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CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N. CONNECTICUT P.o. box 27o HARTFoRo. CONNECTICUT 06141-0270 TELEPHONE 2os-ess sooo March 31,1986 Docket No. 50-213 B12020 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen: | |||
Haddam Neck Plant Probabilistic Safety Study - Summary Report and Results Northeast Utilities Service Company (NUSCO), on behalf of Connecticut Yankee Atomic Power Company (CYAPCO), has recently completed a plant-specific Probabilistic Safety Study (PSS) including a Best Estimate LOCA Analysis for the Haddam Neck Plant. The studies have been and will be utilized for many purposes. Among the near-term applications for the PSS is for assistance in making backfitting decisions in the context of the Integrated Safety Assessment Program (ISAP) currently being implemented at the Haddam Neck Plant. | |||
===Background=== | |||
During the past several years, NUSCO has been implementing a Living PRA Frogram for the nuclear plants within the Northeast Utilities (NU) system. The major element of this program is the development, maintenance and use of PRA models, of each of the system's nuclear power plants, for assistance in evaluating potential plant backfits and operating procedures modifications. The Living PRA Program affords us the flexibility to quickly and accurately analyze the impact on plant safety of changes to the plant's design configuration and significant changes to operating procedures. | |||
NUSCO has completed the development of a computerized PRA model of the Haddam Neck Plant. This model will be periodically updated to incorporate plant design changes, significant operational changes and relevant updated equipment performance data. | |||
The Haddam Neck PSS utilized current state-of-the-art computational and modeling techniques. Included in the study were a comprehensive analysis of best estimate plant specific LOCA and transient response, Haddam Neck specific operating, reliability and maintenance data, the latest plant design 8604180083 860331 3 PDR ADOCK 0500 P | |||
gg | |||
'lI9 4 | |||
changes and draft symptom-oriented . Emergency Operating Procedures.. The | |||
, scope of the PSS also . included support system failure initiated events , | |||
l interfacing system LOCA events and fire initiated events.(1) | |||
As noted in SECY-85-160,(2) theLintegrated assessment of the Haddam Neck j Plant is scheduled to be completed during 1986. The schedule was based, in part, i on our schedule for completion of and documentation of a summary report of the Haddam Neck PSS .during March 1986. In keeping with the schedule for completion of the integrated assessment, our submittal of the.Haddam Neck PSS Summary Report and Best Estimate LOCA Analysis will enable the Staff to begin l the probabilistic safety analysis review phase of the integrated assessment of | |||
+ | |||
Haddam Neck. ! | |||
i 1 j Results It is Northeast Utilities Corporate Policy to provide dependable and economic power to its customers, without endangering the health and safety of the public. | |||
As one element of achieving this policy, NU has set forth quantitative safety- | |||
, goals for the design and operation of our nuclear power plants. If it is l determined that any one of our nuclear power plants does not meet the | |||
, acceptance criteria as established by our safety goals, our policy is to take the corrective action necessary to meet the criteria, on a schedule commensurate with the level of deficiency found. | |||
The Haddam Neck PSS calculated a mean core melt frequency of 5.5 x 10-4 per l reactor-year for the Haddam Neck Plant which exceeds one of our safety goals. l As a result, we have committed resources to work to decrease the overall core l melt frequency at the Haddam Neck Plant. We have begun evaluating some of | |||
: the engineering insights obtained from the PSS, in order to determine .the alternatives available to reduce the calculated core melt frequency. A summary- . | |||
! of the engineering insights obtained from the PSS and items we are evaluating to reduce the calculated core melt risk at the Haddam Neck Plant are contained in | |||
, Enclosure 2. | |||
! As outlined in Enclosures I and 2, the PSS and Best Estimate LOCA Analysis i | |||
yielded many insights into the operation of the plant including: | |||
o Containment heat removal is available for most core melt sequences. As the availability of containment heat removal is--' considered to be an | |||
~ | |||
effective means of maintaining containment integrity and removing fission ~ | |||
products to prevent large scale radioactive releases if core melt occurs, . | |||
the public risk impact of a majority of the dominant core me't se_quences is reduced by having containment heat removal available. ' | |||
(1) The fire initiated events section of the PSS has not been completed at the present time. NUSCO will document and docket the updated results'to the | |||
, Staff following completion of the analysis. . | |||
(2) SECY-85-160, " Integrated Safety Assessment Program - Implementation Plan," dated May 6,1985. | |||
i | |||
.-----+~~_+,.--+_-~,.-_.,,.,~,,<,a ,,..-,,,,,-,n+, w,,,_,n.. ,,,,.,,,..e_,n,,,,--,.n ,..,,n, , w n,., ~,-r-n,.-. | |||
3-4 o . An AC independent means of providing containment spray during a station blackout is provided by- the diesel-driven fire pump. This feature is important in minimizing the offsite public consequences resulting from a | |||
, core melt accident due to station blackout, o Approximately 40% of the calculated core melt frequency at Haddam Neck ' . | |||
is attributable to small and medium break LOCA scenarios. The PSS | |||
. utilized LOCA frequencies which were derived from generic PWR data as presented in WASH-1400. As the Haddam Neck Plant utilizes austenitic stainless steel which has superior ductility toughness .and resistance to l brittle fracture compared to ferritic carbon steel used in many PWR-plants, the actual Haddam Neck Plant LOCA frequency is expected to be lower. | |||
o The Haddam Neck . plant-specific control room simulator has -been completed and is currently being used for operator training. This affords i us the opportunity to improve the training of operators to respond to plant transient situations. l i In light of the above, and our continuing efforts to identify, evaluate and when necessary implement procedural or hardware modifications to upgrade the operation of the plant, CYAPCO has concluded that continued operation of the Haddam Neck Plant does not pose any undue risk to the public. | |||
; Summary Report , | |||
I l | |||
; NUSCO has completed summary reports of the Haddam Neck PSS and Best . l Estimate to this letter.13 LOCA) Analysis which we are providing to the Staff as an attachment The PSS summary report is a comprehensive summary of the core melt frequencies for various sequences calculated in the ; study and the , | |||
calculational methodologies and analysis techniques utilized in the development ! | |||
of the PRA model. The PSS summary report contains the items outlined below: | |||
4 o Determination of Initiating Events j i - | |||
System investigations Initiator frequency calculations o Accident Sequence Analysis Classification of event sequence outcomes Plant systems event tree models Plant support system event tree models | |||
'; o Plant Systems Reliability Analysis Plant component reliability data collection and analysis l Plant systems reliability modeling. I I | |||
( | |||
(3) The summary reports consist of the Haddam Neck Probabilistic Safety Study (Volumes 1-4) and the Best Estimate LOCA Analysis (one volume) for I the Maddam Neck Plant. | |||
t | |||
-- .- --, ..-*-y. ,,-,,e m~ . . - , , ~ , , - ,,,y-% -,.my.,1rw-- we* ,,--,.%c,n ,,,w-,. .,--%< -.-.y-e ,--y,- - | |||
,-<w,,pr e e- = e -y* | |||
-4 o Human Reliability Analysis | |||
- Introduction and methodology | |||
- Screening analysis | |||
- Detailed representation of operator action - | |||
- Summary of results o Accident Sequence Quantification - | |||
- Matrix quantification | |||
- Core melt accident sequence quantification results | |||
- Containment heat removal reliability consideration , | |||
In support of the PSS, a Best Estimate.LOCA Analysis was performed concurrent-with the development of the PSS models. ~ The Best Estimate LOCA Analysis , | |||
summary report describes the results of comprehensive plant specific accident-analyses used to determine best estimate system success criteria, best estimate plant response to failed equipment and components, and time frames for operator recovery of failed systems. These calculations were performed using the NULAP 5 Code with best estimate input values. The Best Estimate LOCA - | |||
Analysis contains analyses of the scenarios described below: | |||
o Station Blackout A | |||
; o incore Instrument Tube Rupture i o Steam Generator Tube Rupture 1 o Large Break LOCA 4 o Medium Break LOCA o Small Break LOCA o Total Loss of Main and Auxiliary Feedwater with Feed and Bleed Cooling o Anticipated Transient Without Scram 3 | |||
o Total Loss of DC Power In accordance with your request we are providing 26 copies of the PSS and Best-Estimate LOCA Analysis to the ISAP Project Directorate for distribution within , | |||
the NRC (including NRR, Region 1, ACRS, etc.). | |||
i | |||
! In addition to the summary reports, we are providing the Staff with additional mformation on the scope and results of the PSS as outlined below, o Enclosure 1 A summary of the results of the PSS is provided herein. | |||
o Enclosure 2 1 l | |||
Discussions of the major engineering insights NUSCO has obtained from the PSS including the following two plant design changes already implemented at the plant are provided herein. | |||
o Plant Design change to eliminate the common dependence of i emergency diesel generator cooling on Motor Control Center-5 (MCC-5). This dependence, found while doing the analysis, could | |||
, be significant during a loss of offsite power event as the diesel i generators are dependent on MCC-5 to maintain diesel generator cooling. However, the diesel generators are not dependent upon i | |||
o | |||
._ _ _ , , _ _ _ - . , _ . _ , _ . - . - _ . - - _ . , _ . . .__,_~.__.--.a,_.-._,-,_.-__ _._._,a._... | |||
_3_ | |||
MCC-5 to start up following a loss of offsite power event. If one or both diesel generators start-up, accept their emergency loads - | |||
and properly reenergize MCC-5 ,as designed, the original configuration would result in long-term diesel generator cooling . | |||
availability. | |||
I o Plant design change to mitigate the effects of the loss of MCC-5 e | |||
(as an initiator) by tripping the charging pumps. | |||
These design changes were credited in our quantification of the PSS as a result of our preliminary evaluations of the PSS models which highlighted i these design conditions as being potential significant contributors to the j total core melt frequency at the plant. | |||
Upcoming Activities In order to facili. tate Staff. review and understanding of the Haddam Neck PSS and to ensure that maximum utilization of the PSS is achieved, in ISAP and other applications, we are willing to meet with your Staff to review the PSS. Our efforts to date have focused on completing and documenting this study, and on a | |||
