both qualified anid unqualified.
PRA....
method' is changed,:
.i 'PA-'3:-i wiring nthMO'B. How¢eVer, the respdnse* d oes not state how th M(CB' 8l/13/2015 consistent wvt the refrerenced RAIo.
analysis will be Updated., Describe (or reference a descrilption Of) the response i MOB analysis and clarify how the revised, treatment of qua lification is i*
"i'.;,:
- ii..:
i:!
consistent witli or bounlds the actual MOB wiring configuration..
- ,...o 37
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Credit for MCR Abandonment Actions - Describe how MCR abandonment was modeled for loss of habitability in both the post-transition and the compliant plant. Include identification of the actions PRA method is changed PRA-14A required to execute safe alternate shutdown and how they are modeled 4/13/2015 consistent with the referenced RAI in the FPRA, including actions that must be performed before leaving response the MCR. Also, include an explanation of how the CCDPs and CLERPs are estimated for fires that lead to MCR abandonment.
'* ~Explain how the COOps and CLERPs estimated for fires that lead to
,,ii abandonment due to loss of habitabilityaddress varioUs possible fii~e-
.. PRA method is changedJ PRA14Bi) induced failures. Specifically, provide a discussion of how the following.4/13/2015 consistent wi~th the referenced RAI PRA-4B~)
'scenarios are addressed: Scenarios where fire fails only a few fu~nctions
- response *:
i' asidle fromf forcing MOR abandonment and successful alternate' shutdown is straightforward.
Explain how the CCDPs and CLERPs estimated for fires that lead to abandonment due to loss of habitability address various possible fire-PRA method is changed PRA-14B(ii) induced failures. Specifically, provide a discussion of how the following 4/13/2015 consistent with the referenced RAI scenarios are addressed: Scenarios where fire could cause some recoverable functional failures or spurious operations that complicate response the shutdown, but successful alternate shutdown is likely
~Explain how the CCDPs and CLERPs estimated for fires that lead to abandonm~ernt du~e to loss of habitability address 'various possible fire-
- PRA mfethod is cihanged PRA-14B(iii)Y induiced failures. Specifically, :provide adiscussion of how the following 4/13/20:15 consistent withi thereferenced RAi scenarios are addressed: Scenarios where the fire-induced failures rspns
, ° :,
cause great difficuitY for shutdown by failing multiple functions and/or repns complex spurious operations that make successful shiutdown Unlikely Explanation of the timing considerations (i.e., total time available, time PRA method is changed PRA-14C until cues are reached, manipulation time, and time for decision-41/05 cnitn ihterfrne A
PRA-14C making) made to characterize scenarios in Part (b). Include in the 41/05 cnitn ihterfrne A
explanation the basis for any assumptions made about timing.
response PRA method is changed P -1D Discussion of how the probabilitY associated with failure to transfer 4/13/2015 consistent with the referenced* RAI PA1D control to the Auxiliary' Shutdown Panel is taken into account in Part (b).
response 38
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe sseRsubmitta submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
MCR Abandonment on Loss of Control - Clarify whether the above fire PRA method is changed PRA-15A areas (or other non-MCR areas) contain fire scenarios for which primary 4/13/2015 consistent with the referenced RAI command and control is not retained in the MCR (i.e., the MCR isrepne-aseePA1.0 abandoned), and if so, explain how this decision was reached.
rsos lo e R-A0 C*fr htteefcsof :inivdl fires~
i ri tie CSR areevalua:ed and S*8hut'dow *Panel, for both the post-tr~ansitio~n and complia'ntpln If primary command and control is retained in the MCR, then RG 1.205 states, "Operation of dedicated or alternative shutdown controls while the MCR remains the command and control location would normally be considered a recovery action." If actions taken at the PCS are not PRA method is changed PRA-15SB considered RAs for scenarios in which primary command and control 4/13/2015 consistent with the referenced RAI are retained in the MCR, assess the impact of treating such actions rsos losePA1A0 consistent with RG 1.205 on both the delta risk and additional risk ofrepne-assePA1 A0 RAs as part of the integrated analysis performed in response to PRA RAI 03. Additionally, discuss the results of the feasibility and reliability evaluation of any new RAs in accordance with FAQ 07-0030.
For:
scnai:
in w*:'*:hich primary comman andcontrl is nhot: retaine i
the MC a
d i
taf tranerrd tohe PSthe*r action i*taki*:*
at the 39
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR At., W update NRC RAIaccompanying this RAI RAI Issue
Response
submittal, and in the PRA that Number Submittal will be used at the beginning of Date the self-approval of post-transition changes)
State of Knowledge Correlation - Section 4.7.3 of the LAR explains that the sources of uncertainty in the FPRA were identified, and specific parameters were analyzed, for sensitivity in support of the NFPA 805 FRE process. It is further explained that during the FRE process, the uncertainty and sensitivity associated with specific FPRA parameters were considerations in the evaluation of the change in risk relative to PRA method is changed the applicable acceptance thresholds. Based on these explanations, it consistent with the referenced RAI PRA-1 6 appears that the risk results presented in Attachment W of the LAR are 2/9/2015 response - a revised uncertainty point estimates and do not include parameter uncertainty. Explain how analysis is perfomred in support of the SOKC was taken into account in the FPRA quantification, including the PRA-03 integrated response fire ignition frequencies, circuit failure likelihood and hot short duration, and non-suppression probabilities. If the SOKC for these parameters was not addressed in the FPRA quantification, then include the impact of the SOKC for these parameters in the integrated analysis performed in response to PRA RAI 03.
- sensitivity for 6850 Frequencies Versus Supplemnent 1, Frequencies -
"The April 13, 2015, response PRA The licensee's analysis appears to indicate that generic fire,ignition,
RAI 17 stated that that the frequencies were 'based upon those provided in Sul llemernt 1 :to sesiivt sud*dsriedi NUREG/CR-6850. Chapter,10 of this supplement, however, states that.,
NUREG/CR-6850 will be a sensitivity analysis should be performed when Using the fire ignition promdi ojnto ihte S
frequencies in the supplement instead of those provided in Table 6-1 of perf ormpedtion'oh cnuantiiction withth NUREG/CR-6850: As part of the response to PRARAI 03, provide the com/01 spletont of the quapntifiction inRA PRA-17 results (i.e., CDF, LERF, ACDF and ALERF) of a sensitivity analysis 4/321 suprofteepnetoPA that evaluates the impact of using the supplement frequencieS, 0.Terslso h estvt consistent with Chapter 10 of Supplement 1ito NUREG/CR-6850. If RG studY requested in PRA-17 but not
- 1.174 risk acceptance guidelines are exceeded, (1) discuss which ones provided in the April 13, 2015; are exceeded, (2) describe the fire protection or related measures that
,aumne responsetsp ovided As 17an will be taken to provide additional DID, and (3) discuss conservatisms inaumnerspsetPA-7o the analysis and the risk significance of these conservatisms.
as pato
- h epnet R-3 Calculation of VFDR ACOF and ALERF - Provide a detailed definition of both the post-transition and compliant plant models used to calculate the reported change-in-risk, including any special calculations for the MCR and other abandonment areas (if applicable). Include description PRA-1 8A of the model adjustments made to remove VFDRs from the compliant 2/9/2015 No change to PRA methods plant model, such as adding events or logic, or use of surrogate events.
Also, provide an explanation of how VFDR-and non-VFDR-related modifications are addressed for both the post-transition and compliant plant models.
40
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI RAI ssueResp nse subm ittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Provide justification for the assumption in the licensee's analy~sis PRA method is changed PRA-18B (Section 8.0 of EPM Report R2215-008-024) that the risk associated 4/113/2015 consistent with the referenced RAI
~~~with the post-transition plant model is considered equivalent to that ofrepneadttOfPARI1 the Compliant plant model for scenarios requiring MCR abandonment.repneadttofPAAI1 Provide a description of how the reported additional risk of RAs was calculated, including any special calculations performed for the MCR and other abandonment areas (if applicable). If non-VFDR-related modifications are credited to reduce delta risk, equating the additional PRA-18C risk of RAs (as discussed in W.2.1) to the sum of the delta risks of the 2/9/2015 No change to PRA methods VFDRs that are resolved by crediting an RA may be non-conservative.
In this case, the additional risk of these RAs should be re-calculated consistent with FAQ 07-0030 as part of the integrated analysis performed in response to PRA RAI 03.
Provide a summary of the types of VFDRs that were identified but not PRA-1 8D modeled in the FPRA. include any qualitative rationale for excluding 2/9/2015 No change to PRA methods these from the change-in-risk calculations.
Provide a clarification of whether the DID RAs listed in Attachment G of PRA-1 8E the LAR are quantified in the FPRA. Also, explain whether credit for 2/9/2015 No change to PRA methods such DiD RAs is necessary for the change-in-risk to be acceptable.
41
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Art. W update RAI ssueRespnse accompanying this RAI Number Submittal submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Attachment W Inconsistencies - In Table W-6, Unit 1 Fire Areas 2, 8, 13, 18, 18A, 22, 23, 25, 26, 27, 28, 31, 38, 40, and 2CNMT are indicated as Determiinistically Compliant (4.2.3.2); however, they are indicated as having VFDRs (i.e., there is a "Yes"' under the "VFDR" column and sometimes under the "RAs" column) as well as very small risk values (i.e., Fire Area 18) or epsilon for ACDF/ALERF. Similarly, in Table W-7, Unit 2 Fire Areas 3, 4, 6, 14, 15, 19, 19A, 21, 30, 33, 39, and 1CNMT are noted as Deterministically Compliant (4.2.3.2);
response PRA-19A(i) however, they are indicated as having VFDRs and very small risk wloiththsdA No change to PRA methods values (i.e., Fire Areas 19 and 30) or epsilon for ACDF/ALERF.
subiththiAl Attachment C does not identify any of the above deterministic fire areas as having VFDRs. Furthermore, while for most of these fire areas the ACDF/ALERF and additional risk of RAs is reported to be epsilon, actual (very small) numerical values are reported for ACDF!ALERF for Unit 1 Fire Area 18 and for Unit 2 Fire Areas 19 and 30, and actual (very small) numerical values are reported for additional risk of RAs for Unit 1 Fire Area 23.
In Table W-6, Unit 1 Fire Areas 12, 14, 15, 19A, 21,30, 32, 33, 35, 36, 39, 1 CNMT, and IS are indicated as Performance-Based (4.2.4.2) and as having an RA credited in the FPRA (i.e., there is a "Yes"' under the "RAs" column); however, no RAs are described in the VFDR dispositions presented in Attachment C or listed in Attachment G for these areas. Similarly, Unit 2 Fire Areas 12, 13, 18A, 20, 26, 27, 28, 32, response PRA-19A(ii) 34, 35, 36, 40, 2CNMT, and IS are indicated as Performance-Based proided NochngstPARetod (4.2.4.2) and identify a "Yes"' under RA; however, no RAs were described in the VFDR dispositions presented in Attachment C or listed sumta in Attachment G for these areas. Furthermore, while for most of these fire areas the additional risk of RAs is reported to be epsilon, actual (very small) numerical values are reported for Unit 1 Fire Areas 21 and 36 and for Unit 2 Fire Area 13.
The LERF (9.49E-08/year (yr) reported in Table W-4 for scenario PAU CC-1A-C (Complete Burn of Vertical Cable Chase 1A) is greater than the total LERF (4.04E-08/yr) reported in Table W-6 for Fire Area 20 (Cable Chase 1A). For two scenarios reported in Table W-4 (PAU response provided N
hnet R
ehd PRA-1 9A(iii) 230E-C and PAU 230W-C), which represent fires in Unit 1 with this RAI N
hnet R
ehd Containment, the summation of their LERF (1.99E-07 /yr) is greater submittal than the total LERF (1.91 E-07/yr) reported in Table W-6 for Fire Area 1 CNMT (Unit 1 Containment). These inconsistencies also exist between Tables W-5 and W-7 for the same scenarios in Unit 2.
