ML20235T502
| ML20235T502 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 07/17/1987 |
| From: | Hodgdon A NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | NEW ENGLAND COALITION ON NUCLEAR POLLUTION |
| References | |
| CON-#387-4043 OLA, NUDOCS 8707220161 | |
| Download: ML20235T502 (323) | |
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d~ UNITED' STATES OF AMERICA NUCLEAR REGULATORY COMMISSION {'
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
VERMONT YANKEE NUCLEAR ) Docket No. 50-271-OLA POWER CORPORATION ) (Spent Fuel Pool Amendment)
.)
(Vermont Yankee Nuclear Power )
Station)
NRC STAFF RESPONSE TO NECNP'S SECOND SET OF' INTERROGATORIES AND DOCUMENT REQUESTS TO THE NRC STAFF On June 23, 1986, New England Coalition on Nuclear Pollution (NECNP) filed its Second Set of interrogatories and Document Requests to the NRC Staff. The Staff notes. that while interrogatories to parties other than the Staff are ' governed by 10 C.F.R. 5 2.740b, answers to interrogatories to the Staff, pursuant to 10 C.F.R. 5 2.720(h)(2)(ll), are required only on a finding by the presiding officer that answers to the interrogatories are necessary to a proper decision in the proceeding and that answers to the interrogatories are not reasonably obtainable from any other source. The Commission's regulation concerning production of NRC records and documents, 10 C.F.R. 5 2.744, requires that a request to j
the Executive Director of Operations for the production of an NRC record or document not available pursuant to 6 2.790 by a party to an initial licensing proceeding state, among other things, why the requested record or document is relevant to the proceeding. Notwithstanding the regulations in 10 C.F.R. 66 2.740b, 2.744 and 2.720(b)(2)(ii), the Staff 8707220161 870717 PDR ADDCK 05000271 G PDR .p.
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- i. h [ j is - voluntarily providing responses to NECN P's Second Set of.
- Interrogatories and Document Requests.1I INTERROGATORY 1 Please provide all documents in NRC's possession concerning ' the !
Vermont Yankee Containment Safety Study, including but not limited to, all NRC Staff / consultant documents commenting on or otherwise evaluating the study, and documents by othcrs commenting on or otherwise evaluating the study. ,
RESPONSE
Documents other than publicly available documents responsive to the request 'are provided in Enclosure 1.
INTERROGATORY 2 Please provide copies of all documents which contain and/or present .the results of calculations of core melt progression, containment loads due to severe accidents, and/or fission product release estimates (source terms) for the Vermont Yankee Nuclear Power Plant.
RESPONSE
The Staff has not been able to identify any documents that are directly responsive to the request.
INTERROGATORY 3 Please provHe all copies of documents which evaluate the likelihood and/or magnitude of containment pressurization resulting from phenomena
-1/ On July 16, 1987, Staff counsel received a letter from NECN P's counsei, dated July 14, 1987, in which mention is made of an
" extension of time (the second) for the Staff's responses." As noted ,
above, the Staff's responses to interrogatories unless required by the . Licensing Board pursuant to 10 C.F.R. 9 2.720(h)(2)(ll) are I voluntary. In a telephone conversation, Staff counsel had indicated to NECNP's counsel that the Staff would voluntarily respond to the interrogatories. Apparently, NECNP's counsel misunderstood what Staff counsel said. Staff counsel agreed to provide responses; they were at no time " overdue".
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. known as high pressure melt ejection and direct containment heating in ,
BWR Mark i nuclear power plants.
RESPONSE ]
i With regard to this matter, the Staff of the Office of Nuclear Reactor Regulation (NRR) is awaiting a response from the Office of Nuclear Reactor Research (RES). Any documents that RES might have other than public documents will be made available at a later date.
INTERROGATORY 4 )
Please provide all seismic PRA studies which have been conducted on BWR Mark I nuclear power plants. <
RESPONSE
The Staff is unable to provide a comprehensive response to this request at this time as persons with knowledge of these matters are currently unavailable. The Staff wlIl provide any documents responsive to this request that are not public documents at 6 later date.
INTERROGATORY 5 Please provide all estimates of the probability of seismic ground acceleration (i.e., seismic hazard curves) at the Vermont Yankee Nuclear Power Plant.
RESPONSE
The Staff has searched its records and has not been able to identify any documents that are directly responsive to thl:s request.
INTERROGATORY 6 Please provide all documents which evaluate the likelihood and/or mechanism of failure of reactor bulidings (i.e., secondary containments) i in BWR Mark i nuclear power plants as a result of loadings imposed by I
or during severe accidents.
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. RiiSPONSE The S'taff of. NRR is ' not aware of documents .'other than public documents that .are responsive to this request. However, NRR is awaiting RES verification o? this information and will respond further' at a later. l
'date.
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( INTERROGATORY 7 Please provide all documents which evaluate the likelihood and/or mechanism of failure of spent fuel pools in BWR Mark i nuclear power
. plants as a result of loadings imposed by or during severe . accidents.
. RESPONSE
' Studies addrer. sing the subject matter of this interrogatory have been undertaken' by Lawrence Livermore under contract to the NRC as a part of the effort to resolve Generic issue No. 87. Documentation of that effort is not yet available.
t INTERROGATORY 8 Please provide all documents which evaluate the fragility of spent fuel pools for seismic ground acceleration-induced failures for BWR Mark I nuclear power plants.
RESPONSE
See response to interrogatory 7.
Respectfully submitted, n.
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1 Ann P. Hodgdon CounceI for NRC Staff Dated at Bethesda, Maryland this 17th day of July,1987
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i UNITED STATES OF' AMERI'CA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD in the Matter. of )
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VERMONT YANKEE NUCLEAR ') Docket No. 50-271-OLA POWER CORPORATION -- ) (Spent Fuel . Pool Amendment)
)
(Vermont Yankee Nuclear Power ) ;
' Station ) . !
c AFFIDAVIT OF VERNON L. ROONEY, JR. l I, Vernon L. Rooney, Jr,- being duly sworn, state as 'follows: .
I am' employed by the U.S. Nuclear Regulatory' Commission as a I
Senior Project Manager: in ' the Division of Reactor Projects-1,II. I have !
the responsibility for management and coordination of the actions concerning the NRC staff review of ~ Vermont Yankee Nuclear Power l l
Corporation's application for an amendment to expand the capacity of its i spent fuel pool. l 1 have provided the responses to NECN P's Second Set of i
Interrogatories and Document . Requests to the NRC Staff, specifically !
Interrogatories 1 through 8 and am duly authorized to do so.
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I attest that these responses are true and cor;ect to the best of my knowledge and belief.
MN NM bi Vernon L. Rooney, Jr. (
Subscribed and sworn to before me tj>1s/g day of July,1987 WW/
Notarp Public
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My commission expires: July 1,1990 i
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UNITED STATES OF AMERICA
- s NUCLEAR REGULATORY COMMISSION 87 JUL P4:17 BEFORE THE ATOMIC SAFETY AND LICENSING,BOAg 00;u ,
"I in the Matter of )
)
VERMONT YANKEE NUCLEAR ) Docket No. 50-271-OLA POWER CORPORATION ) (Spent Fuel Pool Amendment)
)
(Vermont Yankee Nuclear Power ).
Station)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO NECNP'S SECOND SET OF INTERROGATORIES AND DOCUMENT REQUESTS TO THE NRC STAFF" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or as indicated :
by an asterisk through deposit in the Nuclear Regulatory Commission's internal mall system, this 17th day of July,1987:
Charles Bechhoefer, Esq. Mr. Glenn O. Bright Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555* Washington, D.C. 20555*
Dr. James H. Carpenter George Dana Bisbee Administrative Judge Senior Assistant Attorney General i Atomic Safety and Licensing Board Environmental Protection Bureau U.S. Nuclear Regulatory Commission 25 Capitol Street Washington, D.C. 20555* Concord, NH 03301-6397
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Atomic Safety and Licensing Board Ellyn R. Weiss, Esq.
U.S. Nuclear Regulatory Commission Harmon & Weiss ;
Washington, D.C. 20555* 2001 S Street, N.W. l Washington, D.C. 20009 David J. Mullett, Esq. Carol S. Sneider, Esq.
Special Assistant Attorney General Assistant Attorney General Vermont Depart. of Public Service Office of the Attorney General 120 State Street One Ashburton Place,19th Floor Montpeller. VT 05602 Boston, MA 02108 -
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Thomas C. Dignan, Jr., Esq. Jay Gutierrez Ropes and Gray Regional Counsel 125 Frankl!n Street USNRC, Region i Boston, MA - 02110 631 Park Avenue King of Prussia, PA 15406*
Atomic Safety and L'4 censing Appeal Docketing and Service Section-Board Panel Office of the Secretary U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555* Washington, D.C. 20555* i
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%nn 'P. Hodgdon Counsel for NRC Staff 1
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MAY o r g f Docket No. 50-271 DISTRIBUTION-Docket 'ile- t. Nerses NRCPDR ,
V. Rooney a local PDP E. Trottier DGC ACRS(10) _ _ i Mr. J. G. Weigand PDI-3 Rdg. W. Kane (RI).
Resident & Chief Executive Officer . S. Verga J. Craig Vemont Yankee Nuclear Power B. Boger M. Rushbrook Corporation R.D. 5 Box 169 Ferry Road i Brattleboro, Yemont 05301
Dear Mr. Weigand:
SUBJECT:
VERMONT YANKEE CONTAINMENT SAFETY TNITIATIVE - STAFF REVIEW OF PLANS We are in receipt of your series of letters sunrnarizing thee ma.ior analyses and engineering studies you have voluntarily undertaken to improve containment performance. In_ particular, we appreciate your most recent correspondence of February.5, 1987,.in which you provide er updated status, the expected completion date_ for your engineering design effort (July 1987),andthe commitment to provide specific desigS details. l The staff is very interested in the (iens, design details, procedures and training associated with your containment safety' initiative. In particular, we would like to review all aspects of your design and implementation plans. :
Accordingly, please provide design details and implementation schedule at your i earliest convenience as they become available, but no later than July 15, -l 1987. Your cooperation in this matter is appreciated.
Should you have any questions regarding this, plea $e contact the NRR Project !
Manager, V. Rooney (301-49?-80?4).
Sincerely, I
-Steven A. Varga, Director Division of Reactor Projects, 1/11 cc: See next page
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J Mr. R.'W. Capstick' Vemont Yankee Nuclear. Power Corporation b Vemmont Yankee Nuclear Power Station CC:
Mr.' W. - G. Weigand W. P. Murphy, Vice President Prctident & Chief Executive Officer and Manager of Operations Vemont Yankee Nuclear Power Corp. Vermont Yankee Nuclear Power Corp.
R.D. 5, Box 169 R.D. 5. Box 169- '!
Ferry Road. .
Ferry Road Brattleboro, Vermont 05301 Bratticeboro, Vemont 05301 Mr. Donald Hunter, Vice President Mr. Gerald Tarrant, Commissioner Vement Yankee Nuclear Power Corp. Vemont Department of Public Service 1671 Worcester Road, 120 State Street Framingham, Massachusetts 01701 Montpelier Vermont 05602 New Enoland Coalition on Nuclear Public Service Board Pollution State of Vermont Hill and Dale Fam 170 State Street R.D. 2 Box 223 Montpelier, Vermont 05602 Putney, Vcmont 05346 Mr. Walter Zalurny Vemont Yankee Decommissioning Chaiman, Board of Selectmar Alliance Post Office Box 116 Box 53 Vernon, Verment 05354 Montpelier, Yemont 05602-0053 J. P. Pelletier, Plant Manager Resident inspector Vemont Yankee Nuclear Power Corp. U.S. Nuclear Regulatory Comission .i Post Office Box 157 Post Office Box 176-Vernon, Vermont 05354 Vernon Vermont 05354 Rayrond' N, McCandless Vemont Public Interest Research Yemont Division of Occupational Group, Inc.
and Radiological Hemith 43 State Street Administration Building Montpelier, Vermont 05602 10 Baldwin Street .
l Montpelier Vemont 05602 Acting Regional Administrator Region 1 Office l Honorable John J. Easton U.S. Nuclear Regulatory Comission Attorney General 631 Park Avenue i State of Vemont King of Prussia, Pennsylvania 19406 109 State Street Montpelier, Vemont 05602 Mr. R. W. Capstick .
Vemont Yankee Nuclear John A. Ritscher Esquire Power Corporation Ropes & Gray 1671 Worcester Road 225 Franklin Street Framingham, Massachusetts 01701 Boston, Massachusetts 02110 -
L . f Y Occket No: 50-271 t
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MEMORANDUf! FOR: Chairman Zech Commissioner Roberts Commissioner Asselstine-
. Commissioner Bernthal u Commissioner Carr L FROM: Victor Stello, Jr.
Executive Director for Operations
SUBJECT:
MODIFICATION OF EMERGENCY RESPONSE CAPABILITY.0RDER -
VERMONT YANKEE NUCLEAR POWER STATION-In mid-1986 Vermont Yankee at.the urging of the Governor of Vermont, undertook c Containment Safety Study addressing concerns related to the capability of Vermont Yankee's Mark I containment to withstand severe accidents. The conduct of this study was in part in res,ponse to the staff's initiative in improving-the performance of BWR Mark.I containments; but also has been coordinated with the State of Vermont through the state's Vermont State Nuclear Advisory Panel 0n several occasions Vermont Yankee met with and formally responded
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(VSNAP) to questions on the study from the staff and VSNAP. Based on this study Verment Yankee identified several improvements that might be implemented (improved drywell spray and improved venting capability). In order to provide resources for these activities, Vermont Yankee has requested (Enclosures 1 and
- 2) schedular relief from certain NUREG-0737 Supplement I activities (Require-nents for Emergency Response Capabiitty). The request is based on VermSnt Yankee's evaluation of the relative safety importance of delaying
. implementation of some NUREG-0737 Supplement I activities versus delaying engineering necessary for the containment modifications.
Specifically Vermont Yankee requested modification of the Confirmatory Order confirming its schedular comitments for the Safety Parameter Display System (SPDS) and for Regulatory Guide 1.97. The SPDS is presently required to be fully operational and operaters trained prior.to startup for Cycle 14 (approximately winter 1988/89). Vermont Yankee proposed that the SPDS would be functional at Cycle 14. startup, but testing and training would be completed during Cycle 14. Implementation of Regulatory Guide 1.97 requirements is presently required prior to startup from Cycle 13 (approximately fall 1987).
Vermont Yankee has proposed to meet this requirement for all equipment except for the Local Power Range Monitor power supplies, which would be unqualified with respect to Regulatory Guide 1.97 until prior to startup for Cycle 14, at which time they would be qualified. In addition, Vermont Yankee requested CONTACT:
Vernon L. Rooney, NRR 28024 l i
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relief from a comitment regarding the Detailed Control Room Design' Review which was not confirmed by Order.
-Vermont Yankee proposed to use engineering resources, which would otherwise b'e devoted to the above NUREG-0737 Supplement 1 activities for engineering-necessary to implement containment safety improvements. The improvements are an improved alternate drywell spray path,~ improved capability of the diesel fire pump for drywell spray and core cooling , and an improved containment wet-well vent path. Provided the schedular extensions are granted, Vemont Yankee has committed to:
- 1. Implement a modification to the plant to insure a reliab'le and controllable alternate containment spray path. . Generally, this modification will involve providing an alternate sourre of power to four motor-operated-valves to allow remote operation of these valves in a station blackout type of scenario.
- 2. . Continue analyses and calculations addressing the performance characteristics of the diesel fire pump when used to provide water via the service water,,RHR service water, and RHR system to either the reactor vessel or containment spray header. These results will provide the criteria to detemine modifications, if required, necessary to ensure the containment spray headers perform when required.
- 3. Implement a modification to the plant to ensure a reliable and controllable vent path from the containment wetwell. . Engineering studies
. presently under evaluation involve an independent' pipe from the wetwell airspace with the appropriate valving and manual control circuitry to allow for initiation and positive termination of venting, if required.
Vermont Yankee will provide to the staff a schedule for implementation of these modifications, and specific designs, upon completion of the engineering design effort in July 1987.
The staff considers that the delay for one fuel cycle of these particular Emergency Response Capability comitments is compensated for by the importance of the planned containment safety activities, and therefore has granted the requested relief by issuing the enclosed letter (Enclorure 3). Because the Comission consented to the dates in the Confirmatory Orders dealing with Emergency Response Capability (SECY-83-484), we are informing the Commission of the relief we have granted.
It should be noted that Vermont Yankee's participation in the Mark 1 Containment Safety activities was voluntary and therefore any plant modifications undertaken would not be subject to backfit considerations by the staff. The staff's action in granting the requested relief does not relieve
r-Ve .:. Yarkee f rom implementing any additional requirements which may result !
frce, the ongoing generic Mark 1 Containment Safety progran.
Origirel sit:ne.1 by Victor Stell o l
Victor Stello, Jr.
Executive Director for Operations
Enclosures:
DISTRIBUTION As stated Dociet File TSpeis JRoe NRC POR WRussell TRehni cc: SECY LocL1 PDR DMuller JSniezek OGC PD#2 Vemo File LJhy EDO r/f VStello SNorris HDenton MThadani RVollmer Glainas RBernero JHulman Director of PWR Licensing A. WHodges FMiragila l
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OFFICIAL RECORD COPY
- Previously Concurred i DBL:PD#2 DIR:DIV DD:NRR DIR:NRR EDO DBL:PDfJ DBL:PD#2 ShoWis VRooney:cb' DMuller* RBernero* RVollmer* HDenton* VStello 3 / 10/87 2/17/87 2/17/87 2/18/87 2/24/87 2/26/87 / /87 l
- s. .i 60-d7 /
VERMOST YANKEE rneiosure 1 NUCLEAR: POWER CORPORATION FVY 86-122
. . } AD 5. Box 169. Ferry Road. Brahleboro VT 05301 jy'2 g ,,, ,o ENGINEERING OFFICE U _i M) s 1671 WORCESTER ACAD FRAMINGHAM. MASS ACHuSETTS Ct% -
TELE **sD*vf $1?4724'DC December 19, 1986 U.S. Nuclear hegulatory Commission Washington, D.C. 20555 Attn: Office of Nuclear Reactor Regulation Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
References:
a) License No. DPR-28 (Docket No. S0-271) b) Letter, USNRC to VYNPC, NVY 85-187, dated 8/29/85 c) Letter, VfNPC to USNRC, FVY 86-30, dated 3/31/86-d) Letter,.VYNPC to USNRC, Fvy 86-81, dated 9/2/86 e)- Letter, USNRC to VYNPC, NVY 86-218, dated 10/24/86
Dear Sir:
Subject:
Request to Modif y Confirmatory Order and Provide Schedular Extension for Emergency Response Capability (Supplement I to NUREG 0737) Requirements This purpose of this letter is to request modification of the order con-firmirg Vermont Yankee schedular commitments on Emergency Response Capability (Suppiiement 1 to NUREC 0737) for Safety Parameter Display $ystem (SPDS) opera-bility and Regulatory Guide 1.97 implementation. Additionally, we are indi-cating our revised schedule for certain specific commitments associated with the Vermont Yankee Detailed Control Room Design Review (DCRDR).
Previously, Vermont Yankee established with the Nuclear Regulatory Commission (NRC) en integrated plan'and schedule for addressing the issues detailed in NUREG 0737, Supplement 1, " Requirements for Emergency Response Capability". This overall integrated approach resulted in mutually acceptable program plans for the'Esergency Operating Procedures (EOPs), the DCRDR, the Regulatory Guide 1.97 assessment, a new Emergency Response Facility (ERF), and the SPDS. 89 letter, dated August 29, 1985 [ Reference b)), the staff ;
transmitted its most recent license modifying order confirming Vermont Yankee l NUREG 0737, Supplement 1, commitments. This order additionally stateo that, {
" Extension of time for complating these items may be granted by the Director, Division of Licensing, for good cause shown."
B612300287 B61219 PDR p
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VERMONT YANKEE NUCLEAR POWER CORPORATION l.i-
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U.S. Nuclear Regu13ttey Commission Decemoer 19, 1966 Page 2 )
In accordance with Vermont Yankee's integrated NUREG 0737, Supplement 1, plan and effort, commitments associated with EOPs, DCRDR, and the ERF have been .
completed (except that TSC data acquisition requirements will not be fully func- f tional until SDDS completion). The two remaining requirements to be completed involve SPDS operability (scheduled for completion prior to startup for Cycle 14), and Regulatory Guide 1.97 implementation (scheduled for completion prior to-startup for Cycle 13). Additionally, although Vermont Yankee's confirmatory order requirements for DCADR are completed, our letter of March 31, 1986 l
[ Reference c)), provided a proposed schedule for implementing all identified i DCRDR Human Engineering Discrepancy (NED) modifications during the'next two suc- l eessive outages (i.e., prior to Cycle 13 and la startup). For the reasons and causes discussed below, Vermont Yankee requests schedular. extension for commit-ments associatec with NUREG 0737, Supplement 1, pertaining to SPDS operability, anc Regulatory Guice 1.97 implementation; and provides our revised schecule for DCPDA modifications.
In mio-1966, Vermont Yankee undertook an extensive ef fort in response to NRC vitiatec Marn I concerns following tne Chernoby1 accident regarding the capar. ty cf containments such as Vermont Yankee's to witnstand severe occi-cente.
- orce- to properly assess the performance of Vermont Yankee's desig*,
are its noility to mitigate severe accioents, Vermont Yankee uncertook a stuoy wnicr. 3rcorporateo recent aevances in analytical technicaes and accounted for tne signi+1 cant cesign features specific to Vermont Yankee which affect the olent's apility to responc to a severe accident. On September 2, 1986
[Re'erence o, Vermont Nankee submitted the result of this study to the NRC. During a subsecuent September 11, 1986 meeting at the NRC's headquarters in Bethesos, Marylano, it was noted that Vermont Yankee's efforts constituted a sample care for NRC's generic containment requirements activities. Consistent with tnat purpose, the staf f stated its intention to review and comment on the 4 Vermont Yankee study and to provide questions in late October 1986 [ Reference I e)). While agreeing to provide timely responses to the comments and questions, ) Vermont Yankee stated that continuation of the level of effort associated with i the containment safety initiative might conflict with certain other scheduled commitments to the NRC. Subsequently, the staf f advised Vermont Yankee to ] icentify such cases to NRC Project Management for discussion. Accordingly, J Vermont Yankee has reviewed its schedular commitments to NRC for the next several refueling outages and determined that certain commitments associated with NUREG 0737, Supplement 1, requirements will require schedular extension. At the time Vermont Yenkee established its commitments for NUREG 0737, Supplement 1, requirements, the extensive ef fort associated with the contair, ment safety initiative was unanticipated. As a result of the resources expended Dy Vermont Yankee associated with this effort to date and those anticipated in the f uture, Vermont Yankee has concluded that the relative safety importance of tne planned containment safety efforts warrant a modification to the remaining NUREG l l
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VERMONT YANKEE NUCLEAR POWCR CORPORATION s U.S. Nuciear Regulatery Commission Decemoer 19, 1966 Page 3 0737, Supplement 1, commitments involving SPDS, Regulatory Guide 1.97 and DCRDR. Specifically, these resources included approximately 1,000 manhours in the PRA disciplines, 1,000 manhours in the System Engineering discipline, and 400 ; manheurs each in the I&C and Electrical Engineering disciplines to perform the initial study and support the follow-up questions and meetings. Resource estimates for the planned containment safety ef forts include the followingi 1. An engineering scoping study which examines 9bt feasibility of improving
- the reliability of the valves located in the Reactor Building associateo witn the alternate spray path has been estimated to require A00 manhours in the Electrical ana Systems Engineering disciplines.
2. Analysis ano calculations are being performed to prove the viability of the ciesel fire pump to provide water via the service water anc RHR containment spray heacer for core cooling and drywell spray. This analysis is esti- i matec to recuire ECO manhours in the Systees Engineering disciplines anc is seneduleo for completion in mio-198't. 3. An engineering scooing study has been initiated to evaluat_e the feasibility of previong a heroeneo. reliable vent path from the containment wetwell to the. plant Vent stack. AP Proximately 900 canhours in the Systems, Eiectrical, anc 16C Engineering disciplines as well as 200 manhours from tne Nuclear Analysss discioline will be required. This study is s,cheduieo for-completion in tne spring of 1987. The basis for each senecular extension request is discussed below. REGULATORY GUIDE 1.97 Vermont Vankee's present commitment regarding the Regulatory Guide 1.97
- requirement to implement (install or upgrade) post-accident monitoring instru-mentation is prior to startup for Cycle 13' (approximately summer 1987). During the.1986 refueling outage, Vermont Yankee completed all Regulatory Guide 1.97.
installation / upgrades associated with our Environmental Qualification Program. The previously planned work scope for the 1987 refueling outage included comple-tion of all remaining Regulatory Guide 1.97 installation / upgrades. Recently we have determined that the LPRM power supplies should have been included in the work scope for Regulatory Guide 1.97 modifications. However, due to the dif-ficulty of including this recently determined upgrade in the 1987 outage work scope in view of the containment resource commitmer.ts previously discusseo and based on the fact that, with the exception of this one item, all Regulatory Guide 1.97 installation /upgraces will be completec in accordance with the i
. ,f VERMONT YANKEE NUCLE AR POWER CORPORATION 6
3 U.S. Nuclear Regulate.y Commission , Decemoer 19, 1986 Page 4
. .t existing schedular commitments we reovest that the subject order be modified to extend the Regulatory Guide 1.97 schedular commitment for this specific item one accitional outage cycle (prior to startup for Cycle 14, approximately winter 196b/89).
DCRDR As stated in the subject confirmatory order, Vermont Yankee has completed the NUREG D737, Supplement 1, commitment relating to DCRDR requirements. By letter, cated March 31,19E6 [ Reference c)), Vermont Yankee provided r, proposed scheoule for implementing all DCRDR HED modifications, assuming no " unforeseen oifficuities in the oesign, procurement and installation efforts associateo with each of the respective modifications," during the next two successive outages (i .e. , prior to Cycle 13 and 14 startup). It is our. firm belief that the HED modifications proposed from the DCRDR effort are of relatively minor operational significance and generally relate to modifications necessary to conform to specific human f actors criteria. This has been sLostantiated by tr;e results of our Control Room Dynamic Evaluation per-fornec ir the fall of tr.is year. Due to the higher safety significance paced on tne Cor.tainrent Sa'ety Stucy, a finite resource base to support Vermont Yankee's continJing Mark I Containment effort [ resource commitments previcusly
.ciscuss,ec), anc completion of all DCRDR tauks (including Control Room Dynamic Evaluatio*. and Expanceo Task Analyses for all Vermont Yankee Specific E0P's), we I have ioentified a one cycle schedular extension to the Commitment for DCRDR HED l modifications contained in Reference c). I Accordingly, our present intention is to complete these modifications curing tne two refueling outages prior to startup for Cycles 14 and 15, suDject to no unforeseen difficulties in the design, procurement and installation efforts associated witn each of the respective modifications. We will provide advanced written notification of items, if any, that are later determined to require extension.
SPDS Vermont Yankee is presently committed to having an SPDS fully operational prior to startup'for Cycle 14. As our design efforts end concurrent installa-tion scheduling activities progress, we have identified significant resource , shortages in concurrently developing containment initiative tasks, and SFDS l design and Installation and Test Procedure (!&T) packages. Additionally, cue to ; further developed design details, we can now identify significant amounts of l SPDS work which can be completed during plant operation. We therefore reouest [ that the schedular commitment for SFDS be modified to state that SPDS operabi-l lity and operator training will be completed prior to the end of Cycle 14 To 1 1
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VERMONT YANKEE NUCLE AR POWER CORPORATION b.S, Nu: lear Aegulat6ry Comnission December 19, 1986 Fage S alleviate this resource burden and permit more levelized resource allocation, we propose to cefer commencement of SPDS installation activities until following
. startup f or Cycle 13 (approximately Winter 1987/88) instead of our existing plan to commence with the Cycle 13 outage.
We propose a part-cycle deferral of SPDS as follows. We will proceeo with SPDS development and design activities to the extent achievable with our ewisting resources. Installation of non-outage driven modifications will com-pence and continue during Cycle 13. Outage driven modifications will be per-formed 'in the outage prior to Cycle 14 (approximately February-March 1989). We toen anticipate having SPDS functional at startup for Cycle 14 but necessary j sta-tup testing, system verification and validation, and operator training would l centinue into Cycle la prior to the SPDS being decle' red fully operational. Full I SPDS operability would De achieved prior to coepletion of Cycle 14 Vermont Yankee. believes that the foregoing provides good Cause jur,tifica-tien f or:eacn of the requested schedular extensions. Therefore, based on the aoove, we recuest your approval of the proposed revised scheoules and mooific6-t or of tr.e orcer cenfirn.ing verment Yankee schedular commitments on Emergency Respense Operability (Suppiement 1 to NUREG 0737) for SPDS ano Repuistory Guice
;.97 arc toproval of seneouler extension for certain commitments associatec witn tne Vere. ort vanhee DCRCA.
Snocic you nave any cuestsons or require aoditional information concerning this reovest, please contact us. I' Very truly yours, A gg Warren P. Murphy Vice President and Manager of Operations
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1 i i i l
A s e: .- Enclosure _2 VERMONT YANKEE , e NUCLEAR POWER CORPORATION
) FVY 87-18 RD 5. Box 169, Ferey Roac, Brattleboro. VT 05301
;' , , ,,, , o ENGINEERING OFFICE S
1611 WORCESTER RC A D f RAMINGHAM, M ASS ACHUSETT$ C1701 sne. oww 3.se February 5, 1987 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Office of Nuclear Reactor Regulation Danie? R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
References:
a) License No. DPR-28 (Docket No. 60-271) b) Letter, USNRC to VYNPC, NW 85-187, dated 8/29/85 c) Letter, WNPC to USNRC, FW 86-30, dated 3/31/86 d) Letter, VYNPC to USNRC, FW 86-81, dated 9/2/86 e) Letter, USNRC to WNPC, NVY 86-218, dated 10/24/86 f) Letter, WNPC to USNRC, FVY 86-117, dated 12/19/86 g) Letter, WNPC to USNRC, FW 86-122, dated 12/19/86
Dear Sir:
Subject:
Vermont Yankee Containment Safety Initiatives - Status Update The purpose of this letter is to provide you with an updated status of Vermont Yankee's Containment Safety Study initiatives. In our letter of December 19,1986 [ Reference f)), we provided you with a summary of several of the major analyses and engineering studies initiated by Vermont Yankee to address containment performance concerns associated with severe accidents. Vermont Yankee's progress to date and current plans [provided the schedular extensions requested in Reference g) are approved) ralated to these initiatives are as follows:
- 1. Vermont Yankee proposes to implement a modification to the plant to insure a reliable and controllable alternate containment spray path.
Generally, this modification will involve providing an alternate source of power to four motor-operated valves to allow remote opera-tion of these valves in a station blackout type of scenario. Vermont Yankee will notify you of our schedule to implement this modification and the specific design upon completion of the engineering design effcet (July 1987). G702180326 DR 870205 p ADOCK 05000271 1 PDR 00l
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. VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission.
February 5, 1961 Fage 2'
- 2. Vermont Yankee is continuing analyses and calculations addressing the !
performance characteristics of the diesel fire pump when used to pro-Vide water via the service water, RHR service water, and RHR system to either the reactor vessel or containment spray header. These results will provide the criteria to determine modifications necessary to ensure the containment spray headers perform when required. Vermont Yankee will notify you of our schedule to implement modifications if required and the specific design upon completion of the engineering design effort (July 1987).
- 3. Vermont Yankee proposes to implement a modification to the plant to ensure a reliable and controllable vent path from the containn.ent wetwell. Our engineering study evaluating the details of this vent path is progressing on schedule. The concept presently under eval-uation involves an independent pipe from the wetwell airspace with the appropriate valving and manual control circuitry to allow for ini-tiation and positive termination of venting, if required. Vermont Yankee will notify you of our schedule to implement this modification and the specific design upon completion of the engineering design effort (July 1987).
We trust that this information is of assistance regarding your con-siderations of our letter of December 19, 1986 [ Reference g)) which requested modification of certain schedular commitments impacted by our past and con-tinuing safety initiatives. Should you have any questions or requ4re additional information concerning these matters, please contact us. Very truly yours, VERMONT YANKEE WUCLEAR POWER CORPORA 110N l 4 ans. M Warren P. Murpny Vice President and Manager of Operat ns
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- Enclosure 3
'c . UNITED STATES l l 7, y T ;'e NUCLEAR REGULATORY COMMISSION usmoto9.o. c.rosss
%, [ Tebruary IF, 1967
~'
Docket No. 50-271 - l. Mr. R. W. Capstick Licensing Engineer Vemont Yankee Nuclear Power Corporation 1671 Vercester Road Framinghem, Messechusetts 01701-
Dear Mr. Capstick:
SUBJECT:
SUFPLEMENT 1 TO NUREG-0737. REQUEST FOR FURTHER MODIFICATION-0F COMMISSION ORDER DATED JUNE 12,1984 Re: Vermont Yankee Nuclear Pcwer Station The Comission Confirmatory Order dated June 12,198a confirmed the implementation cates for certain items relating to Supplement I to NUREG-0737.
-The Comission Order dated August 29, 1985 modified the June 12,'1084 Order to-confirm edditional commitments you had made pertainin,1 to the schedules for the Safety Parameter Display System (SPDS) and Regulatory Guide'1.97 requirements.