; limited number of high priority issues which were identified during the conduct of the PSS. In the near term, we plan to complete and document our analysis of fire initiated events and to evaluate the report in more detail to identify | |||
, additional issues which warrant further study. Consistent with the overall ISAP | |||
! framework, these issues will be identified, new ISAP topics will be proposed and i they will be evaluated in a manner consistent with the process in place for all | |||
; other ISAP topics. | |||
I Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY - | |||
1F Cd J. F. Opeka U Senior Vice President cc: T. E. Murley, Region I (with Enclosures 1 and 2 only) 1 1 | |||
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Docket No. 50-213 B12020 i | |||
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1- - | |||
i Enclosure 1 i | |||
. Summary of the Results of the Haddam Neck Probabilistic Safety Study-1 | |||
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1 i | |||
i S. | |||
1 i ; | |||
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March,1986 - | |||
i Summary of Results The Haddam Neck PSS calculated a mean value core melt. frequency of 5.5 x - | |||
j 10-4 per reactor-year. The core melt frequency is based on the internal- | |||
- initiators examined in the study which include anticipated transients, special initiators, as well as hypothetical design basis accidents. A summary of the | |||
; dominant core melt sequences is provided in Table I. | |||
The dominant core melt sequences involve a loss of high pressure recirculation during small-break and medium-break LOCAs. The high pressure recirculation system at Haddam Neck utilizes the charging system in conjunction with the residual heat removal system. 'The high pressure safety injection system is not currently used for high pressure recirculation. The charging system has only one injection path designed for use during recirculation, into Loop 2 cold leg. | |||
j Failure of the charging system results in loss of high pressure recirculation. | |||
t l' In addition to loss of high pressure recirculation due to random component failures, the Haddam Neck PSS considered the loss of recirculation due to a break in the charging injection path. If a small- or medium-break LOCA occurs | |||
, in Loop 2 cold leg, connecting piping or in the charging line downstream of the check valves and is sized sufficiently to degrade the charging flow rate but not i | |||
I depressurize function and we thehave system, high pressure assumed its failurerecirculation will(' l)fail in this analysis. Theto loss perform its - | |||
of high pressure recirculation either due to component failure or an adverse break size l and location, has been identified as an important contributor to core melt | |||
; frequency (14% of total core melt frequency). | |||
i In addition, sequences caused by the loss of offsite power and/or Motor Control Center-5 are also found to be significant. Under general plant transients, consequential LOCAs resulting from an unisolated, stuck-open PORV are important contributors. | |||
, The remaining core melt sequences listed in Table I show that, besides those | |||
; sequences already mentioned, no other single accident sequence dominates core i melt frequency. The table also shows that containment heat removalis available l for most of the core melt sequences. For the station blackout sequence, the i availability of the diesel-driven fire pump for containment spray is an important consideration. The availability of containment heat removal is considered in the analysis as a means of maintaining containment integrity and as an active means i of removing fission products to prevent large scale radioactive releases should core melt occur. The public risk impact of the dominant core melt sequences is - | |||
i reduced by having containment heat removal available. | |||
A listing of the various initiating events and their contributions to the core melt l frequency is presented in Table II. Small-break LOCA and loss of offsite power i events are the largest contributors, representing 24.45 % and 22.45 % | |||
respectively, of the calculated core melt frequency. | |||
i i | |||
(1) The implications of this situation with respect to 10CFR50.72 and 50.73 are being evaluated separately and are the subject of separate correspondence. | |||
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Table I (Con't) | |||
DON!NANT CORE del.T ACCllENT SEQUt NCES CONTP9FT. | |||
FAII.ED ltEAT FPEO./ 3 0F Stirr0RT SEQUENCE WOTRS REMITAL TEAR _ TOTAL CW SISTstt5 DESCRIFTION 195T18708 S t uck Dr e n, 3.18E-5 2.15 4 o Reactor Trly Falle Avall. Unisolat e s ronv/ | |||
General Branch to ET22 SRV durtra AIW3 j | |||
Flant Hanual Trip Falls, Transient Fressuriser Beller Falls OR 0 37 Emergency poration Avall. 2.02E-6 Falls OR 0.21 A vall . 3.55E-6 Hain Ieedwater Unavall, AFW Falls 3.55E-5 2.50 ET26 1s 14entl. | |||
LDSte o Bus 4 or 7 Avall or Avell. cat to EI2) See Loss or Falls, Conseq. ISCL ET23 (DCleEAC8 Offsite | |||
+DC2*EAC9) Branch to E126, Seal Fower Fallura LOCA, Rc5 Depress. Falla, OR HFSI Injeo. Falla Avell. l.0$E-5 f 92 OR 0.44 2.40E-6 1 Low Pressure Seelro. Avell. | |||
Falls I | |||
o AFW Falls, Fall to Avell. 2.45E-6 0.45 Decover of fstte Power, Bleed and Feed Falle OR Avall. 9.02E-7 0.17 Long Term Cooling Falls a) Ava ll . l.02E-5 5.86 All o AFW Falls 5.85E-7 0.11 Support Recover Power Win. b)Unavall. | |||
Systees 30 Hinutes,liigh IComb. of Fressure injection 2 Losses of Falls DC. AC, j Serv. Water 4 | |||
l l . | |||
Table I (Con't) | |||
TAM E 6.3-5 DONlWANT CORE El.T ACCIDENT SEQUENCFJ IXNITittf. | |||
FAILED nEar FsEo.i s Or Note 3 surr0nt stauEnCe 1UTM. CW BE8DfAL _TEAa_ | |||
SISTEMS_ BE3CRIPTICII IslTIA10e 8.54E-6 1.56 o AFW Falla a) Avall. II 86E-7 0.09 | |||
' Loss or Fall to Recover b) Unavail. | |||
Offslte Fower within 30 Min. | |||
Fower Recover Fower within 100 Min. | |||
Illgh Fressure i | |||
Injection Falls 9 05E-6 1.65 . | |||
o AFW Avallable a) Avell. 5. lie-7 0,09 Fall to Recover b) Unavail. | |||
within 15 Minutes, Recover Fower f | |||
' within 8 Mours, High Pressure Injection Falla Avell 5.78E-7 0.11 o AFW Avellable Fall to pecover Power within 15 Min, Fall to Recover within 8 Moura Onantified usins 2.00E-5 3.65 results of IACge o AFW Avellable, Fall Avall. | |||
to Recover Fower Section 4.3 17 EAC9 (Station within 15 Min, Fall Slackout) to Secover Power within 8 Houra | |||
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Table I (Con't) | |||
DOHtWANT CORE E LT ACCIDENT SEQUENCES CowitelT. | |||
FAILED nEAr r no./ : Or StoUENCe NOTE 3 | |||
! SUrPott REPENAL TEAR _ TUTAL CW I SISTEMS DE3CalFit0N 1Ntf tkt0E_ 0.5) ET26 is ident. | |||
Avalt. 2.89E-6 to El?), See o Csnseq. LOCA - | |||
Loss of Branch to ET23, ET 2) | |||
Offstte Seal Fatture LOCA. | |||
Power RCS Depres. Falls | |||
* OR 0.17 Avall. 9.1)E-7 t.ow Pressure Rectro.Fatts | |||
* OR 0.09 Avall. 4.69E-7 High Pressure - | |||
Injection Fatts Avalt. 7.69E-6 1.40 LOSF'HCC5 o Fett to pecover (DC1 EACS. MCC-5 AFW Avell., | |||
DC2*EAC9) Frteary Integrity Fatts Avatt 3.80E-6 0.69 o Fett to Secover NCC-5, AFW Fatts 0.99 Mot Evalunta4 | |||
-- 4.90E-6 Using Event free None o Unisolated LOCA in Unisolated Letdown Line LOCA in Letdown Line Mot Evaluated | |||
-- 2.80E-7 0.05 Using Event f ree None o Interfacing Systees Interfacing LOCA Systems LOCA Mot Evaluated | |||
-- 2.70E-7 0.05 Using Event Tsee None o Catastrochte Reactor Cats. Tessel Rupture strophie Reactor vessel Septure I | |||
= | |||
l 1 | |||
Table I (Con't) | |||
Dr1MlW4WT CDitE El.T ACCIDENT 3ROUEWCT3 CONTNff. | |||
FAILED FREO./ $ OF SEQUENCE IIEAT NOTES SOFFORT 3EA8_ TUT AL CSF DESCRIPTION _ NJDfAL INITIANE SYSTEM 5 Avell. 6. TOE-6 1.22 Loss or HCC-5 o Fall to Recover HCC-5, 8tW Avellable. | |||
NCC-5 Primary Intes. Falls Avell, 7. 30E-6 1 33 o Fall to Recover HCC-5 Af W Falls Avell. 4.27E-6 0.78 o NCC-5 secovered AfW Avelletite, Frimary Intes. Falls, Bleed and Feed Fetts, OR l A vell . 4.0fE-7 0.07 Leng Term Cooling Falle 0.29 Total Less or DC t | |||
DCI'DC2e o MCC-5 secovered, Avell. l.5BE-6 la Not Recow. | |||
l HCC-5 Frimary Intes. Fette, SS6 result s in Bleed and Feed Falle Loss of both OR l.59E-6 0.29 se=I-vital and AfW Tells, Avell. vital Instrset n. | |||
Bleed and Food Felle A.9fE-6 0.91 o Fall to Secover Avell. | |||
HCC-5, AFW Avellable, Primary Intes. Falle Ave 11 5.0IE-6 0.98 o Fall to Recover HCC-5, AFW Falls I.62E-5 2.96 Loss of DCleDC2 o Frimary Intes. Felle, Avell. | |||
BC eve i Bleed and Feed Falls OR A vell . 5.0PE-6 0.93 Long Term Cooling i | |||
Falle l | |||
i l | |||
i | |||
4 4 | |||
Table I (Con't) i peten44T Coat teI.T ACCittui Su@m.NCF.5 I. EXMITMIT. | |||
FAILED FREO./ $ OF 389UEWct NEAT | |||
* SUFF00tf TEAR _ 701 AL CW j DE3CRIrfl01f M9CTAL 18111A108 STSTSNB o AFW Falls, Fell Avell . 5 35E-6 0.98 Loss oF None to Recover ifW, Feedwater Bleed and Feed Falls OR Long Term Cooling Avall, 1.46E-6 0.27 l Falls 4 | |||
o Conseq. LOCA - Avell. 3.60E-6 0.66 Branch to ETP), | |||
FORT l.0CA, High Fressure injee. Avell, High Pressure Seelro. | |||
Falls o Fall to Trly ( ATW5) Avell. l.35E-6 0.25 Branch to ET22, Manuel Trly Falls, Presserleer BelleF Falls OR Emergency Doretion Avall. 7 31E-7 0.04 Falls o Low Pressure Injee. Avell. 8.06E-6 l.47 Large More j Avall, l.ow Pressure j Dreek Beelreulation Falls LOCA o Low Pressure Injec. Avell. 3.05E-6 0.56 Falls 1 | |||
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== | |||
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a .J C =00 C e COI 0C 9 m a. m | |||
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30 e a == m e 9 4 c | |||
e | |||
.Fae a - | |||
E pe s | |||
ee s | |||
as - 6 o | |||
. C .e -~,.e. | |||
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eee . | |||
e. | |||
E == G, | |||
: e. e.e es b & S en | |||
: e. . se e G. as m. | |||
es ye 5 a= == K e gal & - a= ae W as E == maC 0hO eec0 0 3 e se .* C eO O ea e == > == 0 m 0 > es e OOCe e me == 0e == > = = 06 0 ad W E e. U 88 em M MMS e.8 ta. E E ed u O a.8 E 4 eJ U M O O O O O sue | |||
*E e . | |||
8 4 | |||
IR. b . | |||
8 E | |||
as | |||
-: I. | |||
e=* ** | |||
96 t | |||
e b | |||
M to eX e a es EXe eeC e633 eee3 OC& sh e | |||
se beu ** e 3 2 3 ebO ** b OS M O De a M o& Es ad MS&S 4 l i i i | |||
I d ' | |||
I I | |||
l | |||
Table I (Con't) | |||
DOMINANT CORE El.T ACCIDENT SEQtENCFJ Miff t9ff. I DF F4 FLED IERT FREO./ llofs3 SequtWCE TEAR T_OTAL M 30FFORT metNAI. | |||
ST3tspel 9_ ESCRIFTICIl_ | |||
lulflat0B_ 5 5cE-7 0.10 o peactor 1rly Falls Avail. | |||
Stees Loop Isolation Avat! | |||
Cenerator Tube BCS 1.evel Rupture Avell. 5.1TE-6 l.05 Recovery A4#ee-None o Charging Falls, sed as Blee1 and ontsolated 3G Cooling Avati, Feed RCF Seal (ifW or AF2 Avall) | |||
Leak RCS Level Recov. Fet! | |||
OR 1.89E-6 0.35 Long Term Cooling Avett. | |||
Fatts Results in Ioss 6.18E-6 1.13 or 30 Coollns o Feedline Isolation Avell. | |||
None tietn Falls, Bleed and Feed!!ne Feed Fetts Break OR 2.93E-7 0.05 Long feesa Cooling A ve ll . | |||
." 7.72E-7 0.14 Feedline Isoletion Avell . | |||
o Avall, 30 Cooling Falls (tF2 + AF3 Fell) i Steed and Feed Fatte 4.82t-4 er., | |||
forAs.: | |||
Table II CORE MEI.T FRtuutJom BT INITIATORS DEPCEf,T CO'TF:n' TION DESEPID IOf! CS' rREOUENCIESND.? | |||