42
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
The Table W-1 Unit 1 fire LERF of 3.2E-06/(chemical reactor (rx)-yr) response PRA-1A~iv) does not match the corresponding value reported in Table W-6.
provided N
hnet R
ehd PRA-1 A~iv)
Similarly, the Table W-1 Unit 2 fire LERF of 4.4E-06/(rx-yr) does not with this RAI N
hnet R
ehd match the corresponding value reported in Table W-7.
submittal Describe what is meant by the use of "E," or epsilen, in Columns for Fire
.... ~Area CDF/LERF, ACDF/ALERF, and additional risk of RAs. Address if.
PR-B epsilon is defined by a specific cut-off value(s). Also, clarify how an.
2/9/2015 No change to PRA methods PRA-19B actual value for LERF can be reported while epsilon is reported for the
.. "ii
" ' corresponding ODE (i.e., Unit 1 Fire Area 24 for additional risk of RAs, Unit 2 Fire *Areas 8 and.10 for CDF/LERF and ACDF/LERF).
Describe what is meant by the use of "N/A" in columns for Fire Area CDE/LERF, ACDF/ALERF, and additional risk of RAs. In doing so, PRA-1 9C clarify the basis for not reporting Fire Area CDE/LERF values (or 2/9/2015 No change to PRA methods epsilon) for Unit 1 and Unit 2 Fire Areas 44, AB-1, AB-3, ABEL, DGB1, DGB2, and TB FL.
- Tables W-6 and W-7 include a risk reduction credit for internal events rspn that. is described in a footnote to these tables as covering random res.
pronsded NchnetPRmtos PRA-1 9D failures and internal floods. This risk reduction credit is Used to offset prvdd NchnetPRmtos
- with this RAI the increase i n.fire risk reported in these tables. Explain how the risk reduction from iinternal events reported in these tables is calculate'd.
submittal Implementation Item Impact on Risk Estimates - Table S-3, Implementation Item 12 of the LAR commits to updating the FPRA and verifying the risk results after "risk related" plant modifications have PRA-20 been incorporated. However, it is unclear to which modifications the 2/9/2015 No change to PRA methods implementation item refers. Update Implementation Item 12 to reflect completion of both the Table S-2 modifications and Table S-3 implementation items before this verification.
43
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR Att. W update NCRAI IseRsos accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that Numbe Submttal will be used at the beginning of Date the self-approval of post-transition changes)
Internal Events Peer Review - Attachment U of the LAR indicates that the full-scope IEPRA peer review was performed against ASME/ANS PRA Standard, RA-S-2008a. In light of this observation, if RG 1.200, PRA-21 Revision 2, and ASME/ANS PRA Standard, RA-Sa-2009, were not 2/9/2015 No change to PRA methods used as the basis for the peer review of the IEPRA, then discuss whether any differences between SRs were evaluated and whether they had any impact on the application.
PRA Upgrades - The LAR does not indicate whether any changes made to the IEPRA or FPRA since their most recent full-scope peer reviews are consistent with the definition of a "PRA upgrade" in ASMEIANS-RA-Sa-2009, "Standard for Leveil/Large Early Release Frequency for Nuclear Power Plant Applications,' as endorsed by RG PR-2 1.200, Revision 2. In light of this, identify any such changes. If a 2/9/2015 No change to PRA methods PRA-22 focused-scope peer review has not been performed for the identified changes, describe what actions will be implemented to address this issue. If a focused-scope peer review has been performed, confirm whether it was done consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by RG 1.200, and provide any findings and their resolutions.
Deviations from Acceptable Methods - Section 4.5.1.2 of the LAR States that the FPRA model uses "a methodology consistent with the guidance provided in NUREG/CR-6850 and subsequent clarifications documented in responses to NFPA 805 FAQs" and that "[n]o unreviewed methods or deviations from NUREG/CR-6850 were utilized in the FPRA model development." Indicate if any other methods were No change to PRA methods is PRA-23 employed that deviate from other NRC-accepted guidance (e.g.,
2/9/2015 expected. Issue is addressed subsequent clarifications documented in FAQs, interim guidance consistent with RAI response.
documents, etc.). If so, describe and justify any proposed method that deviates from NRC guidance, or replace the proposed method with an accepted method. Also, include the proposed method as a method "currently under review" as part of the integrated analysis in the response to P.RA RAI 03.
Defense-in-Depth and Safety Margin - Provide further explanation of the PA2A method(s) or criteria used to determine when a substantial imbalance 2/9/2015 No change to PRA methods PRA-24A between DID echelons existed in the FREs, and identify the types of plant improvements made in response to this assessment.
44
ATTACHMENT (1)
REQUEST FOR ADDITIONAL IN FORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST Impact Assessment ( on final CCNPP aggregate analysis, results NRC RAI provided in LAR At., W update RAI ssueRespnse accompanying this RAI Numbe IseRsubmitta submittal, and in the PRA that will be used at the beginning of Date the self-approval of post-transition changes)
Provide further discussion of the approach in applying the NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire PRA-24B Protec*tion Program Under 10 CFR 50.48(c)," Revision 2 (ADAMS 2/9/2015 No change to PRA methods Accession No. ML081130188) criteria for assessing safety margin in the FREs.
Site Procedures for Storage of RAM - The radioactive material (RAM) described in the CENG [Constellation Energy Nuclear Group]
Calculation No. CA07953 provides a quantification of-the maximum RAD-01 amount of RAM that may be stored in various areas. Provide 31/05 N
hnet R
ehd information, if any, on site procedures that are (or will be) established to 31/05 N
hnet R
ehd limit the amount of RAM in storage containers to the levels identified in the analyses (e.g., West Road Cage area, Warehouse #3, Pre-Assembly Facility, and Upper Laydown Area).
Controls for Opening RAM Containers - Provide information,' if any, on site procedures that establish operational controls to restrict the, RAD-02 opening of storage containers in open, uncontained areas (e.g., West 3/11/2015 No change to PRA methods Road Cage area, Warehouse #3, Pre-Assembly Facility, and Upper Laydown Area).
Upper Laydown Area "Sealed" Containers - In the Upper Laydown Area, there are "sealed" Sealand containers, casks, and other containers. Describe what is meant by "sealed" (e.g., are the containers RAD-03 locked and access is not allowed, and do site procedures prevent the 3/11/2015 No change to PRA methods opening of these ccntainers?). Also, describe how potential effluent will be contained based on the "sealing' of containers and concluding that there will be negligible RAD.
Compensatory Measures for FFSM - Describe any compensatory actions that may be taken during fire suppression activities to minimize RAD-04 RAD (e.g., diking of liquid effluent, use of storm drain covers, 3/11/2015 No change to PRA methods radioactive monitoring, or use of other gaseous effluent controls (e.g.,
use of eductors, effluent filtration)).
3c. See Enclosure 1 for updated LAR Attachment W demonstrating that all RG 1.205 risk acceptance guidelines are met. None of the fire areas result in increases that exceed the RG 45
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST 1.205 guidelines. Modifications and recovery actions in support of the Attachment W results are specified in Attachments S and G.
3d. No unacceptable methods were used to support the CCNPP NFPA 805 LAR submittal or the post transition, self-approval model.
PRA RAI 19-Attachment W Inconsistencies:
Several inconsistencies were noted within Attachment W as well as between its tables and those in Attachments C and G for particular fire areas. In light of this:
a) Provide clarification on the following inconsistencies, and discuss their significance to the risk results reported in Tables W-6 and W-7:
i.In Table W-6, Unit 1 Fire Areas 2, 8, 13, 18, 18A, 22, 23, 25, 26, 27, 28, 31, 38, 40, and 2CNMT are indicated as Deterministically Compliant (4.2.3.2); however; they are indicated as having VFDRs (i.e., there is a "Yes"' under the "VFDR" column and sometimes under the "RAs" column) as well as very small risk values (i.e., Fire Area
- 18) or epsilon for ACDF/ALERF. Similarly, in Table W-7, Unit 2 Fire Areas 3, 4, 6, 14, 15, 19, 19A, 21, 30, 33, 39, and 1CNMT are noted as Deterministically Compliant (4.2.3.2); however, they are indicated as having VFDRs and very small risk values (i.e., Fire Areas 19 and 30) or epsilon for ACDF/ZILERF. Attachment C does not identify any of the above deterministic fire areas as having VFDRs. Furthermore, while for most of these fire areas the ACDF/ALERF and additional risk of RAs is reported to be epsilon, actual (very small) numerical values are reported for ACDF/ALERF for Unit 1 Fire Area 18 and for Unit 2 Fire Areas 19 and 30, and actual (very small) numerical values are reported for additional risk of RAs for Unit 1 Fire Area 23.
ii.
In Table W-6, Unit 1 Fire Areas 12, 14, 15, 19A, 21, 30, 32, 33, 35, 36, 39, 1CNMT, and IS are indicated as Performance-Based (4.2.4.2) and as having an RA credited in the FPRA (i.e., there is a "Yes"' under the "RAs" column); however, no RAs are described in the VFDR dispositions presented in Attachment C or listed in Attachment G for these areas. Similarly, Unit 2 Fire Areas 12, 13, 18A, 20, 26, 27, 28, 32, 34, 35, 36, 40, 2CNMT, and IS are indicated as Performance-Based (4.2.4.2) and identify a "Yes"' under RA; however, no RAs were described in the VFDR dispositions presented in Attachment C or listed in Attachment G for these areas. Furthermore, while for most of these fire areas the additional risk of RAs is reported to be epsilon, actual (very small) numerical values are reported for Unit 1 Fire Areas 21 and 36 and for Unit 2 Fire Area 13.
iii.
The LERF (9.49E-O8/year (yr) reported in Table W-4 for scenario PAU CC-lA-C (Complete Burn of Vertical Cable Chase 1A) is greater than the total LERE (4.04E-O8/yr) reported in Table W-6 for Fire Area 20 (Cable Chase 1A).
For two scenarios reported in Table W-4 (PAU 230E-C and PAU 230W-C), which represent fires in Unit 1 Containment, the summation of their LERF (1.99E-O7/yr) is greater than the total LERF (1.91 E-O7/yr) reported in Table W-6 for Fire Area 1CNMT (Unit 1 Containment). These inconsistencies also exist between Tables W-5 and W-7 for the same scenarios in Unit 2.
46
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST iv.
The Table W-1 Unit 1 fire LERF of 3.2E-O6/(chemical reactor (rx)-yr) does not match the corresponding value reported in Table W-6. Similarly, the Table W-1 Unit 2 fire LERE of 4.4E-O6/(rx-yr) does not match the corresponding value reported in Table W-7.
b) Describe what is meant by the use of "e," or epsilon, in columns for Fire Area CDF/LERF, LICDF/ALERF, and additional risk of RAs. Address if epsilon is defined by a specific cut-off value(s). Also, clarify how an actual value for LERE can be reported while epsilon is reported for the corresponding CDF (i.e., Unit 1 Fire Area 24 for additional risk of RAs, Unit 2 Fire Areas 8 and 10 for CDF/LERF and A CDF/LERF).
c) Describe what is meant by the use of "N/A" in columns for Fire Area CDF/LERF, ACDF/ALERF, and additional risk of RAs. In doing so, clarify the basis for not reporting Fire Area CDF/LERF values (or epsilon) for Unit 1 and Unit 2 Fire Areas 44, AB-1, AB-3, ABEL, DGB1, DGB2, and TBFL.
d) Tables W-6 and W-7 include a risk reduction credit for internal events that is described in a footnote to these tables as covering random failures and internal floods.
This risk reduction credit is used to offset the increase in fire risk reported in these tables. Explain how the risk reduction from internal events reported in these tables is calculated.
19a -
- i.
The areas identified for Table W-6 are deterministically compliant for Unit 1 and there are no VFDRs associated with Unit 1.
Table W-6 has been revised to specify a compliance strategy of 4.2.3.2 with no VFDRs for these areas. The areas identified for Table W-7 are deterministically compliant for Unit 2 and there are no VFDRs associated with Unit 2. These areas comply with 4.2.4.2 for Unit 1. Table W-7 has been revised to specify a compliance strategy of 4.2.3.2 with no VFDRs for these areas. The updated Attachment W-6 and W-7 tables enclosed replace those provided with RAI SSA-06C.
ii.