Subsequently, by letter ' dated December 19, 1986, you requested an additional modification of the August 29, 1985 Order to extend the implementation date fer the SPDS and Regulatory Guide 1.97. You are presently required to have the SPD5 fully. operational prior to startup for Cycle 14 (approximately winter 1988/89) and required to implement Regulatory Guide 1.97 requirements (install or upgrade) prior to startup from Cycle 13 (approximately fall 1987). You rer;uested extension of the schedule such that, although SPDS would be functional at Cycle 14 startup, testing and training would be accomplished during Cycle 14, with full SPDS operability achieved prior to completion of Cycle 14 With respect to Regulatory Guide 1.97 requirements, you requested extension of your comitment for Local Power Range Monitor (LPRM) power supply qualification in accordance with Regulatory Guide 1.97 for one additional outage cycle (prior to startup for Cycle 14), although other Regulatory Guide 1.97 installation / upgrades would be completed prior to startup for Cycle 13. You also requested that a scheduler comitment for implementing all Human Engineering Discrepancy (HED) modifications resulting from the Detailed Control Room Design Review (DCRDR) be changed from prior to Cycle 14 startup to prior to Cycle 15 startup (approximately mid-1990). The above commitment is contained in your letter dated March 31, 1986, and was not confirmed by Order. Enclosed is our evaluation of your request for schedular extension. Based on this evaluation, in accordance with the terns of the June 12, 1984 and August 29, 1985 Comission Orders, we hereby grant the requested delays for implementing the SPDS requirements and for implementing the Reguletory Guide 1.97 requirements. Also, based on this evaluation we find your proposed schedule for implementing DCRDR modifications to be acceptable.
\
Finally, approval of the foregning does not comit the Com.ission to any specific apprpvel or disapproval of the Vermont Yankee plans relative to the Containment Safety Study. A sumery of the status of our review will be previded ir the spring in response to your September 2,1980, December 29, 19Bf ,.tnd February 5,1987 submittels. Sincerely,
) ^
Robert M. Bernero Director Division of BtlR Licensing Office of Nuclear Reactor Regulation
Enclosure:
As statec cc w/ enclosure: See next pege i I 1 1 I i
~ ^ . . , 4 . . .2.... ...s..... - . . . .. .- .
I;r. R. W. Capstick Vem:nt Yankee Nuclear Power vernont Yankee Nuclear Power Corporation Statio., cc* -
- Mr. J. G. Weigand Mr. W. P. Murphy, Vice President &
President & Gnief Executive Officer Manager of Operations Vemont Ycnkee Nuclear Power Corp. Vemont Yankee Nuclear Power Corp. R. D. 5. Box 169- R. D. 5, Box 169 Ferry Road Ferry Road Brattleboro, Vement 05301 Brattleboro, Yemont 05301 Mr. Donald Hunter, Vice Fresident Mr. Gerald Tarrant, Commissioner Vermont Yankee Nuclear Power Corp. Vemont Department of Public Service 1671 Worcester Road 120 State Street Fresingham, Massachusetts 01701 Montpelier, Yemont 05602 New England Coalition on Public Service Board huclear Pollution State of Yemont Hill and Dale Farm 120 State Street R. D. 2, Box 223 Montpelier, Yemont 05602 Putney, Vermont 05346 Vemont Yankee Decommissioning Mr. Walter 2aluzny Alliance Chaiman, Board of Selectman Box 53 Post Office Box 116 Montpelier, Vermont 05602-0053 Vernon, Vermont 05345 Resident inspector Mr. J. P. Pelletier, Plant Manager U. 3. Nuclear Regulatory Comission Yemont Yankee Nuclear Power Corp. Post Office Box 176 Post Office Box 157 - Vernon, Vermont 05354 Vernon, Yemont 05354 Vemont Public Interest Mr. Raymond H. McCandless Research Group, Inc. Vemont Division of Occupational 43 State Street
& Radiological Health Montpelier, Yemont 05602 Administration Building 10 Baldwin Street Regional Administrator, Region 1 Montpelier, Vemont 05602 U. S. Nuclear Regulatory Comission j 631 Park Avenue Honorable John J. Easton King of Prussia, Perr.sylvania 19406 Attorney General State of Vemont Ellyn R. Weiss 109 State Street- Hamon & Weiss Montpelier, Vermont 05602 2001 S. Street, N. W.
Suite 430 Washington, D.C. 20009-1125 John A. Ritscher, Esquire Ropes & Gray Office of the Attorney General 225 Franklin Street 1 Ashburton Place Boston, Massachusetts 02110 Igth Floor Boston. Massachusetts 02108 l
'e UNITED STATES l' # e
.y NUCLEAR REGULATORY COMMISSION j- '
i WASHWGTON, D. C. 2D65$ p.:/ EVALUATION
- l-BY THE OFFICE OF NUCLEAR RECTOR REGULATION FOR THE VERMONT YANKEE NUCLEAR POWER STATION VERMONT tANKEE NUCLEAR POWER CORPORATION DOCKET NC. 50-271
1.0 INTRODUCTION
Ey letter dated December 19, 1985, the licensee Vernent Yankee Nuclear Power Corporation (YYHPC), requested relief from'the Comission Confirmatory Order dated June 12, 1984, which confirned the implementation dates for certain items relating to Supplement I to NUREG-0737 The Comission Order dated August 29,19E5 modified the June 12,198a Order to confirm additional comitments the ifcensee had made pertaining to the schedules for the Safety Parameter Display System (SPDS) ano Regulatory Guide 1.97 requirements. The licensee's Decent >er 19. 1986 letter requested an additional modification of the August 29, 1985 Order to extend the. implementation date for the SPDS and Regulatory Guide 1.97. VYNPC startup forisCycle presently 14 req (approximately winteruired 1988/89) to have and required to the SPDS fully o implement Regulatory Guide 1.97 requirements (install or upgrade) prior to startup from Cycle 13 (approximately fall 1987). YYNPC requested extension of the schedule such that, although SPDS would be functienal at Cycle 14 startup, testing and training would be accomplished during Cycle 14, with full SPDS operability achieved prior to completion of Cycle 14. With respect to Regulatory Guide 1.97 requirements, VYNPC requested extension of its comitment for Local Power Range Monitor (LPRM) power supply qualification in accordence with Regulatory Guide 1.97 for one additional outage cycle (prior to startup for Cycle 14), although other Regulatory Guide 1.97 installation / upgrades would be completed prior to startup for Cycle 13. VYNPC elso requested that a schedular commitment for implementing all Human Engineering Discrepancy (HED) modifications resulting from the Detailed Control Room Design Review (DCROR) be changed from prior to Cycle 14 startup to prior to Cycle 15 startup (approximately mid-1990). The above comitment is contained in VYNPC's letter dated March 31, 1986, and was not confirmed by Order. 2.0 _ EVALUATION , i In the December 19, 1986 submittal, the licensee stated that in mid-1986 VYNPJ undertook an extensive and unanticipated effort in response to NRC initiated concerns regarding the capability of containments such as Vemont Yankee's to withstand severe accidents. YYNPC perfomed a Containment Safety Stucy (CSS) which incorporated recent advances in
2 enelytical techniques and acc0unted for the design features specific to Vermont Yankee which affect the plant's ability to respond to a severe accident. VYNPC subsequently met several times with the NRC staff and representatives of the State of Vermont and provided responses to NRC steff questions and comments thereby laying the groundwork for Vermont Yankee Containment Safety improvements while advancing the NP.C generic Mark 1 Containment Safety activities by providing a sample case for a typical Mark I reactor containment. The licensee stated that as a result , of the rescure.es expended on this effort to date and those anticipated in j the future it has concluded that the relative safety importance of the planned containment safety efforts warrant a modification to the , remaining NUREG-0737, Supplement 1, commitments involving SPDS, i Regulatory Guide 1.97 and DCRDR. Specifically, resources expended to i date on the CSS include approximately 1,000 manhours in the PRA l disciplines,1,000 manhours ir. the System Engineering discipline, and ADO manhours each in the I&C and Electrical Engineering disciplines. VYNPC described planned CSS activities and estimated that resource requirements associated with these planned activities totaled an additional 2,000 manhours. In the December 19, 1986 submittal VYNPC stated that it has identified significant resource shortages in concurrently developing containment initiative tasks, and SPDS design and Installation and Test Procedure packages. Additionally, due to further developed design details, it can now identify significant amounts of SPDS work whi.ch can be completed during plant operation. The licensee therefore requested that the schedular comitment for SPDS be modified. The licensee stated that, as previously required, the SPDS would be installed and functional at the end of Cycle 14. The change in SPDS requirements which the licensee proposed is that startup testing, system verification and validation, and operator training be conducted during Cycle 14 rather than prior to the initiation of Cycle 14. We find that the delay of these testing and training activities is compensated for by the importance of the planned containment safety activities. Therefore, the staff concludes that there is adequate justification for modification of the Commission Order. The licensee stated that during the 1986 refueling outage, VYNPC completed all Regulatory Guide 1.97 installation / upgrades associated with their Environmental Qualification Program. The previously planned work scope for the 1987 refueling outage included completion of all remaining Regulatory Guide 1.97 installation / upgrades. Recently the licensee
- determined that the LPRM power supplies should have been included in the work scope for the remaining Regulatory Guide 1.97 modifications.
However, due to the difficulty of including this recently determined upgrade in the 1987 outage work scope in view of the containment resource commitments previously discussed and based on the fact that, with the exception of this one item, all Regulatory Guide 1.97 installation / upgrades will be completed in eccordance with the existing schedule commitment; the licensee requested that the subject Order be modified to " :.' ' extend the Regulatory Guida ~ #7 schedular comitment for this specific T
item by one additional fuel cycle (prior to startup for Cycle 14, approximately winter 1988/89). The staff finds that the licensee's proposal is reasonable in view of the safety significance of the containment rafety activities. Therefore, the staff concludes that there is adequate justification for modification of the Comission Order. a In the licensee's submittal of August 29, 1986 it states that Vermont Yankee has completed the NUREG-0737, Supplement 1, comitment relating to-DCRDR requirements, but that by letter, dated March 31, 1986 Vermont Yankee provided a proposed schedule for implementing all DCRDR HED modifications, assuming no " unforeseen difficulties in the design, procurement and installation efforts associated with each of the respective modifications," during the next two successive outages (i.e., prior to Cycle 13 and 14 startup). In the licensee's htter dated December 19, 1986, it stated that VYNPC believes that the HED modifications proposed from the DCRDR effort for Vermont Yankee are of relatively minor operational significance and generally relate to modfice'Ns necessary to conform to specific human factors criteria and that it:, belief has been substantiated by the results of VYNPC's Control Room Dynamic Evaluation oerformed in the fall of 1986. Due to the higher safety signi'icance the licensee has placed on the Containment Safety Study, the licensee's limited resources to support vermont Yankee's continuing CSS effort and completion of all DCRDR tosks, the ifcensee foresees a one cycle schedular extension to its comitment for DCRDR HED modifications. This will result in completing these modifications prior to startup for Cycle 15. The staff finds the proposed schedele for implementing DCRDR modifications is acceptable. The Containment Safety activities which VYNPC described involve engineering studies preparatory to implementing the alternate drywell spray path, improving the capability of the diesel fire pump for drywell spray and core cooling, and hardening the containment wetwell vent path. By letter dated February 5,1987, the licensee has committed to installation of these containment safety improvements, as appropriate, 1 based on the results of the engineering studies, if the requested relief 1 is granted, and has comitted to provide a schedule for implementation I upon completion of the engineering design effort in July 1987.
3.0 CONCLUSION
With respect to the relief requested in the letter dated December 19, 1986: '
- 1. The staff finds the request for extending the schedular requirement for the SPDS to be fully operational and operators trained to be i acceptable. The SPDS is to be functional at startup for Cycle 14 ;
The SPDS is to be fully operational prior to the end of Cycle 14 l (including startup testing, system verification and validation, and I operators trained).
t
- 2. ' The staff finds the request for extending the schedular requirement fod tegulatory Guf de 1.97 implementation with' respect to LPRH power supplies to prior to startup for Cycle 14 to be acceptable.
Regulatory Guide 1.97 schedular requirements for other items are to remainunchanged(priortostartupforCycle13). ,
- 3. The staff finds the licensee's proposed schedule for implementing DCRDR modifications to be acceptable.
Dsted: February 28, 1987 l t i, a J
* /,q / r 1 CHRONOLOGY JUNE-16, 1986': MEETING.WITH BWROG/IDCOR PROPOSED A GENERIC LETTER, PRESCRIPTIVE SOLUT!ON, BY.BACKFIT JUNE 30,'1986: VERMONT. YANKEE COMMITS To GOV. KUNIN TO DO A SPECIAL 60-DAY CONTAINMENT STUDY
. SEPTEMBER 1, 1986: VERMONT YANKEE CONTAINMENT STUDY COMPLETE
^l SEPTEMBER 11, 1986: MEETING WITH BWROG TO COMPARE BACKFIT NOTES AND STRAWMAN GENERIC REQUIREMENTS SEPTEMBER 11,.1986: MEETING WITH VERMONT YANKEE TO REVIEW CONTAINMENT STUDY SEPTEMBER 23, 1986:
MEETING WITH ACRS SUBCOMMITTEE NOVEMBER 3, 1986: COMMISSION BRIEFING NOVEMEER 17, 1986: LWR OWNERS GROUP MEETING NOVEMBER 17, 1986: MEETING WITH VERMONT YANKEE NOVEMBER 18, 1986: MEETING WITH THE STATE OF VERMONT NOVEMBER 20, 1986: NUMARC MEETING WITH THE COMMISSION l l l n
- a. -
DECEMBER 3, 1986:. DRAFT G;L. TO ACRS DECMEBER .f4, 1986: DRAFT.G.L. TO CRGR
' *- DECEMBER 9, 1986:
ACRS' SUBCOMMITTEE MEETING DECEMBER 12, 1986: ACRS FULL COMMITTEE MEETING DF.CEMBER 22, 1986: .CRGR MEETING 1 1 r PROJECTED CHRONOLOGY FEBRUARY 27, 1987: DRAFT COMMISSION PAPER MARCH 13, 1987: OFFICE CONCURRENCES MARCH 20, 1987: CRGR MEETING APRIL 10, 1987: COMMISSION PAPER TO EDO-APRIL .17, 1987: COMMISSION PAPER TO THE' COMMISSION (RW bJ
- MAY 15,- 1987:' PROPOSED GENERIC LETTER FOR PUBLIC COMMENT-AUGUST 1987: ISSUE GENERIC LETTER A
i i 1
/
ON t, I : f[... ...,ik, UNITED STATES y ws -g NUCLEAR REGULATORY COMMISSION o l c
' WASHINGTON.D.C.20sss February 6, 1987
%.,....f TO: Harold R. Denton, Director -
Office of Nuclear Reactor Regulation FROM: Robert M. Bernero, Director Division of BWR Licensing
SUBJECT:
DIVISION OF BWR LICENSING COMMENTS ON NUREG-1150 On January 30, 1987, you asked'the NRR Divisions to comment on the draft results of NUREG-1150 which'are available to (.s in five respects:
- 1. Technical comment on the severe core damage frequency (SCDF) estimates presented.
- 2. Whether these SCDF estimates appear to be representative of each riass of plants.
- 3. Technical comment on the conditional containment failure prob-abilities (CCFP) presented.
- 4. Whether these CCFP estimates appear to be representative of each class of plants.
- 5. What regulatory action should be undertaken in view of these results?
The comments'in this memorandum speak to the final draft of NUREG-1150 as represented by the briefing slides. We have not reviewed or even seen the final draft. , i
- 1. The SC0F for the two BWR's analyzed, Peach Bottom and Grand Gulf, are 8.2x10 -6 /yr and 2.8x10 -5 /yr respectively, virtually all from station blackout in both cases. These frequencies are so low that only guarded use of these numbers should be made. At these low frequencies even minor l
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omission or simplifications could significantly affect the results. We recognize that Peach Bottom is the most PRA-analyzed BWR because of the Reactor Safety Study. Nevertheless, we believe that most PRA experts, including those who worked on NUREG-1150 would not be surprised-would even expect-that an independent team performing a Level 1 PRA on Peach
-6 Bottom today would calculate a mean frequency higer than 8.2x10 /yr.
- 2. The SCDF estimates for the two BWRs in NUREG-1150 appear too low to reasonably represent typical'BWRs. The attached table lists.all the available PRA results for BWRs. Note that the well reviewed efforts for Millstone-1, Shoreham, and Limerick show that the SCDF values are -
an order of magnitude higher,
- 3. The CCFPs presented in NUREG-1150 are perhaps the most important result.
For all but the large dry the report seems to say that uncertainties and even close to best estimates make it impossible to say that, given a core melt, there is' reasonable assurance that the containment will substantially mitigate its consequences. This is a profound challenge to the adequacy of the final barrier for defense-in-depth, and for the defense-in-depth philosophy as well. It has long been recognized that the oft cited inner barriers of defense-in-depth, namely the fuel form, the cladding, and the primary' pressure boundary offer little defense in severe accidents. And . now the containment is suspect. It will take extensive review to under-stand and perhaps concur with these results. The Grand Gulf performance appears to be poor but strangely precise, CCFP = 0.25 0.35. The Peach Bottom performance is extremely poor, CCFP = 0.1 - 0.9, but appears to be driven upward to an unknown degree by the dominant station blackout sequence, resulting in virtually no drywell spray or venting during the containment response."
*We think it would be useful to introduce an alternate definition of core melt mitigation performance. Rather than using the conditional early failure of the containment pressure boundary, we suggest using the fraction of core melts which result in a large release (FRLR) of radioactive materials from the plant. In this way the mitigating effects of secondary containments or reactor buildings as well as the accident sequence distribution are taken into account. This would clearly be realistic, and technically correct.
l l
- 4. As noted in 3. preceding, we have not reviewed and understcod the basis for the NUREG-1150 CCFP results for BWRs. However, from our own review of BWR Mark I and Mark III containments in severe accidents we have con-cluded that these might well be representative analyses fry all contain-ments of each type but not for the plant, since the entire plant involves systems' reliabilities which vary from plant to plant. The Peach Bottom Mark I results may well represent all Mark I's if no further improvements are made. On the other hand the Grand Gulf Mark III results are far worse than our results for the similar GESSAR II plant, and we do not uncerstand the reason for the difference.
- 5. We recommend the following regulatory actions:
- a. NUREG-1150 Brief the Commission and publish the document for comment with a vigorous disclaimer of the five plants analyzed as a represent-tative sample of the risk of nuclear plants in the U.S. today. The core melt frequencies do not appear to be representative and, espe-cially with the very low BWR numbers, might give a falsely reassuring picture of the industry average. In this regard, we are aware of a widely circulated estimate of core melt likelibct'd for the entire U.S. -
reactor population in the next 20 years, an estimate of 12% vs. the controversial earlier estimate of 45%. We believe that the 12% esti-mate is based on NUREG-1150 SCDFs, using the 8x10-6 Peach Bottom estimate as representative of all 24 Mark I BWRs and 9 Mark II BWR's etc. Such an estimate can only be represented as what might be achieved and is not an honest representation of the present state of reactor reliability. Further, the combination of these f requencies with the containment results to obtain risk estimates is dangerous. It implies that these are representative risk estimates and leads to false conclusions, as in the slide comparing all five plants to the possible safety goal
-6 for a large release. People have already begun criterion of 10 /yr
i 4 to conclude from that slide that Ice Condenser PWRs are clearly un-acceptable and Mark I BWRs are already acceptable - a cenclusion
. completely driven by core melt frequency since this report shows the -
Mark I containment to be even worse than the Ice Condenser. The at-tached comment paper is furnished to support our suggestion that this definition of large release,1 early fatality, should never be used.
- b. _ Severe Accident Core Melt Frequency Since the efficacy of contain-ment is challenged on tbout half of U.S. plants by HUREG-1150 the likelihood of core melt is far more important and therefore the .
Individual Plant Examination process for the front end should be pursued with a11' deliberate speed to search for outliers and implement improvements as needed as soon as r.ossible.
- c. Severe Accident Containment Performance We believe that, for the state-of-the-art LWRs defense-in-depth with at least a modicum of
" containment" performance is mandatory. We believe that a practical variation of our current generic Mark I proposal, if reviewed in the NUREG-1150 process, would be found by the RES scientists and others to provide reasonable assurance that only a small fraction of core melts in a Mark I BWRs, would result in a large release from the plant. ,
We believe that this modified proposal for Mark Is should be prepared, published for coment with Commission consent, and explicitly analyzed in NUREG-1150 (final) during the comment period. Similar proposals should be prepared for the other containments, to be treated in the same way.
. w Robert M. Bernero, Director Division of BWR Licensing
Enclosure:
As stated cc: NRR Division Directors DBL SES Staff i _ _ _ _ _ _ .--___- _________ _ __ _ a
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1 RN R R R R R R R R R R R R R R R R OO W W W W W W W W W W W W W W B W B W G E CC B B B B B B B B B B C B B L B A T T RR OA 5 4 6 3 6 2 6 1 3 3 5 6 G 1 4 PE 7 8 8 8 8 8 9 8 8 8 8 8 8 8 8 EY 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 R 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 A y A A E y R r 1 R R P r P a P P I a n 0 e - - - a I 0 m R n R k R m R m R s I m 4 u1 C o C c C a C a C u1 T 1 S2 /5t /1 ir /8 h /0h /9S2 R R - G8sS G0 5 G2 e G5e G5 A O H hk E01S E8 5 e E0 r E0r E6hk S R2 C R3 o R4o E R1 cs S P E S A csE eaP R31 U 1P U Y im U h U h s P U M ea E G R W TTI M M M V L M s M I TT M P A O A R RE R O 5 O GM O P C P 5 C C M O C OA S C E E S E C P L L L E E S RN S 0 P R U R Y E M I N P P S D E P R 1 I I N I V P B L B I I R I G y r a r n f f e e e n l l n n om o n n F e k k m ma m a u u o o o t e c a a h G G t t t t t s nk i icr h h h e u R A T hthtt.t s s n on r e er e d d N cococo 1 1 w o wa e e r o o r o q s n a n a S S A L aBaBaB e e e 1 l 1 i l r Bjh rY e im L miL h s h S S h S u r G G r E G P I P P P ,f P P 1 l
I. t ENCLOSURE COMMENTS ON INTERPRETING . A SAFETY GOAL' In the 'implementat' ion of the Safety Goal the staff is considering .the per-formance guideline:
" Consistent with the traditional defense-in-depth approach and the accident mitigation philosophy requiring reliable performance of containment systems,--
the overall mean frequency of a large release of radioactive materials to .the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation." Current thinking, as expressed to the Comission in the January 2,1987,
- memo to the Commission on Safety Goal Implementation Status, and as analyzed in the~ forthcoming NURES-1150, Reactor Risk Reference Document, uses a definition of "... a large reletse of radioactive meterials to the environment..." as a release which causes even 1 early fatality. The use of fch a definition poses a serious' logical problem.
In 1982 an NRC-sponsored report, NUREG/CR.2239 Technical Guidance for Siting , Criteria Development, presented standard consequence analysis for 91 reactor sites in the U.S. as a way to develop siting criteria. For these analyses three fission product releases were considered SST.1*, SST-2*, and SST-3*. They were described as: COMPOSITION OF SITING SOURCE TERMS Released fractions (to atmosphere) SST Xe, Kr 1, Br Cs, Rb Te BaSr Ru La 1 1.0 .45 .67 .64 .07 .05 9x10~3 3 2 0.9 3x10~3 9x10 3x10-3 1x10-3 2x10~3 3x10-4
~ ~ 6 3 6x10 2x10 1x10' 2x10 1x10' Ix10' 1x10
i SST-1 Severe core damage. Essentially involves loss of all in' stalled safety features. Severe direct breach of containment. SST-2 Severe core damage. Containment fails'to isolate. Fission product release mitigating systems (e.g., sprays, suppression
' pool, fan coolers) operate to reduce release.
SST-S Severe core damage. Containment fails by basemat melt-through. All other release mitigation systems function as designed. , I SST-1 is in essence the laroest core melt accident release ever calculated for a U.S. reactor. Among other things the siting Study calculations determined the conditional probability of early fatality at each site give an SST-1 release. Some interesting results were obtained: CONDITIONAL PROBABILITY NUMBER OF OF 1 EARLY FATALITY SITES WITH GIVEN SST-1 RELEASE THIS RESULT 0.5 - 0.6 2 0.1 - 0.5 30' O.05 - 0.1 22 0.01 - 0.05 24 0.005 - 0.01 5 I 0.001 - 0.005 8 I l l
A fair reading of the results is that the conditional probability of 1 early fatality at a typical site given an SST-1 release is 0.1. In NUREG/CR-2239 an absolute probability context was needed and a reading of then available PRA results was used to state the following nominal release frequencies: l SST - 1 1 x 10-5/yr SST - 2
-5 2 x 10 /yr
~4 SST - 3 1 X 10 /yr --
PRA results available today indicate that these are reasonable frequencies. Consida.r now the following syllogism: The frequency of SST-1 is 1x10 -5 /yr.
~1 The conditional probability of 1 early fatality is 1x10 /yr.
- Therefore, the frequency of 1 early fatality is 1x10-6/yr Therefore, such a plant meets the safety goal for a large release. .
Therefore, an SST-1 is not"... a large release of radioactive materials to the environment..." This patent absurdity indicates that another definition of a large release is needed, based on releases, not effects.
l I J OFFSITE DOSE ESTIMATES FOR CONTAINMENT VENTING January 21, 1987 Robert M. Bernero V.S. NRC
l, NOTES
'- ALL CALCULATIONS ARE FORL A LARGE BW'R (3412 MWT.)
I EXPECTED VALUES ARE THOSE CALCULATED FOR TYPICAL WEATHEP. CONDITIONS-LFOR BAD WEATHER DOSES CAN BE AS MUCH AS'10 TIMES HIGHER FOR GOOD WEATHER ~ DOSES CAN BE AT LEAST 10 TIMES LOWER DOSE ESTIMATE FOR CORE MELT FOLLOWED.BY-DIRECT BREACH OF CONTAINMENT MAY BE MITIGATED BY SOME DELAY IN RELEASE OR BY
.THE SPECIFIC PATH OF CONTAINMENT BREACH, DOSE ESTIMATES FOR VENTING ARE FOR NOBLE GAS ONLY
- 1 HOUR HOLDUP IS MINIMUM TIME, LONGER DELAYS CAN REDUCE DOSE BY UP TO A FACTOR OF 'F 30.
,i_._..
- EXPECTED WHOLE BODY RADTATION DOSE (REN) i ,i i i u . 4,4
...... o_
- - - , i ,
s..... i r=- y r-,I +- = - m . t-i..- ^ NOTES:
.' ._.,3_3 s..... - -
...... ; " 1) The dose values used are expected mean values adjusted W
( .m . . -. 1 for a 3412 MWT BWR. Actual doses received could be di
'"-" k m;_:3. _.f- L approximately an order of magnitude higher or as low M.&
n.n . -:- U.-L y 8---- - i~~ 7_ Ti'~ . ~"~~ 5 7-- 6 as zero, primarily depending on meteorological and accident conditions. M ' g: _ _ _a :EV : : :
- 2) The estimated doses from noble gases released at P'
~~ - ' -
-l' ground level or elevated assume one hour holdup and --
D. (-e . ~_ ~- q. decay prior to release. The release is assumed to be _,___,: p;~-~mv. n tfl?= G . W over a five hour period. Greater delay in release cans;2
, c i==.5:=~=-M produce lower doses (e.g., as much as a factor of .m b
- --V ===C := about 30 at one mile for 12 hours of in reactor s....
g'~
..... d-Qbg@@
=n-n w w.,w holdupcomparedtoonehour).
8---- MfM7kN 3) The direct breach of containment dose values assume o :=_i_-t Z -
.:= (a) essentially the loss of all installed safety
- - ~~ ~ -y~ 4-- ;r_~ _.=features _e = d:at the plant, (b) no emergency response
~
j r. -- .__ e pc ___ actions taken, and (c) one day exposure to
-- A ;_
m. radionuclides deposited on the ground. . g i___.N - _ _ _ _- _ _ _ _ _ . .
~ . ,,., , w .,; .a ... ,~ - . s. . w. . ] a .,.
g,::::- : m. ,~ . _=. , p._, x .-
= 0 :.
- =~
2 .=cw=e u-n Ir r= wM N -M- - =: .1.- .= _;r s.v m; r w+' e =n - a:=c r - -n -:v = . - :*n,s-. ,. . .
., . ., m. .\e.a. ,3
- s. =
- ~ ""'
~
~
~~F'^'\~' I"' '
" "' ] ~O - -
" I"'" "# ~~ 1' ~^ I = = =
,--- * ' . ff2 M~ ~'f _'EE'Es=-5=Ef=NT/ . e . ~% n . , s.
(*CWr_E-t *i . ' - .~sTisr. W.-E 4-~':=t_E9.=1
= + '-i=#r.f-M --as -
w q u..~ vx- 1- + . , ... ->
~.~.x. n,. . + . _w=-
~**-=~U
- I E =:~~- '
=' E '- - M4
.._.Mo -r -r b . C-'i d
- Y Si^
".+ w' = + -_E +.+.m = w w' 'n' + N :w v- w .w._ m . ."u z'=: = .
* =. a.. a., n- , -
m x wn=.wn=; .
,... c w--me=& 1 := ==r+. - x =:T
. . m = = _=. .: .===w=.= =
- k ! - $-E- - jr=D [i-i - L .H - '
~
$ =IM-d "~=E 2: -i 5~- ; .'. i=k"- f_ L- 9. M_ ; . ;, I -- I @ b-Q M.f--i~ ]= =-- :.. =- : :. ::r .g_. , ] ~y ~ i :,=-_5[i ;: Ed ~. -
. . ~ __
;' (A:-gg_"._
fs "g,. 5 CORE MELT FOLLOWED BY DIRECT BREACH OF CONTAINMENT g==-E..--_
. g ._.,-
x_ . _.. ._g .___u_. ,
..._..c_.___._..._-T.
- 4-_
5, g --;
...s _-.;2_.u-
_ 4__ , _. ..,_ y._ . M
, , jg,=, i v. . _ . ..,_m.,_._, s ; y=c p.=
- = .;.
.x,,gsQ
.- _r-g,,, ,
,c3 T e g r-y, yp ; F:=y v - - --
c: os amw w. m-m s mv : 9.v s-m m.w, . = . .. .+, n 7 .,_ +,-m e
,' ] $g=m.fiL~ hfi- mA '-V iiPMb~.3. r-=.=s} f-::.} i.~ bk[:}=.W
. . .= = t * :-M='.sL b=i. .. W: ='
=4-l=4_==
q
. . . ._ m - .. ~ ..
, - . _ = = = -.-_
... =a
=_s- _ . =
.. m. _ .m. =m
. - _: c + :- -- v , - = - m - =
_ - . -=.=* 5, r Ng g,, d3 1 '-4.d EN D I s e i .: '- 2 i'# E*:=I- ' !Eilk :. ' 5:H *. I' 13-#.-M < M .I t * / .. I r 8' -' rb1. h-
=. ,- w'"' E E.m A' emi,,se y*
i:#W - :5-EM v =-+ . 4 U5:6.' C = =Mm-m m m .-rem : : ~9 =~=; m . E +r ' =F ' ~= H
+
. =. i n.c.m . _ - - .
J + <";=- ; W +1 W
== '"
- -- -i"..-N N=~c E. m .+- o c
T ~. _; ~~C_ - _:-_ y.-: p; g i=_%L, GROUND LEVEL RELEASE OF, NOBLE GASES
,,, ..:--_.. z _- .r- , _
.n. _____._.___n.____,.__ _ _. . . . . . . . . . . . . __ . ._. o O
==
x " a_%
-. . _ _. . . _ . .) . _ .. ._ . 8 =;
x , x ._m . ._.. ...__ _ . .
- w. , __g -
- x i r .r. 2.:. __...=.__o 9-: _
- ,7 m,- = r, .
t
.w._w
- w. - ,- 4 ~ .._._.. . ~ . _
, x
- 8. d '~[' ".YM ,I"..~.' YN -m ! .== ,%
i# + [=illf,I
=~ f=--'.
'.I .x _:n=+.- '2 3=5r-==D U 5"= m =i'm = W=$ ~I'i=rMX .. . ..
i p. 2. c= s=r.m :1 = w.- 2 m --. t s a 'i l
. avsm= =
e : n:s ==m2%2-mfma c==.+=11 y;.exs 5 wm c. . g1 t
; ___ 35-EEEEP. _a ; i -3_ g_ ___;_yEEs=_rgg-=q=-6.;; ; =_ .' Wig =g= _.i g.;;_; ;.J_:__ ;.g.+- - uj m.m.m,c.m,3 m m.mu . ym uw wm. ;
^
4 ELEVATED RELEASE OF NOBLE GASES (100 METERSFe b- = -. vW mM'-T en A.,. m
,"w , 1
?":-:r-E-:'E #.=1=";~;~i- t_= .: ==r . ..jl2TEFK?Ay .i:- et s -W
.== a:.d 2.;-w ~f iE J
;~ ' =E 1 -
- - _. : = - ~~6:~i? E.- 3 ) '. : ..== .:?:-
1
- s. _~l=C-'.~.
- ..:=-_=-. - - -
.- _._= - ==-
= = = ,_ =
= = =; .c_z_:
- . -.=:.= .: x.=__=.u-u
=
.-- r- .u. - w.: _ :.- .a' -. a= :- -
l {,*. _' T='$' , , , __ , _
. _ - . _ _ _ . .-_ - . = . = - .
.-- - - _ _ . .- . -_=._=_=t==:., _ +- __
y _ + r_-2:. ,
)
-- DISTANCE FROM RELEASE POINT IN DOWNWIND DIRECTION (MILES)----- _
_~ _ _ - , , _ =i
, e m .- . . . em. m .e es =..l g, . u ,1 ,I 2
i i I i I i I
,f ,- . , ,
i e e a.
> g r ( 7 5 i a u it o 4 l l
l L. . . _ _ _ . . . . _ _ _ _ _ _ _ __.____________________________)
..
- 1, PM E PC
,q u g. ;$3i g ,
4 4 2 - r a __..=- IiW \ 4 = 4 i e a s - - m _ _- ._ q:- a e _ __ : __.,m, _ y n T kw : L-a : _i g i ai s: " = - 7 __4 ._ _ _ .