$$1 1.34E-4 24.45 E703 Small-Break LOCA l 1.23E-4 22.u5 ET12 Loss of Offsite Power 8.13E-5 14.84 ET02 Medium-Break LOCA 5.34E-5 9.74 ET09 General Plant Transient 3.36E-5 6.13 ET18 Loss of MCC-5 2.94E-5 5.36 ET20 Loss of DC Bus 1 1.36E-5 2.48 ET10 Total Loss of Main Feedwater 1.22E-5 2.23 ETC1 Large-Break LOCA 1.10E-5 2.01 ET06 Steamline Break Down-stream of NRV | |||
/ | |||
Steam Generator Tube 1.04E-5 1.90 ET04 Rupture 8.45E-6 1.54 ET08 Unisolated RCP Seal Leakage Main Feedline Break 7.48E-6 1.36 ET07 Steamline Break Upstream 7.00E-6 1.28 ET05 j of NRV l | |||
5.93E-6 1.08 ; | |||
ET16 Insufficient Flow of Service Water LOCA Outside Containment 5.45E-6 .99 V, V1 4.46E-6 .81 ET19 Total Loss of DC Power Loss of Control Air 3 40E-6 .62 ET17 2.66E-6 48 ET11 Loss of DC Bus 2 Ei21 Loss of Semi-Vital AC 1.TTE 6 25 TOTAL 5.48E-4 100 I Includes Station Blackout (ET 15), LOSP and Loss of MCC-5 (EI13), LOSP and Loss of one Dnergency Bus (ET14). | |||
1 i | |||
i | |||
e Docket No.'50-213 B12020 Enclosure 2 Haddam Neck Probabilistic Safety Study NUSCO Evaluation of PSS Results I | |||
March,1986 | |||
- __------- ._- -----se__ - _ - - - _ _ _ ----_----__--__-.----_--------.------_--_--------.----------_--------__---------,,aa----,--.-_-_-_, .a_- - _ s-- | |||
During the development of the Haddam Neck Probabilistic Safety Study (PSS), , | |||
Northeast Utilities Service Company (NUSCO) obtained many new engineering | |||
. insights into. the expected performance of the Haddam Neck Plant during | |||
~ | |||
y | |||
: transient situations. Discussions of the major insights developed from the study - | |||
} are presented in this section which serves as the cornerstone for Connecticut - | |||
; Yankee Atomic Power Company's (CYAPCO) efforts -to reduce the overall | |||
[ calculated core melt frequency at the Haddam Neck Plant. | |||
4 l 1.- Plant Changes Implemented as a Result of the PSS L | |||
; 1) Removal of the Interdependence of the Emergency Diesel Generator i on Common Motor Control Center-3 (MCC-5) 3' | |||
!' During efforts to integrate the PSS system reliability models, an - | |||
! interdependency between the emergency diesel generators (ED/G) and y ' | |||
l safety-related MCC-5 was highlighted. . This interdependence was j | |||
i discussed ing{{letter dated|date=November 8, 1985|text=letter dated November 8, 1985}}(1) and was reported as j LER 85-029. l 2 | |||
i Specifically, the air-operated valves (AOV) which allow cooling water to the ED/Gs are operated by solenoid-operated valves (SOV). | |||
: Although the AOVs fall open on loss of air, the SOVs require AC power | |||
! from MCC-5 to operate and op(n the ~AOVs. A loss of offsite power i coincident with a sustained interruption of power .to MCC-5 could , | |||
i cause both diesels to overheat and subsequently fall if operator - | |||
! recognition and action does not take place. However, loss of offsite | |||
] power and MCC-5 does not affect the ability of the diesel generators | |||
; to start up. If one or both diesel generators start up, accept their ; | |||
i emergency loads and properly reenergize MCC-5 as designed, the | |||
! original AOV/ SOY configuration would result ' in long-term 4 diese! | |||
* generator cooling availability. | |||
i Following discussions with the diesel manufacturer, immi diate l corrective actions were implemented to maintain one of the ED/G j AOVs in the open position (air supply removed), monitor the effect on t tube oil temperature of centinuous service water flow. to the jacket j water heat exchangers and perform a service test after tube oil | |||
; temperature equalized (testing was in accordance with the-1- requirements of Technical Specification Section 4-5.a). Subsequently, l the remaining AOV was also modified to permit continuous cooling | |||
; water to the diesel engine. Permanent modifications were planned for the 1986 refueling outage to remove this interdependency and allow | |||
: the AOVs to remain closed during normal plant . operation.- The | |||
] equipment / parts that will allow CYAPCO to install permanent | |||
; modifications to this system are on order 'and are expected to be | |||
! (1) 3. F. Opeka letter to C. I. Grimes, "Probabilistic Safety Study Finding," | |||
l dated November 8,1985. | |||
j (2) Licensee F. tent Report 85-029 for the Haddam Neck Plant, 2 | |||
"Interdepen Ic. ice of Emergency Diesel Generators and Common. Motor | |||
; Control Center," dated December 2,1985. , | |||
l I | |||
delivered during the'latter part of April 1986. Presently, the Haddam L Neck Plant is scheduled to start-up from the 1986 refueling outage during April 1986. As such, the permanent modifications to the | |||
~ | |||
emergency diesel generators may not be fully implemented before the end of the current outage. However, as the planned modifications can | |||
. be implemented with the plant in operation, CYAPCO will install the permanent modifications . upon receipt of the necessary equipment / parts. This information supersedes the schedule for - | |||
!. resolution presented in the previously mentioned LER. In the interim, ! | |||
the -short-term corrective actions (continuous service -water flow to the Jacket water heat exchangers) will remain .in place, ensuring the above-mentioned interdependency will not adversely affect the- ! | |||
emergency diesel generator operation. | |||
At the time of the discovery of this interdependency, all of the PSS l system's models were being integrated to provide an overall i | |||
assessment of core melt frequency. A preliminary assessment of this l particular sequence identified it as the dominant contributor to station | |||
. blackout frequency and a major contributor to core melt frequency. | |||
As a result, the above-mentioned short-term modifications were | |||
: Implemented to remove the interdependency. Thu's the core melt frequency value of J.5 x 10-4/ reactor-year does not include the contribution from this scenario. | |||
? | |||
l' 2. Mitigation of the Effects of the Loss of MCC-5 as a Transient Initiator | |||
! As part of the overall PSS effort, the frequency and consequences of - | |||
o the loss of MCC-5(3) as an initiating event were also considered. | |||
, System reliability calculations were performed using fault tree analysis to estimate the frequency of occurrence of loss of MCC-5. | |||
. Failure modes and effects analyses were performed to determine the | |||
{ consequences of the loss of MCC-5. Walkdowns of the plant with the i operating staff at Haddam Neck identified the means of identifying the transient as well' as possible means of recovering the system | |||
: and/or mitigating the effects of its loss. | |||
! The scenario is initiated by a de-energization of MCC-5 resulting from | |||
; any of a number of causes. A large fraction of the de-energization of | |||
; MCC-5 events (95 to 99 percent) are potentially recoverable. | |||
{ However, as a result of the loss of MCC-5 there is a loss of semi-vital i AC power along with a trip of both main feedwater pumps on sensed j low-suction pressure. Subsequently, the reactor trips on feed flow - | |||
: steam flow mismatch. The loss of MCC-5 also causes the loss of the I | |||
control air compressors, loss of closed cooling water pumps for i compressor cooling and loss of most of the important motor-operated i valves' including safety injection, core deluge, containment spray, i pressurizer relief, loop isolation,: and chemical and volume contro! | |||
. systems valves. | |||
4 i | |||
i | |||
! (3) This component is not required to meet the single failure criterion as | |||
: described in the " United States Atomic Energy Commission Safety Evaluation i by the Division of Reactor Licensing", dated July 1,1971. | |||
i l | |||
3-j l 1 | |||
. The loss of semi-vital AC power, in-addition to causing the loss of.- ' | |||
i important instrumentation such as steam generator pressure and I auxiliary feedwater flow, automatically starts the stand-by charging q pump and isolates letdown makeup to the volume control tank (VCT). : | |||
The charging pump main lube oil pumps stop on loss of MCC-5, although the auxiliary lube oil pumps would .be available. Automatic l switch-over of charging pump suction from the VCT to the refueling i water storage tank (RWST) falls as the VCT level goes down. Within i several minutes of initiation, the VCT is drained. . Without operator j | |||
intervention, both charging pumps would cavitate and be damaged soon , | |||
thereafter. With the loss of MCC-5, many alarms would be received in | |||
: the control room. Because of the eventual loss of control air, all l AOVs would fall in their safe position, but automatic control would be i lost. Auxiliary feedwater would be available because the AOVs would | |||
, fail open, but their position could not be controlled from the control | |||
; room. Likewise, the positive displacement (metering) charging pump j would not be controllable. Component cooling water flow to the j' reactor coolant pump (RCP) thermal barrier would be lost because the AOV falls closed on loss of air or loss of semi-vital AC power. Hence, | |||
) all RCP seal cooling is potentially lost. t On the secondary side, the loss of air results in the eventual closing of the main steam trip i | |||
valves. The turbine bypass valves (steam dump to condenser) as well l as the atmospheric steam dump valve would not be available. | |||
i | |||
: Preliminary evaluation of the above scenario identified this transient | |||
{ as one of the dominant contributors to the core melt frequency. As a 1 result, a project assignment using " emergency". work-order procedures I was initiated to effect plant design changes necessary to trip the j charging pumps on loss of MCC-5 (specifically loss of semi-vital AC power). Manual realignment of charging suction to the RWST would j ensure reactor coolant system make-up capability during this scenario. | |||
The PSS identified possible means of mitigating the accident including l the potential for recovering MCC-5, and the cross-tie of service air to | |||
; control air along with compressor cooling using the well water system. | |||
} With air restored, auto-control of the charging metering pump as well 1 | |||
as control room operation 'of auxiliary feedwater flow would be ' | |||
] available. The draft symptom-based Emergency Response Guidelines are also undergoing evaluation to incorporate means of identification | |||
; and mitigation of this transient. | |||
1 l As -of this writing, the above described plant modifications and j associated procedure changes are in the process of being implemented. | |||
! Our objective is 'to complete this effort prior to plant start-up. | |||
3 However, depending upon the actual date of plant start-up, completion of the work may occur early during the upcoming cycle. Since their work will be completed either prior to start-up or shortly thereaf ter, i we have elected to credit these modifications in the PSS calculations. | |||
1 I The calculated core melt frequency of 5.5 x 10-4 per reactor-year j does include some residual contribution from this accident scenario, i but the plant design / procedural changes described above have removed l most of the contribution of this scenario. | |||
I i | |||
1.,.-.... _ , _ . _ _ . - . .. . _ ,,._ _ _ _ _ _ -, - . - , , . _ - _ _ _ _ , _ - - , , _ . _ . - | |||
~ | |||
i II. Engineering Insights Obtained from the PSS | |||
: 1) Adverse Break Size and Location in the Reactor Coolant System (RCS) 1 | |||
!- During performance of the "Best Estimate LOCA Analysis" (NUSCO- . | |||