The areas do not have credited recovery actions for the unit in question. Table W-6 and W-7 have been revised accordingly.
The updated Attachment W-6 and W-7 tables enclosed replace those provided with RAI SSA-06C.
iii.
The updated LAR Attachment W tables, (Enclosure 1) resolve this issue.
iv.
The updated LAR Attachment W tables, (Enclosure 1) resolve this issue.
1 9b - Response provided in Reference 1.
1 9c - Response provided in Reference 1.
19d - The same plant improvements used in the fire model (i.e. modifications specified in NFPA 805 LAR, Attachment S) are applied to the internal events model and the associated reduction in non-fire risk is specified as a risk offset for the Fire PRA risk.
References 47
ATTACHMENT (1)
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 LICENSE AMENDMENT REQUEST
- 1. Letter from G. H. Gelirich (Exelon) to Document Control Desk (NRC), dated February 9, 2015, Request for Additional Information Regarding the National Fire Protection Association Standard 805 License Amendment Request 48
ENCLOSURE 1 Section 4.7 Pages and Attachments C, G, S, and W Calvert Cliffs Nuclear Power Plant February 24, 2016
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 4.7 Program Documentation, Configuration Control, and Quality Assurance CENG Configuration Management (CM) Procedures listed in section 4.7, have or will be transitioned to Exelon procedures with similar elements.I 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805 In accordance with the requirements and guidance in NFPA 805 Section 2.7.1 and NEI 04-02, CCNPP has documented analyses to support compliance with 10 CFR 50.48(c).
The analyses are being performed in accordance with CENG's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analysis.
Analyses, as defined by NFPA 805 Section 2.4, performed to demonstrate compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy. Note that these analyses do not include items such as periodic tests, hot work permits, fire impairments, etc.
The fire protection design basis information described in Section 2.7.1.2 of NEPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 will be created as part of transition to 10 CFR 50.48 (c) to ensure program implementation following receipt of the safety evaluation.
See Attachment 5, Table S-3, IMP-7.
Appropriate cross references will be established to supporting documents as required by CCNPP processes. Figure 4-9 depicts the planned post-transition documentation and relationships.
The post-transition NFPA 805 FPP design and licensing basis for specific fire areas will be documented in design controlled basis documents and databases. For generic design basis, a design basis reconciliation Engineering Change Package will be developed to transition CCNPP to an NFPA 805 FPP licensing basis. The Engineering Change Package will consist of multiple analyses that will comprise the basis for the post-transition NFPA 805 FPP. The specific documents and databases include, but are not limited to:
- Transition of fundamental FPP and design elements review results (B-i Table)
- NSCA review results (B-2 Table)
- Nuclear safety transition fire area assessment review results (B-3 Table)
- NPO Modes transition review results
- Cable routing database Radioactive release review results
- Code compliance review results
- Combustible loading calculation
Constellation Eneraly Nuclear Group 4.0 Compliance with NFPA 805 Requirements V*
m W
m NFPA 805 DOCUMENTS I
Equipmetvi Technical Databsase I
PR Equipment I o-oe I
Data I l EqulpmentoDatal I
NSCA ANALYSIS I Comp&
Cale IFAAssesamenit MI
[
- --u I so*,~
L------o~l SRevised License Condition SRevised UFSAR NSCA SUPPORTING INFO I
-=
I A o n=, I ICalculations I 1Descrlptonshat MHIF Taient SuportNSCJ Non-Power Mode NSCA Treatment N on-Powr Operations Analyasas r--------------------------------
Design Basis Informat~on (091)
- On a Fire Area Basis Fire Ag'es D escripti*on Fire Hazards Summary Infomiation Nuclear Safety Performance Criteria Compliance Summary (NE!104-0284-Table Reaults)
Combustion Loading Calculation Non-Power Evaluations Resulta Radioactive Release Summary
- On a Generic Basis
- B-I Table Results Fire Code of Record Report Radioactive Release Report Monitoring Program Description NFPA S0S FIRE RISK EVALUATIONS I,
Fire Risk Evaluation(s)
FHA DATA Loading FetrsDa Cal[uation L-----------e FIRE PRA FHA SUPPORT DOCUMENTATION ICode Complance FP Drawings IDescriptions Equivalency I
[ Pre-Fr Plans Release Review I
- t..
.A..
ly...i I
Bold text indicates new NFPA 805 documents "includes the entire population of equipment I cables (NSCA, NPO, PRA)
Figure 4 NFPA 805 Planned Post-Transition Documents and Relationships CCNPP Page 49 CCNPP Page 49
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805 Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48 (c) is subject to CCNPP configuration control processes that meet the requirements of Section 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes impacting the FPP are reviewed appropriately.
The RI-PB post transition change process methodology is based upon the requirement of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-2.
Table 4-2 Change Evaluation Guidance Summary Table Document Section(s)
Topic NFPA 805 2.2(h), 2.2.9, 2.4.4, A.2.2(h),
Change Evaluation A.2.4.4, D.5 NEI 04-02 5.3, Appendix B, Appendix I, Change Evaluation, Change Appendix J Evaluation Forms (Appendix I)
RG 1.205 0.2.2.4, C.3.1, C.3.2, 0.4.3 Risk Evaluation, Standard License Condition, Change Evaluation Process, Fire PRA The plant change evaluation process consists of the following 4 steps and is depicted in Figure 4-10:
- Defining the change
- Performing the preliminary risk screening.
- Performing the risk evaluation
- Evaluating the acceptance criteria Change Definition The change evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the design basis and licensing basis (NFEPA 805 licensing basis post-transition).
- 1. The baseline is defined as that plant condition or configuration that is consistent with the design basis and licensing basis (NFPA 805 licensing basis post-transition).
- 2. The changed or altered condition or configuration that is not consistent with the design basis and licensing basis is defined as the proposed aftemative.
Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the FPP. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-03 process. This CCNPP Page 50 CCNPP Page 50
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).
The characteristics of an acceptable screening process that meets the TUassessment of the acceptability of risks requirement of Section 2.4.4 of NFPA 805 are:
- The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk.
,* The screening process must be documented and be available for inspection by the NRC.
- The screening process does not pose undue evaluation or maintenance burden.
If any of the above is not met, proceed to the risk evaluation step.
Risk Evaluation The screening is followed by engineering evaluations that may include fire modeling and risk assessment techniques. The results of these evaluations are then compared to the acceptance criteria. Changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition. The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature.
The risk evaluation involves the application of fire modeling analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below.
Acceptability Determination The change evaluations are assessed for acceptability using the ACDF (change in core damage frequency) and ALERF (change in large early release frequency) criteria from the license condition. The proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Defining the Change (5.3.2) y-~o Preliminary Risk Screening (5.3.3)
Risk Evaluation (5.3.4)
PRA Capability Category Assessment Acceptance Criteria ILrim rr¶ Drie t---I-i(5.3.5)
Figure 4-10 Plant Change Evaluation [NEI 04-02 Figure 5-1]
Note references in Figure refer to NEI 04-02 Sections CCNPP Page 52 CCNPP Page 52
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements The CCNPP FPP configuration is defined by the program documentation. To the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses, and FPP license basis reviews will be utilized to maintain configuration control of the FPP documents. The configuration control procedures which govern the various CCNPP documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.
Several NFPA 805 document types such as: NSCA supporting information, NPO Mode treatment, etc., generally require existing processes to be revised or new processes and documentation to be developed since they are new documents and databases created as a result of the transition to NFPA 805. Design and PRA documents will be revised to reflect the new NFPA 805 requirements.
The process for capturing the impact of proposed changes to the plant on the FPP will continue to be a multiple step review. The first step of the review is an initial screening for process users to determine if there is a potential to impact the FPP as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used. Reviews that identify potential FPP impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, PRA) to ascertain the program impacts, If any. If FPP impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:
Deterministic Approach: Comply with NFPA 805 Chapter 3 and Section 4.2.3 requirements.
- Performance-Based Approach: Utilize the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process would be used to determine if the proposed change could be implemented "as-is" or whether prior NRC approval of the proposed change is required.
The process follows the requirements in NFPA 805 and the guidance outlined in RG 1.174, which require the use of qualified individuals, procedures that require independent review and verification of calculations, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when discrepancies are discovered.
,Speeif-ieally 1-CCNPP evaluates potential impact on the FPP during transition to NFPA 805 through the implementation of a configuration management process such as CNG-CM-1.01-1003 (Reference 6.37). This document controls design changes to the facility. It contains evaluation criteria that must be reviewed to determine if the design change can have potential FPP impact.
Changes affecting the design of the plant are performed in accordance with procedures such as -.CNG-CM-1.01-1003.
The ie-procedures h:c boon ro.-icod to reflect the requirements of NFPA 805,.
CNG-CM-1.01. 1.003 and -ensures that reviews are performed to determine if plant changes impact the FPP documentation.
The evaluation criteria are reviewed during the development of the plant design change. If CCNPP Page 53
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements this evaluation identifies potential impact, then a more detailed review is performed by qualified fire protection, safe shutdown, and PRA personnel who are involved with the ongoing NFPA 805 transition activities.
Engineering standards CN.G-FES-007 (Rofroc 38), whi\\;:.ch comp.,i.,nt,-
CNG C'M*_~_ 1.01 n,
1003, hac
,-a boon--,r,..,ic,, t,.
ad'd a coction for address NFPA 805 applicability criteria and required actions.
The CCNPP NFPA 805 NSCA (Reference 6.39) was generated as part of the transition to NFPA 805. Evaluation of plant changes will be performed during the NFPA 805 implementation period;,.,,and' CNG CM,_r 1a4.......011,
,.an-ndtv.
...,,G" FES.=oT.,..-.
007 change evaluation criteria will be utilized to provide assurance that plant changes will be properly integrated into the FPP documents developed under the NFPA 805 project.
A review of implemented plant changes is also performed by the PRA organization to determine potential model impacts. A review of plant changes will be performed during the NFPA 805 implementation period to ensure that changes are appropriately evaluated for potential impact on the PRA model. Thi i don par CM* 1.01 3003'*tr, In conclusion, the plant processes described above have been in place during the NFPA 805 transition to identify changes that may impact the FPP. Additionally, maintenance of the NFPA 805 analyses will be performed during the NFPA 805 implementation period to reflect the current plant configurations. The update will include review of plant configuration changes along with changes that may have occurred from RAI responses, updates from industry groups for Multiple Spurious Operating configurations, new or revised FAQs (e.g., FAQ 12-0061), and development of modifications. This ensures current plant configurations are appropriately reflected and evaluated in the NFPA 805 documentation prior to and after full implementation of NFPA 805. See Attachment 5, Table S-3, IMP-B.
4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality Current QA Program The existing CCNPP fire protection QA program requirements are contained in the following documents:
- Updated Final Safety Analysis Report (UFSAR) Section 9 and Appendix 1 B.
- Quality Assurance Program Topical Report QA Program Utilized During Transition During the transition to 10 CFR 50.48 (c), CCNPP performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805 and the existing fire protection QA program described above. This included requirements that each analysis, calculation, or evaluation performed to support compliance with 10 CFR 50.48 (c) be independently reviewed.
Post Transition QA Program The QA program for the existing CCNPP FPP will be utilized with the following changes:
Constellatlon Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Coselto nryNcerGop40 opinewt FA85Rqieet
- For the post NFPA 805 transition, the fire protection QA program requirements contained in the UFSAR and Quality Assurance Topical Report will be revised to reflect the new NFPA 805 licensing basis and..C..r*nG....QL 4-1 n.0 1001.* (Ro~fo.r.,n...o...
6A-1-) will be revised to contain the audit requirements.
Consequently, the components and systems currently considered within the scope of the QA program for the CCNPP FPP will be expanded to include those components and systems that are in the power block and are required by Chapter 4 of NFPA 805. This means that certain fire protection systems and features in some buildings not currently in the program but are required by NFPA 805 Chapter 4 will now fall under the QA program. As such, any future modifications to these systems will be conducted under design controls which are audited by the QA program.
The changes to the QA documentation will be completed as needed for implementation of NFPA 805. Documents will be revised, as they pertain to Quality Assurance for the FPP, to reflect the new licensing basis for NFPA 805:
- Updated Final Safety Analysis Report. The change is required to address FAQ 12-0062.