-==- -
-=, FIGURE 2 3
ll.: L i(l l! @ H d I I I I ! 9 ! l ! EXPECTED WHOLE BODY RADIATION DOSE (REM) FROM RELEASE OF 100i NOBLE GASES 5= :::i = M i _ f . _ _ . l (1 HOUR DECAYED AND 5 HOURS DURATION 0F g RELEASE) FROM 3412 MWt LWR VS. DISTANCE -- a : r-e NOTES: i: li;;;;;; ((:::'i gog 'tc 1.release.Graphs Great assume one hour holdup and decay prior l l ' l l l l l ;; i l 1 10wer doses (e.g..er delay in release can prodace 000 as much as a factor of about y == u ie 30 at one mile for 12 hours of inreactor holdup
. . ; _= __
- ,=:
comparedtoonehour). 1 aa z= -
- = =
b
- 2 8
%. ]i,,,{
gi ,, j ] g,[ Dose estimates are based upon MCCS computer -
\
ut i:$__
=
2 codecalculationsusingrevised(relativeto l 1
),
, y.= =~
== I!L _. CRAC and CRAC2) meteorological sampling models.
_ - n_ .- { 55 =~ _ - 0 5
~~
- 3. The likelihood of exceeding the - i? .- _-m --
s estimated 95 percentfle dose is less 4 - : *
- than Si given r
__._4 noble gases as'eleast ofabove. 2001of g' ' :g' ', t specified
,l i ll; h
gy 5 gj l i { j l iI I J l I ii( l { l1 ll il 1 I
, .' < , ,,,,';; TFFQ o I
l 1 l I h [g* @ M l I I I I I I I iOO I I I I I I I I N g
*" T
,. :S~ 1- -
siand!!n,= : g =b b b b======ssissam 11 . m 1 n a m _ _ , GROU 10 METERS i ite: g(j N t
==e 1
k .B-
]2 BBRIMERREf 1- 2 9
=s NW ESU {fAsrw-00 METERS N E li F a= 2 E_
3 -- -
* '= -
-; e
_5- :- E :5- .: 42
*3 -r -- 5= _
.2 5 .i- 5 -- 'i ~ =? ?
I !! a ;-+
. c. -
-u .:
5 :.; 5 m .:
- ; 5- g E E E w
9
- -2 =: :_ : : :. : :: a :
- =
.i
=- - '
u. g: r
,: = - .. . - w .-
3lg .- Jj q ;A ; ; :M
- ' ;K:: : .:.:. n .:: ,.:.='.:: : u:
I 5F= ;q ; ;: % :; :
; .L.
- W ;: . :::: C;
- ,,.A.m; i;i.>;;;, ,; ; ,;; .=.,
- ...; ., >>.m>>>'
f $. h
< iiii ii - i>>
I _ 'n 4 v; . -.I_ _ t s. J
,g3 t-==ta us == n=ag u :w s4 x :
.r".=
1;p1-=
= " "
- u
"" - 4; w:-
1 o .- t
!?-
m
* >=.-s
== - _ - -
'g =u=7 ?_" -J-=
s s: = iE i i .i ii Ez ii r! ! ~ ' _M GROUND m: _ LEVEL RELEACE 10 METERS =7 ~~
=
i=i 5 ggg e:--
- E-.
=E E = :-- =- : 2~
E-m . E
.g____
'3 g - ;:i=? -s
~
^__ j w===
7_=9
= --- - Ei; ; i ? - s-- - - -
i inm n ed,-e, _ Mn cw w g J gA4;;q M;;ggg; ; 7
-- =C_=-
Sis n
'~ '
= _x.
g ELEVATE 0 RELEAS[100 METERS : == =: ==- = ~= = _ _ _ _--_ _ :=
=
== %.__-___.- - - . _ - - - . - -
E :E: E EE a =w -:::* :::: ::: == ::: =:-
- =: ::: =: ::: == :::== ==: :=a:: ::: ==
==
_5_g-4}-:$__ . _ r. __ _ . w_,.__-_.r :: :: ::: :: ::: :-: ::P: _y _ _-- F
.!: ll . 4 - - _
11 l DISTANCE FROM RELEASE POINT (MILES) ;,;,,7,;;l;;;;;; i e i ,
- i n i in;; ;;;; ;;
i, ;,m;; H'g N' t a~3 i ii i+ i i ii5i i~ i Ge i i i7 ii i i8i ii i 9n i at i s iti m 'ito< < i= ' ' 15 M _ - . . . . - ~ ~ . - - ~ ~ - - - - - -' i . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
k VERNONT YANKEE NUCLEAR POWER CORPORATION RD 5. Box 169 Ferry Roa1 Brattleboro. VT 05301 % ,o
/ ENGINEERING OFFICE 1671 WOACEST ER ROAD FRAMINGHAM. MASS ACHUSET15 01701
} T E L E *MONE617 8'2 8100
,,/
December 19, 1986 EYY 66-117 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Office of Nuclear Reactor Regulation Mr. Robert M. Bernero, Director Division of BWR Licensing
References:
(a) License No. DPR-28 (Docket No. 50-271) (b) Letter, VYNPC to USNRC, FVY 86-81, Vermont Yankee Containment Safety Study, dated September 2, 1986 l (c) Letter, USNRC to VYNPC, NVY 86-218, dated October 24, 1986 1
Subject:
Vermont Yankee Containment Safety Study - Responses to NRC . Request for Additional Information
Dear Sir:
In accordance with your letter of October 24, 1986 [ Reference (c)) _ regtesting additional information concerning the Vermont Yankee Containment l Safety Study, responses have been prepared and are enclosed herein. These responses substantially reflect the discussions with your staff at a public j meeting in your offices on November 17, 1986. Comments and questions received at that meeting have been incorporated into the enclosed responses for completeness. Vermont Yankee is continuing its efforts aimed at closure of containment J perfomance concerns during severe accidents. Toward this end, we have summarized below the current follow-up analyses and engineering studies which are already in progress:
- 1. An engineering scoping study has been completed which examines the feasibility of improving the reliability of the valves located in the Reactor Building associated with the alternate spray path. The proposed modification is currently under review, and a final decision regarding i. implementation and schedule will be made in January 1987.
- 2. Analysis and calculations are being performed to prove the viability of the diesel fire pump to provide water via the service water, RHR service water, and RHR System to either the reactor vessel or containment spray header for core cooling. This analysis is i scheduled for completion in mid-1987.
j ) l I
l'. 1 i United States Nuclitar Regulatory Commission December 19, 1986 Attention: Mr. Robert M. Bernero page 2 l [ .. 4
- 3. An engineering scoping study has been initiated to evaluate the l feasibility of providing a hardened, reliable vent path from the containment wetwell to the plant vent stack. This study is scheduled for completion in the spring of 1987, and a final decision on implementation of any resulting modifications will be made by July of 1987.
Vermont Yankee is committed to the pursuit of the highest level of containment safety and perfomance at the plant that is practical. We believe that our Containment safety Study efforts to date and continuing actions will !. assure that this commitment is achieved. We trust that this information is sufficient; however, should you have any further questions, we are ready to meet with your staff again or discuss any aspect of this information at your convenience. Very truly yours. VERMONT Y,ANKEE NUCLEAR POWER CORPORATION
.f-R. W. Capstick Licensing Engineer RWC/bam Attachments 4 l
t a I i 1
. 4
Iy r 9
-1 ENCLOSURE 1
. VERMONT YANKEE CONTAINMENT STUDY RESPONSE TO NRC REQUEST FOR-ADDITIONAL INFORMATION g ' QUESTION'1 What is year estimate of the overall uncertainty of conditional containment failure probability and its basis?
RESPONSE
The scope of the Vermont Yankee conditional' containment failure-probability evaluation was defined to provide a "best estinate" value for the containment conditional failure probability. Because of this scope, a quantitative uncertainty evaluation was not performed. -In fact, a comprehensive uncertainty evaluation of the containment conditional failure probability has .not been performed in any pubnished BL'R PRA. Current on-goit.g work by the NRC in the NUREG-1150 Program and other programs (e.g., MELCOR) may provide greater insights into this area. The NRC's tatement from Page '3 that "the licensee's estimates appear optimistic considering the unc*rtainties..." is not appropriate. The containment conditional failurt probability point estimate was derived using a realistic "best estimate" analysis, reflecting the best information available. This estimate, therefore, is considered neither optimistic nor pessimistic. The statement also implies an attempt to chose optimistic outcomes during the core melt progression. On the contrary, an attempt was made to use both IDCOR and NRC developed codes and assumpticos to provide a balanced assessment. The outcome of that assessment is presented in the Vermont Yankee Containment Safety Study as. the "best estimate" which may be characterized as a mean value, net a median value. 5104R 1
QUESTION 2
- What is the effect on core' damage frequency'when accident sequences TPUV, TPQUV,' TE. 0
- E PQ ..T,PQUV, TFW are included in the dominant accident sequences based on reduced battery life, the number and type of SRVs' compared to Peach Bottom, and on the-CCFP7 (Note: The. question originally included'T,QUV and not T,PQUV;-however, since the question appears to be directed at stuck open valves, we believe it should have ens been T,PQUV.)
RESPONSE
-There are a number of considerations associated with stuck open relief valves leading to accident sequences. These sequences could potentially
~
be of sufficiently high frequency to result in a change in the character of the dominan* sequences in a plant-specific analysis.' However, many of these considerations were examined in the Vermont Yankee evaluation. A discussion of each item follows. o Type of SRVs Both Vermont Yankee and Feach Bottom nave Target Rock three-stage relief valves. This is identified on Page A-4 of the Corstainment Safety Study. Because the designs are the same no change was made to the surrogate plant quantification for stuck open relief valves (SORVs). In addition, the Vermont Yankee experience with their, modified three-stage SRVs is substantially better than that for Peach Bottom. Therefore, the Vermont Yankee plant is less susceptible to the postulated SRV failures than those from the surrogate plant. o Number of SRVs The number of SRVs is different between, Vermont Yankee and Peach Bottom. This is identified on Page A-L of the Containment Safety Study as follows: 5104R
i$ l q; < No. SRVs No. ADS
-Total Valves
-Peach' Bottom 11 5 Vermont Yankee 4 -4 There are two effects which could impact. the sequences identified by the NRC: ,
- The smaller number of SRVs results in a lower likelihood of an-
. Inadvertent Open Relief. Valve (IORV) given all other considerations are equal. This would reduce the frequency of sequences 'such as, T gQW ,
- The smaller number of SRVs should-also have a.similar impact on the'SORV conditional failure probability. This is because more SRVs could be called upon to operate following a transient-at c Peach Bottom than at Vermont Yankee.
Therefore, the smaller number of SRVs would appear to reduce the impact of the sequences identified by the NRC. However, these considerations were not factored into the Verment Yankee cuantification. o- Secuence-Considerations
?
n The IPE trethodology treats similar sequences together. Specifically, the IPE methodology bins together accident sequences caused by an 10RV and those which result from a transient induced SORV. Therefore, the Vermont Yankee Containment Safety Study includes sequences from both these categories. The sequences are labeled as T QW y
'.in the Vermont Yankee study and appear in Class ID. The Ty QW sequences in the Vermont Yankee Study include:
TyQW TC 0 f TgPQW j +- _3 5104R
T TPQUV T FPOCv Therefore, these sequences are addressed'as suggested by the NRC. Please refer-to the BWR IPE methodology for further discussion of this binning scheme.
~
- o. IStation Blackout SequencejLT E
- E The BWR IPE methodology addresses these specific issues on Pages 4-42 through 4-44 The conclusion from the Bk1 IPE is that' sequences of this nature have frequencies in the range of 3E-7/yr. qThe NRC is correct that these sequences should be reassessed for Vermont Yankee' and shown in the tables summarizing the accident-sequences. This reassessment for Vermont Yankee indicates that a frequency;of approximately SE-8/yr should be added to thatireported in the Vermont Yankee Containment Study for the total core melt frequency and for Class %B. This increase represents an increase of:
E 0.16% for total frequency
'O.81% for Class IB This would not alter the qualitative or quantitative conclusions of .{
the evaluation. l o Reduced Battery Life ll The battery life used for Vermont Yankee ic judged to be equal to or f l greater than that used for Peach Bottom and not to be a factor in the I assessment of any of the sequences identified by th:: NRC. -l
.i
-)
i o Recovery of RHR Pumps (TPW) j TPW sequences are long-term loss of containment heat removal sequences. These sequences can be effectively mitigated by any containment heat removal pathway, i.e.:
-L-5104R
RHR Main condenser
- Containment venting It can be shoan that whether an SORV exists or not would have little or no impact ca the course of this.particular sequence. Therefore, the consideration of TW sequences is sufficient to account for the frequency and source term impacts associated with TPW sequences.
l l l 5104R
Ol'ESTION 3 - Given that the. national' average value for frequency of loss of off-site power.is on the order of 0.22/yr, justify on the basis on the Bayesian estimste that the frequency of loss of off-site power at Vermont Yankee is 0.07/yr. RESPONyE s of ff te po e s 0 07 e n yea ba ed n e m nt Yank e s record of having no total loss of of f-site power events in over 14 years-ofioperation. In Appendix D of the study, " Evaluation of Vermont' Yankee's Electrical Power System Capability Relative to Station Blackout " we estimated the frequency of total loss of off-site power using the NLC's recommended methodology in NUREG-1032, " Evaluation of Station Blackout Accidents at Nuclear Power Plants." Using the NRC's methodology, the frequency is 0.05 events / year. i The NRC states that industry average value for lots of off-site power is 0.22 events / year without providing the reference. NUREG-1032 states that the frequency of loss of off-site power is about 0.1 events per year and in Table 3.1 provides a summary of the data on total loss of off-site power events through 1983. Table 3.1 indicates that the frequency of occurrence of loss of off-site power is 0.088 events / year. We also disagree with the statement that the Vermont Yankee frequency of loss of off-site power should be adjusted using Bayes theorem. Again, NUREG-1032 states " design characteriscies, operational features, and the location of nuclear power plants within different grids and meteorological areas can have a significant effect on the likelihood and duration of loss off-site power." Thus, site-specific analysis and site experience provides a more accurate estimate of frequency of less of off-site power than generic or average data. The estimate used in the Containment Safety Study of 0.07 events / year'is more conservative than the f requency as calculated using NUREG-1032 and does not deviate significantly from industry experience of about 0.1 events / year. 5104R
t
.t; -
b -QUESTIONf Verify'that the total battery capacity available at Vermont Yankee'is greater than 2,175 ampere-hours, and that it could be_ maintained at a
~
voltag'e greater than.l.75 volts / cell in high ambient temperature during the accident for six to eight hours.
, RESPONSE .
The Class 1E 125 volt station batteries consist of two 2,175 ampere hour station batteries'and one 330 ampere hour station battery (all ratings at the eight hour discharge rate). In addition, a non-Class IE Alternate Shutdown System battery rated at 495 ampere hours is available to supply the RCIC System loads. Therefore, the total 125' volt system battery capacity is 5,175 ampere hours. We have conservatively calculated the de load requirements and have included motor inrush currents. Our calculations indicate that the required load can be supplied for eight hours with the available sources of battery capacity with cell voltage remaining greater than L
- r o
'e 1.75 volts / cell.
The ambient temperature for the batteries is not expected to-increase at all during the station blackout scenario; however, if the temperature did m increase, the effect would be to increase battery capacity not reduce it. ..
.. l l}
In addition, note that a battery rated at 2,175 ampere hours at the eight
]
hour discharge rate can supply 271 amperes for eight hours, 603 amperes ] for three hours, 1,221 amperes for one hour, or 2.660 amperes for one minute. In a battery calculation a load profile is created for the .i requireil duty cycle. For each interval in the duty cycle, the battery-discharge is calculated and the calculation verifies that the battery can support the load profile for the entire duty cycle. ) l-5104R I 1
l 0UESTION 5 How often are the RHR/RHRSW interconnecting valves actuated to assure that the valves work properly? RESPONSE q The RER/RRRS =' interconnecting valves are stroked monthly ("RHR Valve Operability Surveillance," OP-4124) to assure that the valves work properly. l 5104R
QUESTION 6 How often are the interconnecting valves between the RERSW and the Fire Protection Syt. tem (fire pumps) actuated to assure that the valves work properly?
RESPONSE
The interconnecting valve, SW-8. between the Fire Protection (FP) System and Service Water (SW) System is tested annually (" Surveillance of FP Equipment," OP-4020. Page 45). All other interconnecting valves between the SW Syttem and the RHF.SW/RHR crosstie valves are normally open. A review of our maintenan:e records has indicated that there has been no maintenance required on this valve since the plant has been operating. i i l 5104R
QUESTION 7 j How readily can the MSIVs be reopened following closure at operating conditions? What interlocks must be bypassed and how complicated are the procedures (e.g., Must the differential pressure across the MSIVs be reduced for the valves to be reopened?)? EESPONSE 1
- a. How readily the MSIVs can be reopened is determined by both the cause for MSIV isolation and by present overall plant conditions. The MSIVs are closed upon any of the following conditions:
n 1
- 1. Rx Low-Low Water Level - 82.5 Inches
- 2. Main Steam Line High Radiation - 3x Normal Full Power Background ,
(NFPB) I
- 3. Main Steam Line High Flow - 140% (Rx Mode Switch in RUN) or 40%
(Rx Mode Switch in STARTUP, SHUTDOWN or REFUEL) 4 Main Steam Line Tunnel High Temperature - 212 F
- 5. Main Steam Line Low Pressure - <800 psi (Rx Mode Switch in RUN) ,
o
- 6. Low Main Condenser Pressure - <12" HgA
- b. Under emergency conditions, the operator is permitted by the Eraergency Operating Procedures (EOPs) to bypass all valid isolation signals e
exceot 2 and 3 provided the main condenser is available as a heat sink. Under normal operating conditions, the operator is directed to wait until the cause of the isolation is cleared. Conditions 5 and 6 may be cleared by the operator by placing the reactor mode switch out of RUN and by use of a keylock bypass switch
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.t< v .
E - on:a Control Room back panel,'respectively. All other isolation conditions can'be' bypassed by initiating lifted lead'and jumper-procedures'(Appendix, OE-3100). The following five-step procedure n must then be carried out: e . E '
'1.
Place valve control switches (total of 11) in the CLOSED or SHUT position to satisfy the inadvertent opening / reset permissive relay logic.
- 2. ' Reset the PCIS Group I isolation logic'(three-position switch).
- 3. Open the outboard MSIVs (4) via control switches.
4 Open MSL drain valves (2).to equalize upstream and downstream pressures to within approximately 50 psig. (50 psig'is a procedural requirement, allowing a safety margin from the 200 psig, design value recommer.ded by the manufacturer. The reason for equalizing pressures is to minimize the force that must be overcome (spring and delta-P) to reopen the valve since it is designed to close.in the direction'of flow.) , 1 1
;5. Open the' inboard MSIVs (4) via control switches.
We believe that this five-step procedure can be carried out, on average, in approximately two (2) minutes. 5104R _ _ _ _ _ - _ _ _ - _ - .w
l QUESTION 8 It is not clear how the CCFPs given on Page 74 of the report were obtained. Please explain. R_ESPONSE The basis of conditional probability of early containment failure (CI) for each class of accident is as follows: Accident Class CI IA 10'3" IB 10-3a IC b
.31 ID ~"
10 1 II 1.0 C III 10
-3a l IV 1.0*
Notes
- a. Failure mode was estimated to be hydrogen burn. The probability is based on a Shoreham PRA study.
- b. Failure probability was estimated as the fraction of accident sequences in Class IC that would result in elevated containment pressure (>40 psia) before reactor vessel failure,
- c. Given the large probability of drywell failure by overpressure (see containment event tree for Classes II and IV).
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I N QL'ESTION 9 What SLCS modifications are proposed for Ver'mont Yankee? Page 86 discusses the advantages of two differe:.t possible modifications, but gives no commitment to either.
RESPONSE
At the present time, the preliminary design proposed for Vermont Yankee is to utilize two pumps to inject the VY equivalence of the 86 gpm, 13% sodium pentaborate solution required by the regulation. However, a final design has act been approvet as yet. Our present commitment to address Item 50.62(c)(4) of 10CTR50.62 is outlined in our letter, dated September 29, 1985, titled " Generic Letter 85 ATWS Compliance Schedule (10CERSO.62)." The letter states that Vermont Yankee will implement a design or operational modification of its Standby liquid Control System during its second refueling outage after July 26, 1984 (summer 1987 outage). 5104R
QUESTION 10 - Identify the testing and maintenance requirements you use for the diesel-driven fire pump. Do these requirements conform to those contained in the National Fire Codes? Also, identify any reliability information for the a,ystem such as outages and failures to start on demand. ht outage time limitations do you use for the systera while at power?
RESPONSE
- a. The testing and maintenance requirements for the diesel-driven fire pump are:
- 1. Diesel Fire Pump (DFP) starting battery - weekly - electrolyte level, voltage.
- 2. DFP operational check - monthly - lube and fuel oil, auto and manual starting.
- 3. DFP operational performance and capacity - annual - pressure, flow rate, pump protection, and elarm circuitry.
4 DFP starting battery - once per cycle - cleaning and inspection.
- 5. DTP - A0 rounds (each shift) - battery charger and fuel oil level. I
- b. Although the above testing and maintenance requirements do not conform i
to the 1985 edition of the NFC for diesel-driven fire pumps, they are consistent with Vermont Yankee's maintenance and surveillance practices for safety-related equipment. We believe this is a sound alternative and assures an acceptable amount of reliability.
- c. The FP System is considered operable with:
l l i l 1. Two fire pumps operable and lined up to the fire suppression loop (Note: one electric, one diesel-driven). l I i 1 l 5104R ( I r l
- 2. Water available from the Connecticut River.
- 3. An operable flow path capable of taking suction from the Connecticut River and transferring the water through the distribution piping with operable sectional 4. ping control or isolation.
.The table on the following page indicates those periods when the diesel fire pump has not been available.
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._ ___ ____ ______-_____ _ u
n..
-)' ;. *
'Date Duration Reason 5/14/80? 2.0' Calibration IT 6/23/B0 6.0 hrs Preventative Maintenance'(PMs) 10/06/81 .1. 5 Check operation of. generator t
,4/2i/82 8.0 PMs 5/21/82. 7.0 Battery charger 7/29/82 6.0 PMs' 3/03/83 3.5 Ex repair 8/23/83 7.0 PMs 12/08/83 1.5 (?)
7/24/84* 8.0 PMs 12/06/84 8.0 PMs 6/28/P,5 6.5 Replace battery terminals 8/09/85 1.0 DFP battery 10/29/85* 2.0 Replace exhaust mufficr 2/10/86* 2.0 (?) 6/11/86* 1.5 (?) TOTAL DURATION: 71.5 hrs NOTES
- 1. This information pertains to the period 6/28/B0 to 6/11/86.
- 2. Dates and durations for occasions DFP is 00S due to failure-to-start (2) are not included here, but are in the following table.
- 3. The dates asterisked above indicate those times that the DFP was unavailable while the plant was not on-line. The total duration of DFP unavailability is thus reduced to 58.0 hours if those periods when the plant is not operating are excluded.
! l :
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l l 4 The following information indicates the reliability of the diesel fire pump: l A. FAILCRE TO START 1 Date Reason l 8/10/76 Starting solenoid relay 9/30/65 Poor battery connections B. STARTING BATTERIES DEGRAD_ED D6te Reason 5/13/80 End cells on one bactery; battery charger timer 9/30/85 Bad connection to one of two starting batteries 1/07/66 Low voltage both batteries; blown battery charger fuse l
- d. With the Fire Suppression Water Supply System inoperable, a backup Fire Suppression Water System must be established within 24 hours, or j the reactor must be placed in het standby within the next six hours and cold shutdown within the following thirty hours.
1 5104R
QUESTION 11 Identify the scope of modifications required to the spray system, or increases in the pumping capacity, to assure a uniformly distributed spray with proper droplet size (as opposed to a dribble) if the diesel fire pump were used in a core melt event. Approximately what would the costs be of ! such modifit:ations?
RESPONSE
At this time. Vermont Yankee has not determined that any modifications would be required to assure an adequate spray pattern. We are reviewing the test data (pictures) from Monticello and will further evaluate the adequacy of Vermont Yankee's spray pattern to ensure that use of the diesel fire pump will be beneficial in a core melt event. l 5101.R
QUESTION 12 Can portable AC generators be used effectively to power vital valves and/or small pumps for station blackout accidents? If so, what modifications would be required, and what would be their approximate costs?
RESPONSE
- a. The most " vital" valves under station blackout conditions will be those which are required by steam-driven pumps (HPCI, RCIC) supplying makeup to the RPV, and those valves (SRVs) which provide 4 means for removal of the bett generated by the core (e.g., decay heat, assuming a successful reactor scram). Of next importance are those valves which would permit venting or spraying of the primary containment to preclude overpressurization from the heat load imposed on it as it performs the function of primary beat sink throughout this accident scenario.
j Primary containment vent valves are air-operated wita ac solenoid valve control (fail close upon loss of either ac or low air pressure). Local manual operation of these valves is considered the most effective of various options considered provided it is performed precore melt. It has been targeted for further study, as discussed in the Containment Safety Study. l The containment spray valves are ac powered motor-operated valves. A conceptual study is underway to investigate options to aid the operators in aligning the diesel fire pump for vessel / containment injection.
- b. Modifications and inventory requirements to support these options will be developed if design change options are pursued.
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. QNESTION 13
'In Section 2.2.1 you conclude' that the containment can be " expected to withstand
- pressures approximately two times design prior to f ailure."
Provide t'he. bases for:your conclusion.
RESPONSE
The allowable stress limit used in the primary containment design (Reference A) was 17.5 ksi. The minimum yield strength of the shell , material (SA516 G70) is 38 ksi at 100 F and 33.9 ksi at 281 F (design temperature). Therefore, the factor'of safety between yield stress and allowable stress is 2.17 at ambient temperature and l'.94 at design temperature. When primary yielding is' conservatively used as a failure . criteria, it can be stated the primary containment can'be " expected to withstand pressures approximately two times' design prior to f ailure." In acdition, the BWROG Subcommittee on Mark I Containment Ultimate i Strength Analysis is presently involved in an effort that will attempt'to 1 l determine the actual vessel failure pressure for Mark I containments. The subcommittee will also attempt to show that early overpressure containment failure under severe accident conditions is unlikely. l Reference A: Vermont Yankee Project - Containment, Contract 7-6202, General Design Criteria, Revision 1, Chicago Bridge and Iron Company, February 25, 1969. l l 5104R
QUESTION 14 In Section 2.2.10.3. is the water supply from at least a portion of the cooling powers also available? 3
RESPONSE
i o The deep basin under the west cooling tower is available to parts of the Service Water (SU) System via the RHRSW System. This system would provide cooling for the following emergency cooling loads: o RHR Heat Exchangers o Emergency Diesel Generators o RECCW: Recire Pumps
- CRD Pumps Fuel Pool Heat Exchangers
- PC Air Compressor RHR Pumps Rx Building Sump Coolers o ECCS Corner Room Coolers o Station Air Compressor (s)
In addition, the deep b6 sin is an available source of water for vessel injection and drywell spray via the RHR service water pumps. l l l l l 5104R { l l I . l
)
OUESTION 15 The use of the Vernon Hydroelectric Station is referenced in Section 2.2.11.1iand discussed in more detail in Section 4.4.2.3. Reliability estimaterd are presented on Page 62. Please provide the basis for the - reliability estimates with reference to both the historical operation oft Vernon Hydro, and the transmission line and substations to Vermont Yankee.
RESPONSE
In addition to Vermont Yankee's diesel generators, a direct line from the nearby Vernon Hydroelectric Station can be aligned manually from the Main Control Room to either of the emergency buses. The loads of either emergency bus can be met with this supply. The ten-unit Vernon Hydroelectric Station is located less than one mile from Vermont Yankee. A dedicated, normally energized, insulated tie line can be connected directly to either Emergency Bus 3 or 4 via remote manual breaker operations from the Vermont Yankee Control Room. There is a direct telephone circuit between the Main Control Room and the Vernon Hydroelectric Station. Use of the Vernon tie is part of operator training and is well known to the operators. The unavailability of the Vernon tie to supply either of the emergency buses can be written as: U=H+0+V+C where: H - represents the unavailability of the hydroelectric stations and l any required active breaker transfers at the hydroelectric plant. 0 - represents operator errors defeating the successful connection. l l V - represents active and passive electrical system hardware failures l at Vermont Yankee. 5104R
C - represents those common events, such as e::treme wind conditions, I that cruwe loss of power at Vermont Yankee and defeat the Vernon tie line. Determinatica of H 1 The best available information indicater that the Vernon Hydroelectric Station was unavailable for a total of 2 hours and 24 minutes in a 21-year
* -5 period making the average unavailability 1.3 x 10 ,
According to the past history of the Vernon Hydroelectric Station, the tie-line was taken off-line for approximately two days as a result of scheduled maintenance at the substation switchyard. (This scheduled maintenance occurs about once every three years.) There were two other nonmaintenance-related events which affected the tie-line availability: one related to the construction of the fish ladder at Vernon Hydroelectric Station and the other related to the regrading of soil in the vicinity of the ductbank near the Vermont Yankee cooling tower. In summary, the total unavailability of Vernon Hydroelectric Station tie-line due to reintenance or other events is about 2.7 x 10~ (approximately 15 days in 15 years). In response to a grid collapse, the hydroelectric station would have to separate from the grid to allow the tio line to remain available. The only active action identified is the automatic opening of a single normally closed feed breaker. The probability of failure associated with the opening of this normally closed feed breaker is 6.5 x 10 '. This is
~
based on the Seabrook Probabilistic Safety Assessment. Therefore, H can be approximated as:
-5 + 2.7 x 10 -3 l H = 1.3 x 10 + 6.5 x 10~
= 3.4 x 10
-3
- It should be noted that only two events were recorded: one of 2 hours and 20 minutes and one of 4 minutes. The station recovered quickly from the latter event which was initiated by a lightning strike.
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Determination of 0 The estimates for operator inappropriate action are as follows: 4 _0_*
- Phase I 2 0-2 hours .1 -
Phase II' 2-4 hours .05 Phase III 4-10 hours .01 } Phase IV 10-24 hours .01 The information above had to be taken from the referenced material because no Vermont Yankee-specific human error probability was estimated for this study. Determination of_y Two breakers have to close to feed either bus. In addition, it is assumed that the diesel breaker must open (this is conservative since failure of this breaker to close may have been the cause of " diesel failure to supply emergency bus"). Therefore, V can be approximated as 1.7 x 10~ . Determination of C l The factor C reflects those loss of off-site power events that would also render the Vernon Hydroelectric Station unavailable. A review of 114 off-site power events identified 24 that were caused by extreme external phenomena (e.g., lightning, ice storms, heavy snow, tornadoes, etc.). Events such as saltwater spray and Florida grid instabilities were assessed not be be applicable to the Vermont Yankee site. C could range from 4 x 10~ to .1. The values above are taken from the following documents: 4 (1) BWR Individual Plant Evaluation Methodology. I l
* (2) A. D. Swain and H. E. Guttman, Handbook of Human Reliability Analysis With Emphasis on Nuclear Power Plant Applications, NUREC/CR-1278.
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The upper estimate (0.1) is used for the point estimate quantification in this analysis. Suma ry 4 The'unavdilabilityof.VernonHydroasaneffectiveACpowersourcetothe emergency buses given e station blackout is: . U=H+0+V+C U = 3.4 x 10
-3 + 0 + 1.7E-3 + IE-1 Vernon Hydro Unavailability For Extreme External Phenomena Events (H + 0 + V + C)
Phase I 0-2 hours .2 Phase II 2-4 hours .15
. Phase III 4-10 hours .11 Phase IV 10-24 hours .11 5104R
Q QSTION 17-In Section 4.1;4 MARCH /RMA and'MAAP code' package for Vermont Yankee is. ref erencegd. Were calculations made for Vermont Yankee, or were the cresults of computation for other reactors evaluated for the Vermont Yankee 2
' design? % at calculations were made?. .
i s
RESPONSE
3 a). Calculations were made specifically for Vermont Yankee utilizing the-above codes. The different calculations and the code utilized for. each are detailed below. b) The following2 accident sequence analyses were performed specifically for Vermont' Yankee using the MARCH /RMA code:
- 1. T,C,C2 (ATWS sequence).
- 2. Station blackout and further failures to the HPCI and RCIC Systems
'or.their support systems.
- 3. Station blackout for more than six hours, coupled'with failure of-HPCI, RCIC, and the diesel fire pump.after.six hours.
The following accident Ftquence analyses were performed specifically for Vermont Yankee using he MAAP code:
- 1. Station blacknut fot more than six hours, coupled with failure of HPCI and RCIC and the diesel fire pump after six hours.
- 2. Station blackout and further failures to the HPCI and RCIC Systems or their support systems. Water injection became available at the time of vessel failure.
- 3. Similar to (2) but without the addition of water on the debris after vessel failure.
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2 I'
. - QUESTION 18 l4 1
In-vessel and ex-vessel steam explosions' were not considered credible
~
(Page 55,i first paragraph)' based upon research. Identify the research that'forqs the basis'for this cocclusion. - i e
RESPONSE
IDCOR has studied in-vessel and ex-vessel steam explosions [1], [2] and found that the energy transfer mechanisms sufficient to threaten the l' reactor vessel or the containment of current LWRs do not exist. 'Also, y information published in NUREG-1116 [3] tupports the conclusion that the
' loads from steam' explosions which potentially might fail the containment (Alpha mode. failure), are thought to be sufficiently unlikely to be neglected in source term determination.
a)' In-Vessel Steam Explosion
'The energy release from large scale steam explosions has been hypothesized.to be sufficient to cause containment failure.
l Generally..this has been conceived to result from in-vesst, steam explosions which fail the primary system and subsequently the containment. The issue to be addressed is whether large scale steam explosions of this magnitude could occur durin'g degraded core accidents. A review of the literature regarding in-vessel steam explosion was performed to allow quantitative characterizatier. of this potential containment failure mode [1-10]. The principal sources of information used in the Vermont Yankee assessment and the conclusions drawn from ') those sources are as follows: o IDCOR Reports [1, 11, 12] o Steam Explosion Review Group (SERG) [3] 5104R 1
)
)
1
- . - - - - ______-__-____________-______________a
),
u
- i. . .