f 150), a small range of break sizes in loop 2 cold leg of the RCS was l identified for which safety injection flow in the high-pressure | |||
; recirculation mode may be insufficient to prevent core uncevery in the absence of modifications to facility operating procedures and/or J | |||
system (e.g., valve) realignment. The specific break sizes and locations involved are breaks of 0.045 ft.2 to 0.02 ft.2 in the cold leg pump discharge line for RCS Loop 2 and breaks less than 0.045 ft.2 in i the Loop 2 charging line and connected piping. | |||
! For Loop 2 breaks greater than or equal to 0.045 ft.2 (equivalent | |||
! diameter = 2.9 incheCthe RCS, with all accident mitigation pumps | |||
; activated in the injection mode, will depressurize to 165 psia (150 psig) and below prior to,)pitiation of core uncovery and also prior -to the depletion of 100,W0 gallons of RWST inventory. Low-pressure recirculation using the residual heat removal (RHR) pumps would follow. | |||
For Loop 2 cold leg breaks less than or equal to 0.02 ft.2 (equivalent t i | |||
diameter = 1.9 inches), one charging ' pump and ' only one of two charging flow control valves is sufficient in the high pressure recirculation mode to prevent core -uncovery. (The high pressure recirculation mode commences by' aligning the charging pump suction - , | |||
i to the discharge of the RHR pumps when RCS pressure is above 165 psia. The sole delivery point is to the Loop 2 cold leg.) | |||
i Breaks located directly in the charging line or connected piping i downstream of the check valves for sizes 0.045 ft.2 and smaller may l also produce results analogous to the Loop 2 cold leg breaks discussed i | |||
! above since for these breaks the RCS pressure also remains above 165 psia while the charging flow provides insufficient flow to prevent | |||
; core uncovery. | |||
t | |||
) The frequency of these breaks was estimated in the PSS by multiplying i the (generic) frequency of various size LOCAs times the conditional j probability that a break of a given size occurred in the adverse location. This conditional probability was derived by totalling up the j number of pipe segments in the affected location and dividing by the l total number of pertinent pipe sections ' of the given size range i throughout the RCS. ' A pipe segment is defined as a section of pipe | |||
{ between major discontinuties (e.g., valves, reducers, etc). In the PSS, | |||
; approximately 9% of the medium break LOCAs and 1% of the small i break LOCAs are estimated to occur in adverse locations. The definition of a pipe segment and the ' methodology used for quantification is consistent with WASH-1400. | |||
! As a result of this discovery, evaluation of the feasibility of utilizing 1 additional sources of borated water to extend the injection phase and/or using other available injection paths, such as the loop-f.ill i | |||
i | |||
4 . | |||
+ | |||
header or auxiliary preswrizer spray, are being evaluated. 'This issue | |||
; is being addressed in separate correspondence. | |||
l 2) Best Estimate LOCA Analysis | |||
! In support of the PSS, plant-specific thermal-hydraulic analyses were performed for a wide spectrum of potential core uncovery events. The results are summarized in the " Connecticut Yankee (Haddam Neck Plant) Best Estimate LOCA Analysis," . NUSCO-150. The . major findings of the analysis are summarized as follows: | |||
! o For large breaks (area > 0.2 ft.2), success of emergency- core cooling in the injection phase requires only one low-pressure | |||
! safety injection pump providing flow through one core deluge , | |||
; valve. | |||
I o For medium breaks (0.2 ft.2 2 area 20.02 ft.2), success requires only one high-pressure safety injection (HPSI) pump providing i flow to 3 of 3 unfaulted loops (or 2 HPSI pumps to 2 of 3 unfaulted loops). | |||
l l o For medium-small breaks (0.02 ft.22 area 3 0.003 ft.2), success .t requires only one charging ' pump or one HPSI pump for j | |||
; mitigation. ; | |||
o For small breaks (area < 0.003 ft.2), success requires steam generator heat removal capability in addition to either one HPSI pump or one charging pump for mitigation. | |||
i o " Feed-and-bleed" cooling requires one charging pump and 4 one pressurizer PORV, or one HPSI' pump and two PORVs. | |||
l Initiation of feed-and-bleed is required approximately 40 minutes following the loss of all steam generator feedwater. | |||
o For station blackout, core damage would not occur for over i eight hours, even assuming catastrophic failure of all reactor j coolant pump seals. i i | |||
! 3) Additional Findings The following additional insights are also of interest: | |||
i i o There are eight potential paths for high pressure / low pressure l Interfacing systems LOCA; however, the overall contribution to | |||
; core melt frequency is low, i | |||
j o Pipe ruptures in the letdown line' outside containment but upstream of the flow control orifices, although relatively low in | |||
: frequency, have only one potential means of isolation (LD-MOV- ! | |||
j 200). Failure of the MOV to close either automatically or. | |||
manually could potentially result in an unisolated LOCA outside containment; the sequence could possibly be risk dominant. As part of the Integrated Safety Assessment Program (ISAP), | |||
l; l s. | |||
i | |||
1 - . . _ _ . _ ._ . . _ _ _ .- | |||
,s , | |||
i evaluations are underway to determine whether a proposed design change to add an additional containment isolation valve outside containment for other reasons might be more appropriate if relocated inside containment. | |||
o A reliability analysis of the auxiliary feedwater system has found- | |||
: i. the unavailability of automatic initiation of the system to be 3 | |||
dominated by failures of air-operated (and particularly solenoid-operated) valves. Changes to periodic' testing frequencies of the valves are under consideration. In addition, a ' potentially , | |||
significant common cause failure of the system due to the | |||
~ | |||
nonrestoration of manual valves following . preventive , | |||
maintenance has been identified and is under evaluation. | |||
o Failure of SW-MOV-5 and 6 to open on demand are found to be dominant contributors to RHR system unavailability. In the LOCA recirculation mode of core cooling. | |||
. o Based on current procedures, the testing of the main steam trip , | |||
valve transmitters was found not to be systematic in that testing ' | |||
. of all four channels is not assured. | |||
o The fans for cooling the charging pump lube oil coolers offer a ; | |||
potential means of providing cooling independent of component cooling water; however,'no_ information on the design adequacy i | |||
has been found nor. are there any testing or maintenance procedures for the fans. | |||
1 o The diesel-driven fire pump provides an AC independent means 2 | |||
of containment spray during station blackout should containment | |||
! heat removal be necessary; however, a single motor-operated 4 | |||
valve (MOV-31) outside containment would have to be opened | |||
, manually. This feature of an AC-independent containment spray pump is important in minimizing the offsite public consequences resulting from core melt due to station blackout. | |||
o On a best estimate basis, the RCS loop isolation valves have been found to provide a reasonable means of mitigating the | |||
; effects of steam generator tube ruptures. | |||
I o Due to the diversity and redundancy of containment heat removal features (four fan coolers, 'six pumps for spray),- the overall availability of containment heat removal systems for mitigating the impact of core darnage accidents was found to be | |||
; exceptional. | |||
o The overall reliability. of main .feedwater/ auxiliary feedwater post-trip (non-loss of feedwater events) was likewise found to be i exceptional. | |||
> III. Discussion of RCS Piping Ductility i | |||
The PSS results are based on LOCA frequencies derived from the 6 experience of pressurized water reactors, as quantified in WASH-1400. " | |||
1 | |||
, I i | |||
The actual LOCA frequencies at Haddam Neck are expected to be lower than the values used in the PSS since the piping material is austenitic stainless steel while the PWR population includes plants which use ferritic carbon steel. Stainless steel provides superior ductility and toughness and is not susceptible to brittle fracture at low temperatures. Thus, the PSS results include an unquantified conservatism in terms of the actual LOCA frequencies. | |||
IV. Discussion of Haddam Neck Plant-Specific Simulator Experience The plant-specific control room simulator for the Haddam Neck Plant became available during the last month of the study. Additional benefits will be realized in the future as the results of the PSS are incorporated into the training of operators on the simulator. | |||
V. Future NUSCO Actions As stated previously, NUSCO is currently reviewing the results of the PSS to determine the best altr natives available to reduce the core melt frequency at the Haddam IV . Plant. Completion of these analyses will result in either a clarificatt J the system success criteria, identification of weaknesses or recommerd aons fo possible hardware and/or emergency operating procedure changes When these are completed we will requantify the Haddam Neck PSS models reflecting the net effects of these changes. | |||
As new issues are identified during our review of the PSS, we plan to either: | |||
a) Implement a project assignment corresponding to the identified issue, if the . issue is deemed to have a high level of importance to plant safety, or , | |||
I b) Evaluate the issue and any resultant projects within the framework of l the ISAP, and implement modifications to the plant / procedures on a schedule commensurate with the results of the ISAP evaluation. | |||
4 DISTRIBUTION PER MIKE BOYLE LTR ENCL REG FILE 1 1 EBE McCABE REGION I 3 3 RRAB 2 2 ERNST MALCOLM RESEARCH 2 2 AEOD 1 1 j ICE 1 1 NRC PDR 1 1 LPDR 1 1 NSIC 1 1 24X 1 1 14 14 SEND ALL EXTRA COPIES TO: MIKE BOYLE | |||
i DISCLAIMER The information contained in this topical report was prepared for the specific requirements of Northeast Utilities Service Company (NUSCO) and its affiliated companies, and may contain materials subject to privately owned rights. Any use of all or any portion of the information, analyses, methodology or data contained in this topical report by third parties shall be undertaken at such party's sole risk. NUSCO and its affiliated companies hereby disclaim any liability (including but not limited to tort, contract, statute, or course of dealing) or warranty (whether express or implied) for the accuracy, completeness, suitability for a particular purpose or merchantability of the information. | |||
/""~%. | |||
'd l | |||
l l | |||
I O | |||
g *. -- | |||
d A m, | |||
* i CONNECTICUT YANKEE PROBABILISTIC SAFETY STUDY}} |
Latest revision as of 18:25, 10 April 2022
ML20155E741 | |
Person / Time | |
---|---|
Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
Issue date: | 03/31/1986 |
From: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO. |
To: | Charemagne Grimes Office of Nuclear Reactor Regulation |
Shared Package | |
ML20155E746 | List: |
References | |
REF-GTECI-A-44, REF-GTECI-EL, TASK-A-44, TASK-OR B12020, NUDOCS 8604180083 | |
Download: ML20155E741 (26) | |
Text
'
t.