- Quality Assurance Topical Report.
The change is to remove the existing commitment to an NRC guideline (Appendix A to Branch Technical Position APCSB 9.5-1) and replace it with the NFPA 805 licensing basis.
See Attachment 5, Table S-3, IMP-9.
Fire PRA Quality Configuration control of the FPRA model will be maintained by integrating the FPRA model into the existing processes used to ensure configuration control of the internal events PRA model. This process complies with Section 5 of the ASME/ANS RA-Sa-2009 and ensures that CCNPP maintains an as-built, as-operated PRA model of the plant. The process has been peer reviewed. FPRA quality is assured via the same processes being applied to the internal events model.
This process follows the guidance outlined in RG 1.174, which requires the use of qualified individuals, procedures that require independent review and verification of calculations, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when discrepancies are discovered. Although the entire scope of the formal 10 CFR 50 Appendix B program is not applied to the PRA models or processes in general, parts of the program may be applied as a convenient method of complying with the requirements of RG 1.174. CNG-CM-I.01-3003, CNG-CM-1.01 -
3004 (Reference 6.42), and CNG-CM-2.01 (Reference 6.43) or equivalent Exelon processes address the review and verification process applied to the CCNPP FPRA.
With respect to QA program requirements for independent reviews of calculations and evaluations, the existing requirement for FPP documents will remain unchanged.
CCNPP Page 55
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements CCNPP specifically requires that the calculations and evaluations in support of the NFPA 805 LAR, exclusive of the FPRA, be performed within the scope of the QA program which requires independent review as defined by CCNPP procedures. As recommended by NUREG/CR-6850, the sources of uncertainty in the FPRA were identified and specific parameters were analyzed for sensitivity in support of the NFPA 805 fire risk analysis process.
The uncertainty and sensitivity analysis is in the Uncertainty and Sensitivity Notebook (Reference 6.44). in addition, sensitivity to uncertainty associated with specific FPRA parameters was quantitatively addressed in the same notelbook.
While the removal of conservatism inherent in the FPRA is a long-term goal, the FPRA results were deemed sufficient for evaluating the risk associated with this application.
While CCNPP continues to strive toward a more urealistic" estimate of fire risk, use of mean values continues to be the best estimate of fire risk. During the fire risk analysis process, the uncertainty and sensitivity associated with specific FPRA parameters were considerations in the evaluation of the change in risk relative to the applicable acceptance thresholds.
Specific Requirements of NFPA 805 Section 2.7.3 The following discusses how the requirements of NFPA 805 Section 2.7.3 were met during the transition process. Post-transition, CCNPP will perform work in accordance with NFPA 805 Section 2.7.3 requirements.
In accordance with NFPA 805 Section 2.7.3, the Quality Assurance Topical Report addresses the CCNPP FPP.
The following items list specific aspects of NFPA Section 2.7.3 and describe the controls currently in place to assure that NFPA 805 related activities are performed correctly and in conformance with applicable requirements for this LAR and in the future:
NFPA 805 Section 2.7.3.1 - Review Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with CCNPP procedures that require independent review.
Review of PRA documents and,,,,,,,,.;, are o ""m b... procod....
o CMr 1.01 3003. R"o'.i" of Design Engineering activities, analyses, and calculations are governed by configuration management procedures.CN1.01..-..
100v3.
NFPA 805 Section 2.7.3.2 - Verification and Validation Calculation models and numerical methods used in support of compliance with 10 CFR 50.48 (c) were verified and validated as required by Section 2.7.3.2 of NFPA 805.
Configuration management Pfedwprocedures, for example CNG-CM-1.01-2001 (Reference 6.45), requires design verification to be performed on quality related calculations in accordance with procedural requirements. in. p..ro.-.----.du...ro..
C
,N_..,G C
.1'.01 1006rl (Reforonc-6.-,46).
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements NFPA 805 Section 2.7.3.3 - Limitations of Use Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were applied appropriately as required by Section 2.7.3.3 of NEPA 805. Exelon PRA procedure series ER-AA-600 (e.g. ER-AA-600-1012,-1014,-01 15,-
1016,-1021) or procedure CNG-CM-1.01-1006, (Reference 6.46), provides the methodsI and requirements for performing design verification of quality related and augmented quality related documents as they relate to engineering methods and numerical models utilized *for NFPA 805 activities.
NFPA 805 Section 2.7.3.4 - Qualification of Users Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.
For personnel performing fire modeling for FPRA development and evaluation, CCNPP develops and maintains qualification requirements for individuals assigned various tasks. Position specific guides were developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work.
Qualification earls--records provide evidence that Design Engineering and PRA personnel have the appropriate training and technical expertise to perform assigned work, including the use of engineering analyses and numerical models.
Qualification requirements are contained in procedures such as CNG-TR-1.01-1014, (Reference 6.47). CCNPP will maintain qualification requirements for the performance of NFPA 805 related tasks. Position specific qualification records ead identify and I document required training and mentoring to ensure cognizant individuals are appropriately qualified to perform assigned work per the requirements of NFPA 805, Section 2.7.3.4.
Procedures C,*NTDR 1011011 is tho governing pr...ur to m.. in.t..;..in and, enhanc,,.,
the Engineering Support Personnel (ESP) training and qualification program. The procedures have C.NG-TR-1.01-1011 has been revised to list the qualifications for NFPA 805, and changes have been made to the ESP program to include the qualification of cognizant individuals. Training will be provided to ESP. See Attachment 5, Table S-3, IMP-i10.
NFPA 805 Section 2.7.3.5 - Uncertainty Analysis Uncertainty analyses were performed as required by Section 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and FPRA development. Note: 10 CFR 50.48(c)(2)(iv) states that NFPA 805 Section 2.7.3.5 is not required for the deterministic approach because conservatism is included in the deterministic criteria.
Uncertainty evaluations for the various tasks used to develop the FPRA model (specifically those outlined in NUREG/CR-6850) were completed. A summary of the uncertainty analysis is provided in C0-UNC-001 (Reference 6.44). While many of the uncertainty analyses were qualitative, some quantitative uncertainty results are CCNPP Page 57
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements presented. The EPRA work performed as part of the transition or needed to maintain the program (i.e., FPRA model updates) will include uncertainty evaluations. CCNPP will follow the necessary guidance (NUREG/CR-6850, FAQs, PRA Standard) in the development of FPRA related uncertainty analyses.
Based on the above provisions, future NFPA 805 analyses associated with CCNPP will be conducted in accordance with the requirements of NFPA 805, Section 2.7.3.
4.8 Summary of Results 4.8.1 Results of the Fire Area Review A summary of the NFPA 805 compliance basis and the required fire protection systems and features is provided in Attachment C, Table 0-2. The table provides the following information from the NEI 04-02 Table B-3:
- Fire Area / Room: Fire area/room Identifier.
Description:
Fire area/room description.
- Required Fire Protection System / Feature: Detection / suppression required in the fire area based on NFPA 805 Chapter 4 compliance. Other required features may include electrical raceway fire barrier systems, fire barriers, etc.
The documentation of required fire protection systems and features does not include the documentation of the fire area boundaries. Fire area boundaries are required and documentation of the fire area boundaries has been performed as part of reviews of engineering evaluations, licensing actions, or as part of the reviews of the NEI 04-02 Table B-i process. The basis for the requirement of the fire protection system / feature is designated as follows:
o S - Separation Criteria: Systems/features required for Chapter 4 separation criteria in Section 4.2.3 o
L - Licensing Action Criteria: Systems/features required for acceptability of NRC approved licensing action (i.e.,
exemptions/deviations/safety evaluations) (Section 2.2.7) o E - EEEE: Systems/features required for acceptability of EEEEs (Section 2.2.7) o R -
Risk Criteria: Systems/features required to meet the risk criteria for the performance-based approach (Section 4.2.4) o D -
Defense-in-depth Criteria:
Systems/features required to maintain adequate balance of defense-in-depth for a performance-based approach (Section 4.2.4)
Attachment W contains the results of the FREs, additional risk of recovery actions, and the change in risk on a fire area basis.
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase Planned modifications, studies, and evaluations to comply with NEPA 805 are described in Attachment S.
The EPRA model represents the as-built, as-operated and maintained plant as it will be configured at the completion of the transition to NFPA 805. The FPRA model includes credit for the planned implementation of modifications identified in Attachment 5, Table S-2. Following installation of modifications and the as-built installation details, additional refinements surrounding the modifications may need to be incorporated into the FPRA model (the FPRA will verify the validity of the reported change-in-risk on as-built conditions after the modifications are completed).
However, these changes are not expected to be significant.
No other significant plant changes are outstanding with respect to their inclusion in the FPRA model. See Attachment 5, Table S-3, IMP-i12.
4.8.3 Supplemental Information - Other Licensee Specific Issues None.
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements CCNPP Page 60 CCNPP Page 60
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition G.
Recovery Actions Transition 20 Pages CCNPP Page G-1 CCNPP Page G-1
Constellation Energy Nuclear Group Attachment G -Recoverv Action Transition CoseltinEegyNcerIru ttcmn ecvry cinTasto In accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions). The methodology consisted of the following steps:
- Step 1: Define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in the main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the main Control Room are not recovery actions, by definition.
Step 2: Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).
- Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path
- Step 4: Evaluate the feasibility of the recovery actions Step 5: Evaluate the reliability of the recovery actions An overview of these steps and the results of their implementation are provided below.
Step I - Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s)
The first task in the process of determining the post-transition population of recovery actions was to apply the NFPA 805 definition of recovery action and the RG 1.205 definition of primary control station to determine those activities that are taken at primary control station(s).