J* ol .American Physical Society' Study Group,[6] g .
.o Mark 1 Contal'n ment Event' Tree for Application'in NUREG-1150 [9]
? -
. 4 d
U. 1 . E _ Based upon the1information presented in these sources, the conclusion ' l 2 . l . was reached that the probability' of .a steam explosion suf ficient to , ,i 1 cause RPV and' containment failure is small. In. addition, the failurt probability at elevated RPV~ pressure'may be zero and at' low pressure l may be IE-3-to zero depending upon the modeling' assumptions. Most of-the sequences in the Vermont Yankee analysis would be at high pressure i l and therefore a value of 1E-4.is'used consistent with Steam Explosion Review Group (SERG), a flow type melt model (MELPRI). ano the lack'of
' sufficient energy released'in a very short time.
L b) Ex-Vessel. Steam Explosion. l
~
l It has been postulated that energetic steam exp1osions caused by molten material dropping into shallow water pools in the drywell-could lead to containment failure. For the purposes of the Vermont Yankee containment failure probability calculation, a review of published evaluations was made to establish the' current state of knowledge. The following sources were considered in the Vermont Yankee quantification: o IDCOR Reports [1, 11, 12] o American Physical Society [6] o Sandia Preparation of CET for NUREG-1150 [9] There appears to be lets general agreement regarding the possibility and impact of ex-vessel steam explosions as opposed to'in-vessel explosions. It does appeat, however, that the likelihood of a : sufficiently severe ex-vessel steam explosion to threaten containment is very small if the core melt process is a flow type. Therefore, a value'of IE-4 is used in the Vermont Yankee analysis. 4 5104R
L
, . References-
- 1. . Technical Support.for. Issue Resolution, Fauske and' Associates, FAI/85-27, dated July.1985.
~
.1
- 2. IDCOR Task 14.1. f
- 3. Steam Explosion Review Group (SERG), "A Review of the Current
~c . Understanding of the Potential for Containment Failure Arising From' In-Vessel Steam, Explosions," NUREG-1116 U.S. Nuclear Regulatory.
Commission, June 1985. 4 L. S. Nelson and P. M. Duda,' Steam Explosion Experiments With Single Drops of_ Iron Oxide Melted'with a CO Laser: Part II Parametric 2 Studies, NUREG/CR-2718, dated April.1985.
.5. J. B. Rivard, et al., Identification of Severe Accident Uncertainties, NUREC/CR-3440, dated September 1984.
- 6. Richard Wilson, et al... Radionuclides Releases from Severe Accidents at Nuclear Power Plants, Report to the American Physical Society, dated
' February 1985.
- 7. Snyder,'A. M., A Current Perspective on the Risk Significance of Steam Explosion, Vortrag Jehrestagung Kerntechnik, Mannheim, 1982.
- 8. Shoreham Nuclear Power Station Probability Risk Assessment, Long Island Lighting Company Docket No. 50-3, dated June 1983.
- 9. C. N. Ames, et al., Containment Event Analysis for Postulated Severe Accidants at the Peach Bottom Atomic Power Station (Preliminary Draft for Review), SAND 86-1135, dated May 12, 1986.
- 10. Reactor Safety Study, WASH-1400, October 1975.
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i
.r
. 11. IDCOR Technical Summary Report, November 1984
- 12. IDCOR Technical Report 14'.1B, " Key Phenomenalized Models'for Assessing Nonexplcisive Steam Generatien Rates," June 1983.
i
- .m I
5104R . A.
H -a QUESTION 19 As we understand Tables 4.7 through 4.10, th'e designations E. L, and NCF refer to early (E) or late (L) containment failure estimates, and NCF refers to,no containment failure. The second designators H,'M, and L referito'high medium, and low releases. respectively. To what extent can l" . mitigation through manual actions in the time available, and in the
- l l, temperature and radiation environments associated with such accident types 1-be expected to be successful for early failures; for late failures?
.Specifically.-for.the combustible gas control, spray, and venting evaluations, what do you judge the effe:tiveness of the existing plant and procedures to be.versus the possible improvements for early and late sequences? .
- RESPONSE.
1 - The probabilistic analysis results are provided in Tables 4.7'through 4.10 Land'take into account the. existing plant design and procedural guidance. Mitigation through manual actions as presently prescribed in emergency l procedures (based upon BWROG EPG, Revision 3) are designed to maximize the
. plant capability to deal effectively with all severe accident scenarios.
Manual actions for early containment failures are limited to ATWS sequences. For these sequences, the prescribed manual actions to reopen MSIVs, scram rods, and inject SLC are each considered highly successful means of dealing with the ATWS event and preventing early containment failure. In addition, if all of these actions are unsuccessful in the early phases of the event, the symptom based procedures provide guidance on controlling the plant to delay ultimate containment failure which will allow time to succeed in reactor shutdown. All such manual actions for preventing early failures are accomplished from the Control Room, which is a mild environment. Sequences that could result in late failures are equally suited to manual actions that can successfully mitigate the event. Containment spray, venting, and combustible gas control (via N CAD 2 System) can all be 5104R
1 i e. ,)j - i
, accomplished by manual actions outside the.Reacter Building'provided' power-is available.' For the station blackoutiscenario, the existing plant-design does'not'cupport remote operation.' Local operation of containment spray' valves in the Reactor Building is considered the most feasib1' e-containmtint pressure control method for' a sustained station blackout-
- situation.- 'The valves in the Reactor Building'can be operated prior to [
core damage and the ' diesel fire pump can be operated from outside the Reactor Building. . Additionally, Vermont Yankee is presently performing a
-conceptual design study to evaluate options to increase the availability of these valves under station blackout events. The N CAD and vent
~
2 valves are not' currently remotely operable under station blackout events. However,;the N CAD System would-not be expected to;be needed due to our inert containment. Although venting may be an option, we believe containment spray is more desirable. As.our engineering studies progress, we will determine which improvements- O are'to'be recommended. At this time, it is difficult to judge the relative-effectiveness of potential design improvements. 1 F 1 I i
;. 5104R L
o t t 4' g QUESTION 20 - NPSH during' spraying is identified as a concern on Page-ll7. To what extentwik1furtherinvestigationbeundertakentodeterminewhetherNPSR. is an isque? Verify that pro'cedures exist for the operator to line up - i ECCS water sources outside the containment in the event NPSH requirements
, are not met.- If analysis indicates it is an issue,'what-do you propose be done.to eliminate or reduce the level of concern?
P
RESPONSE
The ECCS systems were analyzed at full-rated flow conditions to determine what the available NPSH (ANPSH) for each would be with fruppression pool r water _ temperature' ranging from 60 F to 200 F. The analysis assumed that suppression pool water level was 6.5 feet which is sufficient to ensure submergence of all ECCS suction strainers. Torus
. airspace pressure was assumed to be 14.7 psia.
The results of this analysis were used to generate curve T/L-5 in OE-3104 (Torus Temperature and Level Control). This curve is referenced in Step T/L-14, which directs the operator to line up for injection those systems which take a suction external to primary containment if the combination of torus pressure and water temperature cannot be maintained above curve T/L-5. Our analysis indicates that the suppression pool water temperature has a marked negative effect upon ANPSH, particularly above approximately f 175 F. Efforts to quantify this effect and to provide additional guidance and corrective action to operators will be undertaken consistent with the incorporation of EPG Revision 4 into the Vermont Yankee Emergency ) Operating Procedurcs. For the range of accidents and transients which i have been analyzed and which form the design basis for Vermont Yankee, the maximum torus temperature is less than 175 F. 5104R
- I 0 .
Qt'ESTION 21 Venting is considered for the station blackout sequences only. Please discuss your rationale for not considering other, events when venting may be beneflyial. . 5 H a
RESPONSE
Venting is discussed for both the station blackout and ATWS severe accident sequences in Sections 5.4.3, 5.7.3, 5.7.4, and 5.7.5. Section 5.4.3 summarizes the results of our survey of two reports that provided considerable information relative to' venting (References (21) and (22)), as well as the analysis performed specifically for Vermont-Yankee (Reference (11)). Section 5.7.3 provides the recommendations regarding the NRC's integrated five-elcaent proposal relative to the station blackout event. The conclusion was that wetvell venting could be utilized for containment pressure control in lieu of containment spray, but further study was required to determine the most appropriate procedural / design changes to pursue. Section 5.7.4 provides the recommendations regarding the NRC's integrated five-element proposal relative to the ATWS event. The conclusion here was that a certain limited set of conditions may result in which venting would be beneficial, but it is generally not I advisable during an ATW5 event unless no other method of pressure control is available. In addition, further studies were recommended to determine the uncertainties / risks associated with venting as compared to other containment pressure control nethods such as containment spray. Finally, Section 5.7.5 draws the conclusions for the five issues proposed by the NRC. It states that, "the current industry position on venting suggests that it may be desirable to vent under long-term loss of decay heat removal scenarios, but only if no core damage / source term is involved. Vermont Yankee should perform additional plant-specific analysis to losure any decisions on venting ere based on a sounr; engineering foundation." l In an effort to provide a sound engineering foundation for venting at l Vermont Yankee, we have ir.itiated an engineering scoping study to provide the option for a hardened, reliable vent path from the wetwell to the ventstack. l 5104R
l e , , QCE5 TION 22 l l Since there is a substantial dif f erence between the heights of the VYNPS l ,4 l plant stack (318 feet) and the reference plant stack (500 feet), indicate how this pas taken into account in the comparison of the two plants. l 1 e
RESPONSE
Stack height was not a consideration in our study. The focus of our study was on the containment failure probability. Only plant characteristics that inf1tance containment integrity were evaluated. As noted in our response to Question 23, source term and off-site consequence considerations were outside the scope of this study. However, information on the relative differences between ground level and stack releases has been provided in the response to that question. 510LR
OrESTION 23. Evaluate Jhe' differences >in off-site. dose consequences due to venting at ground level versus through the stack. Using site-specific meteorology
>- addtopogkaphy,provideanestimateoftheoff-site' dose' differences betweenthetwotypesofreleasesasa'functionofdistancefromthesitb.-
RESPONSE
Although not a part of,the Vermont Yankee containment study, the following data is provided as. insight'to release characteristics at the Vermont Yankee site. The data was generated utilizing the 1985 on-site' meteorological data'and the SKIRON-II Accident Dispersion Analysis Computer. Code.
=The differences in off-site' dose consequences due to ground' level venting-versus a primary vent stack release can be estimated by comparing the atmospheric diffusion factors (CHI /Q values) from these two relesse l pathways. Cumulative probability distributions of hourly CHI /Q values at the site area boundary (0.16 to 0.44 miles) and at cistances of one, two, three, and four miles from the site were generated for both release pathways using site-specific meteorological and topographical data. 'The median values from these probability distributions are presented below:
Median CHI /O Value (sec/m ) Distance Ground E,t tack Ratio (Gnd/Stk) 9 Site Boundary 1.4x10~' 7.2x10-l' 1.9x10
-5 -6 1 Mile 2.5x10 2.2x10 11,3
-6 -6 1.6 2 Miles 9.9x10 6.3x10
-6 4.8x10-6 1.2 3 Miles 5.8x10
-6 3.5x10-6 4 Miles 3.9x10 g,7 The above table indicates that doses from a ground level release would be significantly higher out to approximately two to three miles, beyond which doses from both release pathways would be similar.
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l l l l l 1
# 1 QUESTION 24 1 i l On Page 1.25, rapid containment depressurization which could fail the 4
drywell i's offered as an uncertainty relative to containment venting. k' hat anaSysis and/or tests are being conducted to reduce this ' uncertainty? If no analysis or tests are contemplated, what actions are proposed to minimize the uncertainty?
RESPONSE
l The uncertainty referred to results f rom a concern addressed by the WROG on Emergency Procedure Guidelines (EPGs) in their draft report,
" Development of Bk'R Containment Venting Procedures." The failure postulated would originate from a suppression pool swell which would result from the clearing of the drywell vents or downtomers and the !
flashing of the suppression pool liquid. This concern is primarily with the opening of large lines in (greater than approximately 24 inches in diameter) MARK I or MARK II containments where the rate of depressurization is largest.
~his issue has since been transferred to the Decay ifeat Removal Subcommittee. Review of additional information supplied by the Subcommittee indicates the following:
(1) The concern was directed mainly at Mark II containments due to their smaller suppression pool surface area and/or suppression pool air space volume (when compared to Mark I and Mark III containments). l l (2) Venting a typical Mark II containment at 70 psia with a 33.6-inch vent pipe, the maximum pool swelling height is less than 4 feet. This venting capacity corresponds to a heat removal rate of 30% of rated power. (3) Based on comparisons with pool surface area and free airspace volume, the pool swelling height for Mark I containment designs would be less than 4 feet under the same containment venting conditions. 510LR
1 (4) The maximum calculated pool swell height is 4 feet during containment venting. This low swell height and corresponding low swell velocities
- are within the containment design basis. It is thus concluded that l pool.pwell loads are not concerns in formulating the containment venting procedure.
3 5104R
l Q'JESTION 25 Remote manual valve operation is discussed in Section 5.3.5.1.1, primarily with respect to station blackout. To what extent can the remote vent valve and any spray valve alignment be counted on for the other classes of sequences you assessed? That is, if remote manual operation is not j available, would the local environment the operators would encounter allow successful local operation?
RESPONSE
The remote manual valve operation discussed in Section 5.3.5.1.1 deals with RHR valves 183, 184, 26A, and 31A. These valves are necessary to line up the diesel fire pump for spraying the drywell after the initial five hours of a station blackout. The valves are located in the Reactor EuildinC and may not be accessible if severe fuel failure occurred prior to taking action to open them due to area radiation dose rates. AC power independent remo+e manual control for the.e valves is being studied as described in our response to Question 19. For the other classes of sequences assessed in the report, the diesel fire pump would not be necessary for spraying the drywell. The RHR System with the torus water inventory as the primary source of water would be utilized until the torus water temperature exceeded the spe'ified c maximum value. At this point, the Service Water System could be crosstied into the "A" l loop of the RER System to provide an ultimate backup capability to inject water into the reactor vessel and/or containment from the Connecticut River. The valves required to utilize both spray paths are operable from the Control Room; however, should remote manual operation from the Control Room not be available, these valves may not be accessible if severe fuel failure has occurred prior to taking action to open them since they are located in the Reactor Building. The remote vent valves required for the six vent paths discussed in Section 5.4.4 are all operable from the Control Room. However, operation of these valves from the Control Room would not be possible in the event of a station blackout since they are all ac powered. (Operation would be 510LR
1 possible for the postulated ATWS scenario.) In, addition, if severe fuel failure has occurred prior to taking action to open the valves, they valves may not be accessible since they are located in the Reactor Building end the torus area. The capability to manually open these valves, should it be necessary precore melt, may exist depending on vent
~
path integrity. However, this would require manul actions and temporary connections. It is Vermont Yankee's position that the primary method ol containment pressure control in severe accident sequences should be the use of containment sprays, i 5104R
Ol'ESTION 26 Severe accident venting discu,ssed.in Section 5.4 does not include an evaluatio#n of the reliability nf the ADS System. Given the types of severe agtident challenges you have described, provide your estimates of-the reliability of ADS valves; i.e., their potential as a suppression pool bypass path. Can battery packs or portable generators be used to assure high reliability? If so, at what approximate cost? RESPONSE I In the severe accidents discussed in Section 5.4, the ADS valves (main steam safety / relief valves (SRVs)) are relied upon to transfer energy from the reactor to the torus. These valves would be subject to harsh environmental conditions, including high pressure, temperature and j possibly radiation during the postulated severe accidents. DC control power to the solenoids, as well as sufficient instrument nitrogen gas i pressure, is needed to insure proper operation of the SRV valves. The reliability of our ADS valves for severe accident service is primarily influenced by three factors: (1) valve f ailure rate history at Vermont Yankee, (2) de control power availability, and (3) instrument nitrogen gas supply availability. Vermont Yankee's experience with their SRVs is that they have always actuated during "as found" testing. However, there have been five occasions when the valves did not lift within the code required + 1% of the nameplate set pressure. In addition, there have been three instances when the air operators associated with the valves failed inspection testing. The last occurrence was in July of 1976. These three failures were attributed to the diaphrages on the operators. The diaphragm material was changed in 1976 and their replacement frequency was also increaded. Since that time, there have been no additional failures. DC control power is provided from reliable redundant battery banks. The de solenoid valves require very little power to operate. It is expected 510LR
t that even if the battery banks become' depleted in eight hours' post-station blackout, the battery voltage would still be sufficient to operate the SRV. solenoid valves for a long period of time. Therefore, battery packs or
~
. portable generators need not be considered.
3
- The instr'ument nitrogen gas is supplied from an on-site liquid nitrogen-i storage tank. No electric power is needed to maintain this supply systdb.
operable. A review of the maintenance records for the Nitrogen Gas Supply System indicates that there have been only five.(5) instances where: repairs have.been required, indicating very~~ good operating. experience. For a postulated suppression pool bypass path accident scenario, one of the SRVs fails to reclose after being activated and the associated discharge line has ruptured somewhere along its run in the wetwell airspace. The steam and potential radioactivity may bypass the suppression pool and release to the containment. This safety concern was
- studied by the Brookhaven National Laboratory and reported recently in NUREG/CR-4594, " Estimated Safety $lgnificance of Generic Safety Issue 61." 'The best estimated core melt frequency of an SRV line-break accicent reported in that study is less than 1.0 x 10
-13
/ reactor year, which is small compared'to the other accident types reported in the Vermont Yankee Containment Safety Study.
l
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QUESTION 27 What is the approximate cost for improving the valving for the diesel fire pump? 4 2 RESPONSE i e A conceptual design study is underway that will address the cost of improsing certain valves to aid the operators in aligning the diesel fire pump for vessel / containment injection. Our cost estimate is not available at this time. Several options require further review before a realistic cost can be estimated. 5104R
l l QUESTION 28 J For the improvement options you have evaluated, what maintenance and surveillance guidelines would you propose to use?
~
2 . l
RESPONSE
1 e A number of design improvements are being considered by Vermont' Yankee. As our engineering studies progress, we will determine which improvements l I are to be recommended. Maintenance and surveillance guidelines will be [ addressed as part of the detailed design change process to implement the I recommended improvement. Maintenance and surveillance requirements are generally determined with consideration of a r, umber of f actors including importance to safety, equipment reliability, vendor recommendations, etc. I l, a 1 I l l l l
-LS-5104R l
l ________.____._____________O
QUESTION 29 To what extent do you consider the option of drywell flooding to be effective? If effective, would you include the option in future revisions to your Emergency Operating Procedures.
RESPONSE
a) The greatest benefit of drywell flooding is to establish and maintain RPV water level above the Top of Active Fuel (TAF) if RPV flooding / injection (core cooling) capability is lost at some subsequent point in tims. To be effective the following conditions I would need to exist:
- 1. A means to flood the primary containment to the desired water level.
- 2. Communication between the primary containment and the RPV to permit flow of water into the RPV from the containment.
- 3. The elevation and size of the drywell vent must be such that:
- a. With primary containment water level above that corresponding to TAF, the vent is not submerged.
- b. Drywell vent size is capable of flow rates necessary to preclude overpressure failure of the containment due to
{ increasing hydrostatic head and permit removal of heat from ' the RPV (assuming the RPV is venting to the drywell atmosphere). l l 4 Caprbility to vent the RPV outside the primary containment to i J ensure that the RPV will be flooded via containment flooding by reducing the pressure within the RPV to as close to atmospheric j pressure as can be achieved. 1 1 5104R l l ..-
a l b ! 1 k All of these conditions must exist simultaneously for an appreciable amount of time (approximately 12 hours) before core . cooling can be effectively achieved by this'means. If all other means of core cooling are lost prior to this, it still protides a means for cera debris control.and protectien of the primary containment structuro. b) This option is set forth in Contingency No. 6 of Revision 4 of the Emergency Procedure Guidelines. Vermont Yankee is committed to implement Revision 4 of the EPGs following NRC approval. 4 5104R
l
-QUESTION 30
-It'is estimated that th'e maximum debris layer thickness on th'e drywell 1
floor would be approximately 1.1. inches (Page 136). Provide the bases for such a conclusion.
' RESPONSE
.At the time'of vessel melt-through, it was assumed in an analysis of MARCH /RMA that the debria was separated into two layers - the upper layer.
l- being the lighter melted metallic materials and the lower' layer being the l .- . denser oxide. The' debris temperature was calculated to be higher than the' liquidus temperature of the metallic melt and lower than the liquidus l: temperature of the'exide layer. Therefore, it is assumed that metallic melt would spread out on the drywell floor when the debris melts through the vessel. The oxide layer is calculated to be of such a low temperature that_its viscosity would prevent spreading to the drywell wall. The metallic material calculated by MARCH /RMA consists of approximately 53,000 pounds of zirconium (145 ft ) and 56,000 pounds of steel (115 ft ). The latter includes a portion of the core support structures and lower vessel head that are involved in this melting process at the time of vessel failure. The total area of the drywell floor is l approximately 1,240 ft2 (see Figure 5-13). In addition, there are two sumps whose total volume is 156 ft . Therefore, assuming the sumps fill l completely, there will still be 104 ft of metallic material available to spread over the drywell floor. Based on this, it can be shown that the maximum metallic layer thickness would be approximately 1.1 inches. 510tR l
l QUESTION 31 . What is the thickness of the vent duct between torus and drywell (Figure 1. Page F-6)? l
RESPONSE
The vent duct is 1/4 inch thick with a 6-foot 9-inch inside diameter. !
)
5104R
l' QUESTION 32 l It is stated that " additional physical barriers are believed to be counter-productive as they may prevent containment spray from cooling the j , debris while it is confined in the Subpile Room." Please elaborate how 1 such a, barrier would prevent the spray from effectively cooling the debris (Page 137). l
RESPONSE
The basis for the above statement was that the physical barriers will confine the debris in the Subpile Room where little, if any, water can be sprayed on them to provide cooling. If the debris is allowed to sit the Subpile Room and spread over the drywell floor, more water will be available to cool the debris and there will be more debris surface area to be cooled. 5104R 1
QUESTION 33 It is indicated that, with the exception of the one-inch nitrogen CAD line and the six-inch nitrogen purge line, the pipe lines associated with four potential vent paths are likely to fail. ' Provide background information j which led to this conclusion (Page 131).
RESPONSE
The basis for the statement on Pages 131 and 132 that "with the exception of the one-inch nitrogen CAD line and the six-inch nitrogen purge line, these lines are likely to fail in the Reactor Building when venting at elevated pressures" is as follows: j l
- a. Section 5.4.4.1, 18-inch Atmospheric Control System vent path (via RTF-5). This path contains a Reactor Transfer Fan RTF-5 "which was not designed to withstand any significant pressure and would leak and probably totally fail under the estimated venting pressure."
- b. Section 5.4.4.2, three-inch Atmospheric Control System vent path.
This path includes the Standby Gas Treatment System (SGTS) which "is not cepable of handling steam; the manufacturer has stated that the HEPA filters would be blown out if steam passed through them. In addition, the housing for the SGTS has a design pressure rating of only 2 psig positive pressure."
- c. Section 5.4.4.3, 18-inch Atmospheric Control System vent path. This path " ties into duct work prior to exiting the Reactor Building. The duct has been tested to eight inches water gauge." Due to the low structural capability of this ducting, we assumed that it would fail at elevated pressures. In addition, the ducting is not designed as a leak tight installation.
- d. Section 5.4.4.4, 20-inch and 18-inch ventilation supply vent path.
These lines tie 'nto duct work prior to exiting the Reactor Building. The discussion in (c) above applies here as well. 5104R
QUESTION 34 It is implied that a layer of debris (1 1 inch thick) would not penetrate the drywell steel shell and enter the torus (Page 136). If such is the conclusion, pleese discuss why such a burn through is unlikely while core debris is attacking the drywell floor. There is a gap between the drywell steel shell and the concrete shield. If the molten core were to burn through the steel shell at the indicated corium elevation, what would prevent the fission gas from entering the Reactor Building since the concrete shield outside the drywell shell is not designed as a pressure boundary?
RESPONSE
a) Drywell Shell Thermal Attack Civen that a core melt accident has been postulated, the conditional failure probability for the drywell shell due to direct contact with molten material is a function of the type of accident sequence which led to core melt and the available mitigation. The Vermont Yankee CET uses drywell shell conditional failure probability values of .01 and 0.1 for sequences with drywell water injection and without water injection, respectively. The chain of events and phenomena occurring that determine the Vermont Yankee technical evaluation of drywell shell integrity and conditional failure probabilities can be categorized as follows: o Core melt process o RPV bottom head failure mechanism o Heat sinks for rapid heat transfer o Debris spreading o Drywell water injection o Drywell shell temperature rise 510LR ,
. Each ofJthese is discussed'as follows:
o Core Melt Process The core melt process can be postulated to occur in a variety of ways. The best' estimate assessment of the process is that portions.ofEthe core can become molten and move to the RPV bottom head in fractions of core from 10% to 50%; If less than 10% of. the core material is molten, it would most likely be quenched; greater than 50% of the core material being' molten is probably not, possible in view of the rad 5ative cooling.that can occur from the outer fuel assemblies. This type of core melt process is of the slow flow type melt mechanism and is consistent with the model presented in the NRC containment event tree write-up for the NUREG-1150 Mark I evaluation (Preliminary Draft. SAND 86-1135 May 12, 1986). o RPV Bottom Head Failure As soon as molten material reaches the reactor vessel bottom head, it may form an insulated mass that can begin to heat up the RPV bottom head penetrations even with a water overburden. The potential temperature rise can result in attack and failure of some of the multitude of penetrations in the BWR reactor vessel bottom head, e.g., instrumentation or CRD housings. L The best estimate model for this attack mechanism is that individual CRD and instrumentation penetrations would fail due to localized debris temperature in the range of 2500 K. This debris temperature is calculated in MAAF to be at the eutectic formed between UO and zircaloy. The material is also near the 2 melting point of the debris and could be solidified by heat transfer to available heat sinks with changes in temperature of several hundreds of degrees. 1 5104R ; 1 i
Once the debris causes local seal penetration failure, the debris is anticipated to fall or be ejected (depending upon the RPV pressure). I o Mass Transport to the Drywell Floor During the process of material ejection from the RPV, the material must make its way through the maze of CRD housings, piping, and steel support structure below the RFV and fall approximately 30 feet to the drywell floor. As a result of the energy exchange with the structures in the pedestal and radiative heat loss from the debris during the trrnsport process, it is anticipated that the debris would cool more than 300 C. If continuous drywell water injection is available from either drywell sprays or RPV injection and through the RPV breach, then additional quantities of the debris would be quenched and prevent debris contact with the drywell shell wall. o Debris Spreading Because the debris is nearly at its freezing temperature without any water quenching, it is found that the debris would not reach the drywell walls because of its high viscosity and the large amount of radiative heat transfer to the drywell atmosphere. o Drywell Shell Temperature Rise In the unlikely event that molten debris reaches the drywell wall with no drywell injection (e.g. drywell sprays) then the drywell wall temperature could be high, i.e., on the order of 1500 F. If the dryvell shell is not adequately restrained (and the judgment is that the shell in tnis area is adequately restrained) then a high temperature creep-rupture, failure mechanism could result. 5104R l . . . ... . - . _ . .. . i
I l
, However, if a water overburden exists due to drywell sprays, then the temperatures of the drywell shell are substantially less and the failure probability due to creep rupture is judged to be less 1 J
than .01 onditional on the fact that the slurry could even reach n drywell shell wall. t'e In summary, the best estimate evaluation of the conditional failure probability of the drywell shell due to direct thermal attack (using MARCH and MAAP analyses) concludes that:
- 1) With water injection to the drywell: The chain of adverse events that must occur to result in a thermal attack of drywell shell results in an assessed failure probability of this containment failure mode of .01.
- 2) With no water injection: The same chain of adverse events is required to occur, but there is no water quenching from external sources. The failure probability is judged to be a factor of 10 higher.
b) Fission Product Pathway The molten debris direct thermal attack on the containment shell at the interface of the drywell floor and the shield is considered in the Vermont Yankee evaluation. The consequential fission product pathway into the Reactor Building is also included. The following description of the flow path is judged to be the best estimate failure mode and path if it were to occur: o The failure would tend to be a local failure at one drywell location. o It would produce a leakage pathway through sand and air gaps into the Torus Room (the lowest elevation of the Vermont Yankee Reactor Building). 5104R
$n' , L. , h3 'o. The pathway would be restricted and would promote refreezing of the interface debris that could seep through the containment drywe11Lfailure. pathway. Therefore, it can'be modeled as an
~
intermittent gas. flow path through a tortuous path'and.into the Reactor-Building. Based upon both the supporting MAAP and MARCH calculations,-the radionuclides. releases associated with such a leakage pathway into Lthe' lowest Reactor Building elevation are found to be low releases. If p 'the Reactor Building is effective, as:it is judged to be at Vermont. Yankee for.this sequence, the radionuclides releases would be characterized by a containment leakage failure with a Reactor Building effective in reducing releases to the' environment. Both alternatives are addressed in the Vermont Yankee CET. 4 5104R ur
,au 4 .
QUESTION 35; It'is stated that 135' psia is a reasonable value for the Vermont Yankee containment failure pressure (Page'F-5).
, a. What is the uncertainty range associated.with this value? -
- b. What.would be a change in core melt and conditional containment failure probabilities associated with the' uncertainty?
- c. Provide references for the Ames and Sandia calculations mentioned in the Appendix F (Page F-1).
RESPONSE
a) In this study, time did not permit the detailed structural analysis required to produce exact failure pressure along with variance. The-structural capacities shown on Page F-4 were estimated in a simple conservative manner to support the accelerated schedule. Strue: ural
~
characteristics such as actual yield strength, strain hardening, coupling of see: ions, and possible benefits of drywell concrete in limiting deflection was not evaluated. The 135 psia failure pressure was selected based on WASH-1400 published values and other plant reports. After performing simplified calculations, considering the above mentioned conservatism, and considering the results of more detailed analysis by Ames and Sandia (NUREG/CR-3653.) 135 psia was determined to be a reasonable approximation. No uncertainty range calculations were performed. However, as stated in the response to Question 13, the BWROG on Mark I Containment Ultimate Strength Analysis is currently performing a study 1 to determine the actual vessel failure pressure including uncertainty calculations. b) Given that no uncertainty evaluation is performed for the containment
-failure pressure, no formal change in core melt frequency or conditional containment failure can be estimated.
j l 5104R 1
- -- --- -. - _ -_ _ _a
c) The source of data from the Ames and Sandia calculations is from hTREG/CR-3653 : hTREC/CR-3653, SAND 83-7463, " Final Report Containment Analysis Techniques - A State-of-the-Art Summary " March 1984 o Browns Ferry by Ames Laboratory Section 16.1. o Peach Bottom Containment by Sandia National Laboratory, Section 17.1. l l l 1 l l l l-l l l ' I 5104R l l l
l l QUESTION 36 i Your evaluation of deinertinr; indicates a relatively few hours of pcwer l 1 operation while the contairanent is deinerted (i.e., about 1% in the run ] mode). For each such instance, please identify the following:
- a. The number of hours deinerted;
- b. The purpose for the deinerted condition, and whether it was successful, before shutdown was required by your Technical Specification; and
- c. The power level and its corresponding reactor pressure at which containment entry and exit were made.
Please indicate the impacts you would expect for a deinerting Technical Specification to either 16 hours or li hours.