i .
CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N. CONNECTICUT P.o. box 27o HARTFoRo. CONNECTICUT 06141-0270 TELEPHONE 2os-ess sooo March 31,1986 Docket No. 50-213 B12020 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Haddam Neck Plant Probabilistic Safety Study - Summary Report and Results Northeast Utilities Service Company (NUSCO), on behalf of Connecticut Yankee Atomic Power Company (CYAPCO), has recently completed a plant-specific Probabilistic Safety Study (PSS) including a Best Estimate LOCA Analysis for the Haddam Neck Plant. The studies have been and will be utilized for many purposes. Among the near-term applications for the PSS is for assistance in making backfitting decisions in the context of the Integrated Safety Assessment Program (ISAP) currently being implemented at the Haddam Neck Plant.
Background
During the past several years, NUSCO has been implementing a Living PRA Frogram for the nuclear plants within the Northeast Utilities (NU) system. The major element of this program is the development, maintenance and use of PRA models, of each of the system's nuclear power plants, for assistance in evaluating potential plant backfits and operating procedures modifications. The Living PRA Program affords us the flexibility to quickly and accurately analyze the impact on plant safety of changes to the plant's design configuration and significant changes to operating procedures.
NUSCO has completed the development of a computerized PRA model of the Haddam Neck Plant. This model will be periodically updated to incorporate plant design changes, significant operational changes and relevant updated equipment performance data.
The Haddam Neck PSS utilized current state-of-the-art computational and modeling techniques. Included in the study were a comprehensive analysis of best estimate plant specific LOCA and transient response, Haddam Neck specific operating, reliability and maintenance data, the latest plant design 8604180083 860331 3 PDR ADOCK 0500 P
gg
'lI9 4
changes and draft symptom-oriented . Emergency Operating Procedures.. The
, scope of the PSS also . included support system failure initiated events ,
l interfacing system LOCA events and fire initiated events.(1)
As noted in SECY-85-160,(2) theLintegrated assessment of the Haddam Neck j Plant is scheduled to be completed during 1986. The schedule was based, in part, i on our schedule for completion of and documentation of a summary report of the Haddam Neck PSS .during March 1986. In keeping with the schedule for completion of the integrated assessment, our submittal of the.Haddam Neck PSS Summary Report and Best Estimate LOCA Analysis will enable the Staff to begin l the probabilistic safety analysis review phase of the integrated assessment of
+
Haddam Neck. !
i 1 j Results It is Northeast Utilities Corporate Policy to provide dependable and economic power to its customers, without endangering the health and safety of the public.
As one element of achieving this policy, NU has set forth quantitative safety-
, goals for the design and operation of our nuclear power plants. If it is l determined that any one of our nuclear power plants does not meet the
, acceptance criteria as established by our safety goals, our policy is to take the corrective action necessary to meet the criteria, on a schedule commensurate with the level of deficiency found.
The Haddam Neck PSS calculated a mean core melt frequency of 5.5 x 10-4 per l reactor-year for the Haddam Neck Plant which exceeds one of our safety goals. l As a result, we have committed resources to work to decrease the overall core l melt frequency at the Haddam Neck Plant. We have begun evaluating some of
- the engineering insights obtained from the PSS, in order to determine .the alternatives available to reduce the calculated core melt frequency. A summary- .
! of the engineering insights obtained from the PSS and items we are evaluating to reduce the calculated core melt risk at the Haddam Neck Plant are contained in
, Enclosure 2.
! As outlined in Enclosures I and 2, the PSS and Best Estimate LOCA Analysis i
yielded many insights into the operation of the plant including:
o Containment heat removal is available for most core melt sequences. As the availability of containment heat removal is--' considered to be an
~
effective means of maintaining containment integrity and removing fission ~
products to prevent large scale radioactive releases if core melt occurs, .
the public risk impact of a majority of the dominant core me't se_quences is reduced by having containment heat removal available. '
(1) The fire initiated events section of the PSS has not been completed at the present time. NUSCO will document and docket the updated results'to the
, Staff following completion of the analysis. .
(2) SECY-85-160, " Integrated Safety Assessment Program - Implementation Plan," dated May 6,1985.
i
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3-4 o . An AC independent means of providing containment spray during a station blackout is provided by- the diesel-driven fire pump. This feature is important in minimizing the offsite public consequences resulting from a
, core melt accident due to station blackout, o Approximately 40% of the calculated core melt frequency at Haddam Neck ' .
is attributable to small and medium break LOCA scenarios. The PSS
. utilized LOCA frequencies which were derived from generic PWR data as presented in WASH-1400. As the Haddam Neck Plant utilizes austenitic stainless steel which has superior ductility toughness .and resistance to l brittle fracture compared to ferritic carbon steel used in many PWR-plants, the actual Haddam Neck Plant LOCA frequency is expected to be lower.
o The Haddam Neck . plant-specific control room simulator has -been completed and is currently being used for operator training. This affords i us the opportunity to improve the training of operators to respond to plant transient situations. l i In light of the above, and our continuing efforts to identify, evaluate and when necessary implement procedural or hardware modifications to upgrade the operation of the plant, CYAPCO has concluded that continued operation of the Haddam Neck Plant does not pose any undue risk to the public.
- Summary Report ,
I l
- NUSCO has completed summary reports of the Haddam Neck PSS and Best . l Estimate to this letter.13 LOCA) Analysis which we are providing to the Staff as an attachment The PSS summary report is a comprehensive summary of the core melt frequencies for various sequences calculated in the ; study and the ,
calculational methodologies and analysis techniques utilized in the development !
of the PRA model. The PSS summary report contains the items outlined below:
4 o Determination of Initiating Events j i -
System investigations Initiator frequency calculations o Accident Sequence Analysis Classification of event sequence outcomes Plant systems event tree models Plant support system event tree models
'; o Plant Systems Reliability Analysis Plant component reliability data collection and analysis l Plant systems reliability modeling. I I
(
(3) The summary reports consist of the Haddam Neck Probabilistic Safety Study (Volumes 1-4) and the Best Estimate LOCA Analysis (one volume) for I the Maddam Neck Plant.
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-4 o Human Reliability Analysis
- Introduction and methodology
- Screening analysis
- Detailed representation of operator action -
- Summary of results o Accident Sequence Quantification -
- Matrix quantification
- Core melt accident sequence quantification results
- Containment heat removal reliability consideration ,
In support of the PSS, a Best Estimate.LOCA Analysis was performed concurrent-with the development of the PSS models. ~ The Best Estimate LOCA Analysis ,
summary report describes the results of comprehensive plant specific accident-analyses used to determine best estimate system success criteria, best estimate plant response to failed equipment and components, and time frames for operator recovery of failed systems. These calculations were performed using the NULAP 5 Code with best estimate input values. The Best Estimate LOCA -
Analysis contains analyses of the scenarios described below:
o Station Blackout A
- o incore Instrument Tube Rupture i o Steam Generator Tube Rupture 1 o Large Break LOCA 4 o Medium Break LOCA o Small Break LOCA o Total Loss of Main and Auxiliary Feedwater with Feed and Bleed Cooling o Anticipated Transient Without Scram 3
o Total Loss of DC Power In accordance with your request we are providing 26 copies of the PSS and Best-Estimate LOCA Analysis to the ISAP Project Directorate for distribution within ,
the NRC (including NRR, Region 1, ACRS, etc.).
i
! In addition to the summary reports, we are providing the Staff with additional mformation on the scope and results of the PSS as outlined below, o Enclosure 1 A summary of the results of the PSS is provided herein.
o Enclosure 2 1 l
Discussions of the major engineering insights NUSCO has obtained from the PSS including the following two plant design changes already implemented at the plant are provided herein.
o Plant Design change to eliminate the common dependence of i emergency diesel generator cooling on Motor Control Center-5 (MCC-5). This dependence, found while doing the analysis, could
, be significant during a loss of offsite power event as the diesel i generators are dependent on MCC-5 to maintain diesel generator cooling. However, the diesel generators are not dependent upon i
o
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_3_
MCC-5 to start up following a loss of offsite power event. If one or both diesel generators start-up, accept their emergency loads -
and properly reenergize MCC-5 ,as designed, the original configuration would result in long-term diesel generator cooling .
availability.