Results of Step 1:
Based on the definition provided in RG 1.205, and the additional guidance provided in FAQ 07-0030 Revision 5, the following are considered to be the primary control stations, with the associated enabling, control, and indication functions as identified:
- 1C43, Alternate Shutdown Panel, Unit I
- 1. 11 Steam Generator Wide Range Level Indication (1LII114A)
- 2. 11 Steam Generator Wide Range Level Indication (1LI1114B)
- 3. 11 Steam Generator Pressure Indication (IPI1013AA)
- 4. 11 Steam Generator Pressure Indication (1PI1013BB)
- 5. 11IA Reactor Coolant Loop T-cold Temperature Indication (1TI11I2CA)
- 6. 11B Reactor Coolant Loop T-cold Temperature Indication (1TIII2CB)
- 7. 11 Reactor Coolant Loop T-hot Temperature Indication (1TI1 12HA)
- 8. 11 Reactor Coolant Loop T-hot Temperature Indication (1TI11I2HB)
- 9. Unit 1 AFW Turbine Driven Pump to Steam Generator 11 Flow Indication (1 FI450gB)
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition CntlainEegNulear Gru tahetG-Rcvr cinTasto 10.Unit I AFW Motor Driven Pump to Steam Generator 11 Flow Indication (1 FI4524A) 11.11 Steam Generator Atmospheric Steam Dump Valve 1 CV3938 Hand Controller (1HC4056A) see note 2 12.11 Steam Generator AFW Turbine Driven Pump Flow Control Valve 1 CV4511I Hand Controller (1 HC4511I B) see note 2 13.11 Steam Generator AFW Motor Driven Pump Flow Control Valve 1 CV4525 Hand Controller (1 HC4525B) see note 2 14.11 Steam Generator Surface and Bottom Blowdown Isolation Valves 1ICV4010 and 1 CV4011 Hand Switch (1HS4010A) 15.12 Steam Generator Wide Range Level Indication (1LI1 124A) 16.12 Steam Generator Wide Range Level Indication (1LI1124B) 17.12 Steam Generator Pressure Indication (1PI 1023AA) 18.12 Steam Generator Pressure Indication (1PI1023BB)
- 19. 12A Reactor Coolant Loop T-cold Temperature Indication (1TI122CA)
- 20. 12B Reactor Coolant Loop T-cold Temperature Indication (1TI122CB) 21.12 Reactor Coolant Loop T-hot Temperature Indication (1TI122HA) 22.12 Reactor Coolant Loop T-hot Temperature Indication (1TI122HB) 23.12 Steam Generator Atmospheric Steam Dump Valve 1 CV3939 Hand Controller (1 HC4056B) see note 2 24.12 Steam Generator AFW Turbine Driven Pump Flow Control Valve 1CV4512 Hand Controller (1HC4512B) see note 2 25.12 Steam Generator AFW Driven AFW Pump Flow Control Valve 1CV4535 Hand Controller (1 HC4535B) see note 2 26.12 Steam Generator Surface and Bottom Blowdown Isolation Valves 1CV4012 and 1CV4013 Hand Switch (1HS4012A) 27.Unit 1 AFW Turbine Driven Pump to Steam Generator 12 Flow Indication (1FI4510B) 28.Unit 1 AFW Motor Driven Pump to Steam Generator 12 Flow Indication (1 FI4534A) 29.11 AFW Turbine Driven Pump Speed Hand Controller (1 HC3987B) see note 2
30.12 AFW Turbine Driven Pump Speed Hand Controller (1 HC3989B) see note 2
- 31. Condensate Storage Tank 12 Level Indication (1LI5610A)
- 32. Condensate Storage Tank 12 Level Indication (1 LI5611IA)
Constellation Enemy Nuclear Group Attachment G - Recovery Action Transition Constelatio Enra ucea rot Atacmet I-Reovry cinTasto 33.11 Reactor Coolant Pressurizer Level Indication (1LI110X) 34.11 Reactor Coolant Pressurizer Level Indication (1LIllOY) 35.11 Reactor Coolant Pressurizer Pressure Indication (1PI105AA) 36.11 Reactor Coolant Pressurizer Pressure Indication (1PI105B)
- 37. Unit 1 Neutron Power, Logarithmic, Wide Range, % Power Indication (1NI016 from either 1NE002 or 1NE004 via Hand Switch 1HS015B)
- 38. Unit I Neutron Power, Logarithmic, Wide Range, Counts Per Second Indication (1 N1015 from either 1NE002 or 1NE004 via Hand Switch 1HS015B)
- 39. Backup Pressurizer Heater Bank 11 (1 UCC2) Transfer / Control Hand Switch (1HS100-4A) see note I
- 40. Backup Pressurizer Heater Bank 13 (1UHI) Transfer/IControl Hand Switch (1HS100-6A) see note 1 41.Unit I Reactor Coolant Pump Controlled Bleedoff Isolation Valve 1CV505 Hand Switch (1HS2505A) 42.Unit I Reactor Coolant Letdown Isolation Valve 1CV516 Hand Switch (1HS2516A)
- 43. unit 1 Reactor Coolant Sampling Isolation Valve 1CV4564 Hand Switch (1 HS5464B) 1(343 Notes:
Note 1: Enabling of each Backup Pressurizer Heater also requires a local recovery action to verify closed I reclose the associated feeder breaker to the heater Motor Control Center (1 MCCI109PH I IMCCIIIPH) at the 480V Unit Bus (1BUSIB01B I 1BUSIB04B) as identified below:
Backup Heater Feed Breaker Feed Breaker(s)
Room Location Bank II (1UCC(2) 1BKR52-1127 317 Bank 13 (1UH1) 1BKR52-1427 430 Note 2: To enable the 1 C43 hand controller requires a local recovery action to reposition the associated hand valve(s) as identified below:
Hand Controller Hand Valve(s)
Hand Valve(s)
Room Location IH C4056A I1HVM S-3938A 430 1HVMS-3938B 430 1 HC4056B 1 HVMS-3939A 430 1HVMS-3939B 430 1IHC4511B 1HVAFW-4511 226 1 HC4512B 1HVAFW-4512 226 CCNPP Page G-4
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition CoseltinEeg Nula Gru tahetG-Rcvr cinTasto IH C4525B 1 HVAFW-4525 226 1 HC4535B 1 HVAFW-4535 226 1H C3987B 1HVMS-3987 603 1HC3989B 1HVMS-3989 603 Note 3: Enabling of Channel B WRNI at 1C43 requires placing 1HS002B1 to off.
This is considered a PCS action that initiates control of instrumentation at the alternate shutdown panel.
The hand switch is located in the same room as 1C43.
2C43, Alternate Shutdown Panel, Unit 2
- 1. 21 Steam Generator Wide Range Level Indication (2LI11 14A)
- 2. 21 Steam Generator Wide Range Level Indication (2L11ll 14B)
- 3. 21 Steam Generator Pressure Indication (2PI1013AA)
- 4. 21 Steam Generator Pressure Indication (2PI1013BB)
- 5. 21A Reactor Coolant Loop T-cold Temperature Indication (2TI1112CA)
- 6. 21 B Reactor Coolant Loop T-cold Temperature Indication (2TI11I2CB)
- 7. 21 Reactor Coolant Loop T-hot Temperature Indication (2TI11I2HA)
- 8. 21 Reactor Coolant Loop T-hot Temperature Indication (2TI11I2HB)
- 9. Unit 2 AFW Turbine Driven Pump to Steam Generator 21 Flow Indication (2FI4509A) 10.Unit 2 AFW Motor Driven Pump to Steam Generator 21 Flow Indication (2FI4524B) 11.21 Steam Generator Atmospheric Steam Dump Valve 2CV3939 Hand Controller (2HC4056A) see note 2 12.21 Steam Generator AFW Turbine Driven Pump Flow Control Valve 2CV4511 Hand Controller (2HC4511IB) see note 2 13.21 Steam Generator AFW Motor Driven Pump Flow Control Valve 2CV4525 Hand Controller (2HC4525B) see note 2 14.21 Steam Generator Surface and Bottom Blowdown Isolation Valves 2CV4010 and 2CV4011 Hand Switch (2HS4010A) 15.22 Steam Generator Wide Range Level Indication (2LI 11 24A) 16.22 Steam Generator Wide Range Level Indication (2LI1124B) 17.22 Steam Generator Pressure Indication (2PI1023AA) 18.22 Steam Generator Pressure Indication (2PI1023BB)
- 19. 22A Reactor Coolant Loop T-cold Temperature Indication (2TI122CA)
- 20. 22B Reactor Coolant Loop T-cold Temperature Indication (2TI122CB)
CCNPP Page G-5
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition Coselto nrgy Nula Gru tahetG-Rcvr cinTasto 21.22 Reactor Coolant Loop T-hot Temperature Indication (2TI122HA) 22.22 Reactor Coolant Loop T-hot Temperature Indication (2TI122HB) 23.22 Steam Generator Atmospheric Steam Dump Valve 2CV3938 Hand Controller (2HC4056B) see note 2 24.22 Steam Generator AFW Turbine Driven Pump Flow Control Valve 2CV4512 Hand Controller (2HC4512B) see note 2 25.22 Steam Generator AFW Driven AFW Pump Flow Control Valve 2CV4535 Hand Controller (2HC4535B) see note 2 26.22 Steam Generator Surface and Bottom Blowdown Isolation Valves 2CV4012 and 2CV4013 Hand Switch (2HS4012A)
- 27. Unit 2 AFW Turbine Driven Pump to Steam Generator 22 Flow Indication (2FI4510QA)
- 28. Unit 2 AFW Motor Driven Pump to Steam Generator 22 Flow Indication (2FI4534B) 29.21 AFW Turbine Driven Pump Speed Hand Controller (2HC3987B) see note 2
30.22 AFW Turbine Driven Pump Speed Hand Controller (2HC3989B) see note 2
31.Condensate Storage Tank 12 Level Indication (2L15610A)
- 32. Condensate Storage Tank 12 Level Indication (2LI5611IA) 33.21 Reactor Coolant Pressurizer Level Indication (2LI110X) 34.21 Reactor Coolant Pressurizer Level Indication (2LI110Y) 35.21 Reactor Coolant Pressurizer Pressure Indication (2PI105AA) 36.21 Reactor Coolant Pressurizer Pressure Indication (2P1105B)
- 37. Unit 2 Neutron Power, Logarithmic, Wide Range, % Power Indication (2NI016 from either 2NE001 or 2NE003 via Hand Switch 2HS015B)
- 38. Unit 2 Neutron Power, Logarithmic, Wide Range, Counts Per Second Indication (2NI015 from either 2NE001 or 2NE003 via Hand Switch 2HS015B)
- 39. Backup Pressurizer Heater Bank 21 (2UCC2) Transfer / Control Hand Switch (2HS1 00-4A) see note 1
- 40. Backup Pressurizer Heater Bank 23 (2UH1) Transfer!/ Control Hand Switch (2HS100-6A) see note 1
- 41. Unit 2 Reactor Coolant Pump Controlled Bleedoff Isolation Valve 2CV505 Hand Switch (2HS2505A)
- 42. Unit 2 Reactor Coolant Letdown Isolation Valve 2CV516 Hand Switch (2HS2516A)
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition Coselto nrgy Nula Gru tahetG-Rcvr cinTasto
- 43. Unit 2 Reactor Coolant Sampling Isolation Valve 2CV4564 Hand Switch (2HS5464B) 2C43 Notes:
Note 1: Enabling of each Backup Pressurizer Heater also requires a local recovery action to verify closed / reclose the associated feeder breaker to the heater Motor Control Center (2MCC209PH / 2MCC211IPH) at the 480V Unit Bus (2BUS2B01B/I2BUS2B04B) as identified below:
Backup Heater Bank 21 (2UCC2)
Bank 23 (2UH1)
Feed Breaker 2BKR52-21 27 2BKR52-2427 Feed Breaker(s)
Room Location 311 407 Note 2: To enable the 2C43 hand controller requires a local recovery action to reposition the associated hand valve(s) as identified below:
Hand Controller 2HC4056A 2HC4056B 2HC4511B 2HC4512B 2HC4525B 2HC4535B 2HC3987B 2HC3989B Hand Valve(s) 2HVMS-3939A 2HVMS-3939B 2HVMS-3938A 2HVMS-3938B 2 HVAFW-4511I 2HVAFW-451 2 2 HVAFW-4525 2HVAFW-4535 2HVMS-3987 2HVMS-3989 Hand Valve(s)
Room Location 407 407 407 407 205 205 205 205 605 605 Note 3: Enabling of Channel A WRNI at 2043 requires placing 2HS001A1 to off.
This is considered a PCS action that initiates control of instrumentation at the alternate shutdown panel.
The hand switch is located in the same room as 2C43.
1C43 (Unit 1) and 2C43 (Unit 2) are the primary control station for implementation of the alternate shutdown strategy in the event of a fire that requires the evacuation of the main Control Room. NRC approval for the design of the Alternate Shutdown Panel(s),
and for the overall alternate shutdown strategy to meet the requirements of 10 CFR 50 Appendix R, Section lII.G.3, was provided in SER Supplement No. 3, dated September 27, 1982, Item 3.2.1 of the SER, Appendix R to 10 CFR Part 50, Items II1.G.3 and III.L.
Enabling of the Alternate Shutdown Panel(s) involves the transfer of control from the main Control Room to 1C43 (Unit 1) and 2C43 (Unit 2) through an operator action to manually position eight hand switches (HS), and eight hand controllers (HC) which are CCNPP Page G-7
Constellation Enerav Nuclear Groua Attachment G - Recoverv Action Transition located on 1043 (2043). Enabling of each Backup Pressurizer Heater also requires a local recovery action to verify closed / reclose the associated feeder breaker to the heater motor control center at the 480V Unit Bus (see Note 1 above for lC43 and 2043).
Enabling of each hand controller also requires a local recovery action to reposition one or more hand valve from the normal instrument air control loop to the alternate shutdown control loop (see Note 2 above for 1C43 and 2043).
Following activation of the Alternate Shutdown Panel, the plant operator is provided with the capability to control and monitor secondary side decay heat removal capability utilizing the AFW system, the capability to control RCS pressure, and the capability to monitor critical RCS process parameters which are necessary to verify that natural circulation has been established in the RCS and that it is being successfully maintained thereafter.
Step 2 - Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk or defense-in-depth criteria)
On a fire area basis all VFDRs were identified in the NEI 04-02 Table B-3 (See Attachment C).