RESPONSE
See table on the following page. 1 5104R
E y R ~ U r
, t S a n
S i d E s . e ct eni R p P - es pi l 7 7 7 0 7 0 7 0 l d R . . . 2 2 0 2 0 Sl ee O 4 4 4 0 4 0 0 4 0 0 wt T 1 1 l 1 l 1 8 1 1 8 .e yl C ( ( < < < h r r u A ca d s E e - e R T e nr ( l e c h s
%y .
wt 4 c e n l d e enb ov voa oit i e r Z R bt el E aas pl W I ri a O seh d P e i pt e) g o t 5 n R n n- n r8 o O a ori e/ i T R) i e n0 t 1 t wd i 2 C a r A D D D % D % D D M1 D aoe r pd e/ E / / / 0 / 3 / / R < / d9 e R S S S 7 S 2 S S I ( S t u p _ nl l e , O eac h4 r cnri mn t 8
/
e oo g5 w c nt n1 o o i / P nan r6 g e u i s gge ' d , n N r ynr 3 i O u s s s s s s xia s8 r I o r r r r s r r or u/ u ue l T ? ? h u u u u r u u t 4 l A o o o o u e o t dl a/ R 5 h h h h o h h n c t 3 n U h ed y s( i o D 4 6 4 4 9 8 5 2 mec nr t s 1 1 1 1 1 4 1 t_ i r a ne r e au ai t cf l r n nc o pt i e oo n c g ee D t n t eai ag t d d e hh n cn n n e p t t n ii e a t i i dl m a p esg ne n k c f l md e i u i a a i o i t i ob f e d a s ti ee t L n t T e rt r r n I u T l ea u o . r O h ap s e C d o e g e t ogw s s e y e t g g i r er r c a a n H u t no rh a r k i s id pt . a l a ae m , o et n n m p p eg i B l h ri reo i r e e L a T 0 k C t ew ohi R S t 7 a n t t t P pa A - e l sig c a N r l e me 6 V L a aen arr O i a rk uL 8 R g i di eoe rf p S a e ua S . - t d t A p S t e g f n r eyr o E e sL l n d o i a nl e d t R R p i l i n e k P i nn n pl m oe ek a nr s c f Oi aea 4 u Mni wc g ou y a , e e cm
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2 8 3 3 3 8 f' 4 8 8 : 1 2 8 / 8 8 8 / / 8 / / S E / 7 / / / 2 7 / 0 6 E T 8 2 7 2 4 2 2 5 2 1 T A / / / / / / / / / / O D 6 8 1 2 3 6 8 1 1 4 N
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es 7 7 7 7 7 0 pi l d R 2 Sl l e O 4 4 '4 4 4 0 et T 1 1 1 1 1 1 .e wl C ( ( ( ( < h r yu A ca rs E e d e R Te r (l n
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d bt r1 i aas e1 5 i seh ri pa P i pt d) n R o e5 o' n n t8 i O ori r/ t T i e e0 a C t wd n2 r A fi D D D D % aoe i/ e E / / / / / 6 r pd e9 p R S S S S S 5 - t u d O nl l , eac e4 r cmn h8 e nri / w oo g5 o cnt n1 P o i/ g nan e r6 u n I s s gge d , i 5 r r ynr 1 r 0 u u s s sr s xia s8 u 1 o o r r r or u/ D) T h h u u u u ue t 4 d A o o o o t dl a/ ne R 5 5 h h h h n c t3 ou U ed y s( i n D 3 2 2 1 8 1 8 1 8 1 0 2 mec nr t i t s rt i ra ne en au ai no t cf l r - iC nco pt 1 e( ) oo n 6 D t n c g ee - t n t t e eai ag n L m e d n hh n cn e A t t n i i m c a i dl n U a esg ne i Q l r md e i u a E p e om i ob f t R t e ti ee r r n - il rt u o e nb l ea - d n oo ap se _ C_ p Mr t s se - y u i P ogw er r o e P n t no rh a r r oh id pt . m C u c i c et n n i l r t t h ri reo r n i i ai t ew ohi _ P o a c r w n t t t N O S i t c a t F g n R d e bS i Vl e d sig aen di t c arr eoe rf p a A n pv eyr E R i d n f i L i i m ( a me uL i nn nl e dt n pl o L I T P f Oi aea g n A o r d u D g n ei l d e e rxr cm =
.d i 0t o 0 i rO eE o l 8 c r 8 l i s) p w n e - e h - e cr i) u o u Vn S V u eo bt p . o f
e I n x I f Rt n( r dt S e re o 7. at S o - o . R MC R M R BM i orn ta A. s t rr cuu L ru 7. he act ece L
. D3t R or 4 4 4 5 6 ) )
8 4 8 8 8 8
- 1 2
/ 8 / / / / S R E 5 / 8 0 0 0 E 4 T 1 9 1 3 2 2 T 0 A / / / / / / O 1 D 6 8 9 9 9 8 N 5
4 QUESTION 37 Please provide your estimates of pressure and temperatures as a function of time for the accident sequences ;rou analyzed for CCFP estimates. i RESPONSE 37 The following figures show some of key thermohydraulic parameters of the accident sequence analyses performed specifically for Vermont Yankee using the MARCH /RMA Code:
- 1. T,C,C2 (ATWS sequence): Figures 1 through 11.
- 2. Station blackout for more than si:: hours, coupled with failure of HPCI, RCIC, and the diesel fire pump after six hours: Figures 12 through 21.
- 3. Station blackout and further failures to the HPCI and RCIC Systems or their support systems: Figures 22 through 26.
The following figurec show some of key thermohydraulic parameters of the accident sequence analyses performed specifically for Vermont Yankee using the MAAP Code: i
- 1. Station blackout for more than sir. hours, coupled with failure of HPCI and RCIC and the diesel fire pump after six hours: Figures 27 through 30.
l
- 2. Station blackout and further failures to the HPCI and RCIC Systems or their support systems. (Water injection became available at the time !
of vessel failure): Figures 31 through 34 I
- 3. Similar to (2) but without the addition of water on the debris after vessel failure: Figures 35 through 38.
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PRESSURE, PSIA. c0,.00 . 2,0.00 4,0.00 6,0. 00 80.00 i 100.00 i 120.G0 a I I 56 -4 L f. i - l l 3
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, Ol'ESTION 38 It-is not clear from your evaluation why the probability estimates of early failures with higher releases are lower than for Class IV events.
Please explain if venting of ATWS sequences before core melting was ) assumed. RESPONSE 38 a) Comparison of Class IV Low Release Frequarcies Versus High or Medium l Release Frequencies There are many types of accident frequencies which could lead to a core melt accident. For the spectrum of accidents, each accident has a conditional probability of leading to a containment failure that could result in a high or medium release by defeating active and pas:(ve mitigation features such as drywell sprays and the Reactor Building, respectively. Therefore, Class I, II III and IV sequences can all result in higher releases if and only if minimum mitigation is encountered on the release pathway. In particular for Class IV, active and passive mitigation processes also exist. Class IV accidents are one part of the overall accident sequence spectrum. Class IV sequences are those in which an ATWS is not succe.ssfully mitigated and containment could be challenged. However, as noted by the Class IV containment event tree there remain a number of mitigation measures possible to reduce or minimice the radionuclides releases even after containment is failed due to an ATWS event. Class IV accident sequences represent a relatively small fraction of the total core melt frequency, i.e., less than 10%. The remaining 90% of the accident sequences also have the potential to produce large radionuclides releases under special circumstances. There are several alternative paths through the containment event tree which could result following a class IV accident sequence. These . paths vary in their potential consequences from low to high. The I
-101-510tR
l
, ,- .j'
.f mitigation measures which are effective in reducing-the, radionuclides-source terms include: !
I
- o. Passive
-- Wetwell airspace failures (i.e., the containment failure mode identifled in U.'.cH-1400)
- ' Reactor Building decontamination o Active Drywell sprays or coolant injection to drywell Each of the mitigation measures has the capability to reduce the radionuclides source terms to the low category (L). Calculations using both MAAP'and MARCH /RMA indicate that radionuclides releases for severe accidents can be maintained below a 1.0% I equivalent release when an
.. active or passive mitigation measure is available during the core melt and ex-vessel interaction process.
In' summary,'the low release frequency of the AIWS Class IV sequence can be' greater than that of higher releases because the Class IV.
~ containment event' tree has several mitigation measures which tend to reduce.the potential consequence of an ATWS release leading to the high (H) or medium (M) category. This is demonstrated by both MAAP and MARCH /RMA.
-102-5104R
l 1 l
,: 'b)- Venting During TWS .
Venting as'a mitigation measure for.ATWS was not used for.two reasons: 1 1. o .The current Vermont Yanke'e procedures. require that there be no unusua1' radionuclides activity before venting could be' performed.
~For ATWS scenarios thee could be substantial radionuclides present in containment even though the' core is essentially covered and
' effectively cooled (i.e... peak clad temperatures below 2200 F).
o Adequate. venting. capability to vent the excess' steam to the
. environment does not currently exist. There are a multitude of-pathways to vent'at Vermont. Yankee but most are through ductwork
'in the Reactor Building (see Response to Question 33).
4 e
-103-5104R
/0m: i
<v i' .n M DISTRIBUTION 'f NOV 0 3 1986 M L FILE PSB R/F Mr. Terry Pickens M. Thadani
. BWR Owners Group Chairman G. Lainas Northern States Power Company R. Bernero ,
414 Nicollet . Minneapolis, Mn.' 65401 pf " '
~
Dear Mr. Pickens:
, The purpose of this letter is to provide you with our coments on the August, 1986 IDCOR " Evaluation of BWR Accident Mitigation Capability Relative to Proposed NRC Changes". In general, we have found the evaluation useful. Specifically, we are enclosing an assessment of the evaluation for your consideration. The following.sumarizes our present views of the IDCOR evaluation for each area of potential containment perfomance improvement:
- a. Combustible Gas Control - Insufficient information Arc been provided on a plant specific or generic basis to indicate
. that reducing the number of allowable hours of power operation while deinerted would result in decreased safety.
Moreover, no bases have been provided that either a significant decrease in operational flexibility, or power operation, would result.
- b. Filtered ~ Venting - The evaluation of potential vent size guidelines of 6 to 36 inches in diameter, with the laroer size necessary to cope with an ATWS, is useful. What is not clear is whether significant modifications are required at any plant or plants to ensure reliable and remote manual opening and closing of one or more valves that vent up a plant stack. Furthemore, it is not clear what the studies that refer to other means to mitigate an ATWS encompass, nor when and if the staff will be infomed of the results. Lastly, as the evaluation points out, we recognize that venting is a risk tradeoff. As a last resort to prevent a large and uncontrolled release, however, we have tentatively concluded the risks from venting virtually only noble gases through a plant stack significantly outweigh the risks of not venting.
- c. Reliable Drywell Spray - The evaluation points out that sprays could be counted on to achieve goals such as debris cooling, containment cooling, fission product attenuation.
We agree. We conclude that any potential drywell structural problems may not exist after realistic calculations are completed, or can be avoided by proper operator action, spray system modification, or by a combination. Lastly, we also agree that further study is desirable to identify the minimum . flows, rate, system modifications, and related alternative power sources and water supplies necessary to achieve the goals. 3
Mr. Terry Pickens 2
- d. Debris Control - Bases for the IDCOR conclusions have not <'
been provided to allow us to agree or disagree with the conclusion that torus room barriers do "not appear to have significant benefits to justify further consideration." Although more discussion of containment barriers has provided, the same general conclusion was reached. Comparisons of IDCOR evaluations with Brookhaven assessments of core melt progession and core debris were provided which indicated a potential for relatively large uncertainties in debris characteristics. For example, no assessment of the uncertainties in the volume of core debris and steel was provided. Also, evaluations of the benefits and costs of various design options were not discussed. We conclude that this area needs further investigation, since it is just the uncertainties in debris characteristics that indicate debris centrol may be necessary.
- e. Training and Procedures . We agree that emergency procedure guidelines of the type prescntly being proposed by the BWR owners (Revision 4) appear to be of the type necessary to implement satisfactory emergency procedures and related operator training. The staff intends to provide specific comments on the proposed guidelines (i.e., to establish appropriate water level reductions and related power levels to be better able to cope with ATWS type events). What is not clear is the position of every BWR owner with respect to both complete adoption of the guidelines, nor the anticipated implementation dates for emergency procedures.
l Based upon our assessment you may wish to comment further. Sincerely, 4, asw.;, w *1 a mwrt m. hemer* Robert M. Bernero, Director Division of BWR Licensing
Enclosure:
Staff Assessment of IDCOR Evaluation bec: H. Denton K. Campe R. Vollmer J. Kudrick T. Speis S. B. Kim l F. Miraglia A. Notafrancesco l T. Novak P. Hearn W. Rus' sell E. Chow oa tow. 'N. Thadani F. Skopec DBL PD's, BC's, & SL's l "# C. Grimes I le /t/ Ir6 0FC : PSB: DBL I'lD-DBL /BWR : D: DBL : : :
.....:. ...:...g.......:........ ..........:............:............:...........
NAME : n:ye : GL : RBernero : : : :
. . .. . : .. .. . . . .. . .. : .f.W. .na s DATE:10/Jt//86 : 10/f7 /86 : 10/ 3 /86 : : : :
/
1 NRC STAFF ASSESSMENT OF THE IDCOR AUGUST N86 '
" EVALUATION OF BWR ACCIDENT MITIGATION CAPABILITY RELATIVE 70 PROPOSED NRC CHANGES" l <.
The subject report addresses BWR severe accident probabilities and five areas of potential improvement in containment performance. The PRA methodology used ' ' by IDCOR in developing the dominant accident sequences and in identifying the plant " damage bins" is very similar.to that used in the Shoreham Probabilistic Study, and has-been reflected in the IDCOR initial Individual Plant Evaluation for Peach Bottom. In general, we concur with the PRA methodology and the classification of the plant " damage bins" used in the IDCOR report. What has not been addressed are the relatively large uncertainties
- in such assessments, the costs of improvements compared to potential risks averted, and analysis of costs that could be avoided by closure on any further need to consider mitigation (subjecttonoplants ef fic identification of significantly l different core melt probabilities .
Each of the five potential areas of containment performance are discussed below: i' Combustible Gas Control - Typically, Mark I and 11 plants are allowed by Technical Specifications to be deinerted for 24 hours at startup and 24 hours prior to shutdown. The staff has pro)osed the possibility of minimizing the deinerting. Based on the evaluation by IDCOR regarding this initiative, the industry recognizes the increased safety margin of reducing the deinerting '; window. Any plant operation or safety impacts associated with shorter deinerting times were not identified. It would appear that plant specific or further generic assessments need to be perfonned to determine the feasibility of shorter deinerting intervals. This issue should be explored further in order to address any increased assurance that the presence of a combustible mixture inside primary containment is reduced to as low as practicable levels. To this end, future study should l include an evaluation of past practice to determine the impacts on plant I operations and safety due to a shorter allowable deinerting period. Filtered Venting - The IDCOR study considered a spectrum of accidents and identified the rance of the venting sizing requirements to be between 6 and 36 inch equivaient pipe diameters. An ATWS was identified as the most severe event, requiring a vent size up to 36 inches in equivalent pipe diameter. The study that compares vent sfre guidelines to individual plant venting capabilf tfes was not provided. However, it appears from an industry survey conducted by IDCOR that imany plants do not have the large diameter capability needed for an ATWS event, but do have hardware generally avsflable 1
'The staff evaluation of the Yermont Yankee Containment Safety Study identified several uncertainties inherent in methodologies used in the probability calculations (LetterdatedOctober 24, 1986, from NRC to Vermont Yankee Nuclear Power Corporation).
l l l
1 1 i
-2 ,
to support venting under many postulated severe accidents. IDCOR suggested that venting for an ATWS be considered as an option. The evaluation mentions 1 that industry is currently studying an alternative mitigating scheme for this . ( event which may eliminate the need for large diameter vents, but provided l insufficient information to indicate feasibility. The study ddentifie: the need for vents to be remotely and reifably controlled,' and have the ability to be re-closed.~ A conceptual design that meets all the potential venting guidelines is described as requiring two DC-pcwered, fail closed, motor-operated butterfly valves. However - there is no discussion of : the need for a capability of vent signal overriding containment isolation 1 signals, nor was the D. C. power source fully identified. An operability : reliability guideline of .9 to .95 was identified, but no information was ; provided that would indicate that any of the schemes studied by IDCOR would j achieve the guidelines. -IDCOR also indicated that valve opening should be l achieved at pressures of 76 psi or less, but provided insufficient basis for j the guideline, and did not address reclosing. IDCOR ideritified the need to ensure that SRV operability should be a principal { goal for accident mitigation. Such considerations suggested were identified 2 as setting intial vent initiation at I to 1.5 times the design pressure, and vent termination at 30 to 20 psi below the initiation pressure. IDCOR also suggest venting through the SGTS (presumably to insure an elevated release) or, in any case, to outside the reactor building. While we agree I with these conclusions conceptually, IDCOR did not supply sufficient information to indicate bases, nor hbw a utility would convert the guidance into numerical design and performance criteria. One aspect of the IDCOR study is that it depicts containment venting consequences in the light of competing risks. For example, venting strategies intended to avert catastrophic containment failure U.nder severe accident conditions are perceived to lead to a higher frequency of noble gas releases. l Specifically, the study points to cases where containment pressures may rise above design basis levels, but stop short of causing a containment failure. Venting in these cases is perceived to lead to noble gas releases, even though the containment survives the pressure transient. However, the study does not quantify the frequency of such releases and recommends that an additional
- evaluation be made. Another example of competing risks brought out in the study is the possibility of induced core melt due to the loss of water in the suppression pool in the context of an ATWS event. Aside from the' concerns regarding competing risks, the study notes that certain venting strategies can lead to undesirable onsite consequences, even when containment failure is averted by venting. Specifically, the study concludes that venting can lead to' in-plant contamination and inaccessibility to critical plant safety areas and ;
equipment which are vital for post-accident functions and control. The staff in'ttiatives are focused on the reduction of containment vulnerability through debris control; enhanced temperature, combustible gas and source i tern attenuation; and improved procedures and training. That this initiative can lead to competing risks is a possibility also recognized by the staff. The staff has concluded however, that the effect of possible increases in the higher frequency of neble gas releases caused by venting is offset significantly i ____-__-______________:__a
by the dose and economic consequence reductions afforded by containment venting. Moreover, the costs of further plant-by-plant studies of contalment perfomance and risk and potential mitigation improvement evaluations,. could be eliminated. With respect to the possibility of plant contamination, the staff agrees that . venting has the potential for making parts of the plant inaccessible under severe accident conditions. However, compensating measures involving hardware
- and procedure modifications (including "hard pipes" to the plant stack) can be made to mitigate these consequences at what appears to be moderate costs.
Drywell Sprays - IDCOR. identified four objectives for drywell sprays; 1) to lower containment pressures, 2) to cool vulnerable equipment -3) to quench debris and, 4) to scrub aerosols. The staff agrees, and suggests not only would aerosols be scrubbed, but so would certain volatile fission products (particularly if good drywell spray coverage is assumed). IDCOR also indicated that such sprays could provide other benefits in the fom of somewhat lower - primary system pressures, inhibited fission product revolatilization, and as a
" steam generator" mechanism for pushing heat and contaminants into the suppression pool. The staff agrees.
The IDCOR evaluation describes a method for spraying during severe accidents associated with A. C. blackouts by using an onsite diesel fire pump to supply , the containment sprays at a flow rate well below existing design levels. What was not fully discussed by IDCOR was whether such a relatively low flow rate can be expected to achieve all the goals for spraying. The staff concludes that additional power and water sources should be considered to ensure all the l goals are met. Cooling and scrubbing with the suppression pool could require an emergency A. C. power source for the systems necessary to cool the pool for station blackouts. The staff suggests that additional emergency power sources, such as comercial grade portable generators, and local power circuit modifications be considered for such situations. The study points out problems related to creating a significant negative pressure differential on the wall between the drywell and wetwell that might cause a failure. IDCOR fndicated this concern arose from a conservative evaluation of such conditions. The staff concludes that realistic calculations may not indicate such a failure at the containment pressure conditions at which such sprays may be initiated. If failure was indicated, IDCOR described , various A. C. powered pumping configurations for supplying the containment sprays at the rated flow rate, but recommended further study of the potential ! problems associated with spray initiation and the rapidly depressurization of the drywell. Rapid depressurization of the containment may be avoided in the staff's view by modifying the plant emergency procedures and spray system design to specify acceptable initiating pressures, variable initial spray ) ' rates, or a combination. Core Debris / Barriers - The use of core debris barriers to prevent molten core j debris (including steel from reactor internals and the Icwar head) from penetrating the steel containment shell or torus vent pipes, or to provide a water cover for molten material which rasches the torus room floor, were discussed by IDCOR. The 1DCOR report dismisses the torus room barrier with the simple statement "It does not appear to have significant containment related benefits to justify further consideration". Drywell barriers to prevent attack on the shell and torus downcomers are discussed in some detail. The impression I
4 given by IDCOR, however, is that realistic analyses would show that debris. barriers are not needed.
~
The IDCOR analyses performed to date have shown that at the time the loweI portion of the core melts and slumps into the lower plenum of the reactor pressure vessel, approximately 20-30% of the core is molten. This means that at vessel failure approximately 20-30% of the original core inventory could flow out into the containment. This sequence progression could be followed by i melting of the remaining core over the next 3-5 hours. This core melt progression differs from the 60-80% instantaneous melt' discharge used in a Brookhaven (BNL) calculation, and has a very strong. influence on drywell shell failure. Furthermore, after the core material is expelled from the vessel, j IDCOR suggests the water remaining in the lower plenum will flow out and cool ! the core debris. The correct debris temperature history would, therefore, be . one in which the initially expelled debris mass (including related steel) is l cooled by the lower plenum water and then slowly heats up as the water is I boiled away. This scenario influences the calculations performed by Brookhaven and would tend to result in lower estimated and, therefore, more coolable debris temperatures. As shown by BNL, lower debris temperatures tend to result in no shell failure. The IDCOR analyses also suggest that the BNL assumptions on debris mass and temperature for the drywell shell failure analyses are overly conservative and should be replaced with more realistic boundary conditions. The staff concludes that both the IDCOR and BNL analyses reflect the range of , some uncertainties associated with contemporary severe accident assessments. It l 1s just these types of uncertainties that indicate that debris control may be l necessary. l The IDCOR report also identified some negative impacts of debris barriers as follows: o Personnel (ALARA) issues related to installation and maintenance. o Prevention of the collection of normal leakage and it's detection (TechnicalSpecificationRequirement). o Seismic related istues regarding the impact of the barriers on other safety related equipment for deterministic evaluations required by existing regulatory guidance. o Potential interference with the assumed downcomer flow area in LOCA assessments. o Disintegration of the barrier during other severe accidents (seismic, LOCA g ,etc.) which may cause the barrier to be blown into the torus. The IDCOR report gave no indication, however, of attempts to identify design options which would eliminate these negative impacts.
. i 5-1 Lastly, the IDCOR report alludes to bases which would indicate that debris' barriers are not needed. Although many of their arguments appear plausible, they have not been given substance in the IDCOR report. 0 -
Severe Accident Procedures And Training - Although the BWR Owners' Group Emergency Procedure Guidelines (EPGs) are not designed with the specific intent of providing guidance to the operator for an event which has , progressed to the point of vessel failure, the procedural' guidance prior i to vessel failure is thought to be also appropriate following vessel failure. Therefore, we conclude that implementation of Revision 4 to the EPGs should be expedited. The IDCOR report is silent on expedited implementation of the EPGs. A staff review of Revision 4 to the BWR Emergency Procedure Guidelines is underway. The staff expectation is that substantial improvements in Revision 4 will result in several areas. For example, even though it is generally agreed that during an ATWS event, operators need to reduce the reactor water level in order to reduce the core power, it is not clear at what level the reactor water should be maintained and what would be the corresponding core power. It appears that there presently exists large uncertainty associated with estimating the core power for a given reactor water level. For example. Option B among the containment venting options proposed by IDCOR, calls for a wetwell vent (6 - 18 inches) capable of removing approximately 6-8% power associated with controlling reactor water below the top of the active fuel (TAF) for ATWS related events. Option C calls for a wetwell vent (36") capable of removing 18-301 power associated with controlling reactor water level near the TAF for ATWS related events. 4 i 1 l
-)
7 1
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. 1
, ASSURANCE OF BWR PRESSURE SUPPRESSION CONTAINMENT PERFORMANCE NOVEMBER 3, 1986 R. bERNER0, NRR
]
, 1. ;
ISSUES' . L 1 E ' ANALYSES OF BWR: CONTAINMENTS IN CORE MELT O sf. ( ., o REACTOR SAFETY STUDY-(PEACH BOTTOM)z sm'
.c e RSSMAP (GRAND GULF) o f / 3 IDCOR (PEACH BOTTOM + GRAND GULF)
~'*'
e l e -NRC SOURCE TERM AND NUREG-1150 (PEACH BOTTOM AND GRAND GULF) e OTHERS (LIMERICKp }{0REHAL S VERMONT YANKEE) s,m CONCERNS
.e HYDR 0 GEN COMBUSTION e DIRECT HIGH TEMPERATURE ATTACK 0F DRYWELL e OVERPRESSURE FAILURE i
)
L,i p..
- 2 GENERIC ACTION ON BWR CONTAINMENT PERFORMANCE THE SETTING: PLANT EVALUATIONS UNDER THE SEVERE ACCIDENT POLICY STATEMENT n
THE KEY REGULATIONS: GDC 16 AND GDC 50 THE SUBJECTS: 37 BWRS WITH PRESSURE SUPPRESSION CONTAINMENTS z+ H k I.
- 5 THE METHOD: A GENERIC LETTER OF REQUIREliENTS TO I CHANGES BASED ON GENERIC EVALUATION
3 H < NRC SEVERE ACCIDENT
~ '
L POLICY STATEMENT _ . . . l' s THE MOST. COST-EFFECTIVE OPTIONS FOR REDUCING THIS
~
VULNERABILITY SHALL~BE IDENTIFIED AND A DECISION SHALL BE REACHED CONSISTENT WITH THE COST-EFFECTIVENESS CRITERIA 0F THE COMMISSION'S BACKFIT POLICY AS TO WHICH OPTION OR SET OF OPTIONS? ,IF ANY) 'ARE JUSTIFIABLE AilD REQUIRED TO BE IMPLEMENTED e IN THOSE INSTANCES WHERE THE TECHNICAL ISSUE GOES BEYON CURREliT REGULAT0RY REQUIREMENTS, GENERIC RULEMAKING WILL BE IN OTHER CASES, THE ISSUE SHOULD BE THE PREFERRED SOLUTION. DISPOSED OF THROUGH THE C0ilVEllTIONAL PRACTICE OF ISSUIN BULLETINS AND ORDERS OR GENERIC LETTERS WHERE MODIFI ARE JUSTIFIED THROUGH BACKFIT POLICY, OR THROUGH PLANT-SPECIFIC DECISI0il MAKING ALONG THE LINES OF THE INTEGRA SAFETY ASSESSMENT PROGRAM (ISAP) CONCEPTION,
\
I4 1 .. \ J GDC 16: CRITERI0ii 16 - CONTAINMENT DESIGN. --AN ESSENTIALLY ; LEAK-TIGHT BARRIER AGAI!iST THE UiiCONTROLLED RELEASE OF RADI0 ACTIVITY TO THE ENVIR0iiliEl1T AND TO ASSURE THAT THE-CONTAINMENT DESIGN CONDITIONS IMPORTANT TO SAFETY ARE NOT EXCEELED FOR AS LONG AS POSTULATED ACCIDENT C0iiDITIONS REQUIRE." l GDC 50: i CRITERION 50 - CONTAINMEllT DESIGN BASIS. " -AS REQUIRED BY SECTION 50.44, EiiERGY FROM EETAL-WATER A!iD OTHER CHElilCAL REACTIONS TliAT MAY RESULT FROM DEGRADATION BUT NOT TOTAL FAILURE OR EMERGENCY CORE C00LIliG. FUNCTIONING, (2) THE LIMITED EXPERIENCE AliD EXPERIMENTAL DATA AVAILABLE FOR DEFIllING ACCIDENT PHENOMENA AND CONTAINMENT RESPONSES, AND (3) THE CONSERVATISli 0F THE CALCULATIONAL l'.0 DEL AND IllPUT PARA!'.ETERS."
5 U. S. B0ILING WATER REACTORS L l e 24 BWR 2/3/4 WITH f%RK I CONTAli1 MENT (ALL LICENSED) e 9 BWR 4/5 WITH MARK II CONTAINMENT'(7 LICENSED)
.e 4 BWR 6 WITH MARK III CONTAINMENT (4 LICENSED) 1 i
. 1
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. l
)
l- l 1 i L - - - - - - - _ - - - - - _ - - - -_
~
. I
. g WHY DD ANYTHING'
~
FOR MARK I CONTAINMENT? e REQUIRED FOR SAFETY RSS SHOWED 90% L/10% S RELEASES IDCOR SHOWED 10% L/90% S RELEASES
- NOT EN0 UGH ASSURANCE, NEED ABOUT 1% L/99%S J o -x e BACKFIT-COST-BENEF L 9c.a. g
^
~I E- '
~ {a+J /= c
- (1x10-4/YR) x 0.5 x (107 PERSON REM) x ($1000/PR)
= $5x105/YR W f'= # -u OR 4
$9x105 /YR - $1x10 /YR IF LARGE RELEASE FRACTION IS REDUCED TO .~ 1%
" - - ~- h
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7
.e 5 ELEMENT CONTAINMENT STR TEGY_ ,
(MARK I EXAMPLE)
- 1. HYDROGEN
~
STRICT INERTING CONTROLS
- 2. SPRAYS
- TWO BACKUP SUPPLIES FOR SPRAY, EVEN IN BLACK 0UT REDUCE SPRAY N0ZZLE SIZE
- 3. PRESSURE RELIABLE WETWELL VENT TO STACK (ABOUT 8-INCH) EVEN IN BLACK 0UT M
*' 9 CONTROLLED RELEASE OF NOBLE GAS TO AVOID WORSE t 6 @N
~'
- 4. DEBRIS
. Y I
M
'** A DYy KEEP TORUS WATER CONFINED
- 5. PROCEDURES AND TRAINING EPG REV. 4 +
s j
~
8 e 1.l!STITUTIONALPROCESS e CLOSURE OF SEVERE ACCIDENT ANALYSIS FOR CONTAINMENT NO FURTHER ANALYSIS UNLESS EXCEPTION IS TAKEN i e SPECTRUM OF OPTIONS RULEMAKING 50.54F LETTER FOLLOWED BY ORDER GENERIC LETTER FROM DIRECTOR NRR OR DIRECTOR DBWRL INDUSTRY INITIATIVE o PROCESS CHARACTERISTICS
- BASED ON TECHNICAL WORK AVAILABLE, IDCOR, SOURCE TERM, PLANT SPECIFIC WORK, NUREG-1050, NUREG-1150 ;
OPEN TO PUBLIC FOR COMMENT L \ : -- - - - _ - ---- ---
9 _4 , CHRONOLOGY r -e JUNE 16, 1986: MEETING WITH SWROG/IDCOR PROPOSED A. GENERIC LETTER, PRESCRIPTIVE SOLUTION, BY BACKFIT e . JUNE 30, 1986: VERMONT YAWKEE COT 4111S TO GOV. KUllIN TO DO A SPECIAL 60-DAY CONTAllllENT STUDY. e SEPTDGER 1, 1986: VERMONT YANKEE CONTAINMENT STUDY. COMPLETE e SEPTEMBER 11, 1986: tEETINGWITHBWROGTOCOMPAREBACKFIT NOTES AND STRAWMAN GENERIC REQUIREMENTS
-e SEPTEtiBER 11, 1986: MEETING WITH-VERMONT YANKEE TO REVIEW CONTAINIEllT STUDY e SEPTEfEER 23, 1986: MEETING WITH ACRS SUBCOMMITTEE ON C0ii-TAlilMENT PERFORtlANCE PROJECTED: DECEMBER 1956 - ISSUE DRAFT GENERIC LETTER FOR PUBLIC C0ir.ENT APRIL 1967 - ISSUE FINAL GEllERIC LETTER
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PRD CONSULTIN8 P. R. Davis, President - 1935 Sabin Dr. Idano Falls, ID. 83401 - (208)529-2861 3- Oct. 30,1986 Mr. Gerald R. Tarant Commissioner, Dept. of Public Service . I State of Vermont 120 State St. > Montpelier, VT 05602
Dear Commissioner Tarant,
Transmitted herewith is one copy of my report,"A Review of the Vermont Yankee Containment Safety Study. Upon your acceptance, this report constitutes fulf111 ment of the provtstons in Contract No. 0938124. However, I wtil endeavor, as time permits, to examine YAEC-1564 ("BWR Mark l Containment Evaluation-Vermont Yankee, Oct.1986) to determine if the results are consistent with and support the Vermont Yankee Containment Safety Stucy. I will provide you office with a letter 1r i find any important discrepanelet. As you know, I did not receive the report - until Oct. 29, one day before I was obligated to mall the enclosed report in order to meet the Oct. 31 deadilne. As we discussed, I did not consider it mandatory for my review to evaluate YAEC-1564 since my review was based primarily (as stated I,n the attached report) on comparison of the VYCSS with similar contemporary severe acciddnt assessments and related information, with less emphasis on an in-depth review of the VYCSS analysts. . I hope my report is useful to you and the State. I enjoyed working with you and Phil Pault on this project. I particularly appreciated the cooperation Olvan by you sed chil in setting up the arrangements for the routew ad also the objective and competent' evaluation of the material as the project progressed. j Sincerely, l h
- p. R. Davis AO '
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. i A REVIEW OF THE VEFJ10NT YANKEE CONTAIN1ENT SAFETY STUDY l l
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prepard fus: i , l I State of Vermorit Department of Public Service 1 by i P. R. Dav!s principle author PRD Consulting. 1935 Sabin Dr., Idaho Fa:11s, ID, 83401 and I
' Michael L Corra'ini, d PhD ,
contributing author:
. University of.Wiscons'in .
Oct.31,1986 e 9 9 9 0
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. l DISCLAIMER OF RESPONSIBILITY <
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This document was prepared by or for PRD Consulting Company. Neither PRD Consulting nor any of the contributors to this document:
.A. Make5 any Warranty'or representauon, expreu ur.mipneu, wiun npvu to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information disclosed in this document may not Infringe privately owned rights; or l >
B. Assumes any responsibility for liability or damage of any kind which , may result from the use of any information disclorid in this documtnt, l - t I e 9
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. l TABLE OF CONTENT 5 . . .
Sections - __Pagg ,
. 1. Introduction 1-1
- 2. Summary and conclus1ons 2-1
- 3. Procedure 3-1
- 4. Comparison of VY doin10 ant accident sequences 4-1.
.with contemporary BWR4 PRA studies
- 5. Comparison of W conditional containment fatlure 5-1 l
probablitty with contemporary assessments (ATWS and Statton Blackout sequences
- 6. Assessment of issues related to containment failure 6-1 l
modes and effects
- 7. Contalrynent integrity improvements-7-1 l
- 8. Miscellaneous comments and discussion .
8-1
- 9. References 9-1 Appendix A- Vermont Yankee Nuclear Power Corp.. A-1
. Response to Questions :
9 4 [ 6
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trh n, vr unn vinm inwww r, r, ! A REYlEW OF THE YERMONT YANKEE CONTAINMENT SAFF i Y 5iUi)V
- 1. INTDODUCTICH Thlo Poport peccento the ecoulte of a rovtow of tho : .
Vermont Yankee Containment Safety Study prepared by the Vermont Yankee . . Nuclear Power Corporation. The revleiv was undertaken for and T9nded by j the State of Vermont, Dept of Public Service. The objectives of the review ., were as follows: - j
- 1. Perform an overall evaluation of'the Vermont Yankee Containment j Safety Study.