I o Plant design change to mitigate the effects of the loss of MCC-5 e
(as an initiator) by tripping the charging pumps.
These design changes were credited in our quantification of the PSS as a result of our preliminary evaluations of the PSS models which highlighted i these design conditions as being potential significant contributors to the j total core melt frequency at the plant.
Upcoming Activities In order to facili. tate Staff. review and understanding of the Haddam Neck PSS and to ensure that maximum utilization of the PSS is achieved, in ISAP and other applications, we are willing to meet with your Staff to review the PSS. Our efforts to date have focused on completing and documenting this study, and on a
- limited number of high priority issues which were identified during the conduct of the PSS. In the near term, we plan to complete and document our analysis of fire initiated events and to evaluate the report in more detail to identify
, additional issues which warrant further study. Consistent with the overall ISAP
! framework, these issues will be identified, new ISAP topics will be proposed and i they will be evaluated in a manner consistent with the process in place for all
- other ISAP topics.
I Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY -
1F Cd J. F. Opeka U Senior Vice President cc: T. E. Murley, Region I (with Enclosures 1 and 2 only) 1 1
l
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Docket No. 50-213 B12020 i
i i
1- -
i Enclosure 1 i
. Summary of the Results of the Haddam Neck Probabilistic Safety Study-1
-1 l.
1 i
i S.
1 i ;
I i
j 1
March,1986 -
i Summary of Results The Haddam Neck PSS calculated a mean value core melt. frequency of 5.5 x -
j 10-4 per reactor-year. The core melt frequency is based on the internal-
- initiators examined in the study which include anticipated transients, special initiators, as well as hypothetical design basis accidents. A summary of the
- dominant core melt sequences is provided in Table I.
The dominant core melt sequences involve a loss of high pressure recirculation during small-break and medium-break LOCAs. The high pressure recirculation system at Haddam Neck utilizes the charging system in conjunction with the residual heat removal system. 'The high pressure safety injection system is not currently used for high pressure recirculation. The charging system has only one injection path designed for use during recirculation, into Loop 2 cold leg.
j Failure of the charging system results in loss of high pressure recirculation.
t l' In addition to loss of high pressure recirculation due to random component failures, the Haddam Neck PSS considered the loss of recirculation due to a break in the charging injection path. If a small- or medium-break LOCA occurs
, in Loop 2 cold leg, connecting piping or in the charging line downstream of the check valves and is sized sufficiently to degrade the charging flow rate but not i
I depressurize function and we thehave system, high pressure assumed its failurerecirculation will(' l)fail in this analysis. Theto loss perform its -
of high pressure recirculation either due to component failure or an adverse break size l and location, has been identified as an important contributor to core melt
- frequency (14% of total core melt frequency).
i In addition, sequences caused by the loss of offsite power and/or Motor Control Center-5 are also found to be significant. Under general plant transients, consequential LOCAs resulting from an unisolated, stuck-open PORV are important contributors.
, The remaining core melt sequences listed in Table I show that, besides those
- sequences already mentioned, no other single accident sequence dominates core i melt frequency. The table also shows that containment heat removalis available l for most of the core melt sequences. For the station blackout sequence, the i availability of the diesel-driven fire pump for containment spray is an important consideration. The availability of containment heat removal is considered in the analysis as a means of maintaining containment integrity and as an active means i of removing fission products to prevent large scale radioactive releases should core melt occur. The public risk impact of the dominant core melt sequences is -
i reduced by having containment heat removal available.
A listing of the various initiating events and their contributions to the core melt l frequency is presented in Table II. Small-break LOCA and loss of offsite power i events are the largest contributors, representing 24.45 % and 22.45 %
respectively, of the calculated core melt frequency.
i i
(1) The implications of this situation with respect to 10CFR50.72 and 50.73 are being evaluated separately and are the subject of separate correspondence.
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Table I (Con't)
DON!NANT CORE del.T ACCllENT SEQUt NCES CONTP9FT.
FAII.ED ltEAT FPEO./ 3 0F Stirr0RT SEQUENCE WOTRS REMITAL TEAR _ TOTAL CW SISTstt5 DESCRIFTION 195T18708 S t uck Dr e n, 3.18E-5 2.15 4 o Reactor Trly Falle Avall. Unisolat e s ronv/
General Branch to ET22 SRV durtra AIW3 j
Flant Hanual Trip Falls, Transient Fressuriser Beller Falls OR 0 37 Emergency poration Avall. 2.02E-6 Falls OR 0.21 A vall . 3.55E-6 Hain Ieedwater Unavall, AFW Falls 3.55E-5 2.50 ET26 1s 14entl.
LDSte o Bus 4 or 7 Avall or Avell. cat to EI2) See Loss or Falls, Conseq. ISCL ET23 (DCleEAC8 Offsite
+DC2*EAC9) Branch to E126, Seal Fower Fallura LOCA, Rc5 Depress. Falla, OR HFSI Injeo. Falla Avell. l.0$E-5 f 92 OR 0.44 2.40E-6 1 Low Pressure Seelro. Avell.
Falls I
o AFW Falls, Fall to Avell. 2.45E-6 0.45 Decover of fstte Power, Bleed and Feed Falle OR Avall. 9.02E-7 0.17 Long Term Cooling Falls a) Ava ll . l.02E-5 5.86 All o AFW Falls 5.85E-7 0.11 Support Recover Power Win. b)Unavall.
Systees 30 Hinutes,liigh IComb. of Fressure injection 2 Losses of Falls DC. AC, j Serv. Water 4
l l .
Table I (Con't)
TAM E 6.3-5 DONlWANT CORE El.T ACCIDENT SEQUENCFJ IXNITittf.
FAILED nEar FsEo.i s Or Note 3 surr0nt stauEnCe 1UTM. CW BE8DfAL _TEAa_
SISTEMS_ BE3CRIPTICII IslTIA10e 8.54E-6 1.56 o AFW Falla a) Avall. II 86E-7 0.09
' Loss or Fall to Recover b) Unavail.
Offslte Fower within 30 Min.
Fower Recover Fower within 100 Min.
Illgh Fressure i
Injection Falls 9 05E-6 1.65 .
o AFW Avallable a) Avell. 5. lie-7 0,09 Fall to Recover b) Unavail.
within 15 Minutes, Recover Fower f
' within 8 Mours, High Pressure Injection Falla Avell 5.78E-7 0.11 o AFW Avellable Fall to pecover Power within 15 Min, Fall to Recover within 8 Moura Onantified usins 2.00E-5 3.65 results of IACge o AFW Avellable, Fall Avall.
to Recover Fower Section 4.3 17 EAC9 (Station within 15 Min, Fall Slackout) to Secover Power within 8 Houra
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Table I (Con't)
DOHtWANT CORE E LT ACCIDENT SEQUENCES CowitelT.
FAILED nEAr r no./ : Or StoUENCe NOTE 3
! SUrPott REPENAL TEAR _ TUTAL CW I SISTEMS DE3CalFit0N 1Ntf tkt0E_ 0.5) ET26 is ident.
Avalt. 2.89E-6 to El?), See o Csnseq. LOCA -
Loss of Branch to ET23, ET 2)
Offstte Seal Fatture LOCA.
Power RCS Depres. Falls
- OR 0.17 Avall. 9.1)E-7 t.ow Pressure Rectro.Fatts
- OR 0.09 Avall. 4.69E-7 High Pressure -
Injection Fatts Avalt. 7.69E-6 1.40 LOSF'HCC5 o Fett to pecover (DC1 EACS. MCC-5 AFW Avell.,
DC2*EAC9) Frteary Integrity Fatts Avatt 3.80E-6 0.69 o Fett to Secover NCC-5, AFW Fatts 0.99 Mot Evalunta4
-- 4.90E-6 Using Event free None o Unisolated LOCA in Unisolated Letdown Line LOCA in Letdown Line Mot Evaluated
-- 2.80E-7 0.05 Using Event f ree None o Interfacing Systees Interfacing LOCA Systems LOCA Mot Evaluated
-- 2.70E-7 0.05 Using Event Tsee None o Catastrochte Reactor Cats. Tessel Rupture strophie Reactor vessel Septure I
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Table I (Con't)
Dr1MlW4WT CDitE El.T ACCIDENT 3ROUEWCT3 CONTNff.
FAILED FREO./ $ OF SEQUENCE IIEAT NOTES SOFFORT 3EA8_ TUT AL CSF DESCRIPTION _ NJDfAL INITIANE SYSTEM 5 Avell. 6. TOE-6 1.22 Loss or HCC-5 o Fall to Recover HCC-5, 8tW Avellable.
NCC-5 Primary Intes. Falls Avell, 7. 30E-6 1 33 o Fall to Recover HCC-5 Af W Falls Avell. 4.27E-6 0.78 o NCC-5 secovered AfW Avelletite, Frimary Intes. Falls, Bleed and Feed Fetts, OR l A vell . 4.0fE-7 0.07 Leng Term Cooling Falle 0.29 Total Less or DC t
DCI'DC2e o MCC-5 secovered, Avell. l.5BE-6 la Not Recow.
l HCC-5 Frimary Intes. Fette, SS6 result s in Bleed and Feed Falle Loss of both OR l.59E-6 0.29 se=I-vital and AfW Tells, Avell. vital Instrset n.
Bleed and Food Felle A.9fE-6 0.91 o Fall to Secover Avell.
HCC-5, AFW Avellable, Primary Intes. Falle Ave 11 5.0IE-6 0.98 o Fall to Recover HCC-5, AFW Falls I.62E-5 2.96 Loss of DCleDC2 o Frimary Intes. Felle, Avell.
BC eve i Bleed and Feed Falls OR A vell . 5.0PE-6 0.93 Long Term Cooling i
Falle l
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Table I (Con't) i peten44T Coat teI.T ACCittui Su@m.NCF.5 I. EXMITMIT.