Each VFDR not brought into compliance with the deterministic requirements was evaluated using the performance-based approach of NFPA 805 Section 4.2.4. The performance-based evaluations resulted in the need for recovery actions to meet the risk acceptance criteria or maintain a sufficient level of defense'in-depth.
Results of Step 2:
The final set of recovery actions are provided in Table G Recovery Actions and Activities Occurring at the Primary Control Station(s).
Step 3: Evaluate the Additional Risk of the Use of Recovery Actions NFPA 805 Section 4.2.3.1 does not allow recovery actions when using the deterministic approach to meet the nuclear safety performance criteria. However, the use of recovery actions is allowed by NFPA 805 using a risk informed, performance-based, approach, provided that the additional risk presented by the recovery actions is evaluated in accordance with NFPA 805 Section 4.2.4.
Results of Step 3:
The set of recovery actions that are necessary to demonstrate the availability of a success path for the nuclear safety performance criteria (see Table G-1) were evaluated for additional risk using the process described in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205 and compared against the guidelines of RG 1.174 and RG 1.205. None of the recovery actions were found to have an adverse impact on the Fire PRA. The additional risk of recovery actions is provided in Attachment W.
The CCNPP review of recovery actions for additional risk / adverse risk impact is documented in the following:
Fire PRA Human Reliability Analysis Notebook, C0-HRA-001 (Reference V-l0)
C0-FRE-001 (Reference 6.53)
Step 4: Evaluate the Feasibility of Recovery Actions CCNPP Page G-8
Constellation Energy Nuclear Group Attachment G - Recovery Action Transition CoseltonEegulear Gru tahetG-Rcvr cinTasto Recovery actions were evaluated against the feasibility criteria provided in the NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205.
Note that since actions taken at the primary control station are not recovery actions their feasibility is evaluated in accordance with procedures for validation of off normal procedures.
Results of Step 4:
The HRA and R2215-049-O01 (Reference 6.22) have evaluated the feasibility of recovery actions modeled in the FPRA and used to resolve VFDRs identified in the B-3 Table. This includes recovery actions related to AC power, EDGs, and long-term decay heat removal among others. Feasibility of these recovery actions was evaluated in the HRA and in R2215-049-001 against the criteria outlined in NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205, and made extensive use of HEP quantifications (for HRA recovery actions only).
Recovery actions that are required by the FRE but not addressed in the HRA were evaluated for feasibility using the NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205 criteria and documented in R221 5-049-001.
Results of the feasibility assessments in the HRA and R2215-049-001demonstrate that all credited NFPA 805 recovery actions are feasible.
Implementation items resulting from the feasibility evaluation include...
Operator Modify, as needed, the AOP 9 series procedures (Abn Operating Procedures -
Fire) for the recovery actions evaluated.
e training will be completed in LIC accordance with ONG PR 1.01 1011. Sec implementation item IMP-i5, Attachment S, Table S-3. Thic :,,-In,,ud rc..... ng,- AtOD OB* I *tn AOPD OBD 21 to.Addc
÷to
,A FWA (PR13V 03I030vv
& PC 13 03I010)~.
Seeo IMPl 13 & IP 11l.
Step 5: Evaluate the Reliability of Recovery Actions The evaluation of the reliability of recovery actions depends upon its characterization.
- The reliability of recovery actions that were modeled specifically in the FPRA were addressed using FPRA methods (i.e., HRA).
The reliability of recovery actions not modeled specifically in the FPRA are bounded by the treatment of additional risk associated with the applicable VFDR.
In calculating the additional risk of the VFDR, the compliant case recovers the fire-induced failure(s) as if the variant condition no longer exists. The resulting delta risk between the variant and compliant condition bounds any additional risk for the recovery action even if that recovery action were modeled.
Results of Step 5:
The reliability of recovery actions that were modeled specifically in the FPRA were addressed using FPRA methods. The HRA addresses the reliability of these recovery actions, with consideration taken for various performance shaping factors, including cues and instrumentation, timing, procedures and training, complexity, workload pressure and stress, human-machine interface, environment, special equipment, specific fitness needs, as well as crew communications, staffing, and dynamics.
Accordingly, the HRA also evaluates recovery actions depending on whether they correspond or not to main control room abandonment situations.
CCNPP Page G-9
Constellation Enerov Nuclear Group Attachment G - Recovery Action Transition C..n..tel tion Ene rv Nuc.....ar rou Atah etG-Rcv f
cinTasto Recovery actions that are required by the FRE but not addressed in the HRA are evaluated for feasibility and documented in R2215-049-001 (Reference 6.22).
The reliability of these recovery actions is implicitly zero (i.e., not quantified) in the Fire PRA quantification for each specific fire area. Given that these recovery actions have been evaluated for both feasibility (Step 4 above) and for additional risk I adverse risk impact (Step 3 above), their contribution to the quantified risk associated with the respective fire area can only be zero, or a marginal positive improvement in the quantified risk based on an assumed finite value of reliability.
Since actions taken at primary control stations are not recovery actions, no independent reliability evaluation is required.
It should be noted that a reliability evaluation documented in the HRA was made for those actions taken at PCSs that are credited and modeled in the FPRA.
Results of the reliability assessments in the HRA and in R221 5-049-00 1 demonstrate that all credited NFPA 805 recovery actions are reliable.
CCNPP Page G-1O CCNPP Page G-10
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G-t - Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 16 16I1ABUS1 1 4KV UNIT BUS 11 ALIGN 1A EDG TO 4KV UNIT BUS 11 BY LOCALLY 16-16-1 RA BREAKERS CLOSING BREAKER 152-1103 16 16I1AEDG 1A DG AND BUS 17 TAKE LOCAL CONTROL OF BUS 17 BREAKERS AND 16-16-1 RA BREAKERS LOCALLY START 1A DG 16 16IALIGN11BUS 4KV UNIT BUS 11 OPEN FEEDER BREAKER FOR 4KV UNIT BUS 11 BY 16-16-1 RA BREAKERS LOCALLY TRIPPING BREAKERS 16 16IBUSI11BKR 4KV UNIT BUS 11 ALI GN 4KV B US 11 FOR IA DG BY TAKI NG LOCAL 16-16-I RA BREAKERS CONTROL AND TRIPPING BREAKERS 16 16ILOAD11IA 480V UNIT BUS I11A LOCALLY RECOVER 480V LOAD CENTER I1IA.
16-17-1 RA BREAKERS NOTE I 16 16ISTRIP11A 480V UNIT BUS 11A LOCALLY STRIP 480V LOAD CENTER 11A.
16-17-1 RA BREAKERS NOTE I 16 16ILOADI11B 480V UNIT BUS 11B LOCALLY RECOVER 480V LOAD CENTER 11IB.
16-18-1 RA BREAKERS NOTE I 16 16jSTRIP11B 480V UNIT BUS 11B LOCALLY STRIP 480V LOAD CENTER I1B.
16-18-1 RA BREAKERS NOTE 1 16 16I11BUHTR PRESSURIZER BACKUP ENERGIZE 11 PRESSURIZER BACKUP HEATER NONE PCS HEATER BANKS 11 16 16jSWACll SALTWATER AIR LOCALLY ALIGN UNIT1I SWAC-I1 TO THE UNIT1IAFW 16-20-1 RA COMPRESSORS VALVES PRIOR TO DEPLETION OF THE UNIT I AFW VALVE INSTRUMENT AIR ACCUMULATORS.
16 16IAFWP1C43 TDAFWP SPEED CONTROL ALIGN TDAFW PUMP SPEED CONTROL TO 1C43.
16-21-1 RA VALVES NOTE I 16 16IAFWFLOWIC43 AFW FLOW CONTROL ALIGN TDAFW FLOW CONTROL TO 1 C43.
16-22-1 RA VALVES 16-23-1 NOTE 2 16-24-1 16-25-1 16-26-1 16-28-1 16-29-1 16-30-1 16 1 61AFWFLOW1 TDAFWP HAND CONTROL AFW FLOW AT 1 C43 16-23-1 PCS CONTROLLERS 16-28-1 16 16ICSTINV CST LOCAL INDICATION MONITOR CST INVENTORY AT lC43.
16-33-0 RA/PCS NOTE 3 16 1611SO_1ERV4O2 MCC 114R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV 1ERV4O2 16-46-I RA 11449 CLOSED.
16 1611SO_1ERV404 MCC 104R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV IERV404 16-47-1 RA 10449 CLOSED.
CCNPP Page G-I 1
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 16 16IISWGR45VENT PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 1 45' 16-54-1 RA SWGR VENTILATION.
16-55-1 16-56-1 16 16j0CBUS24 4KV UNIT BUS 24 ENERGIZE 24 4KV UNIT BUS BY LOCALLY CLOSING 16-57-2 RA BREAKERS AND 0C BREAKER 152-2406 DISCONNECT 16 16[0CEDG 4KV UNIT BUS 07 LOCALLY RECOVER ONSITE AC POWER TO UNIT 2 4KV 16-57-2 RA BREAKERS UNIT BUS 24 (ALIGN 0C EDG). LOCALLY START 0C EDG.
16 16IOPEN4KVBKRS 4KV UNIT BUS 24 OPEN THE FEEDER BREAKER FOR 4KV UNIT BUS 24 16-57-2 RA BREAKERS AND 0C BY TRIPPING BREAKER 152-2414 DISCONNECT 16 16jLOAD24B 4KV UNIT BUS 24 SUPPLY ENERGIZE 480V LOAD CENTER 24B BY MANUALLY 16-58-2 RA BREAKER TO 480V UNIT CLOSING BREAKER 152-241 3 NOTE 1 BUS 24B 16 16[STRIP24B 480V UNIT BUS 24B LOCALLY STRIP 480V LOAD CENTER 24B.
16-58-2 RA BREAKERS NOTE I 16 1612S WAC VALVES 2-1A-302, 2-IA-303 LOCALLY ALIGN UNIT 2 SWAC-22 TO THE UNIT 2 AFW 16-59-2 RA AND 2-1A-317 VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW VALVE INSTRUMENT AIR ACCUMULATORS.
16 1 6IAFWFLOW2 TDAFWP HAND CONTROL AFW FLOW AT 2C43 16-60-2 RA CONTROLLERS 16 1 6IAFWFLOW2C43 AFW FLOW CONTROL ALIGN TDAFW FLOW CONTROL TO 2C43.
16-60-2 RA VALVES 16 16IAFWP2C43 TDAFWP SPEED CONTROL ALIGN TDAFW PUMP SPEED CONTROL TO 2C43.
16-60-2 RA VALVES 16 16I2SWGR45VENT PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 45' 16-70-2 RA SWGR VENTILATION.
16 16I1C43CONTROL TDAFWP HAND ESTABLISH TDAFW PUMP CONTROL AND AFW FLOW NONE PCS CONTROLLERS CONTROL AT IlC43 16 16IADV1C43 ADV HAND CONTROLLERS INITIALIZE ADV CONTROLLERS ON 1C43 NONE PCS 16 16ICHECKRXSD1 UNIT 1 WIDE RANGE NI OPEN OPTICAL ISOLATOR, ISOLATE CONTROL ROOM NONE PCS DETECTOR CHANNEL B TEST SIGNAL, ALIGN 1C43 TO WRNI CHANNEL B 16 16ICONSERVE1 REACTOR COOLANT CONSERVE RCS AND S/G INVENTORY BY ISOLATING NONE PCS SYSTEM AND STEAM BOUNDARY VALVES AT 1lC43.
GENERATOR VALVES 16 I6IRCSTEMPI ADV HAND CONTROLLERS CONTROL RCS TEMPERATURE AND VERIFY NATURAL NONE PCS CIRCULATION CCNPP Page G-12
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 16 16jSECHTR11_13 PRESSURIZER BACKUP SECURE PRESSURIZER BACKUP HEATER BANKS 11 NONE PCS HEATER BANKS 11 AND 13 AND 13 17 17ICSTINV CST LOCAL INDICATION LOCALLY MONITOR CST 12 INVENTORY AT 1LI5609.
17-02-0 RA NOTE 1 17 17I1AFW_RECHARGE VALVES 1-1A-733, 1-1A-737 LOCALLY ALIGN UNIT 1 SWAC-11 TO THE UNIT 1 AFW 17-03-1 RA AND 1-1A-728 VALVES PRIOR TO DEPLETION OF THE UNIT 1 AFW VALVE INSTRUMENT AIR ACCUMULATORS.