- 2. Determine whether the containment study provides a reasonably
., accurate estimate of the probability of containment failure from severe accidents. If the estimate is judged to be not accurate, determine if it is too high or too low and by what magnitude.
- 3. Identify technical shortcomings in the study and estimate their erfeet on the containment failuit p vbability.
- 4. Determine what changes can be done which would increase the probability that the Vermont Yankee plant containment would not fall in the event of a severe accident. Included will be an evaluation of changes that Vermont Yankee Nuclear Power Corp. has identifled in its report,
, changes suggested by the Nuclear Regulatory Commission but not accepted by Vermont Yankee, and other po,tentially of ferttyp changen un sept. :t, sveo, tne vermont iapr4a Nuclear Aoiuvr coi por stiva submitted I
. ' to,the Nuclear Regulator Commission a document entitled " Vermont Yankee Containment Safety Study".',This Stucy evolunted and estimated tho , ,
- #cbability of a larg6 f 616' Ass of radiesetivity f.=em the Vermont Yonttoo
' Nuclear power Plant as a~ result of a severe ar.r:Ident. The study employed j data and methodology from a technical discipilne commonly referred to 00 probabilistic risk assessment. A brief description of this discipline along with an evaluation of its strengths and limitations 1,s provided in the following Cubcoction. .
IA probabilistic Risk Asseesment - In 1975, the U.S, Nuclear Aegulatory Commission published the Reactor Safety Study (WASH-1400)
' This document provided the results of the first serious attempt in the U.S.
to quantify the risks from the operation of nuclear power plants. The study employed a relatively new technical discipline refered to as probabilistic risk assessment (pRA). In this approach, accident intittating events are postulated and their frequency is estimated based on actual 1 l
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if0V 03_ 'B6 148 31 VTDPS 8028282342 P.7 l i, plant operating experience or application of other data. These initiating . - events are used as starting events on " event trees", which are basically - logic diagrams which consider the plant response to various combinations . -
. of safety system success or failure following the postulated initiating event. . flg.1'provides an example of an event tree prepared for a nuclear ,
power plant PRA study. The headings on the event tree represent safety systems (or safety functions which are sypplied by systems within the plant); The horizontal itnes across the event tree reortsent the various possible accident sequences which can occur following the initiating ; event.! The lines enter each event tree heading from the left and branch : into two segments. The branches represent success or failure o' f the system represented by the event tree heading, with the top branch indicating success and the bottom branch Indicating failure. In some instances, there is no branching, which 1ndicates that the system has . failed as a consequence of an earlier failure of a different system. The , column on the far right of Fig.1 indicates whether each of the various sequences results in successful cooling of the reactor core (s), the core overheates and releases radioactivity to the containment (cm), or the sequence is transferred to another event tree due to a failure (LOCA). The'next step in the PRA process is to quantify.the probability.of the accident sequences delinisted on the event trees. This 'is done by combining the frecuency of the initiating event with the success or failure probabilities, as appropriate, for each branch point of the individual . accident sequences. The failure brobabilittos for each of the event tree headings is g'enerally determined by one of two methods. If the event tree heading represehts a system for which sufficient data exists from testing or actuation, then f ailure probabilities may be obtained from,this source. If insufficient data exists, then fallure probabilities are estabilshed by c the fault tree techoloue. In this instance, fault trees are constructed for
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the event tree system of interest. Tiie fault trees may also be thought of as logic diagrams which subdivide the system into its mechanical, ; electrical, and hydraulic components which are arranged through the use of
]
logic gates such that the manner in which Individual components contribute to system failure are depicted. An example of a fault tree is given in Fig. 2. By assigning tach individual component a failure rate which is obtained from applicable data bases, the failure rate of the system can be calculated. After the' probability of each accident sequence has been computed, the
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e probability of various containment failure modes is estimated for each sequence or group of sequences whIch have utmilar characteristics in the . I context of imposing a threat to the containment integrity. The l containment failure probabilities are estimated by analyzing the
-- containment response to each of the accident sequences, and assessing the likllhood of various f ailure modes. Thls process may utilize containment event trees in which the event tree headings are represented by the
' containment safety systems.- , j ,
Following the assessment of containment fallure modes and associated probabilities, a radioactive source term is estimated for each containment failure mode. This source term defines the amount of each important radioactive species in the core which is expected to be re' leased to the environrnent following each of the containment failure modes. This source term is then used to calculate the public health erfects given a speciff'c containment fallure mode. By corsbining the accident sequence probabilities with the containment failure mode Probability and the public health effects associated with each containment fallure mode, an estimate of public risk can be obtained. f This risk is generally expressed as etther an early fatality which results l fecm large doses of radiation, or a later)t health fatality (cancer) which j can result from lower doses of radiatioh. . 1
' This description is & very simpilfled ov' rvie e hr olf the PRA process. Ih reality, the process can be extremely complex and exhaustive. A major I '
' nuclear plant PRA may evaluate millions of accident secuences. Sucti an '
af fort typically requires several million dollars and requires 40 or more man years of effort. As will be discussed in later sections, the Vermont Yankee Containment Safety Study is a limited PRA study which does not, and was not intended to, include all of the elements of a full scope PRA. I Since the completion of the first major PRA study in 1975, there have been some 35 additional PRA ef forts completed for a variety of plants in the U.S.; and several PRAs nave also Deen completto for foreign plants, in general, these studies have estimated low public risks from the operation of nuclear power plants. Typically, the probability of a serious (core damage) accident nas'been found to be on the order of one chance in
' ten thousand per year (usually written as IX10'4/yr.) or less, and early fatality risks have generally been estimated at less than one chance in s
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,5 million per yr. (1X10-6/yr.) for a person residing within one mile of tne -
plant. Similarly, latent cancer f atalities have been generally estimated at less than one chance in a million per yr. For comparison, the average risk of death ftorn accidents and adverse effects for an individual in the U.S.15 about one chance in 2500 ( 4X10~4/yr.) and the average risk of death from cancer is 1.9X10~3/yr. based on recently published statistics for the year 1982 (most recent year compiled)1S, The advent of pRA in the assessment of public risks from' nuclear power has represented a major and significant advancement tn the assessment of nuclear power safety. Prior to 1975, no realistic and comprehensive - estimate nf the risk of nuclear power Generation existed.! Nuclear' power [ safety was based on designing the plants to withstand a spectrum of , { design basis accidents which were selected in an ef fort to encompass what were thought to be all important accidents. This was' combined with the defense in depth philosophy which required that multiple barriers (including the fuel cladding, the primary system, and the containment) exist to inhibit the transport of radioactive material from the core to the environment. The PRA approach, on the other hand, provides a sys,tematic and Integrated assessment of all conceivable accident sequences coupled with a comprehensive evaluation of plant systems result, trig in a numerical estimate of public risk. In addition to being used to provide overall risk estimates, the PRA ap'proach has been used extensively to provide some rfsk pe'rspective on numerous reactor safety issues, and has been a major factor in several NRC decisions regarding nucleap power plant safety.
. Despite the wide acceptance and extensive use of PRA in nuclear pown plant safety and risk evaluations, there remain shortcornings and limitations to the PRA approach. Further, several important areas extst where controversy and disagreement can be found among PRA researchers.
Some,of the more significant of these problems is discussed below: n Inadeounte Data- The accurate estimate of risk from the PRA approach requires that statistically significant and appilgable data exist to evaluate the frequency of Initiating events, system failure rates, and component failure rates. In numerous Instances, such data is limited, and in some cases, such as .large seismic initiating events, it is lacking altogether. . b Human Error- During the course of a postulated severe accident, 1-4
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opportunttles exist for the plant operators to intervene in an attempt to. Improve the potential for adequate core cooling. It is also possible for these attempts to have a deleterious ef fect if the operator commits an . . error or acts with inadequate or incomplete Information from plant ~ instruments. The potential for and probability of such action Is extremely ; difficult to evaluate quantitatively since human behavfor under vartous ' stress levels is uripredictable. . e system Dependencies- Detailed system analysis performed as part of past PRAs has revealed the existence of important system dependencies, .
,or ns i at nces where a system failure or an inttf ating event can cause {
additional system f ailures by virtue of links between the systems. l Sometimes these links can be subtle and troublesome to analyze properly. It is also diffleult to demonstrate that alt such important dependencies have been found as part of a PRA since there does not exist a validated methodology which is universally accepted for finding dependencies.
- d. External Event Initiators- In some pRA studies, accidents which are initiated by events external to or not associated with the operation of the plant have been estimated to be important contributors to risk. External event Inttlators include earthquakes, fires, and high winds. It has been extremely difficult to assess the frequency of such events because, in general, they have to be more severe than any event recorded at the plant site. Thus, frequency evaluations must rely on extrapolations and judgement. Furthermore, it is not always easy to evaluate the plant response to such events since the important contributors are frequently beyond the design basis of the plant and no experience exists.
- e. Severe Accident phenomena- $1nce there has never been a severe accident in a' nuclear power plant of U.S. destgn which has progressed beyond core: dame.ge with!n the' reactor vessel, the assessment of such progression must depend 'on analysts and a limited amount of small scale, highly idealized experiments. The physical.and chemical processes in these accident progressions are predicted to be extremely complex Interrelated events. In several areas, analytical techniques are limited, and debate and controversy surrounds the most likely accident progression '
scenarios. '
- f. Uncertainties- Due to the lim!tations described in a. through e. '
preceding, as well as additional factors, all PRA assessments contain a ~ considerable amount of uncertainty. The quantification of, and appropriate - acounting for, this uncertainty in the utilfration and Intrepretation of the PRA results has been an area of controversy. It is usually not possible to rigorously quantify the uncertainties involved in a PRA assessment. , 1-5
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'.4 It should.be noted that additional research and anlysis is underway in all . , ,
of the above areas, and'lt is anticipated that PRA assessments will' become more accurate and complete as this work progresses . The foregoing areas describing major limitations and uncertainties with ret,pect to PRA assessments will be described in more detall in subsequent sections of this report as they apply to and influence the results of the Vermont Yankee Containment Safety Study. The next subsection of this . - introduction provides a brief overview of the Vermont Yankee Containment
' Safety Study.. j ,
ILB Overview of the Vermont Yankee Containment Safety Studv- The subject of this report is a review of the document entitled " Vermont Yankee Containment Safety Study" which will be hereaf ter referred to as. : the VYCSS. This document describes the results of an assessment of the Vermont Yankee nuclear power plant to determine the probability of a se'v ere release of radioactivity from the plant. The study was a limited PRA assessment performed principally to examine the integrity of the containment structure under severe accident conditions, it does not, and was not intended, to provide an exoticit estimate of public risk from operation of the plant. The approach taken in the study was to modify the ! analysis and results of the Reactor Safety Study (WASH-1400) in an attempt to render them applicable to the Vermont Yankee reactor. The ' plant used in the Reactor Safety Study was the peach Bottom nuclear'
. power plant, a plant similaf, but not identical to, the Vermont Yankee plant. The modification process in the VYCSS Involved adjusting the probability of the Reactor Safety Study accident sequences to: 1) account for Vermont Yankee plant. specific design features and 2) utilize more current data and accident sequence progression analysis. A further modification involved a more realistic assessment of the containment response to severe accidents based again on Vermont Yankee plant specific design features as well as increased knowledge regarding containment response and failure modes from research and anlysis conducted since ~
1975. A more thorough description of the methods, assumptforas, and - procedures utillted in the VYCSS will be provided in subsequent sections of this report, particularly in those areas that are important to supporting the results. 1-6 4 _ _ _ . . . _ _ _ _ _ _ _ _ _ . _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ ,__ , , _ _ _ _ _ _ , _ _ _ . ___ ,.___,,_,___,m,_______m,,,.,m__ _ _ _ _ _ _ _ _ _ _ _ _ _
me m v 14:as vires nyyyup p,3p . e,- , , The principle results stated in the VYCSS are that the probability of core melt for the'Verniont Yankee plant is 3X10(-5)per yr. or about one chance in 33,333 per year. The probability of having a large release of '. ~ i I radioactivity from the plant was estimated at 2X10(~9)/yr., or about one chance in 500,000 per year. In other words, the VYCSS concluded that. .
.there was a 7% chance of having a large release following a core melt '
BCCident SeqVenct, Conversely, there was a 93% chance that the containment and its support systems would prevent a large release - following a core melt accident. The purpose of the review described in . this report, as , stated in more detail in the first part of this section, was to attempt to ascertain if this result is valld, and, if not, what a more realistic value:might be. The summary and major conclusions from this - review is presented in the following section. . 1 9 e o e f e e l O t 4 1-7 reg a
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- 2. SLM1ARY AND CONCLIEt0t6- The Vermont Y'ankee Containment Safety 5tudy (VYCSS) provides an estimate cf the probability of severe core damage accidents at'the VerTnont Yankee nuclear power plant and the .
; likelihood and mode of containment fallure and resulting radioactive -
release as a result.of such accidents. This review.of the VYCSS consisted .. primarily of comparing the assumptions, analysts, data,.and results wIth , current ava;1#61e information regarding the probability and pr*Fession of severe acetcents m order to ceterame n tne muoy was t.mmlate.t alth this cwrent knowledge. The report was also examined to determine if any
.important omissions mlpt exist. Where' deficiencies or omissions were found, an atlangt ws4 made to evaluate their significance. The primary
~
fitidless and conclusions of this review are as follows: A. 'Overall results - The YYCSS is considered to present a reasonable . estimate of contaloment f ailure probabilities fro'm severe occidents which , could be initiated from internal events only. Although the VYCSS does not' .i
. provide any uncertainty analysis, and this is considered a major deficiency _
.as noted in item C following, based 4 selected sensitivity studies and i contparisons performed as part of this review, the VYCSS probability :
resuits for internal events appear to be generally consistent with present: knowleoge regarding severe accident behevlor and within the range of l large uncertainty associated with suchbehavior. With respect to external event initiators, it is concluded that seismic events have the potential to increase the probabtifty of containment failure in conjunction with severe accidenlu et the Vermont Yanleet plant. B. Inadequate characterfration of results wf th respect to limitations regarding omission of external events- The VYCSS is considered deficient tin not adequately qualifying the,results with respect to omission of external events. An evaluation of the potential contribution from external events is given in Sect.8. . C. Inadequate characterizetten of results with respect to
. uncertainties- The VYESS does not adequately 1) quellfy the results in !
terms of uncertainties 2) discuss the extent and implications'of uncertainty in the results, or 3) provide a basis for ranges given on the results. This aspect is discussed in detall in Sect. 8, item 2. i . D, ihadequate basis for some frequencies and probabilities- The report~ does not prwide an adequete tests for some of the frequencies and - probabilities used. Based en a response from Vermont Yani(ee touclear . Power Corporation, the bests for these frequencies and probabilities is now considered to be adequately documented (Sect. 8, item 7). 2-1 !
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fC/ 03 '96 14:37 VTDPS 8028282342 P.16 9. E. Inadequate description and implication of radioactive release following containment failure- The description of potential consequences , from, and characterization of, radioactive release ranges used in the , j VYCSS are considered inadequate. Details are provided in Sect. 8, item 6. l F. Reasonable assessment.of containment failure issues- The VYCSS ' provides a reasonable assessment of containment failure scenerlos and ) related probabilities. However, the report is c'onsidered deficient in not l exploring the offeet and likellhood of alternative scenarios which would encompass the uncertainty involved evaluating ;Important containment failure modes. . G. Generally adequate assessment of containmentintregrity improvements- With a few exceptions with respect to 1) Operation of drywell sprays (S'ect. 8), and 2) Drywell liner f ailure from molten core interaction (Sect. 6,8), the VYCSS appears to adequately address NRC suggestions regarding improving containment integrity as well as other recommendations identified in the VYCSS. Two additional plant changes and modifications with the potential for reducir)g the. probability of a significant release were identified during the review. These changes were
- 1) Increase capacity and reliability of drywell coolfrs,'and 2) changes to the logic and control for the main steam isolation valves. These aspects-are discussed in detall in Sect. 7. ,
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. 3. PROCEDURE- This section. describes the' procedure utilized in reviewing the VYCSS. It is important to note that resources available for the revie'w were not sufficient to perform an Independent assessment of the VYCSS by performing a comprehensive PRA from basic methodology. Indeed, the - -
VYCSS Itself did not use this approach as noted in Section 1 preceding. , Further, some of the supporting data and analysis referred to in the VYCSS ' was not obtained in time for the review, and a visit to' inspect the plant ,
, was not carrled out. Such a visit would have been of limited value since the plant is operating, rendering access to the containment impossible.
Due to these limitations, the review relled significantly on comparison of
. the VYCSS assumptions, data, and results with the substantial body of
, similar assessments (other than the Reactor Saf ety Study) currently available in an attempt to determine if the VYCSS results were consistent with and supported by this additional Information. It should be emphasized and recogn!!ed that in several areas, to be noted in subsequent s'ections, data and information are lacking to the extent that major i
diff arances of spinien adst' araer,g ressanslBle Pf1A resesPehers. Thuo, it is not possible to determine which approach or result is correct. These , - unknowns do not render the VYCSS results (or any PRA results) invalid. They do, however, contribute to uncertainties in the results, and the significance of such uncertainties can be important in qual!fying and l Interpreting the results. The basic approach employed in tne review consisted of five steps, as follows: ,
- a. Accident sequence probabilities- The VYCSS Important accident
. sequences were compared with several other pRA studies for similar nuclear power plants to determine if the assessed probabilitico woro-consistent, and if not, what vallo reasons might exist to render them .
I inconsistent.
- b. Accident sequence completeness- The VYCSS Important accident ,
sequences were compared with several other PRA studies for slmlbr plants to determine if any sequences fourid to be impnrtant in other studies are missing from the VYCSS and, if any were tound, to determine if the missing sequences would be applicable to VYCSS and what their effect would be on the results if they were included.
- c. Condition'al Containment Failure Probability- The VYCSS conditional containment failure probability from dominant accident sequences were compared with two very receht independent assessments for a similar plant design. Dif ferences were evaluated, and an estimate of the effect of any changes in the VYCSS judged to be valid on the basis of the comparison was made.
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' . c. Containment Fallure issues- A rev!ew was performed of the current .
state of knowledge.regarding traportant containment f ailure modes and processes from severe accidents to determine if the VYCSS assessment - reflects appropriate consideration of this Information, and if the VYCSS results in this regard are properly quallfled and uncertainties . . . acknowledged or accounted for The potential affeet of any discrepancies In this area were evaluated. ,
- e. Containment integrity improvements- A review was made of various sugges.tions and analysis for providing plant modifications to improve the ability of the Vermont Yankee containment to sustain the loads imposed by severe accidents. Such improvements include those 4 suggested by the NRC, the VYCSS, and othere found potentially beneficial as part of this review, j f, General Deficiencies- Any additional deficiencies,'not directly l related to the above areas, were Identified, described, and their
- implication evaluated to the extent practical. .
The following Section provides the results of the first two review areas , (a. and b.), while Section S provides the results of the review of item c. Section 6 considers item d, Section 7, item e, and Section 8, item f, i
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- 4. REVIEW OF WCSS DOMINANT ACCIDENT SE0VENCES i This section presents the results of a comparison between dominant '
1
~ accident sequences from several pR4' studies 'to those found to be
- dominant 1.n th'e WCSS. The main focus of the comparison was to ' -
determine if any important sequences might have been left out of the WCSS, or if any inconsistencies appeared to exist jn establishing accident j probabilities. These two aspects are considered separately ~ln subsections which follow. It was considered important to make this comparison l primarily because the WCSS relles on the sequences found in the Reactor "
- Safety Study which is now some i 1 years old.- Several assessments of plants similar to WCSS have been completed in the interim, and these are ,
Used as the basis for the comparison. - i
. 1 prior to proceding with the results of the comparison, it is useful to .
re-arrange the WCSS results. This will facilitate subsequent evaluations . on the significance of differences which may be found, and will provide - some perspective on the WCSS results which are not currently directly displayed in the report. Table i is a re-arrangement of the WCSS results, and has been prepared based on Information contained in Table 4.7, Pg.108 of the WCSS report. , , Table I lists the accident classes found to be important in the WCSS
' study. These classes contain accident. sequences which have common ;
features in terms of their impact on the containment, and as such, are characterized by similar initiating events af)d system fall,ures which are sumrnarized in the second column. The third column provides th'e computed,
, probability of each accident class. This value.ls followed by a percentage .
contribution (in parenthesis) of each accident clas's to the overall core melt probability.L The next column l's the fract'lon (expressed in percent) of of each accident class that was found to fall in the EH or EM radionuclides " release categories. EH designates a containment failure mode.wl)lch results in an early release with high levels of radioactivity, vihile EM is an early release with medium levels of radioactivity. These designations are important because they represent the two classes of release which were i considered to be high release levels in the WCSS report. All other release I categories were considered to be low (including no containment failure) or much later releases implying much lower public health consequences. The last column in Table 1 is the overall fraction (again expressed in percent) that each accident class contributes to the total probability of either an EH or EM release.
~
Table I serves to illustrate some importa'nt features relative to the l WCSS res0lts in terms of the overall relationship between accident ! i A'
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" classes and 'signifIcant releases. For example, the most significant accident class in terms of contribution to core melt probability,is class .
lA, with a 43% contribution. However, this class is an insignificant contributor to the overall EH and EM categories. On the other hand, class
. IC is a relatively low contributor to core melt probability (84), but a very.
significant contributor to the overall EH and EM categories.;This is because these accident sequences were determined to have a relatively .. high probability,0f damaging the containment and therefore;resulting in a
]
significant release.' This is illustrated by the fractional release columns which Indicate that 20% of the class IC accidents are expected to result in: an EH release. Thus, as illustrated by the table, there is a ~significant 1
. Variation in the relative contribution of each accident cattagory to both core melt probability and contribution to significant release.' These relationships will become important in evaluating the influence of -
potential changes to the WCSS results. For example, if it were found that a more realistic estimate of class IA core melt probability were 1.33E-4, 3
; a very significant ten-fold increase, the core melt probability would be increased to 1.63E-4, a factor of about 5. However, the increase in EH and EM release probalbilty would be very modest, under 205 in.both cases. 1 Conversely, a modest increase in probability of IC accident class - -i sequences, or an increase in the fractional contribution of these sequences to_ EH or EM categories would/esult in a significant increas,e in IH.or EM release probabilities. '
l 4 A[ Revfew of Accident Secuence'Probab111ttes- This review consisted of two phases. In the first phase, the WCSS results were examined to determine if the adjustments and modifications to the accident' sequence .)' ' ~ pro 6 abilities used from the Reactor Safety StudyCI)(RS$i. appeared to be: . valid. As indicated previously', the WCSS did not undertake an independent , assessment of core melt probability, but rather used the RSS results as a ) baseline from which modifications wer,e made. The second phase of the .
)
review consisted of comparing the WCSS results with numerous results {' from other pRA studies completed for plants similar to Vermont Yankee. These results from these two phases are described below dA.a. Modifications to RSS results - The RSS estimated that the core melt probability for the Peach Bottom j reactor is 3.2E-5/yr. The WCSS result is 3.0E-5/yr. These two results J '
. can be considered identical since the uncertainties in the estimates are significantly greater than the difference. However, the accident sequence (
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l probabilities which make up the total core melt probability for the two i studies were not identical in that the WCSS made, in some cases, .
.significant modifications to the RSS sequence probabilities on the basis of plant design dif ferences or what was considered to be more applicable
. data. This dif ference can be important sincs different sequences, as noted previously in this section, can have markedly dif ferent influences on the !
;11kllhood and mode of containment failure leading to large releases. .
Section '4 of the. WCSS provid6s a requantif Ication of the RSS accident sequence probabilities. A review of this section revealed no significant errors or omissions, and it is therefore condluded that the WCSS requentification of RSS accident sequence probabilities is reasonable and .
.no significant changes to any accident sequence probability were found to be warranted. However, two general deficiencies of note were found; 1)
The basis for some of the numerical changes made to the RSS sequences
-were not adequately delinf ated, and 2) There was an inadequate treatement of uncertainties in.the requentification process. The first problem was largely resolved on the basis of answers to questions provided by the
. Utility (Vermont Yankee Nuclear Power Corp.) as given in Appendix A of this report, and the second deficiency is explored further in Sect. 8.
Further comments on Section 4 of the WCSS are also provided in Sect. 8, but none of these are judged to be significant in the context of the overall WCSS results and conclusions. l 4.A.b. Comparison with Recent pRA Studies ' Since the publication of the RSS, as noted previously, there have been numerous additonal pRA studies completed for U.S. plantsJ Several nf these have been _for plants very similar to Vermont' Yankee, which is a Bolling Water Reactor, Type 4, with a Mark I containment. For the purposes of this comparison, four other PRAs for Type 4 BWRs will be used. Table 3 lists the plants for which these PRAs were completed, gives l the organfteHeMvhich sponsored the study, and the performing organization. As notedJn the Table, the four plants are peach Bottom, l.lmerick, Browms Fen y, and Sig vliaro. Two of these plants (Peach Bottom and Limerick) have riark I containments, wn1le Limerick ano Snorenam have Mark 11 containments. However, for purposes of core melt comparisons, the difference in containment design is not considered important. The BWR Type .4 design is essentially identical in terms' of systems included
. which can influence the probability and timing or core melt accidents (7) .
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A9 noted previoucly, thcre h3Ve been two Independent assessments for the Peach Bottom plant since the RSS was published. The f test such Peach . l Bottom results',~as indicated in Table 3, are from an NRC updating of the i , RSS assessment. These results are not yet published, but were presented - at a recent meeting of the Advisory Committee on Reactor Safeguards (2) , The second Peach Bottom assessment is from the industry Degraded Core
. RuleMaktog (IDCOR) program'whte'h 18.e progeom to provido an Independent ,
(from NRC) evaluation of StYtre 3lCQldent response in nuclear power plants. The Limerick and Shoreham PRA studies (references 8 and 9) have '-
. undergone NRC review. These reviews (referencest o and11) resulted tc a requantification of accident sequence probabilities in those instances .
where the reviewers felt the PRA study was deficient.
.The results of all' of th' e assessments described in the preceding are presented in table 3. In this table, the dominant accident sequence :
probabilities have been arranged into the same accident classes, as
. appropriate, that were'used in the WCSS in ~ order to factittate comparisons.1 The table thus shows the accident class probabilities for each study, as well as the total core melt probability (in the last row).
' Co paring the total computed co e melt probability (last row of table 3) shoWs that the WCSS estimate is about in the middle of the results given, with three assessments lov(er than WCSS, and four greater. The WCSS .
res' ult'ts about four times greater than the lowest value (IDCOR-Peach
. Bottom), and about seVen times less than the higest (Browns Ferfy). This comparison 111ustrates two observations; 1) the WCSS results is within <
the range, and therefore cons 1 ster.t wIth, other ' core melt probabi11ty assessments, and 2) there is substantial variation among the results used in this compartson (the range extends to almost a factor of 26 from lowest to highest result). The reasons for these substantial vartations are not considered in depth in this study. The differences are primartly related to credit given for operator action, the use of different data bases, and, to a lesser extent, plant specific design differences and site specific - initiating events (such as los,s of off site power). ,
' In examining individual accident class probab111tfes, it was fourid helpful to combine some of the WCSS classes in order to render the results comparable, since they were considered as single sequence classes in some of the other studies. These include classes l A and ID as well as IC e e
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and IV as indicated in the first column. Further, some of the studies fncluced accident sequence classes which were not considered as e,eparate , in the VYCSS. These include those designated as TPOl and TPOE on the
" table, in these cases, the sequence resulted in a loss of coo! ant accident .
(stuck open reiteNalve) and were apparently considered as LOCAS in the WCSS (and RSS) study lor accident class Ill. In comparing the agcident class probability results from table 3, it is noted that, in som( cases, substantial differences exist. For classes I A, ID (loss of make-up water to the core) the WCSS results are within and
- nearer the higher e'nd of the probab111ty range, and thus are consistent with the other assessments. For class IB, the VYCSS results are also within the probaotlity range of the other assessments, although near the .
lower end of the range. -This sequence is loss of all AC power (plant
- blackout), and was assessed in the WCSS to have a somewhat lower probability than the RSS because of the proximity of, and electrical connection to, the Vernon hydro electrical generating station. The credit assessed in the VYCSS for this additional power source was found to be reasonable in terms of reducing the probability of AC power loss.
For the next combination of classes. IC and IV (anticipated transients
' without scram) the WCSS probability assessment is also within the range of other accessments. being about a factor of 5 higher than the lowest (NRC assessment for Peach Bottom) and a factor of about 12 less than the .
highest (Browns Ferry). The major reason for the difference between the VYCSS ATWS probability estimate an' d that for Browns Ferry as assesped . by EG&G ts credit'taken in the WCSS for implementation of the. ATWS rule (6) This rule'is being implemented at BWR plants and contains features which will have a significant influence in reducing the.
- probability of ATWS. The rule was being formulated at the time the Browns Ferry assessment was published (1982). Based on the review of -
the VYCSS and this comparison of other assessments, it is concluded that the WCSS estimate of ATWS probability is reasonable. L For accident class !! (loss of containment heat removal), the VYCSS assessment is again within, sithough closer to, the higher probability end of the range of other assessments.The VYCSS result is a factor of about 200 higher than the lowest (Peach Bottom, NRC), and a factor of 50 lower than the highest (Browns Ferry). This class therefore has a very large range. The primary reason f.0,r the difference among the assessments is
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.- 1 the credit taken for operator action and use'of plant equipment not
. specifically designed for containment heat removal. For example,.the two . .
assessments for peach Bottom take credit for manual venting of the containment drywell. 'This accident dev'elops very slowly (many hours),- . and time is available for positive operator intervention and alignment of. J other equpiment which may.be available at the plant. The WCSS result is considered reasonable and consistent with other results although - substantia 1 uncertainty, to be considered later, is associated with the j probability estimate for this class. , i . Table 3 shows for accident class ill (loss of coolant accidents) Otat the WCSS result is at the high end of the probability range, but with.In a - factor of 7 of all assessments except the very low Limerick result. This is considered reasonable agreement. The V accident class (containment bypass) was not found to be a ' . significant contributor for any nf the assessments except that the Shoreham PRA assigned this class a probability of 2E-7, which is not a significant contributor (less than 12) to the Shoreham core melt
; probability of 4.4E-5. Since the WCSS did not provide a basis for the conclusion that this sequence would be a negligible contributor, the utility (Vermont Yankee) was asked for the basis. Their response is ini;luded in ' Appendix A as response *10. The response indicates that this accident -
class is estimated to have a probability of IE-7, quite comparable to the Shoreham result, and insignificant in terms of the overall.VCYSS cor,e melt probability (less than a 15 contributor). It should also be noted that,
- based on Table i results, this accide.nt class would not be a significant contributor to the probability of a large release even if a large fraction of .
the accident class. contributed.to high releases. As shown in Table 3, the two. accident classes (TPol and TPOE) not considered separately in the WCSS were not found to be significant contributors (less than 102) for any of the studies. . . From this comparison, it is concluded that the WCSS assessment of core melt probability is consistent with other independent s.tudies and pRA reviews. Further, all important accident classes appear to have been considered. As noted in the Table 3 comparisons, the overall core melt probabilities ranged from a factor of 7 higher than the WCSS result to a factor of about 4 lower. , 4-6
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TABLE l-Distribution of Significan.t Radlonuclide Release per Accident- - Class from the vermont Yankee containment Safety Study ACCIDENT Core Mlt. Fractional 2 overall N . Class Type Prob (R) : .EH EM EH EM
. I lA Loss of Mekeup-High Pressure 1.33E-5 (43) ' 0 0 1 2 x
18 Plant Blackout 6.20E-6 (20) 0 2 1 7 1C ATWS-Loss of Makeup 2.60E-6 (6) 20 11 '96 17 10 Lossof M6keup- Low Press 3.90E-6 ( 13) 0 1 0 ,3 11 Lossof Containment Heat Removal 2.10E-6 (7) 26 0 33 lll Loss of Ccolant Accidents 7.30E-7 (2) 0 0 0 0 IV Anticipated Transtant W/0 Scram 0 30 0 40 , 2.20E-6 (7) . i I g F e 0 0 4-7 . 1 5 4 0
, ce
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~.NOV'03 006 84:44 VTDPS 8028282342 P.26 Table 2. PRA Studies for Bolling Water Reactors, Type 4 Plant 5ponsor Performing Org. Date Pubitshey Reference Path 50tt0m U.S. NRC $6ndle Netionell.ebs . (serly1987) (2)
' PesdiBottom IDCOR IT Corp; veris con- .1986 , (3)
.
- tractors .
'lImnrtri Ph)l. Electric $ctenceAppIlostigns, 1983 '(4)
. Co.
General Electric . Shoreham Longis.: Lighting ScionesAppiteetions 1963 (5)
~ Co. .