FAILED FREO./ $ OF 389UEWct NEAT
- SUFF00tf TEAR _ 701 AL CW j DE3CRIrfl01f M9CTAL 18111A108 STSTSNB o AFW Falls, Fell Avell . 5 35E-6 0.98 Loss oF None to Recover ifW, Feedwater Bleed and Feed Falls OR Long Term Cooling Avall, 1.46E-6 0.27 l Falls 4
o Conseq. LOCA - Avell. 3.60E-6 0.66 Branch to ETP),
FORT l.0CA, High Fressure injee. Avell, High Pressure Seelro.
Falls o Fall to Trly ( ATW5) Avell. l.35E-6 0.25 Branch to ET22, Manuel Trly Falls, Presserleer BelleF Falls OR Emergency Doretion Avall. 7 31E-7 0.04 Falls o Low Pressure Injee. Avell. 8.06E-6 l.47 Large More j Avall, l.ow Pressure j Dreek Beelreulation Falls LOCA o Low Pressure Injec. Avell. 3.05E-6 0.56 Falls 1
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Table I (Con't)
DOMINANT CORE El.T ACCIDENT SEQtENCFJ Miff t9ff. I DF F4 FLED IERT FREO./ llofs3 SequtWCE TEAR T_OTAL M 30FFORT metNAI.
ST3tspel 9_ ESCRIFTICIl_
lulflat0B_ 5 5cE-7 0.10 o peactor 1rly Falls Avail.
Stees Loop Isolation Avat!
Cenerator Tube BCS 1.evel Rupture Avell. 5.1TE-6 l.05 Recovery A4#ee-None o Charging Falls, sed as Blee1 and ontsolated 3G Cooling Avati, Feed RCF Seal (ifW or AF2 Avall)
Leak RCS Level Recov. Fet!
OR 1.89E-6 0.35 Long Term Cooling Avett.
Fatts Results in Ioss 6.18E-6 1.13 or 30 Coollns o Feedline Isolation Avell.
None tietn Falls, Bleed and Feed!!ne Feed Fetts Break OR 2.93E-7 0.05 Long feesa Cooling A ve ll .
." 7.72E-7 0.14 Feedline Isoletion Avell .
o Avall, 30 Cooling Falls (tF2 + AF3 Fell) i Steed and Feed Fatte 4.82t-4 er.,
forAs.:
Table II CORE MEI.T FRtuutJom BT INITIATORS DEPCEf,T CO'TF:n' TION DESEPID IOf! CS' rREOUENCIESND.?
$$1 1.34E-4 24.45 E703 Small-Break LOCA l 1.23E-4 22.u5 ET12 Loss of Offsite Power 8.13E-5 14.84 ET02 Medium-Break LOCA 5.34E-5 9.74 ET09 General Plant Transient 3.36E-5 6.13 ET18 Loss of MCC-5 2.94E-5 5.36 ET20 Loss of DC Bus 1 1.36E-5 2.48 ET10 Total Loss of Main Feedwater 1.22E-5 2.23 ETC1 Large-Break LOCA 1.10E-5 2.01 ET06 Steamline Break Down-stream of NRV
/
Steam Generator Tube 1.04E-5 1.90 ET04 Rupture 8.45E-6 1.54 ET08 Unisolated RCP Seal Leakage Main Feedline Break 7.48E-6 1.36 ET07 Steamline Break Upstream 7.00E-6 1.28 ET05 j of NRV l
5.93E-6 1.08 ;
ET16 Insufficient Flow of Service Water LOCA Outside Containment 5.45E-6 .99 V, V1 4.46E-6 .81 ET19 Total Loss of DC Power Loss of Control Air 3 40E-6 .62 ET17 2.66E-6 48 ET11 Loss of DC Bus 2 Ei21 Loss of Semi-Vital AC 1.TTE 6 25 TOTAL 5.48E-4 100 I Includes Station Blackout (ET 15), LOSP and Loss of MCC-5 (EI13), LOSP and Loss of one Dnergency Bus (ET14).
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i
e Docket No.'50-213 B12020 Enclosure 2 Haddam Neck Probabilistic Safety Study NUSCO Evaluation of PSS Results I
March,1986
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During the development of the Haddam Neck Probabilistic Safety Study (PSS), ,
Northeast Utilities Service Company (NUSCO) obtained many new engineering
. insights into. the expected performance of the Haddam Neck Plant during
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- transient situations. Discussions of the major insights developed from the study -
} are presented in this section which serves as the cornerstone for Connecticut -
- Yankee Atomic Power Company's (CYAPCO) efforts -to reduce the overall
[ calculated core melt frequency at the Haddam Neck Plant.
4 l 1.- Plant Changes Implemented as a Result of the PSS L
- 1) Removal of the Interdependence of the Emergency Diesel Generator i on Common Motor Control Center-3 (MCC-5) 3'
!' During efforts to integrate the PSS system reliability models, an -
! interdependency between the emergency diesel generators (ED/G) and y '
l safety-related MCC-5 was highlighted. . This interdependence was j
i discussed ingletter dated November 8, 1985(1) and was reported as j LER 85-029. l 2
i Specifically, the air-operated valves (AOV) which allow cooling water to the ED/Gs are operated by solenoid-operated valves (SOV).
! from MCC-5 to operate and op(n the ~AOVs. A loss of offsite power i coincident with a sustained interruption of power .to MCC-5 could ,
i cause both diesels to overheat and subsequently fall if operator -
! recognition and action does not take place. However, loss of offsite
] power and MCC-5 does not affect the ability of the diesel generators
- to start up. If one or both diesel generators start up, accept their ;
i emergency loads and properly reenergize MCC-5 as designed, the
! original AOV/ SOY configuration would result ' in long-term 4 diese!
- generator cooling availability.
i Following discussions with the diesel manufacturer, immi diate l corrective actions were implemented to maintain one of the ED/G j AOVs in the open position (air supply removed), monitor the effect on t tube oil temperature of centinuous service water flow. to the jacket j water heat exchangers and perform a service test after tube oil
- temperature equalized (testing was in accordance with the-1- requirements of Technical Specification Section 4-5.a). Subsequently, l the remaining AOV was also modified to permit continuous cooling
- water to the diesel engine. Permanent modifications were planned for the 1986 refueling outage to remove this interdependency and allow
- the AOVs to remain closed during normal plant . operation.- The
] equipment / parts that will allow CYAPCO to install permanent
- modifications to this system are on order 'and are expected to be
! (1) 3. F. Opeka letter to C. I. Grimes, "Probabilistic Safety Study Finding,"
l dated November 8,1985.
j (2) Licensee F. tent Report 85-029 for the Haddam Neck Plant, 2
"Interdepen Ic. ice of Emergency Diesel Generators and Common. Motor
- Control Center," dated December 2,1985. ,
l I
delivered during the'latter part of April 1986. Presently, the Haddam L Neck Plant is scheduled to start-up from the 1986 refueling outage during April 1986. As such, the permanent modifications to the
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emergency diesel generators may not be fully implemented before the end of the current outage. However, as the planned modifications can
. be implemented with the plant in operation, CYAPCO will install the permanent modifications . upon receipt of the necessary equipment / parts. This information supersedes the schedule for -
!. resolution presented in the previously mentioned LER. In the interim, !
the -short-term corrective actions (continuous service -water flow to the Jacket water heat exchangers) will remain .in place, ensuring the above-mentioned interdependency will not adversely affect the- !
emergency diesel generator operation.
At the time of the discovery of this interdependency, all of the PSS l system's models were being integrated to provide an overall i
assessment of core melt frequency. A preliminary assessment of this l particular sequence identified it as the dominant contributor to station
. blackout frequency and a major contributor to core melt frequency.
As a result, the above-mentioned short-term modifications were
- Implemented to remove the interdependency. Thu's the core melt frequency value of J.5 x 10-4/ reactor-year does not include the contribution from this scenario.
?
l' 2. Mitigation of the Effects of the Loss of MCC-5 as a Transient Initiator
! As part of the overall PSS effort, the frequency and consequences of -
o the loss of MCC-5(3) as an initiating event were also considered.
, System reliability calculations were performed using fault tree analysis to estimate the frequency of occurrence of loss of MCC-5.
. Failure modes and effects analyses were performed to determine the
{ consequences of the loss of MCC-5. Walkdowns of the plant with the i operating staff at Haddam Neck identified the means of identifying the transient as well' as possible means of recovering the system
- and/or mitigating the effects of its loss.
! The scenario is initiated by a de-energization of MCC-5 resulting from
- any of a number of causes. A large fraction of the de-energization of
- MCC-5 events (95 to 99 percent) are potentially recoverable.
{ However, as a result of the loss of MCC-5 there is a loss of semi-vital i AC power along with a trip of both main feedwater pumps on sensed j low-suction pressure. Subsequently, the reactor trips on feed flow -
- steam flow mismatch. The loss of MCC-5 also causes the loss of the I
control air compressors, loss of closed cooling water pumps for i compressor cooling and loss of most of the important motor-operated i valves' including safety injection, core deluge, containment spray, i pressurizer relief, loop isolation,: and chemical and volume contro!
. systems valves.
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! (3) This component is not required to meet the single failure criterion as
- described in the " United States Atomic Energy Commission Safety Evaluation i by the Division of Reactor Licensing", dated July 1,1971.
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. The loss of semi-vital AC power, in-addition to causing the loss of.- '
i important instrumentation such as steam generator pressure and I auxiliary feedwater flow, automatically starts the stand-by charging q pump and isolates letdown makeup to the volume control tank (VCT). :
The charging pump main lube oil pumps stop on loss of MCC-5, although the auxiliary lube oil pumps would .be available. Automatic l switch-over of charging pump suction from the VCT to the refueling i water storage tank (RWST) falls as the VCT level goes down. Within i several minutes of initiation, the VCT is drained. . Without operator j
intervention, both charging pumps would cavitate and be damaged soon ,
thereafter. With the loss of MCC-5, many alarms would be received in
- the control room. Because of the eventual loss of control air, all l AOVs would fall in their safe position, but automatic control would be i lost. Auxiliary feedwater would be available because the AOVs would
, fail open, but their position could not be controlled from the control
- room. Likewise, the positive displacement (metering) charging pump j would not be controllable. Component cooling water flow to the j' reactor coolant pump (RCP) thermal barrier would be lost because the AOV falls closed on loss of air or loss of semi-vital AC power. Hence,
) all RCP seal cooling is potentially lost. t On the secondary side, the loss of air results in the eventual closing of the main steam trip i
valves. The turbine bypass valves (steam dump to condenser) as well l as the atmospheric steam dump valve would not be available.
i
- Preliminary evaluation of the above scenario identified this transient
{ as one of the dominant contributors to the core melt frequency. As a 1 result, a project assignment using " emergency". work-order procedures I was initiated to effect plant design changes necessary to trip the j charging pumps on loss of MCC-5 (specifically loss of semi-vital AC power). Manual realignment of charging suction to the RWST would j ensure reactor coolant system make-up capability during this scenario.