17 17[SWAC22 SALTWATER AIR LOCALLY ALIGN UNIT 2 SWAC-22 TO THE UNIT 2 AFW 17-10-2 RA COMPRESSORS VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW VALVE INSTRUMENT AIR ACCUMULATORS.
17 17I0CBKR 4KV UNIT BUS 07 LOCALLY RECOVER ONSITE AC POWER TO UNIT 2 4KV 17-11-2 RA BREAKERS UNIT BUS 24 (ALIGN OC EDG). LOCALLY CLOSE 4KV UNIT BUS 07 BREAKER 152-0701.
17 17[0CBUS24 4KV UNIT BUS 24 ALIGN 0C EDG TO 4KV UNIT BUS 24 BY LOCALLY 17-11-2 RA BREAKERS AND 0C OPENING BREAKER 152-2406 AND CLOSING DISCONNECT DISCONNECT 189-2406 17 1710CBUS24 4KV UNIT BUS 24 ENERGIZE 24 4KV UNIT BUS BY LOCALLY CLOSING 17-11-2 IRA BREAKERS AND 0C BREAKER 152-2406 DISCONNECT 17 17IOCEDG 4KV UNIT BUS 07 LOCALLY RECOVER ONSITE AC POWER TO UNIT 2 4KV 17-1 1-2 RA BREAKERS UNIT BUS 24 (ALIGN CC EDG). LOCALLY START 0C EDG.
17 17IALIGN24BUS 4KV UNIT BUS 24 ALIGN DC CONTROL POWER AND OPEN FEEDER 17-11-2 RA BREAKERS AND 0C BREAKER FOR 4KV UNIT BUS 24 BY TRIPPING DISCONNECT BREAKER 152-2403 17 171BUS24 4KV UNIT BUS 24 LOCALLY STRIP 4KV UNIT BUS 24 LOAD BREAKERS IN 17-11-2 RA BREAKERS AND 0C PREPARATION FOR RECOVERY WITH 0C EDG.
DISCONNECT 17 17IBUS24BKR 4KV UNIT BUS 24 OPEN THE FEEDER BREAKERS FOR 4KV UNIT BUS 24 17-11-2 RA BREAKERS AND CC BY TRIPPING BREAKERS 152-2401 AND 152-2414 DISCONNECT 17 17ILOAD24A 4KV UNIT BUS 24 SUPPLY ENERGIZE 480V LOAD CENTER 24A BY MANUALLY 17-12-2 RA BREAKER TO 480V UNIT CLOSING BREAKER NOTE 1 BUS 24A 17 17ISTRIP24A 480V UNIT BUS 24A LOCALLY STRIP 480V LOAD CENTER 24A.
17-12-2 RA BREAKERS NOTE 1 CCNPP Page G-13
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 17 17ILOAD24B 4KV UNIT BUS 24 SUPPLY ENERGIZE 480V LOAD CENTER 24B BY CLOSING 17-13-2 RA BREAKER TO 480V UNIT BREAKER NOTE 1 BUS 24B 17 17ISTRIP24B 480V UNIT BUS 24B LOCALLY STRIP AND RECOVER 480V LOAD CENTER 17-13-2 RA BREAKERS 24B.
NOTE 1 17 17I22BUHTR PRESSURIZER BACKUP OPERATE 22 PRESSURIZER BACKUP HEATER AT 2C43. NONE PCS HEATER BANK 22 TDAFWP 17 1 7IAFWP2C43 SPEED CONTROL VALVES ALIGN TDAFW PUMP SPEED CONTROL TO 2C43.
17-15-2 RA AFW FLOW CONTROL NOTE 1 17 17IAFWFLOW2C43 VALVES ALIGN TDAFW FLOW CONTROL TO 2C43.
17-16-2 RA 17-17-2 NOTE 4 17-18-2 17-19-2 17-20-2 17-26-2 17 1 7IAFWFLOW2 TDAFWP HAND CONTROL AFW FLOW AT 2C43 17-17-2 PCS CONTROLLERS 17 17JTRIP23AFWP 4KV UNIT BUS 24 OPEN THE MDAFW PUMP LOAD BREAKER FOR 4KV 17-21-2 RA BREAKERS UNIT BUS 24 BREAKER 152-2415 17-22-2 NOTE 1 17 1711SO._2ERV402 MCC 214R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV 2ERV402 17-41 -2 RA 21449 CLOSED.
17 17j1SO_2ERV404 MCC 204R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV 2ERV404 17-42-2 RA 20449 CLOSED.
17 17I2SWGR45VENT PORTABLE FAN UNITS CROSSTIE MCC 104R TO MCC 114R PER OI-27D, AND 17-47-2 RA INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 45' SWGR VENTILATION.
17 17I1SWGR27VENT PORTABLE FAN UNITS CROSSTIE MCC 104R TO MCC 114R PER OI-27D, AND 17-49-1 RA INSTALL PORTABLE FANS FOR TEMPORARY UNIT 1 27' SWGR VENTILATION.
17 17jAFWFLOWI TDAFWP HAND CONTROL AFW FLOW AT 1 C43 17-50-1 RA CONTROLLERS 17 1 7IAFWFLOW1 C43 AFW FLOW CONTROL ALIGN TDAFW FLOW CONTROL TO 1 C43.
17-50-1 RA VALVES 17 17IAFWPIC43 TDAFWP SPEED CONTROL ALIGN TDAFW PUMP SPEED CONTROL TO 1C43.
17-50-1 RA VALVES 17 17I2C43CONTROL TDAFWP HAND ESTABLISH TDAFW PUMP CONTROL AND AFW FLOW NONE PCS CONTROLLERS CONTROL AT 2C43 17 171ADV2C43 ADV HAND CONTROLLERS INITIALIZE ADV CONTROLLERS ON 2C43 NONE PCS CCNPP Page G-14
v_
Attachment G - Recovery Actions Transition Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 17 17ICHECKRXSD2 UNIT 2 WIDE RANGE NI OPEN OPTICAL ISOLATOR, ISOLATE CONTROL ROOM NONE PCS DETECTOR CHANNEL A TEST SIGNAL, ALIGN 2C43 TO WRNI CHANNEL A 17 17ICONSERVE2 REACTOR COOLANT CONSERVE RCS AND S/G INVENTORY BY ISOLATING NONE PCS SYSTEM AND STEAM BOUNDARY VALVES AT 2C43.
GENERATOR VALVES 17 17IRCSTEMP2 ADV HAND CONTROLLERS CONTROL RCS TEMPERATURE AND VERIFY NATURAL NONE PCS CIRCULATION 17 17ISECHTR21_23 PRESSURIZER BACKUP SECURE PRESSURIZER BACKUP HEATER BANKS 21 NONE PCS HEATER BANKS 21 AND 23 AND 23 24 2411ABUS11 4KV UNIT BUS 11 ENERGIZE 11 4KV UNIT BUS BY CLOSING BREAKER 24-01-1 RA BREAKERS 152-1103 24 241BU5ll 4KV UNIT BUS 11 LOCALLY STRIP 4KV UNIT BUS 11 LOAD BREAKERS IN 24-01-1 RA BREAKERS PREPARATION FOR RECOVERY WITH IA EDG.
24 24IBUSllBKR 4KV UNIT BUS 11 OPEN THE FEEDER BREAKERS FOR 4KV UNIT BUS 11 24-01-1 RA BREAKERS BY TRIPPING BREAKERS 152-1101 AND 152-1115 24 24IOPEN1ABKR IA DIESEL GENERATOR TAKE LOCAL CONTROL AND OPEN 1A DIESEL 24-01-1 RA OUTPUT BREAKER GENERATOR OUTPUT BREAKER 152-1 703 24 24ILOAD11IA 4KV UNIT BUS 11 SUPPLY ENERGIZE 480V LOAD CENTER I1IA BY MANUALLY 24-02-1 RA BREAKER TO 480 UNIT BUS CLOSING BREAKER 152-1114 NOTE 1 11A 24 24ISTRIP11IA 480V UNIT BUS I1IA LOCALLY STRIP 480V LOAD CENTER I1IA.
24-02-1 RA BREAKERS NOTE I 24 24I1SWGRHVAC 11 SWlTCHGEAR ROOM RESTORE SWITCHGEAR ROOM VENTILATION 24-02-1 RA VENT FAN HANDSWITCH NOTE I 24 24ICHARGERS1 BATTERY 11 AND 14 PLACE 11 AND 14 BATTERY CHARGERS IN SERVICE 24-02-1 PA CHARGER BREAKERS 24-03-1 NOTE I 24 24ILOAD11IB
.4KV UNIT BUS 11 SUPPLY ENERGIZE 480V LOAD CENTER 11 B BY CLOSING 24-03-1 RA BREAKER TO 480 UNIT BUS BREAKER 152-1102 NOTE I I1B 24 24ISTRIP11IB 480V UNIT BUS 11 B LOCALLY STRIP AND RECOVER 480V LOAD CENTER 24-03-I RA BREAKERS IIB.
NOTE 1 24 24111BUHTR PRESSURIZER BACKUP OPERATE 11 PRESSURIZER BACKUP HEATER AT 1C43. NONE PCS HEATER BANK 11 24 24ISWACllI SALTWATER AIR LOCALLY ALIGN UNIT I SWAC-11I TO THE UNIT I AFW 24-05-1 RA COMPRESSORS VALVES PRIOR TO DEPLETION OF THE UNIT I AFW VALVE INSTRUMENT AIR ACCUMULATORS.
CCNPP Page G-15
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
,Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 24 24I1C43CONTROL TDAFWP HAND ESTABLISH TDAFW-PUMP CONTROL AND AFW FLOW 24-06-1 PCS CONTROLLERS CONTROL AT I1C43 24 24IAFWP1 C43 TDAFWP SPEED CONTROL ALIGN TDAFW PUMP SPEED CONTROL TO 1 C43.
24-06-1 RA VALVES 24-07-1 NOTE 8 24 24IAFWFLOWI1C43 AFW FLOW CONTROL ALIGN AFW FLOW CONTROL TO 1 C43 24-07-1 RA VALVES 24-08-1 NOTE 5 24-09-1 24-10-1 24-11-1 24-12-1 24-1 3-1 24-14-1 24 24IAFWFLOW1 TDAFWP HAND CONTROL AFW FLOW AT 1C43 24-08-1 PCS CONTROLLERS 24 2411S1 ERV402 MCC 1 14R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV 1 ERV402 24-26-1 RA 11449 CLOSED.
24 24jlSOIERV404 MCC 104R BREAKER 52-DE-ENERGIZE BREAKER TO FAIL PORV IERV404 24-27-1 RA 10449 CLOSED.
24 24j0CBKR 4KV UNIT BUS 07 LOCALLY RECOVER ONSITE AC POWER TO UNIT 2 4KV 24-38-2 RA BREAKERS UNIT BUS 24 (ALIGN 0C EDG). LOCALLY CLOSE 4KV UNIT BUS 07 BREAKER 152-0701.
24 2410CBUS24 4KV UNIT BUS 24 ENERGIZE 24 4KV UNIT BUS BY LOCALLY CLOSING 24-38-2 RA BREAKERS AND 0C BREAKER 152-2406.
DISCONNECT 24 2410CEDG 4KV UNIT BUS 07 LOCALLY RECOVER ONSITE AC POWER UNIT 2 4KV 24-38-2 RA BREAKERS UNIT BUS 24 (ALIGN 0C EDG). LOCALLY START 0C EDG.
24 24IALIGNOCBUS24 4KV UNIT BUS 24 ALIGN 0C EDG TO 4KV UNIT BUS 24 BY LOCALLY 24-38-2 RA BREAKERS AND 0C OPENING BREAKER 152-2406 AND CLOSING DISCONNECT DISCONNECT 189-2406.
24 241BUS24 4KV UNIT BUS 24 LOCALLY STRIP 4KV UNIT BUS 24 LOAD BREAKERS IN 24-38-2 RA BREAKERS AND 0C PREPARATION FOR RECOVERY WITH 0C EDG.