Browns ferry U.S. NRC EG&G Idaho inc. 1982 (6) 9 l' 6 i . \ 1
)
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.t' TABLE 3-COMPARl50N OF DOMINANT ACCIDENT SE0VENCE PR0 ABILITIES CLAS$ VYC$$ PEACH BOTTOM BROWNS Lir1ERICK $HOREHAM 1000R NRC FERRY- PRA NRC. 'PRA NRC ,
IA,1D - 1.72E-5 4.1E-8 2.4E-7 5.5E-7 6.0E-5 6.0E-5 6.6E-5 5.0E-5 ,
~
15 6.2C-6 4.5E-7 6.70-6 2.9E-5' 3.1E-5 3.1E-5 1.3E-5 1.3E-5 / 3.7E-6 '3.7E-6 4.5E-5 4.5E-5 IC,lV;, 4.SE-6 7.3E-6 1.0f-65.5E-5 ll 6E-7. 1.5E-7 1.0E-6 1.0E-4 ' 3.2E-6 3.2E-6 i.lE-5 9.0E-6 lli 7.3E-7 1.4E-7 1.lE-7j.(1)- 2.4E-9 (1) .5.5E-7 (1) Y (2) (1) (1) , (1) (1) (1) 2.0E-7 (1) lTPO!W (4) (5) (5) 1.1E 9.0E-7 9.0E-7 1.7E-7 1.7E-7 TPOE(6) (4) (5) '(5) 1.0E-7 ' (5) (5) (5) (5) TOTALCt1P 3.0E-5 7.9E 9.8E-6 2.0E-4 1.5E-5 9.9E .,5 4.4E-5(731.2E-4 (1) Not found to be 4 eminent accident citos (2) Estimated by VermontYankee to tw 1.0E-7/yr., see Appendix A Answer 810 (3) Transient accident with stuck open relief valve followed by loss of containment heat removal (4) Considered, es in the R$$ to be e )oss of coolant accident and included under til ( 5) Not found to be e separate dominent eccident cleos (6) Transient eccident with stuck open relief valve followed by core injection failure (7) This value is actuelly deter 1 bed se a " core vulnerable" condition in the Shorshem PRA and a small factor was appited to estimate the ws melt penhnh111ty given a cart vulnertbit 900@0n, g . es 4-9
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' S. COMPARISON 0F CONTAINMENT FAILURE PROBABILITY WITH RECENT ~ ~ '
RESULTS- This section provides'a comparison between the VYCSS
- assessment of containment Iallure proDaottity ano mose wItn two recent ,
sinillar assessments for BWR Type 4 plants with Mark I containments. ' These essesstnents were chosen because they are the most recent w available, they were done for a reactor and containment design (Peach Bottom in both cases) similar to Vermont Yankee, and they were performed by tivo dif ferent agencies, namely the IDCOR: group and the NRC and their
, contractors. It should be noted that many of the assessments used in the.
pre' ceding section are not suitable for this comparison for various rennns. The Limerick and Shoreham assessments, for example, are for plants with
, Mark Il containment designs and would not be suitable for the Mark i design of VermontYankee. Further, the Browns Ferry assecsment used in '
- Sect 4 did not include an independent assedment of containment f ailure ,
probability. . The IDCOR and NRCiassessments of Mark l containment failure robabliity assessment have been recently summarized in a;draf t report l9 the results of which were presented at a recent ACRS me6 ting (2) These results 'were used, with some intrepretation, to prepare table 4. It should be noted that the results, particularly the NRC results, are preliminary and subject to revision as their assessment is refined. Table 4 considers two accident sequence classes; Station Blackout (designated at Class IB.In Table 3 and in the VYCSS assessment) and ATWS (Classes IV and IC). These sequences have been found in previous PRA assessments to be the dominant contributors to the proability of containment failure resulting in a significant release. Examination of Table 1 reveals that these two sequences were dominant contributors in the VYCSS assessment also. For example, ATWS sequences contributed 962 to the probability of the most serlous release (EH), and ATWS sequences contributed 572 to the EM release probability. The Station Blackott' sequences were less significant,
. contributing only 12 to the EH category, and 72;to EM.
Table'4 shows the' fractio'nal release probability for the two sequences for
- each of the studies. It should be emphasized that the descriptors high, med and low are qualitative and are based on judgement regarding the
- 51mtlarltles of the releases. The footnotes are provided to give additional Information regarding the mechanics assumed for each release. Thus, 5-1
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] > these categories are equivalent only in a rough sense and are used in an .J attempt to assess any significant differences rather than as the basis for ~
I a recuantification of the CYVSS results. The table indicates that foNhe Station Blackout accident, the VYCSS would estimaN a'much lo.wer conditional f allure probability of a high or medium release than either'of 7the IDCOR or NRC studies. As noted on tiie . table, the primary containment failure mode contributing to these releases from the'lDCC4 and NRC studies are overtemperature failure causing - drywhil lea:kage/and drywelliiner failure from contact with molten core debris. The primary rear.on for this difference ppears to stem from .two factors: 1) The VYCSS assumed a lower likllhood of drywell liner fall'ure beceuse of the smaller amount (about 1/2) of molten core material
^
j available due to the lower power level coupled with the fact that the drywells for both the Peach Bottom and Vermont Yankee plants are slmtlar in size, and 2) The VYCSS assumed that:drywell sprays would more likely A be available because of the possibilty of using a diesel driven fire pump
- which can be connected through the RHR system and used to pump water to-the sprays under Station Blackout conditions. As pointed out in the VYCSS,
~
the use of drywell sprays would inhibit both drywell liner / molten core Interactions as well as overtemperature drywell failures. The CYVSS reasoning is considered valid but not conclusive in terms of the ef feet of less core material reducing the 1.lkllhood of molten core /drywell* i' liner interactions.(This issue (s explored in depth in the following _ section). However, the use of the dryWell sprays appears to be - questionable due to the procedures trivolving their use: This issue is' ' i explored in Sect. 8 of this report. Given these considerations, and the further discussion in this report, it is extremely diffleult to assess;the ' conditional probability of drywell failure for the station blackout sequences. .There is very little applicable experimental data, and any p estimate relles heavily on assumptions and analysis, neither of which have i been verified. To assist in evaluating the significance of this issue, a simplified sensitivity study was undertaken. By use of the Information in table.1 in conjunction with table 4.70f the CYVSS.asse'ssment it can be readily shown that raising the VYCSS conditional probability of a high- ' release to .S5 (consistent with the NRC est! mate) would raise the conditional probability of a high or rrWr4m rs!ense (given a core melt) I L from the current CYVSS estimate of 72 to'about 202. This is a rather I
, snodest increase for such a drastic change in the conditional probability of
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X % a high release from the station blackout sequence. The.55 conditional j probability is considered an upper limit for CYVSS because of the features ' mentioned.which should inhibit the drywell liner failure mode. On the other hand, it does not seem possible with present knowledge to rule out . , such a conditional probability value. Thus, while the CYVSS' assessment in - this regard is considered reasonable, the uncertainties are quite large. ,
..' +
For the ATWS sequences, the CYVSS assessment is between the IDCOR and NRC assessments. Given an ATWS s&quence, .the conditional probability of.- , a high or medium release ls .22, .51, and .31,for the IDCOR, NRC, and.CYVSS . '
- assessments respectively. This is considered reasonably good agreement. - ,
To assess the sensitivity of these values ano to provide some insight un the potential range of uncertainty, the CVYSS r,esu,lts were recomputed ' using the NRC results. The;NRC fractions were applied to both types of
- ATWS sequences (accident classes IC and IV in.the VYCSS). The results showed that the conditional release probability for an EH or EM release would be raised to about 105 compared with the 7% for the existing VTCSS .
' estimate. This change is considered insignificant in ylew of the overall ,
4 uncertainties.
;Thus, it is concluded from this comparison that the Vermont Yankee conditional containment failure probability for a moderate or high early release could be as high as about 202, but this is considered an upper 11mlL Further, no def f etences were found in the CYVSS which wer,e judged to lead to a significant potential for changing the best estimate results. .
However, there were some problems encountered in assessing the use of drywell sprays to mitigate drywell containment failures. These are discussed and evaluated in Sect. 7. , c - l _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ - _ _ _ _ - _ _ - - _ - - - - . . _ . . . _ . - - _ _ _ . _ . . . . . . ~ , . . . . _ . .
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o - .. 6.i ASSESMENT OF ISSUES RELATED TO CONTAINMENT FAILURE MODES AND, .
- EFFECTS- in this Section, the important containment fallure modes considered in the VYCSS are evaluated, it should be noted that the time ~
available for this review was limited, and therefore the focus is on the key modes of containment failure (see Sect. 5 preceding for the' Identification of these important modes). . 6.A; ssecif fe Containment Failure Modes- The important containment
-! allure modes are those with potential for causing early failure when the '
fission products might be suspended in the containment. Because the c'omplete details of the particular Vermont Yankee reactor geometry were not availaole, use was made of data provided In the VYCSS Appendices, in , conjunction with the peach Bottom BWR4, Mark i design as a reference. Based on the discussion in the VYCSS, important containment failure modes are those which do not meet the,following criteria: The containment should remain intact without excessive leakage for at least 24 hours end the following limits must be satisfied to insure this i condition: 1) minimize spread of molten core material to the drywell liner, and 2) do nut exceed over-temperature or over-pressure limits of the containment drywell. These two limits should be considered in conjunction with both high , pressure ar:d low pressure core melt sequences. The containment f ailure modes with the potential to violate these criteria are considered separately, for high and low pressure sequences, in the following subsections. 6.A.a' Drywell Liner Melt-thro' ugh- The melt-through of the drywell steel , liner:Is a mechanism for f ailure for the Browns Ferry plant that was identified during the NRC sponsored containment loads Working group meetings held in 1984 In the problem that was considered as,part of this
'ef fort, the whole core of the reactor was assumed to be released from the reactor vessel., The geometry of the Browns Ferry plant is similar to ,
Vermont Yankee in that it has one doorway exiting from the pedestal . region (sub-pile room), and has floor sumps located within the sub-pile room. - The analysis of the thermal attack for this case indicated that the drywell liner melt-through would be assured under conditions of: a) large melt pool contacting the drywell wall,(6" depth), b) relatively high molten pool temperatures (2550K), and c) no significant heat losses from th'e molten pool boundary. The situation appears to be difforent than this for Vermont Yankee as alluded to on page 136 of the VYCSS due primarily to the smaller core providing only a l molten core pool depth for uniform
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spreading conditions out to the drywell wall .However, details of the quantitative basis for the yYCSS evaluation are not clear, and the range of . I uncertainty is not addressed. Three important aspects that affect the actual bahavior will be considered: 1) sp.ecific geometry of the Vermont Yankee floor,2) actual tempe ature of the melt at time of contact with
~
, the drywell liner, and 3) effec,t of pressure in the primary system at time of vessel mett-through. ,
. First, the specific geometry is important because thls will have a,large effect on the material: motion, if there are sumps within the sub-plie -
roo n, they they may be capable of. accommodating a significant melt mass , following molten core melt-through of the reactor vessel. For example, tri ' recent analysis by Sandla National Labs for the NRC, the Peach Bottom sumps were found to have a capacity of 235 cubic feet, which is capable of accommodating the equivalent mass of 20% of the core. Furthermore, the
' melting process is now.not considered to be instantaneous, but a more realistic continuous mel.tdown over tens of minutes. In this case the initial melt mass may be completely accommodated by the pedestal sumps, and as the molten melt continues to exit the reactor vessel, molten core-concrete interaction would provide a cavity minimizing the spread of the melt out of the sub-pile room. This scenario is considered a
- ' reasonable best-estimate scenario at this time for those accidents in .
which the primary system is at low pressure during core melting. As an
,l upper limit, if the effect of the sumps and the progressive melting behavior is ignored, then a molten pool can form, spreading out of the
' sub-plie room, in this case, for even distribution over the floor, the 1" j depth used in the VYCSS is valid, and this is considered insufficient to
" cause liner failure. However, the melt must exit through the sub-pile room doorway, and may preferentially pool up at the doorway exit'and attack the drywell liner adjacent to the doorway. Therefore, a reasonable '
j _ modification.to prevent this scen'ario could be to enhance the uniform spreading of the melt by, for example, small concrete dams on the drywell ., - floor. Given uniform spreading of the melt, drywelliiner melt-through i seems quite imorobable. The second consideration is tiie melt temperature. At this' time, the initial temperature of the melt exiting the core is not well known. ' Therefore, it is prudent to consider relatively high initial melt temperatures (2800-3100K) and then consistently take into account the various ways the melt will lose energy as it proceeds toward the liner .
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wall. Some of the heat sinks would be: 1) heat loss to the water and ste'el in the reactor vessel lower plenum,2) heat loss to the concrete floor, and . .
- 3) heat loss by radiation and convection to the atmosphere and structures .
over the. pool. This final consideration is one of the boundary conditions that must be also considered for temperature conditions in the drywell.
. Given the 1~ depth of the pool from the VYCSS analysis, the conclusion that the melt will be relatively cool at drywell liner contact is considered reasonable. However, due to the uncertainties as noted previously, it is .
considered prudent to consider thicker localized pools to address the '
.' sensitivity of the analysts to these conditions. .
i
'6.Ab. Direct Heating- Finally, the offeet of pressure in the reactor vessel at the time of lower plenum f ailure is important. .When the pressure vessel:Is at low pressures, then the previous qualitative description and considerations apply. If the pressure vessel is at high oressure, the , melt will be ejected at high pressures and dispersed throughout the drywell. In this situation; the melt will contact a number of solid surfaces including the drywell liner wall as dispersed droplets. Therefore, as the VYCSS ,
states, the heat load on structures will be more uniform. However, the specific geometry of the drywell is again important and must be considered to evaluate this case. For example, the location of the pedestal doorway exit should be considered. As discussed in the preceding, there is only one doorway exit from the sub-pile room. This suggests that the melt ; ejected with the high pressure gasses will preferentially impact the drywell liner wall adjacent to the door, it is not clear that this design q j specific situation was considered in the VYCSS analysis. It would be useful to quantitatively exaniine this scenario to determine if it could be a consideration for.a reasonable range of core melt masses and temperatures,'
- l. .
For this particular high pressure case, overpressure is probably not a concern because the pressure rise will be accommodated by the suppression pool. The rise time f or direc,t heating pressurtration is slower than the time it takes to clear the vents in the suppression pool. Therefore, the direct heating energy is transformed into suppression' pool heating (and possibly vaporization) which means that overpressure f ailure . of the drywell is unlikely frous this phenomenon. , I 6.Ac. Over-temperature /Over-pressure Failures- Failures from exnssive pressure and/or temperature may also occur. These failure considerations l Seem to be reasonably considered in the VYCSS. However, the effect of l I l 6-3
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^ -
specifically discussed. With a uniform pool depth, the over-temperature failure potential may be important and is influenced by the effeet of the heat loss from the surface of the pool to the drywell structures above the pool. For example, it is important to consider the presence of miscellaneous structures in the drywell near the pool surf ace. It is not clear to what extent these were j included in,the WCSS analysis, aryd the effect of related uncertainties on , the heat up of the drywell, structures. Also, in the structural analysis , j there was no specif fc mention made of the locations for the critical
. drywell penetrations (except for the vents to the suppression. pool) that .
' could be af facted by the liner t'emperatur6 rise. If these penetrations are low in the drywell, near where the molten pool could be, then the ef fect of heating could be important to drywell Integrity. It would be useful to ,
provide sensitivity calculations to correlate penetration temperatures with heat flux from the molten pool. F in the case of the molten pool concentrated within the sub-pile room, the over-pressure condition would be of concern particularly if the metallic concentration in the pool is high (for example,from molten steel structures .in the reactor vessel). In this case, the molten-core concrete interaction would hold the pool temperatures at initially high levels gausing more noncondensible gas production and fission product release from the pool. In the WCSS, this possible configuration did not seem to be evaluated, and would appear to be a useful additonal consideration. It should b,e noted that the 1;arge containment volume relative to' core volume of the Vermont Yankee plant will tend to delay the over-pressure failure,
- l. .
6 Conclusions- The important containment failure modes in the WCSS were reviewed. The review Indicates that for the cases considered in the study, reasonable conclusions have been reached. However, ther are a number of verlations of the scenarlos that were not considered that could ' have an impact on the three modes of earlp containment failure. Consideration of these scenarios would provide additional support that the WCSS containment failure pr,0babilities are reasonable, and would provide some perspective on the effect of alternative possible scenarlos. 6-4 '
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- 7. .EdNTn!NMENT INTEGRITY II1PROVEMENT5- This section evaluates Imo'rovements to the Vermont Yankee plant which have the potential for decreasing.the probability of containment failure in the event of a severe '
. accident. These improvements fall into three categories Wh1ch are .
discussed in separate subsections which follow. The three categories are:
- 1) Improvements suggested by the NRC,.2) Improvements identifled in the VYCSS, and 3) Improvements identified during this review of the VYCSS. ,
7A. Improvements to containment Inteority Succested by the NRC - Section 5 of the VYCSS discusses recent improvements which have been suggested by the NRC for BWR plants with Mark I containments. These ' improvements were recently (Sept 23,1986) discussed in detall during a meeting between the NRC staff and the Advisory Committee on Reactor ~ Safeguards. A review of the transcripts from that meeting (18) indicates that the VYCSS properly-describes and characterizes all of the NRC . suggestions. The VYCSS evaluates each of the five NRC suggestions and , derives an assessment of their applicability in the context of the Vermont Yankee plant design and severe accident evaluation. The following tdble lists the five NRC suggestions and summarizes the VYCSS response (as described in Sects. 5 and 7) to each: ; NRC SUGGESTION . VYCSS RESPbNSE 1.ft/drogen combustion control No additionel consideration require'd
- 2. Control of molten debris in drywell No a111tfonal consideration required
- 3. Improve reliability of containment sprays Further study recommended 4.Containmentventing Further stud / recommended
- 5. Augment training / procedures for severs accidents further studyrecommended ,
The remainder of this sub-section evaluates the VYCSS response to the NRC suggestions and provides a judgernent on the appropriateness of the response in the context of prov.lding additional measures to ensure containment integrity in the event of severe accidents. Each of.the suggestions Is considered separately. l - 7Aa. Hydrogen Combustion Control- The NRC suggestion for this issue consists of two parts; min!mize time that containment is not inerted , during operation, and minimize the potential for cxygen Ingress during the course of severe accidents. With respect to the former, the VY. CSS Indicates that the Vermont Yankee plant has historically been deinerted only 1.1% of the time with the plant at power, and this is considered . acceptable (pg. I15) by the VYCSS although no basis is provided. In order - to evaluate this conclusion, a sensitivity study was done In'an attempt to 7-1
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.o-determine the potential influence on containmerit failure proballity by having the, containment de-inerled 15 of the time, if it is conservatively '
assumed that a de-Inerled containment Will always cause containment f ailure with either a high or moderate radioactive release, then, using the
- results of Table 4.7 (Pg.108) of the VYCSS It can be seen that this ' .
contribution would rain the early high or moderate relcacco from 7R to-8% of the CMP. This is obtained by multiplying the total CMP by.01 and L adding the result to the existing VYCSS estimate of EH and EM release _ probabilities. This conservative sensitivity study shows that the effect of having a de-inerted containment 1% of the time is negilgible, and a substantial increase would be required to.have a significant ef feet. With respect to oxygen ingress during severe accidents, the VYCSS concludes that due to plant modifications in using nitrogen for those systems that could ir fict air in the drywell, coupled with existing limits for the use of drywell sprays (to prevent depressurization and air ingress through vacuum breakers), oxygen Ingress is not oncern. The VYCSS response to the hydrogen control issue is considered adecuate and valid with one exception: The VYCSS does not provide an adequate description and analysis of when and now orywell spray actuation will be inhibited, and how such inhibition will assure that air ingress will not occur. The references to drywell spray inhibition ( for example, see Appendix A, Answer '9) in the VYCSS Indicate that the drywell spray is inhibited on the basis of excessive drywel' pressure and temperature to avoid the potential containment collapse from negative pressure. The ,
, inhibition is apparently not based on any consideration of air ingress due
,j to opening of vacuum breakers. The issue of drywell spray actuation i durtrg severe accidents at Vermont Yankee is considered further in Sect. 8.
7AD. Control of P10lten Debris in brywell- The NRC concern regarding molten debris control in,the drywell relates to the possibility that the molten debris may flow to the intersection of the drywell floor and steel wall liner. If this occurs, the liner may f all, and a radioactive release may occur which bypasses the suppression pool. The VYCSS concludes that this issue is not pertinent to the Vermont Yankee plant because the small core size will likely prevent the o:currance of molten debris-liner interaction, and the use of drywell sprays will inhibit migration of molten debris to the liner. It is further argued (Pg.136) that barriers to molten debris inigi etica would be counter productivo cince they would inhibit crywell spray flow to the sub-pile room. 7-2
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. P.1 spreading conditions out to the drywell wall. However, details of the quantitative basis for the yYCSS evaluation are not clear, and the range of .
uncertainty is not addressed. Three important aspects that af fect the 4 actual behavior will be considered: 1) specific geometry of the Vermont Yankee floor,2) actual temperature of the melt at time of contact with - the drywell liner, and 3) effect of pressure in the primary system at time of vesrel me!t ,th ough- . .
.. First, the specific geometry is important because this will have a large effect on the material motion. If there are sumps within the sub-plie room, they they may be capable of accommodating a significant melt mass .
following molten core melt-through of the reactor vessel. For example, In ' recent analysis by Sandla National Labs for the NRC, the peach Bot' tom o sumps were found to have a capacity of 235 cubic feet, which is capable of i accommodating the equivalent mass of 20% of the core. Furthermore, the melting process is now_not considered to be instantaneous, but a more , realistic continuous meltdown over tens of minutes. In this case the initial melt mass may be completely accommodated by the pedestal sumps, and as the molten melt continues to exit the reactor vessel, molten core-concrete interaction would provide a cavity minimizing the spread of the melt out of the sub-pile room. This scenarlo is considered a
; reasonable best-estimate scenario at this time for those accidents in r which the primary system is at low pressure during core melting. As an upper limit, if the erfect of the sumps and the progressive melting
, behavior is ignored, then a molten pool can form, spreading out of the
- sub-plie room. in this case, for even distribution over the floor, the 1"
; depth used in the VYCSS is valid, and this is considered Insuf ficient to
; cause liner failure. However, the melt must exit through the sub-pile room doorway, and may preferentially pool up at the doorway exit and .
attack the drywell liner adjacent to the doorway. Therefore, a reasonable ' I
, modification.to prevent this scenario could be to enhance the unif,orm
' 'spresiding of the melt by, for example, small concrete dams on the drywell , q floor. Given uniform spreading of the melt, drywell liner melt-through J
- seems quite improbable. - !
\ .
' The second consideration is the melt temperature. At this' time, the initial temperature of the melt exiting the core is not well known. l Therefore, it is prudent to consider relatively high initial melt temperatures (2800-3100K) and then consistently take into account the various ways the melt will lose energy as it proceeds toward the liner
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NOV G3 '86 14:59 VTDf6 8028282342' " P.2 "The.fssue'of moltNheD315-drywell liner l'nteraction is a controversial asp'ect of severe accident behavior for Mark I plants. As such, the issue is considered separately.In Sect 6. While the VYCSS aryument' s on this issue appear pertinent, two considerations appear incompletely assessed. First, as pointed out in the previous sub-section, the use of drywell sprays under severe accident conditions'is inhibited by high drywell pressure and - temperature. As a result, it is not clear from information in the VYCSS to what extent these sprays can be rolled upon to arrest the progression of molten debris across the drywell floor. Second, It does not appear that the VYCSS has fully explored the alternate placement of physical barrlers (for example, near the drywell liner wall) which would not inhibit drywell soray flow to the sub ph room, or.the use of small local barriers to assure. uniform distribut sn of the molten debris (See Sect. 6). . 7fc. Improve Re,11 ability of Drywell Sprays- The VYCSS concurs with the NRC suggestion that this aspect needs further study, and several afeas are
- Identified, starting on Pace 116, to improve drywell spray reliability.
These areas are considered appropriate and valid. The problem of ' inhibiting the spray actuation when high drywell pressures and temperatures exist is not addressed, howevar (see Sect. 8.for addlttonal disucision) Further, the possibility of low NPSH for the ECCS pumps identified on page 1171s not considered further in the discussion on reliability. It does not appear, however, that this is a major concern for Vermont Yankee since alternate scurces of water (other than the suppression pool) are available. . 7Ad. Containment Venting- The VYCSS Identif tes this suggestion as an issue requiring further study. Various aspects of the benefits and - potential detrimental effects are discussed throughout the VYCSS,land these considerations appear valfd. The discussion regarding furthe.r study of the lssue (starting on Pg.121,) also appears consistent with the.NRC i postion, and encompasses those areas which appear to be important with one exception. The only deficiency noted in the VYCSS with respec,t to this issue'is the lack of any consideration regarding the addition of a 1
. completely separate venting system; the evaluation considers only modification to existing plant equipment.
7Ae. Augment Trafritng/Pr'ocedures for Severe Accidents- The NRC , objective for this suggestion is to assure that the plant operators 'make
- the best use of plant systems for the prevention and mitigation of severe accidents. The VYCSS response to this suggestion (starting on pg.139)-
appears appropriate and complete. The on'y def telency found was th,e lack of any discussion regarding the schedule for the implementation of the changes identified for future consideration. l 7-3 O *
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t . 7B containment Intearity Imorevements identified in the VYCSS . . . . Scattered throughout the WCSS are suggestions and recommendations for additional analysis and study regarding potential plant improvements for - reducing the risk 'from severe accidents. in Sect. 6.2 on page 178, the WCSS explicitly discusses " continuing efforts regarding severe accident analysis". These efforts are grouped into three categories. The first category covers "procedurg changes (which) are recommended for final evaluation and implementation" These changes consist of six Items. The . first five items al.1 pertain to improved recovery from statlpn blackout scenarios, and the sixth pertains to improved procedur'es to respond to the
. ATWS accident. As noted previously in Sect. 5 of this report, the station -
blackout and ATWS sequences are dominant contributors to containment , failure leading to a significant release. Thus, the WCSS recommendations address those sequences of significance, and they are considered appropriate .
' The second category of WCSS improvements are described as
" recommended for further study augmented by detailed analysis to ensure that the positive and negative Impacts of potential changes are well -
understood." Three items are recommended, all of which are considered to be appropriate and potentially important. The three ,ltems are upgrading of the Emergency Operating Procedures, enhancement of conte.Inment spray / reactor injection capability, and enhanced capability for venting of the containment. As the WCSS points out, all of these items have positive and negative aspects which need to be carefully evaluated before any changes are implemented. . : The third category is described as " areas which merit further study", and consists of four items. These include evluation of RHR pump response to ' high suppression pool temperature, evaluation of nitrogen supply to safety / relief vaives, reliability evaluation of the service water system, and reliability evaluation of the standby liquid control system. All of ,
- these items appeair to be appropriate items for consideration. ,
While all of the above items appear appropriate and recognize the L Important severe accident issues, it is not clear which items Vermont Yankee Nuclear power Corporation has committed to accomplish and on
.what time schedule. The WCSS Impiles that all of the items will be addressed, but they are described as " recommendations", not commitments. ,
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'7C. Imorovemenes identif fed Dr.,ina the Revtew of the VYCSS- In
- reviewing the VYCSS, one of the objectives was to identify additional
- Improvements not considered by the study or suggested by the NRC which appeared to have the potential for improving the probability of maintaining containment integrity in the event of a severe accident. Two such improvements were identified, as follows: i
- 1. Upgrade' the capacity an.d reliability of the drywell coolers- At present, according to the VYCSS, the drywell coolers will be Isolated during severe accidents. It also appears that their' l
. capacity is insuffielent to remove decay heat until tens of hours -
af ter core shutdown. However, if the capacity of these Units could be upgraded, and their reliability assured during severe accidents, it appears they could be effective in reducing the liktthood of
' drywell failure from overtemperature or overpressure. They also could rer'n ove some of the heat loading on the suppression pool for transients, and extend the time of, or prevent, overpressure containment failure. ,
- 2. Reduce the frequency of main steam isolation valve (M$1V) closures for transients and/or make provisions to reopen these valves followfng transients. This modification could improve the Ilkellhood that the main condensor would be available as a heet s!nk !
for some accident sequences. The VYCSS does acknowledge (Pg. 151) the possibility of re-opening the MSIVs for the ATWS event. . However', there is no discussion regarding the likely success of such a: strategy, or what operator actions would be required. ! Furthe'r, there appears to be no consideration for changing the - MSIV closure logic, or re-opening the v81ves for other, more likely,. transient events. It should be emphasized that neither the cost, feasibility, nor quantitative . risk reduction potential was evaluated in formulating the ebove suggestions for pptential improvements. Further, it is not meant to imply by making the suggestions that any plant Improvements are necessary to decrease severe accident risks. . 7-5 1
+ .sx , _ ~ w - a s. . _= L.: , ._ _ v.a ; _ , . . . m,, , :PCV 03 ?86'15:01 VTDPS 8028282342 P.5 8.~ Miscellaneous Comments and Discussion of issues . .
This section presents a listing of comments which were developed in ' conjunction with the review of the VYCSS report which are not directly related to items discussed in previous sections. Some of these comments were considered significant and an expanded discussion is provided. Those coristdered significant in terms of impacting the .important results are - identified and an evaluation of their impact is addressed where , appropriate. Editorial and other minor comments not releyant to the . interpretation er validity;of the results are not included.' The comments and a discussion, where ' appropriate, follows:
.1) Characterization or resulth The conclusions given in the VYCSS (page 178) state that "...the best estimate single-point conditional containment failure probability is 75". Also, the cover letter transmitting the CYVSS ,
(Weigand to Denton, Sept. 2,1986) indicates (third paragraph) that
" Vermont Yankee has concluded that the best estimate of probability of -
containment failure is once every 500,000 years, which corresponds to a 7 percent containment failure probabil'ity in the unlikely event of a serious eccident resulting in core melt." These statements appears to be improper character 12ations of the VYCSS results for two reasons. First, the probability of containment f ailure given a core melt is not 72, but actually 325 based on the results presented on Pg.108 of the report. The 72 value is the probability of containment failure resulting in en early radioactive release of high or medium magnitude. Second, the one in 500,000; (2E-6/yr.) probability quoted in the cover letter appears to be based directly on combining the VYCSS core melt probability result (3.0E-5) L times the containment failure. probability with significant release, or 75 .
. (.07). However, on page 58 of the VYCSS it is stated that "As noted
- earlier (Section 4.1), this approach (referring to the approach used in the l
LVYCSS) is not capable of supporting a " bottom line" value for the core melt frequency at Vennunt Yankee" While the meaning and intent of this
,, statement is not entirely clear (there is nd corresponding admonition "noted" in Section 4.1, contrary to the statement contention) it seems to imply that overall core melt probability numbers calculated in the study are insufficiently accurate to use in a bottom line sense. While this statement on page 58 may be too severe a characterization of the overtil ,
core melt probability results, it seems.to undermine the unqualified statement of results in the transmittal letter. This apparent discrepancy j is closely reisted to the uncertainty discussion which follows. l 8-1
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- 2. Uncertainties- The report is. considered def tetent in not providing a .
reasonable assessment of uncertainties. The assessment of uncert'ainties is said to be beyond the available resources. However, this feature of PRA , analyses has been found to be so important and controversial that omitting any discussion detracts from the usefulness and credibility of the assessment. Even a judgemental estimate of uncertainties with appropriate discussion of the basis (as opposed to.a rigorous statistical . based assessment) would have been an important addition. Instead, the report does not even include much qualitative insight on the' significance of uncertainties which would add some perspective ~to the results; To. , compound the problem, there appear to be several inconsistent and confusing references to the results in the context:of their perceived-accuracy. For example, on page 37 it is stated that 'Because this parameter (conditional failure probability of the containment given a cevere accident as calculated in the VYCSS) was riot the intended purpose of earlier analyses, the containment conditional failure probability that can be inferred from such analysis is a conservative unoer bound estimate (emphasis added) adequate only for use in the integrated public risk estimates." The VYCSS assumptions, analysts, and data do not appear to support the contention that the results are a conservative upper bound e::ttmate, nor is it clear how these results are judged to be adequate only for use in the (which7) integrated public risk estimates. Further, on page 39 it is conceded that "Because of the limited time available for this Vermont Yankee spectftc analysis, this uncertainty was not explicitly quantif ted; rather a 'best estimate' value was developed. However, despite the. attempt at a best estimate calculation, there:may still be excessive conservatism in the evaluatton. ! i Another characterization of the results in this co'ntext appears on page 48, to wit;"In the text which follows the term "best ' estimate
- ts used to characterize the processes and results (for the containment fat!ure probablitty evaluation). This term is intendedio Indicate that the evaluation was perfornied in an objective manner that attempted to balance uncertainties... The analysis prcvides a formal means for investigating possible challenges leading to core melt or containment failure and hence characterizing in approximate terms the probabillt'y of containment f ailure." Again, on page 81, it is stated that "The process has 8-2 l
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led to a number of procedural insights which could be enhanced to preserve . the best-estimate nature of the numerical assessment." 'Also on page 81, . .
.there appears to be a range of uncertainty provided in some of the results,
. ,but there is no basis provided.' For example, "The best-estimate of successful containment performance for such a definition (early or intermediate failure times with high or moderate release) is'in the range )
of 90 to 985..0ther ranges for various measures of containment i performance appear on page 81, and'they have ranges of 102, encompassing f the best estimate value. Further, on page 178 (6.0, CONCLUSION) It is . stated that "For'all severe accident types involving core damage and having the potential of causing a significant redlonuclide release, the conditional containment failure probability is 2-102." These results do . not seem to support the earlier cont'ention (" conservative upper bound , estimate"), and no discussion is provided indicating how such ranges were derived. :
; j in the comparisons provided in Sect. 5 of this report, some discussion of -
the ranges of results found in pRAs for BWR4s was given. On the basis of - that evaluation, and utilizing other judgements, it is estimated, roughly, that the VYCSS core melt probability has an uncertainty of about a factor of 5, and the containment conditional failure probability for an early release of high or moderate magnitude has a range of perhaps 45.to an upper bound of 20% (vs. the VYCSS "best estimate" of 75). -
- 1. External Events- The VYCSS excludes consideration of external events (Dg. 54). While this limitation do'es'not invalidate the results for the events considered, it does open the question of the potential Influence of external event considerations on the results. Further, the results stated in the report (particularly in Sect!o.n 6.0, CONCLUSIONS), and In the cover letter, do not express this limitation, nor is there any qualitative judgement offered regarding the potential offeet of this limitation, in '
order to provide some perspective on the possible significance of external events for the Vermont Yankee plant risks a literature search weis
' undertaken to determine if risk assessments for piants in the region of the Vermont Yankee site (i.e. Northeast U.S.) found external events to be significant. It should be emphasized that such a survey.cannot conclude that external events are or are not important for Vermont Yankee because
- 1) As po!nted out in Reference 10, external event risks are very plant and 8-3
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site speelfic, and therefore external event risks from one site cannot be ~ extrapolated, in general, to another,2) external event risk estimates are exceedingly uncertain and controversial, and not much confidence can be . placed in the resultsN, and 3) no pRA has been published for a BWR4 ~
. with a Mark I containment in the Northeast U.S. Notwithstanding the,se limitations, it is useful to assess the overall significance of external event risks from other pRA studies .in order to obtain some perspective _ .
on trends and potential significance as applied to the Vermont Yankee Cast. , - LThe results of six published pRAs for plants in the Northeast U.S. which have included an assessment of external event risks were examined in this ; .