The PSS identified possible means of mitigating the accident including l the potential for recovering MCC-5, and the cross-tie of service air to
- control air along with compressor cooling using the well water system.
} With air restored, auto-control of the charging metering pump as well 1
as control room operation 'of auxiliary feedwater flow would be '
] available. The draft symptom-based Emergency Response Guidelines are also undergoing evaluation to incorporate means of identification
- and mitigation of this transient.
1 l As -of this writing, the above described plant modifications and j associated procedure changes are in the process of being implemented.
! Our objective is 'to complete this effort prior to plant start-up.
3 However, depending upon the actual date of plant start-up, completion of the work may occur early during the upcoming cycle. Since their work will be completed either prior to start-up or shortly thereaf ter, i we have elected to credit these modifications in the PSS calculations.
1 I The calculated core melt frequency of 5.5 x 10-4 per reactor-year j does include some residual contribution from this accident scenario, i but the plant design / procedural changes described above have removed l most of the contribution of this scenario.
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i II. Engineering Insights Obtained from the PSS
- 1) Adverse Break Size and Location in the Reactor Coolant System (RCS) 1
!- During performance of the "Best Estimate LOCA Analysis" (NUSCO- .
f 150), a small range of break sizes in loop 2 cold leg of the RCS was l identified for which safety injection flow in the high-pressure
- recirculation mode may be insufficient to prevent core uncevery in the absence of modifications to facility operating procedures and/or J
system (e.g., valve) realignment. The specific break sizes and locations involved are breaks of 0.045 ft.2 to 0.02 ft.2 in the cold leg pump discharge line for RCS Loop 2 and breaks less than 0.045 ft.2 in i the Loop 2 charging line and connected piping.
! For Loop 2 breaks greater than or equal to 0.045 ft.2 (equivalent
! diameter = 2.9 incheCthe RCS, with all accident mitigation pumps
- activated in the injection mode, will depressurize to 165 psia (150 psig) and below prior to,)pitiation of core uncovery and also prior -to the depletion of 100,W0 gallons of RWST inventory. Low-pressure recirculation using the residual heat removal (RHR) pumps would follow.
For Loop 2 cold leg breaks less than or equal to 0.02 ft.2 (equivalent t i
diameter = 1.9 inches), one charging ' pump and ' only one of two charging flow control valves is sufficient in the high pressure recirculation mode to prevent core -uncovery. (The high pressure recirculation mode commences by' aligning the charging pump suction - ,
i to the discharge of the RHR pumps when RCS pressure is above 165 psia. The sole delivery point is to the Loop 2 cold leg.)
i Breaks located directly in the charging line or connected piping i downstream of the check valves for sizes 0.045 ft.2 and smaller may l also produce results analogous to the Loop 2 cold leg breaks discussed i
! above since for these breaks the RCS pressure also remains above 165 psia while the charging flow provides insufficient flow to prevent
- core uncovery.
t
) The frequency of these breaks was estimated in the PSS by multiplying i the (generic) frequency of various size LOCAs times the conditional j probability that a break of a given size occurred in the adverse location. This conditional probability was derived by totalling up the j number of pipe segments in the affected location and dividing by the l total number of pertinent pipe sections ' of the given size range i throughout the RCS. ' A pipe segment is defined as a section of pipe
{ between major discontinuties (e.g., valves, reducers, etc). In the PSS,
- approximately 9% of the medium break LOCAs and 1% of the small i break LOCAs are estimated to occur in adverse locations. The definition of a pipe segment and the ' methodology used for quantification is consistent with WASH-1400.
! As a result of this discovery, evaluation of the feasibility of utilizing 1 additional sources of borated water to extend the injection phase and/or using other available injection paths, such as the loop-f.ill i
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+
header or auxiliary preswrizer spray, are being evaluated. 'This issue
- is being addressed in separate correspondence.
l 2) Best Estimate LOCA Analysis
! In support of the PSS, plant-specific thermal-hydraulic analyses were performed for a wide spectrum of potential core uncovery events. The results are summarized in the " Connecticut Yankee (Haddam Neck Plant) Best Estimate LOCA Analysis," . NUSCO-150. The . major findings of the analysis are summarized as follows:
! o For large breaks (area > 0.2 ft.2), success of emergency- core cooling in the injection phase requires only one low-pressure
! safety injection pump providing flow through one core deluge ,
- valve.
I o For medium breaks (0.2 ft.2 2 area 20.02 ft.2), success requires only one high-pressure safety injection (HPSI) pump providing i flow to 3 of 3 unfaulted loops (or 2 HPSI pumps to 2 of 3 unfaulted loops).
l l o For medium-small breaks (0.02 ft.22 area 3 0.003 ft.2), success .t requires only one charging ' pump or one HPSI pump for j
- mitigation. ;
o For small breaks (area < 0.003 ft.2), success requires steam generator heat removal capability in addition to either one HPSI pump or one charging pump for mitigation.
i o " Feed-and-bleed" cooling requires one charging pump and 4 one pressurizer PORV, or one HPSI' pump and two PORVs.
l Initiation of feed-and-bleed is required approximately 40 minutes following the loss of all steam generator feedwater.
o For station blackout, core damage would not occur for over i eight hours, even assuming catastrophic failure of all reactor j coolant pump seals. i i
! 3) Additional Findings The following additional insights are also of interest:
i i o There are eight potential paths for high pressure / low pressure l Interfacing systems LOCA; however, the overall contribution to
- core melt frequency is low, i
j o Pipe ruptures in the letdown line' outside containment but upstream of the flow control orifices, although relatively low in
- frequency, have only one potential means of isolation (LD-MOV- !
j 200). Failure of the MOV to close either automatically or.
manually could potentially result in an unisolated LOCA outside containment; the sequence could possibly be risk dominant. As part of the Integrated Safety Assessment Program (ISAP),
l; l s.
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1 - . . _ _ . _ ._ . . _ _ _ .-
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i evaluations are underway to determine whether a proposed design change to add an additional containment isolation valve outside containment for other reasons might be more appropriate if relocated inside containment.
o A reliability analysis of the auxiliary feedwater system has found-
- i. the unavailability of automatic initiation of the system to be 3
dominated by failures of air-operated (and particularly solenoid-operated) valves. Changes to periodic' testing frequencies of the valves are under consideration. In addition, a ' potentially ,
significant common cause failure of the system due to the
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nonrestoration of manual valves following . preventive ,
maintenance has been identified and is under evaluation.
o Failure of SW-MOV-5 and 6 to open on demand are found to be dominant contributors to RHR system unavailability. In the LOCA recirculation mode of core cooling.
. o Based on current procedures, the testing of the main steam trip ,
valve transmitters was found not to be systematic in that testing '
. of all four channels is not assured.
o The fans for cooling the charging pump lube oil coolers offer a ;
potential means of providing cooling independent of component cooling water; however,'no_ information on the design adequacy i
has been found nor. are there any testing or maintenance procedures for the fans.
1 o The diesel-driven fire pump provides an AC independent means 2
of containment spray during station blackout should containment
! heat removal be necessary; however, a single motor-operated 4
valve (MOV-31) outside containment would have to be opened
, manually. This feature of an AC-independent containment spray pump is important in minimizing the offsite public consequences resulting from core melt due to station blackout.
o On a best estimate basis, the RCS loop isolation valves have been found to provide a reasonable means of mitigating the
- effects of steam generator tube ruptures.
I o Due to the diversity and redundancy of containment heat removal features (four fan coolers, 'six pumps for spray),- the overall availability of containment heat removal systems for mitigating the impact of core darnage accidents was found to be
- exceptional.
o The overall reliability. of main .feedwater/ auxiliary feedwater post-trip (non-loss of feedwater events) was likewise found to be i exceptional.
> III. Discussion of RCS Piping Ductility i
The PSS results are based on LOCA frequencies derived from the 6 experience of pressurized water reactors, as quantified in WASH-1400. "
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, I i
The actual LOCA frequencies at Haddam Neck are expected to be lower than the values used in the PSS since the piping material is austenitic stainless steel while the PWR population includes plants which use ferritic carbon steel. Stainless steel provides superior ductility and toughness and is not susceptible to brittle fracture at low temperatures. Thus, the PSS results include an unquantified conservatism in terms of the actual LOCA frequencies.
IV. Discussion of Haddam Neck Plant-Specific Simulator Experience The plant-specific control room simulator for the Haddam Neck Plant became available during the last month of the study. Additional benefits will be realized in the future as the results of the PSS are incorporated into the training of operators on the simulator.
V. Future NUSCO Actions As stated previously, NUSCO is currently reviewing the results of the PSS to determine the best altr natives available to reduce the core melt frequency at the Haddam IV . Plant. Completion of these analyses will result in either a clarificatt J the system success criteria, identification of weaknesses or recommerd aons fo possible hardware and/or emergency operating procedure changes When these are completed we will requantify the Haddam Neck PSS models reflecting the net effects of these changes.
As new issues are identified during our review of the PSS, we plan to either:
a) Implement a project assignment corresponding to the identified issue, if the . issue is deemed to have a high level of importance to plant safety, or ,
I b) Evaluate the issue and any resultant projects within the framework of l the ISAP, and implement modifications to the plant / procedures on a schedule commensurate with the results of the ISAP evaluation.
4 DISTRIBUTION PER MIKE BOYLE LTR ENCL REG FILE 1 1 EBE McCABE REGION I 3 3 RRAB 2 2 ERNST MALCOLM RESEARCH 2 2 AEOD 1 1 j ICE 1 1 NRC PDR 1 1 LPDR 1 1 NSIC 1 1 24X 1 1 14 14 SEND ALL EXTRA COPIES TO: MIKE BOYLE
i DISCLAIMER The information contained in this topical report was prepared for the specific requirements of Northeast Utilities Service Company (NUSCO) and its affiliated companies, and may contain materials subject to privately owned rights. Any use of all or any portion of the information, analyses, methodology or data contained in this topical report by third parties shall be undertaken at such party's sole risk. NUSCO and its affiliated companies hereby disclaim any liability (including but not limited to tort, contract, statute, or course of dealing) or warranty (whether express or implied) for the accuracy, completeness, suitability for a particular purpose or merchantability of the information.
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- i CONNECTICUT YANKEE PROBABILISTIC SAFETY STUDY