DISCONNECT 24 24IBUS24BKR 4KV UNIT BUS 24 OPEN THE FEEDER BREAKERS FOR 4KV UNIT BUS 24 24-38-2 RA BREAKERS AND 0C BY TRIPPING BREAKERS 152-2401 AND 152-2414.
DISCONNECT 24 24IOPEN2BBKR 2B DIESEL GENERATOR TAKE LOCAL CONTROL AND OPEN 2B DIESEL 24-38-2 IRA OUTPUT BREAKER GENERATOR OUTPUT BREAKER 152-2403.
CCNPP Page G-16
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 24 24ILOAD24A 4KV UNIT BUS 24 SUPPLY ENERGIZE 480V LOAD CENTER 24A BY CLOSING 24-39-2 RA BREAKER TO 480V UNIT BREAKER 152-2402.
NOTE 1 BUS 24A 24 241STRIP24A 480V UNIT BUS 24A LOCALLY STRIP AND RECOVER 480V LOAD CENTER 24-39-2 RA BREAKERS 24A.
NOTE 1 24 24ICHARGERS2 BATTERY 21 AND 24 PLACE 21 AND 24 BATTERY CHARGERS IN SERVICE.
24-39-2 RA CHARGER BREAKERS 24-40-2 NOTE 1 24 241LOAD24B 4KV UNIT BUS 24 SUPPLY ENERGIZE 480V LOAD CENTER 24B BY MANUALLY 24-40-2 RA BREAKER TO 480V UNIT CLOSING BREAKER 152-2413.
NOTE 1 BUS 24B 24 241STRIP24B 480V UNIT BUS 24B LOCALLY STRIP 480V LOAD CENTER 24B.
24-40-2 RA BREAKERS NOTE 1 24 24I2SWGRHVAC 22 SWITCHGEAR ROOM RESTORE SWITCHGEAR ROOM VENTILATION.
24-40-2 RA VENT FAN HANDSWITCH NOTE 1 24 24I23BUHTR PRESSURIZER BACKUP OPERATE 23 PRESSURIZER BACKUP HEATER AT 2C43. NONE PCS HEATER BANK 23 24 241SWAC22 SALTWATER AIR LOCALLY ALIGN UNIT 2 SWAC-22 TO THE UNIT 2 AFW 24-42-2 RA COMPRESSORS VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW VALVE INSTRUMENT AIR ACCUMULATORS.
24 24I2C43CONTROL TDAFWP HAND ESTABLISH TDAFW PUMP CONTROL AND AFW FLOW 24-43-2 PCS CONTROLLERS CONTROL AT 2C43.
24 241AFWP2C43 TDAFWP SPEED CONTROL ALIGN TDAFW PUMP SPEED CONTROL TO 2C43.
24-43-2 RA VALVES 24-44-2 NOTE 7 24 24IAFWFLOW2C43 AFW FLOW CONTROL ALIGN AFW FLOW CONTROL TO 2C43.
24-43-2 RA VALVES 24-44-2 NOTE 6 24-45-2 24-46-2 24-47-2 24-48-2 24-49-2 24-50-2 24-51-2 24-45-2 24-63-2 24-64-2 24 24IAFWFLOW2C43 2411SO_2ERV402 241ISO_2ERV404 TDAFWP HAND CONTROLLERS MCC 214R BREAKER 52-21449 MCC 204R BREAKER 52-2fl449 ULUULU-CONTROL AFW FLOW AT 2C43 DE-ENERGIZE BREAKER TO FAIL PORV 2ERV402 CLOSED.
DE-ENERGIZE BREAKER TO FAIL PORV 2ERV404 PCS RA RA CCNPP Page G-17 CCNPP Page G-17
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 24 24ICSR EM VENT MCRICSR EMERGENCY ESTABLISH CSR EMERGENCY VENTILATION.
24-77-0 RA VENTILATION SYSTEM NOTE 1 24 24I1MDAFW 4KV BUS 11 BREAKER 152-LOCALLY START 13MDAFW PUMP BY CLOSING 24-07-1 RA 1116 BREAKER 152-1116.
24 24j2MDAFW 4KV BUS 24 BREAKER 152-LOCALLY START 23MDAFW PUMP BY CLOSING 24-44-2 RA 2415 BREAKER 152-241 5.
24 24IADVl C43 ADV HAND CONTROLLERS INITIALIZE ADV CONTROLLERS ON 1lC43 NONE PCS 24 241ADV2C43 ADV HAND CONTROLLERS INITIALIZE ADV CONTROLLERS ON 2C43.
NONE PCS 24 24ICHECKRXSD1 UNIT 1 WIDE RANGE NI OPEN OPTICAL ISOLATOR, ISOLATE CONTROL ROOM NONE PCS DETECTOR CHANNEL B TEST SIGNAL, ALIGN 1C43 TO WRNI CHANNEL B 24 24ICHECKRXSD2 UNIT 2 WIDE RANGE NI OPEN OPTICAL ISOLATOR, ISOLATE CONTROL ROOM NONE PCS DETECTOR CHANNEL A TEST SIGNAL, ALIGN 2C43 TO WRNI CHANNEL A 24 24ICONSERVE1 REACTOR COOLANT CONSERVE RCS AND SIG INVENTORY BY ISOLATING NONE PCS SYSTEM AND STEAM BOUNDARY VALVES AT lC43.
GENERATOR VALVES 24 24ICONSERVE2 REACTOR COOLANT CONSERVE RCS AND SIG INVENTORY BY ISOLATING NONE PCS SYSTEM AND STEAM BOUNDARY VALVES AT 2C43.
GENERATOR VALVES 24 24ICSTINV CST 12 LEVEL INDICATORS MONITOR CST INVENTORY AT 1C43 AND 2C43.
NONE PCS AT 1 C43 AND 2C43 24 24IRCSTEMP1 ADV HAND CONTROLLERS CONTROL RCS TEMPERATURE AND VERIFY NATURAL NONE PCS CIRCULATION 24 24IRCSTEMP2 ADV HAND CONTROLLERS CONTROL RCS TEMPERATURE AND VERIFY NATURAL NONE PCS CIRCULATION.
24 24ISECHTRI1_13 PRESSURIZER BACKUP SECURE PRESSURIZER BACKUP HEATER BANKS 11 NONE PCS HEATER BANKS 11 AND 13 AND 13 24 24ISECHTR21._23 PRESSURIZER BACKUP SECURE PRESSURIZER BACKUP HEATER BANKS 21 NONE PCS HEATER BANKS 21 AND 23 AND 23.
17A 17AI2AFW_RECHARGE VALVES 2-1A-302, 2-1A-303, LOCALLY ALIGN UNIT 2 SWAC TO THE UNIT 2 AFW 17A-01-0 RA 2-1A-314 AND 2-1A-317.
VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW 17A-02-2 VALVE INSTRUMENT AIR ACCUMULATORS.
CCNPP Page G-I 8
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
- Fire Area Recovery Action ID Component Description Action VFDR RA/PCS 17A 17AI1AFWRECHARGE VALVES 1-1A-733, 1-1A-737 LOCALLY ALIGN UNIT 1 SWAC-11 TO THE UNIT 1 AFW 17A-01-0 RA AND 1-1A-728 VALVES PRIOR TO DEPLETION OF THE UNIT I AFW 17A-03-1 VALVE INSTRUMENT AIR ACCUMULATORS.
TB/NSB/ACA TB/NSB/ACAI1AFWREC VALVES 1-1A-733, 1-1A-737 LOCALLY ALIGN UNIT I SWAC TO THE UNIT I AFW TB/NSB/ACA-RA HARGE AND 1-IA-728 VALVES PRIOR TO DEPLETION OF THE UNIT 1 AFW 01-1 VALVE INSTRUMENT AIR ACCUMULATORS.
TB/NSB/ACA TB/NSB/ACAI2AFWREC VALVES 2-1A-302, 2-1A-303, LOCALLY ALIGN. UNIT 2 SWAC TO THE UNIT 2 AFW TB/NSB/ACA-RA HARGE 2-1A-314 AND 2-1A-317.
VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW 02-2 VALVE INSTRUMENT AIR ACCUMULATORS.
TB/NSB/ACA TB/NSB/ACAI1SWGR27V PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 1 27' TB/NSB/ACA-RA ENT SWGR VENTILATION.
04-1 TB/NSB/ACA TB/NSB/ACAI1SWGR45V PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 1 45' TB/NSB/ACA-RA ENT SWGR VENTILATION.
05-1 TB/NSB/ACA TBINSB/ACA[2SWGR27V PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 27' TB/NSB/ACA-RA ENT SWGR VENTILATION.
06-2 TB/NSB/ACA TB/NSB/ACAI2SWGR45V PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 45' TB/NSB/ACA-RA ENT SWGR VENTILATION.
07-2 YARD YARD[1AFWRECHARG VALVES 1-1A-733, 1-IA-737 LOCALLY ALIGN UNIT I SWAC-12 TO THE UNIT 1 AFW YARD-06-I RA E
AND 1-1A-728 VALVES PRIOR TO DEPLETION OF THE UNIT I AFW VALVE INSTRUMENT AIR ACCUMULATORS.
YARD YARDI1SWGR27VENT PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 1 27' YARD-11-I RA SWGR VENTILATION.
YARD-I12-1 YARD YARDI2AFWRECHARG VALVES 2-1A-302, 2-1A-303, LOCALLY ALIGN UNIT 2 SWAC TO THE UNIT 2 AFW YARD-i15-2 RA E
2-1A-314 AND 2-1A-317.
VALVES PRIOR TO DEPLETION OF THE UNIT 2 AFW VALVE INSTRUMENT AIR ACCUMULATORS.
YARD YARDj2SWGR27VENT PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 27' YARD-18-2 RA SWGR VENTILATION.
YARD YARDI2SWGR45VENT PORTABLE FAN UNITS INSTALL PORTABLE FANS FOR TEMPORARY UNIT 2 45' YARD-I18-2 RA SWGR VENTILATION.
CCNPP Page G-19 CCNPP Page G-19
Constellation Energy Nuclear Group Attachment G - Recovery Actions Transition Table G-1 Notes:
Note 1: The recovery actions that address VFDRs 16-17-1, 16-18-1, 16-21-1, 16-58-2, 17-02-0, 17-12-2, 17-13-2, 17-15-2, 17-21-2, 17-22-2, 24-02-1, 24-03-1, 24-39-2, 24-40-2 and 24-77-0 are credited for defense-in-depth.
Note 2: Portions of this recovery action that address VFDRs 16-22-1 and 16-23-1 are required for risk reduction. Portions of this recovery action that address VFDRs 16-24-1, 16-25-1, 16-26-1, 16-28-1, 16-29-1, and 16-30-1 are credited for defense-in-depth.
Note 3: Monitoring of CST inventory at 1C43 is a POS action for Unit 1, but is considered a recovery action credited for defense-in-depth for Unit 2.
Note 4: Portions of this recovery action that address VFDRs 17-16-2 and 17-17-2 are required for risk reduction. Portions of this recovery action that address VFDRs 17-18-2, 17-19-2, 17-20-2, and 17-26-2 are credited for defense-in-depth.
Note 5: Portions of this recovery action that address VFDRs 24-07-1 and 24-08-1 are required for risk reduction. Portions of this recovery action that address VFDRs 24-09-1, 24-10-1, 24-11-1, 24-12-1, 24-13-1, and 24-14-1 are credited for defense-in-depth.
Note 6: Portions of this recovery action that address VFDRs 24-44-2 and 24-45-2 are required for risk reduction. Portions of this recovery action that address VFDRs 24-43-2, 24-46-2, 24-47-2, 24-48-2, 24-49-2, 24-50-2, and 24-51-2 are credited for defense-in-depth.
Note 7: Portions of this recovery action that address VFDR 24-44-2 are requried for risk reduction. Portions of this recovery action that address VFDR 24-43-2 are credited for defense-in-depth.
Note 8: Portions of this recovery action that address VFDR 24-07-.1 are requried for risk reduction. Portions of this recovery action that address VFDR 24-06-1 are credited for defense-in-depth.