.Garvey. These include Millstone 3W , LimerickW', Shoreham(8h
$2abrookM, and Indian point Units 2&3M. The Millstone and Indian point plants are pWRs (with reactors bull _t by Westinghouse), while the Limerick and Shoreham plants are .BWRs, Type 4, with Mark 11 containments. In .
Summary, in all of these pRAs, external events were found to be . significant (in most cases dominant) in terms of contribution to the~ probability of serious release in most cases the dominant external event contributor in this regard was seismic events. In all cases, however, the . overall risks computed were very low. For example, in the Shoreham case, the estimated frequency of a significant release (capable of causing one or more early fatalaties) was 2.5E-7/yr., and was dominated by seismic , events. The only valid conclusion from this survey is that external events have been found to be important contr'lbutors to nuclear plant risks In the regton of the U.S. In which Vermont Yankee is located. This implies that such risk contributions could be important for Vermont Yankee. l 4 Use of Brvwell Sorsva- The VYCSS assessment of the potential use and ef festiveness.of drywell sprays is somewhat confusing and apparently incomplete. For example, there does no.t appear to be an. adequate. consideration of the potential for a steam overpressure spike when the spray water may contact the pool of molten debris late in some accident sequences.. If this contact does not occur, as might be the case earlier in the sequence, it may be possible for the sprays to condense and cool the dyrywell atmosphere to an extent that the reduction of pressure could open vacuum breakers and allow oxygen to flow into the drywell, raising 8-4 .
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ToveD8615!05VTDPS8028282N2 < P.9 the potential for hydrogen combustons. While this possibility is .
- recognized, it does not seem to be evaluated for some of the accident sequences. On Page 115, it is stated that drywell spray actuation limits ,
- would actually preclude operator actuation in cases where excessive I negative drywell pressures could result. However, it is not clear for-
' which. accidents, and at what times this inhibit condition would occor, nor is it clear how this factor has been considered in evaluating the '
containment response for various sequences. 5 Release Maenttude Perspective -The VY SS divides release magnitudes Into 6 categories (pg.108), and these magnitudes are briefly described in
- terms of. time Intervals and fraction of " equivalent" . lodine release.
.However, there in no discussion provided concerning why those particular ,.
time periods are signific' ant, or why the particular lodine fractions were chosen to characterize different release, it would be particularly helpful If the report could discuss the significance of these releases and release - times in terms of public health consequences. Without this:Information, it is not possible to evaluate the detailed significance of the r,esults in terms of release parameters vs. probability provideo'in table 4.7. Many other radionuclides (i.e., Cs, Te, Ba, Sr, ) can be important contributors to public risk and their release potential is not explicitly identified in describthg the releases. 6 Benefits of Smaller Core powerlevel than Reference pl$nt- On page'48 and 49 several benefits are listed which are said to derive from the fact ' that the Vermont Yankee plant has a lower core power level than the
- WASH-1400 plant (peach Bottom) while the containment volume is .
essentially identical. Indeed, the Vermont Yankee power level of 1593MWt ' Is only about half that of Peach Bottom (3293MWt), and the containment . drywell volumes are comparable (134,000 cu. f t. for Vermont Yankee vs. . 159,000 for Peach Bottom according to Appendix A of the VYCSS). These
. design parameters mean that the core power to containment volume ratto -
.Is, considerably less for Vermont Yankee than for Peach Bot)om, and this dif forence is said (Pg. 68) to result in much delaytj! containment f allures by overtemperature or overpressure compared to Peach Bottom. However, the primary energy absorbing. media for all severe accidents is the water in the suppressloa pool. The ability of this wattr volume to absorb decay heat (or core power for ATWS) will be the primary determining factdr in temperature and pressure rise rate. From Appendix A of the VYCSS lt can 8-5 4
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~
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- power to suppression pool volume ratto. Thus, while some benefits would . j definitely accrue from the relatively low Vermont Yankee core power level, they may not be as significant as implied in the VYCSS. l 1
- 7. Basis for Frecuency and Probability Values - In several Instances i throughout the report, it was~ judged that an inadequate basis was provided l for the frequency and probability values. The more signifIcant of these 1
instances were as follows: , . ! l i PAGE NO. . DESCRIPTION 62 Unavailability for Vernon Hydro given station blackout 'G8 Calculated times for containment failure 74 Conditional failure probabilities of containment failure j 75 Unavailability of water injection i
\
i Th'e basis for the values used in each of these four instances were requested from the utfif ty during the review. The basis was promptly supplied in all Instances, and they appear reasonable and adequate. The utility response to this request is provided in Appendix A to this report. I
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- 9. REFERENCES- . ,
j 1
- 1. Reactor Safety Study, An Assessment of Accident Risks in U.S. !
Commercial Nuclear Power Pitnts, WASH-1400, U.S. Nuclear Regulatory . . , Commission, Oct.1975. i
- 2. Presentation by Brookhaveh National Laboratories to ACRS Subcommittee on Class 9 Accidents, Sept. 24,1986. ,
- 3. Peach Bottom Atomic Power' Station-Integrated Containment Analysis, -
IDCOR Technical Report T23. l P.8, Mar.1985.-
. 4. Probab!!!stic Risk Assessment-Limerick Generating Station, '
' Philadelphia Electric Co.,1984. . .
S. Probabilist'ic Risk Assessment-Shoreham Nuclear Power Station, Long Island Lighting Co., June 1983.
- 6. Interim Reilability Evaluation ProgFam;' Analysis of the Browns Ferry, Unit 1, Nuclear Plant, EG&G Idaho, Inc., NUREG/CR-2802, July,1982.
- 7. Additional Information Required for NRC Staff Generic Report on Bolling Water Reactors, NEDO-24708, Aug.1979.
8.10CFR50.62, NRC Rule on Anticipated Transtents Without Scram for Nuclear Power Reactors.
- 9. Prevention' and Mitigation of Severe Accidents in a BWR4 with a Mark l Containment;(DRAFT), Bro'o khav,en Nat1onal Laboratory, Aug.1986.
- 10. Probabilistic Risk Assessment (PRA) Reference Document, NUREG-1050l U.S. Nuclear Regulatory Commission, Sept,1984.
I 1. Millstone Unit 3 Probabillstic Safety ' Study, Northeast Utillties, Aug. 1963. - 12* Indian Po' int Probablitstic Safety 6tudy, Consolidated Edison Co. and New York Power Authority, Mar.1982., l
- 13. Seabrook Station Probabilistic Safety Assessment, Public Service of New Hanipshire, Dec.1983 -
9- l' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - . _ _ _ - _ :: : _ _ _ _L
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- 14. Severe Accident Risk Assessment, Limerick Generating Station, Philadelphia Electric Co., April 1983.
- 15. Statistical Abstract of .the United States-1986,106th edition, U.S.
Dipt. of Commerce, Bureau of Census, Sept.1986. .
- 16. Interim Reliability Evaluation Program: Analysis of the Arkansas Nuclear One-Unit i Nuclear Power Plant, NUREG/CR-2787, Sandla National 1. laboratories,' June 1982.~'- >
- 17. Interim Reliability Evaluation Program: Analysis of the M111 stone' '
Point Unit 1 Nuclear Power Plant, Science Applications, Inc., NUREG/CR-3085, Jan.1983.
- 18. Transcripts from the Advisory Committee on Reactor Safeguards meeting of the Subcommittee on Containment Requirements aN Reactor Safeguards, Sept. 23,1986.
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a APPENDlX A Responses from Vermont Yankee Nuclear Power Corporation t0 0uestions Developed During the Revtew of the Vermont Yankee Containment Safety Study O e l l 4 4
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. . 1 PREt1MINARY QUESTIONS ON THE VERMONT YANKEE CONTAINMENT C/JETT STUDY l
1' OUESTTON 1 . page 12 - The paragraph at the t.op of the page statas that failure of the NPCI l in the case'of a small LOCA would require operation of.the ADS to reduce vessel pressure. The first paragraph on Page 11 states t.nat the ACIC may 1,. considered as a backup to HPCI in the event of a very small LOCA. Please indicate for which break sizas the RCIC, acting alone, can be considered a viable means of core cooling. ANSVER 1 The Reactor Core Isciation Cooling (RCIC) system can deliver 400 Sailons par minute makeup water to t.he reactor. The Migh pressure Coolant Injection (HFC!) S,vstem can deliver 4,250 gallons par minute of coolant injection to the reactor. Althcush RCIC's design requirement is to provide het shutdown cora coolin5 for transients in which the main condenser is unavailable, its capacity of 4C; spr (or sprroxire.atoly one. tenth that of HPCI) c.an provide sakeup to maintain adequate reactor level for very small leaks in the esseter coolant pressure boundary. The actual leak size is a function of the decay heat level, the leak location, and the leak geometry. Typically, steam lina , leaks of less than two inches in 6iameter are within RCIC System capacity. 6 e
. . . _ _ . _ _ _ , . . . _ _ , . . , _ ,_ _ .,.2
- NOV 04 '86 16:53 VTDPS 8028282342 p,g
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4 CUESTIONJ Page 62 - plaase provide the basis f or the Va'rnen Hydro unavailability values listed on this'page. ANSWER 2 a 1 The best available infomation indicates that the plant was unavailable for a-total of 2 hours and 24 minutes in a 21-year period. The average
~
unavailability is, ther4 fore,1.3 x 10 . In response to e grid collapse, the hydroelectric station,must separate from the grid to allow the tieline to remain ava,ilable. The only active action identified was the automatic opening , of a single normally closed feed breaker (b). Therefore, H can be approximated as: H = 1.3 x 10 4- b (fails to open)***
= 6.6 x 10' l The ert,8rster for operator inappropriate action are as fo11ews:
. O _*
- Phase 1 0 - 2 hours .1
. Phasa II .2 - A hours .05 Phasa III 4 - Id hours .01 1 Phase IV 10-24 hobra .01 i
- It should be noted that only two events were recorded: one of 2 hours I and 20 teinutes and one nf 4 minutas. The station recovared quickly from the latter event, which was initiated by a lish'ning strike.
** (1) .BWR Individual Plant Evaluation Methodology.
(2) A. D. swain and H. E.,Cuttman, Handbook of Human Reliability . Analysis With tinphasis on Nuclear Power Plant Applications, NUREC/CR-1278.
*** b (fails) to open) = 6.5 x 10-4/ demand, seatrook PRA.
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bOV 04 '86 16:53 VTDPS 8028282342 P.6 Assuming (hat required' ection (opening) of the normal feed breaker to each amergency bus is addrassed in the electric power system model, than two . breakers have to closa,,to feed either bus. In addition. it is assumed that the diesel breaker must open (this is conservative since failure of this
- breaker to close may have been the cause of " diesel failure to supply emergency bus"). So that V con be approximated as:
V = 1.7 x 10' i The factor C reflects those loss of off-site power events that would also render the Vernon Hydroelectric station Unavailable. A review of 114 off-site power events identified 24 that were caused by extreme externsi phenomena (e.g. lightning, ice storms, heavy snow, tornadoes, etc.). Events such as saltwater spray and Florida grid instabilities wara assessed not be be applicable to the Vermont Yankee sita. More detailed analysis of tha particular arrangement at Vernon and Vermont Yankee is needed to assess C. However, 11 it is assumeo that not all at the evanca ars applica* sit to Vermen'.
~
Yankee, then C could range f rom 4 x 10 to .1. Tha upper estimste (0.1) is used for the point estimate cuantification in this analysis. In summary the unavailability of Vernon Hydre as an affective AC power source to the emergency buses given a statien black,out is U=H+0+V+C Uu 6.6E-4 + 0 + 1.7E-3 + 1E-1 Vernon Hydro Vernon Hydro Unavailability Unsvallability For Events othar For Extreme External Than Extreme Externet Phanomens_ Phenomena Events _
.2 .1 Phase I O - 2 hours
. 05
- Phase II 2A hours 15
.11 .01 Phase III 4 - 10 hours .
.01 Phase IV 10-24 hours .11 The Vermen Hydro una' availability for extreme external phenomena events has been incorporated into th's quantification of Vermont Yankee station bisckout sequences in 3.3.
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00tSTION_3' _ Page 63 - Provide a roterence for tho' " previous evalu=Lluism" sefesced to in the last paragraph. - - A125WER 3 Re arencest y
- 1. SLI-8218. " Inadequate Core Cooling Detection in Bellins Water Reactors,"
prepared for BWR owner's Group, s. Levy, Inc., November 19R2.
- 2. SLI-8221. " Review of Shoreham Water Level Measurement Systems." 8. Levy.
Inc., November 1982,
- 3. Shoreham PRA Reper't:
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pg I 00EST70N_4 ., < i Page 68 - The first paragraph refers to " deterministic calculations" to . i supportlatecontainnaNtfailuras. 'Please supply or referance there
' calculations. l AMSWER 4 .
The severe accident phanomenology was analyzed for several accidant sequencas using both MARCH /RMA and MAAP 3.0 codes, performed by Risk Management Associates and rauske and,Associstes Inc., respectively. For the station blackout nacident sequences, the result.s of both MARCH /RMA . ,J MAAP 3.0 analyses indicated that the centsinment failure by overpressure would not occur within 24 hours after the initiation of an accident.
Reference:
YAEC-1564, Appendices C and D. t-g e , sd 9 6 0 4 O O e e
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e 00ESTIow 5 , Page 74 - Please elabortta on the bas'L2.for the conditional containment failure probabilities p.rovided at the bottom of the paga. . i A.nsutE S The bases for the conditional probabliity of early containment failure (CI)- for asch class of accident are sa follows: i { Accident Class CI 54114, IA 10-3 a ., IB 10-3 a IC .31 b ID 10-3 a II *' 1.0 c AAA 104 a IV 1.0 a h' Notss
'ine appropriate failur niode, was judged to be hydroger burn The a.
probability is based on t.be Shoraham PRA study.
- b. The failure probability was estimated as the fraction of eccident ,
sequances in class Ic that would result in elevated containment pressure (2.40 psia) befora reactor vessel failur's.*
- c. For these sequences, containment is assumed to have f atiac prict to cors l
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- s QUESTION & .
Page 75 - Pisase provide the basis for the ustimates of unavailability for water injection given "on this page, and Indicate why Classes IE. IV and V art not included. AN5CER 6. , The bases for the estimates of unavailability of affectiva active mitigation. . water injection (T) for each class of accident, are as follows: C,1ven to early cents tnsnent f ailure) 3 asis Accident Clasr a IA 0.01 b IB 0.14 a IC 0.01 c ! ID O .'1 a IE 1.0 e 11 1.0 d 0.0$ III e IV 1.0 a
*V . 1.0 E911.F.
- a. Credit for drywell spray, Low Pressura System, Teodwater System, Cors spray system, and Centrol Rod Drive Systam.
- b. Failure probability.of off-site /on-site power recovery and unavailability of diesel fire pump, weightef by the core nelt frequency o.? each phast in l
Class IB.
- c. Credit for Control Rod Drive Systam.
d ., Credit for Control Rod Drivs Systee and Taedwater system,
- e. So credit takant failure probability of 1.0 was assigned.
". Credit for Feedwater System.
- g. No credit taken. .
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J '* , oursTTow 1 Page 77 - Please provide the basis for the 50% probability for leakass ! failures given in the last paragesph. - ANSWER 7 > 3sded en the PRA studies of Shoreham and Limerick, s 50% probability is assigned to the leakage failure given a slow overpressure c.ha11snse to the '\'Q
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, Pege Please provide a ref erence for the "recent investigations" refstred to in the firs,t' full p'ersgreph. '
MEE8 8
Reference:
NUREG/CR-4550, " Plant Accident Sequence Likelihood ,, Characterization - Peach Botto:n Unit," Draft, Volume III. Hay 1986. 46*
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/ OUEETICV 9 .
l , Pas 115 - Item 4 indicates that, proc'edures include limits to prevent initiation of drywell sprays to avoid excess negative drywell pressures. .
,Elsewhers in the report, t,he use of drywell sprays is shown to, te an effective s vere occident mitigation measure, please elaborate on when procedures would cllow use of drywell spre.ys during severe accident progressions.
,ANSWE'4 9 Dryvell sprays ar2 initiated whenever
- 1. Drywell temperature cannot be meintained below the drywell temperature -
design ilmit, or
- 2. Torus airspace pressure exceeds t.he suppression cha.ther spray Initiation pressure (SCS1p), or
?
- 3. Containment pressure cannot be maintained below the containment pressure design lir.it.
These actions ars predicated on the torus airspace temperature and pressure being below the drywel* spray initiation pressure limit curve. This curve assures that the containment will not collapse or otherwise fail due to negative pressure resulting from the spray initistion and the subsequent evaporative and convective cooling of the containment. atmcsphere. , 6 e e i .
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- CLUESTION 10 ,
Pese los - Table 4.5 indicates that " interfacing LOCA" (Class V) avants are estimated to have a "n'agligible" frequency. There does not seem to' be any - , justification provided in the report for this conclusion. Ple'ase Provida .
. cdditional inferraation.
6FSL*tR 10 , R0ference: " Peach Nottom Individus1 Plant Evaluation." The frequency of interfacing LOCA (Class V) was estimated as 10 / year in FBIPE. This frequancy for Class V in assume,4 to be applicable to Vermont Yankee, and it is negligibis for the purposa of evaluating containment conditional failura probability. . 6 k* e e e 9
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CUESTION__11 Can:ral - Please provide Ret erent.e 11 ('fAEC-1364). AJf5WER 11 - YAEC-1564 was developed as a detailed description of the analyses used and raruits obtained in the Vermont Yankee Containment safety study. The methodologies selected and their specific application to the Vermont Yankes pitnt are considered by Vermont Yankoo to be new work in the field, and have potentially *lr.n4ficant commcrcisi value. Vermont Yankes would be plassed to Prrvide t.his material, if*neesssary, for review provided its confidentiality c n be assured through formal agrammant. .We are availabin, to discuss the , datatis of such an agreement at your convenience. 1 eie .
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l jo,, UNITED $TATES ' !q ') 3 g NUCLEAR REGULATORY COMMISSION
- 5. /nE WASHINGTON, D. C. 20555
....+ October 24, 1986 Mr. J. Gary Weigand, President -
and Chief Executive Officer l Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, VT 05301
Dear Mr. Weigand:
SUBJECT:
VERMONT YANKEE CONTAINMENT STUDY This is in response to your letter dated September 2,1986, enclosing the Vermont Yankee Containment Study for our review. We acknowledge the extensive effort in generating the report in a short interval of 60 days. We have reviewed the report in two parts. The first part relates to the review of your comparison of the Vermont Yankee design features to those of the reference plant in WASH-1400 and calculation of a Vermont Yankee specific containment conditional failure probability (CCFP). The second part d? sis with our review of your response to five generic NRC staff concerns related to Mark I containments for Boiling Water Reactors (BWRs). Your approach was to quantify the Vemont Yankee CCFP using the Peach Bottom analysis as a surrogate and modifying the accident sequence frequencies to { reflect the specific design features and 14 year operating data base of Vennont Yankee, supplemented by industry experience where Vermont Yankee data was not available. The range of plant specific conditions were determined by phenomenological analysis using Industry Degraded Core Rulemaking (IDCOR) group Modular Accident Analysis Program (MAAP), deterministic structural { capability calculations made for Browns Ferry and the Peach Bottom plants, and your staff's engineering judgement. Using the above approach, you have j provided a best estimate CCFP value for Vermont Yankee containment of 7% for all sequences where containment may fail within 24 hours' and where radionuclides releases include all noble gases and greater than 0.1% Iodines. You have, however, not provided an analysis of uncertainties in your methodology, in the phenomenological analysis using MAAP, and your engineering judgments. Based on the staff's experience with other BWR Probabilistic Risk Assessmant (PRAs) we believe that the CCFP of 7% may be fairly representative but is quite uncertain. It is our judgement that the CCFP based on your estimating techniques would have associated uncertainties as discussed in Enclosure 1. We also believe that your evaluation has provided the staff with sufficient insights to conclude that the CCFP for Vermont Yankee may be lower than the 90% estimated from results of WASH-1400. Our assessment of the uncertainties leads us to believe that the Vermont Yankee CCFP is probably less than 50%. That conclusion, that the CCFP is probably less than 50% and may be fairly estimated to be about 10%, brings out the very reason for the five generic { NRC staff concerns related to Mark I containments. We believe that greater
- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . \
. _ . . . .__ . . _ _ _ . . _ . . - - . . ~ ._ - .. _.-.. . . . - -
-2 October 24, 1986
~
assurance of containment perfomance in the face of core melt is desirable and echievable. Therefore, our attention now is shifted to the second part of your study. Our review of the second part of your report dealing with containment perform-ance enhancements has resulted in some preliminary conclusions and several questions. We find that the kind of' enhancements proposed in your study are consistent with the type of improvements considered by the staff. You have. , indicated that further analysis should be performed to test the feasibility l and effectiveness of the concepts C scussed in your study. We agree. I Enclosure 1 is our preliminary assessment of your study. Our evaluation proposes further studies regarding combustible gas control and core debris barriers. We request that you reevaluate these two issues in the light of the staff comments. Enclosure 2 lists detailed staff questions. We request your response to the staff questions and connents by November 17, 1986. Sincerely, 1 m v Robert M. Bernero, Director Division of BWR Licensing cc: See next page 9 .___ _ _ ____i1_.________________.______ _i--__-__--_-__ 1--" - - - - - - - - - - - - - - - - - - _ - - _ _ _ - - - - - - - - - l------'- - - - - - - - - - - - --
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- :. .- a...x . . - .u., --- .---v 1 ' , a Mr. R. W. Capstick Yemont Yankee Nuclear Power Vermont Yankee Nuclear Power Corporation Station CC* - '
Mr. J. G. Weigand Mr. W. P. Murphy, Vice President & President & Chief Executive Officer Manager of Operations Vermont Yankee Nuclear Power Corp. Vermont Yankee Nuclear Power Corp. R. D. 5. Box 169 R. D. 5, Box 169 Ferry Road Ferry Road Brattleboro, Yement 05301 Brattleboro, Vermont 05301 Mr. Donald Hunter, Vice President Mr. Gerald Tarrant, Commissioner Vermont Yankee Nuclear Power Corp. Vermont Department of Public Service 1671 Worcester Road 120 State Street Frar.inghan., Massachusetts 01701 Montpelier, Vermont 05C02 New England Coalition on Public Service Board Nuclear Pollution State of Vemont Hill and Dale Farm 120 State Street R. D. 2 Box 223 Montpelier, Vermont 05602 Putney, Yemont 05346 Vermont Yankee Decommissioning Mr. Walter Zaluzny Alliance Chairman, Board of Selectman Box 53 Post Office Box 116 Montpelier, Vermont 05602-0053 Vernon, Vermont 05345 Resident inspector Mr. J. P. Pelletier, Plant Manager U. S. Nuclear Regulatory Comission Yemont Yankee Nuclear Power Corp. Post Office Box 176 Post Office Box 157 Vernon, Vermont 05354 Vernon, Vermont 05354 Vermont Public Interest Mr. Raymond N. McCandless Research Group, Inc, Vermont Division of Occupational 43 State Street
& Radiological Health Montpelier, Vermont 05bO?
Administration Building 10 Baldwin Street Regional Administrator, Region I Montpelier, Vemont 05602. U. S. Nuclear Regulatory Comission 631 Park Avenue Honorable John J. Easton King of Prussia, Pennsylvania 19406 Attorney General 1 State of Vermont 109 State Street Montpelier, Yemont 05602 John A. Ritscher, Esquire Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110
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-.u . ...~....~.a,u ENCLOSURE 1 PRELIMINARY EVALUATION OF CONTAINMENT SAFETY STUDY , ,
BY THE OFFICE OF NUCLEAR REACTOR REGULATION , FOR THE VERMONT YANKEE NUCLEAR POWER STATION VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271
1.0 INTRODUCTION
By letter dated June 30, 1986 from J. G. Weigand to H. Denton, the licensee, Vermont Yankee Nuclear Power Corporation, committed to perform a containment safety study. By letter dated September 2, 1986 from J. G. Weigard to H. Denton, the licensee transmitted the study for staff review. On September 12, 1986 representatives of the licensee met with the staff, summarized the study, and answered questions. On September 28 and 29,1986, staff representatives visited the plant and the neighboring Vernon Hydro Station to review plant features and portions of the study. The purpose of this evaluation is to provide the staff's preliminary comments on the licensee's study. Final comments will be based on the licensee response to staff's questions and comments. A regulatory position on generic implementation of improve'ments must await review of backfitting implications under the requirements of 10 CFR 50.109. The following evaluation should not I be considered a Nuclear Regulatory Conmission position on the issue of Vermont I Yankee Nuclear Power Station containment safety. 2.0 EVALUATION 4 i The licensee's study basically consists of two parts. The first part is ; an evaluation of containment failure probabilities in severe accidents. The 4 i _ - _ _ - _ - - - _ - _ _ _ _ _ _ - _ - . - . - - - - _ _ _ _ . . - _ . _ ~
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,.2-second part is an evaluation of potential improvements in five areas (combusti-ble gas control, drywell spray capability, containment venting, core debris control, and training and procedures).
2.1 Containment Failure Probability Evaluation The licensee evaluated the design and operation of Yemont Yankee Nuclear Power Station (VYNPS) with respect to core melt probabilities and containment system response. The evaluation used detailed assessments for a similar reactor, adjusted for specific VYNPS features. The principal conclusion by the licensee was that a best estimate of the conditional probability of containment failure was .07. That is, if a core melt accident were to occur at VYNS, there l 1s a 7 per:ent chance that the containment systems will fail and result in a large release of fission products. The bases for the licensee's estimate include a number of assumptions, evaluations and judgements. One important judgement pertains to the use of results from an industry' developed computer code called Modular Accident AnalysisProgram(MAAP). The staff has not reviewed this code, at present has no plans to do so, but will assess the results and conclusions from its applications to VYNS. The accuracy and application of the code is a subject of considerable debate however. Therefore, uncertainty must be attached to the use of the code at this time. A number of important conclusions have been drawn by the licensee based on the containment safety study. The more important ones are:
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i A conditional containment failure probability of .07. A maximum containment system failure pressure of about 135 psia. The best estimate core melt probability was 3x10-5 per reactor year (one chance in 30,000 per year of reactor operation). One accident class (loss of cooling capability with the l pressure vessel at high pressure) dominates the core melt , probability (about40 percent). Although station blackout and ATWS are significant contributors to the core melt probability at Vennont Yankee, as for other plants, the largest contributor is loss of makeup with the vessel at high pressure. The staff's experience with core melt and containment failure estimates indicates that large uncertainties exist in these estimates. In addition, a high pressure core melt entails greater uncertainty regarding containment response than prevails with low pressure core melt. Pending receipt and analysis of the response to staff questions, the staff concludes that the i licensee's estimates appear optimistic considering the uncertainties inherent 1 in failure rate data, modeling of systems, human responses, accident initiator i identification, as well as the physical processes that follow degraded core conditions. Vermont Yankee has four SRVs which serve as the automatic depressurization system (ADS), compared to eleven SRVs at Peach Bottom, five of which serve
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as the ADS. Further, at Vermont Yankee the four SRVs are three-stage Target I
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Rock valves which have relatively high failure rates based on reported failures. $ When the the failure experience of SRVs is considered,' accident sequences TPUV, TPQUV, 1 PQX, T PQUV,gT QUV and TPW need to be added to the dominant E E accident sequences identified by the licensee. As a result, the total core damage frequency of 3x10-5/R-yr could be increased by about a factor af 2. WhenSRVfailures(spuriousopeningandstuckopenreliefvalves) rented accident sequences are considered, the release type for Class II accidents (i.e.,lossofcontainmentheatremovalandsubsequentlossofmakeup)may change. This is due to possible early failure of the RHR capability after loss of NPSH as a result of stuck open relief valves and subsequent pool heatup. As a result, the containment conditional failure probability for Vermont Yankee could exceed 7% for all sequences and result in containment failure within 24 hours. The frequency of Loss Of Offsite Power (LOOP) at Vemont Yankee is estimated to be 0.07/yr. When compared with 0.22/yr for the national average grid, the Vermont Yankee value is deemed too low based on the conventionally used Bayesian estimation' technique. The failure rate of the nearby hydropower plant is estimated to be higher than 0.07/yr. In the Bayesian estimate, the plant data are used as inputs modifying the a priori distribution, i.e., the national average LOOP frequency. Additionally, the estimated 6-8 hour battery capacity at Vermont Yankee was based on 2175 ampere-hours rated capacity discharged at a constant load of 1.75 volts / cell, with the load estimated to be 27 amperes. This does not include the excessive loads likely to occur during the startup of motors, and the abnormally high ambient temperatures likely during the accident. Thus, a more
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- 5 likely estimate for the battery capacity is on the order of 3-4 hours when all those factors are considered. This will, in turn, affect the core damage frequency and the conditional containment failure probability of the complete-station blackout sequence. If there is additional battery capacity not
. included in the Vemont Yankee Containment Safety Study, the Licensee should state so for clarification.
Other important areas of considerable uncertainty are the manner in which the vessel may fail (and the amount of resulting steel in the core debris), the effectiveness of the reactor building for fission product attenuation, iodine may not be a good surrogate for risk estimation (refractory products from core concreteinteractionmaybebetter),andthecontainmentloadingsandresponse following vessel head failure. Based on the above, we conclude the CCFP for VYNP is probably less than 50%. 2.2 Current NRC Concerns For U.S. Containments Severe accidents dominate the risk to the public associated with nuclear power plants. A fundamental objective of the Commission's Severe Accident Policy is to take all reasonable steps to reduce the chances of occurrence of a severe accident, and to assure the capability to mitigate the consequences of such an accident should one occur. The Reactor Safety Study report issued in 1975 found that for BWRs the probabilities of accidents resulting in core melt wtre low, but the containment perfomance following a severe accident was L poor and tended to offset the benefits of low BWR core melt probabilities. l Subsequent actions resulting from the TMI Action Plan have led to several l m ____- - - _ _ _ - _ _ _ _ _ _ - _ _ . _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _
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plant modifications and required improvements in plant procedures to further ) 1 reduce the likelihood and consequences of severe accidents. In concert with l the Comission's policy to further reduce the chances of occurrence of severe I accidents and to mitigate their consequences, an industry initiative is underway to develop a methodology for Individual Plant Evaluations (IPEs) for use in the search for risk outliers. The resulting approach will be epplied on a plant-specific basis. The staff may find that, while the IPE approach may l satisfactorily address system reliability and containment performance for' f each plant specifically, the process will necessarily be a slow one. The staff has, therefore, identified several potential deterministic containment enhancements which lend themselves to generic implementation and have the i l potential to significantly mitigate the consequences of most severe accidents. I The generic approach has the advantage of expeditious implementation on all plants and will be responsive to Commission's policy regarding mitigation of the consequences of severe accidents. J' Based on the insights g'ained from PRAs, the staff has;, identified the following five areas of potential BWR containment enhancements.
- 1. Provisions should be made for reliable operation of drywell containment sprays for a broad spectrum of accident sequences, including blackout i 1
sequences. The reliability of containment sprays should be enhanced
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l by providing independent water and power sources. Backup water i sources and pumps, hose connection and use of fire mains should be considered. The provisions to be implemented should minimize i i
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1 , . occupational exposures that could result from manual actuation, and procedures lshould be explicitly developed and expeditiously. implemented as part of the BWR Owners Group development of the Emergency Procedures Guidelines.
- 2. Provisions should be made for symptomatic response and reliable actuation of containment wetwell purge and vent' valves. They should open and close under accident conditions as a means to assure that thebeyond-design-basiseventsdon8tleadtooverpressurefailure of the containment and the selected vent paths. They should provide
- a. path for releases which will maximize the use of the suppression pool as a condersing and filtering medium.
- 3. Emergency procedures and training should be reviewed and modified as necessary to assure that operators are able to recognize severe accident conditions and use plant equipment to best advantage under such conditions. Revision 4 of the BWR Emergency Procedures guidelines should be implemented promptly following the staff's review and approval.
- 4. Paths for core debri's travel should be evaluated for conditions repre-sentative of a large scale core melt. Where the expected path cf debris travel indicates a substantial likelihood of loss of the suppression pool as a release filtering or debris quenching medium in' BWR containments, the torus room under the suppression pool should contain a 3 foot high barrier to trap water and core debris.
. 5. Combustible gas control provisions should provide substantial assurance that containment failure due to hydrogen combustion is not likely in the
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-potential severe accident sequences, including blackout sequences. The period of operations while containments are deinerted while at power, particularly during potential preshutdown conditions, should be minia.ized by reducing the present Technical Specifications permitted value of 24 hcurs.
The licensee has evaluated the five areas as appropriate to Yemont Yankee and has found that with the exception of combustible gas control and debris control enhancements, the proposed enhancements could be beneficial for Vemont Yankee severe accident mitigation capability. The staff recommends that the licensee should reconsider the combustible gas control and debris control issues in the light of the following staff evaluation. The licensee should al!.o complete the further studies proposed in its containment study report, propose specific modifications that it plans to undertake, and provide a tentative schedule. 2.2.1 Combustible Gas Control Vemont Yankee is allowed by its Technical Specifications to be deinerted for 24 hours at startup and 24 hours prior to shutdown. The staff has proposed the approach of minimizing the deinerting' time, thus reducing the vulnerability of hydrogen combustion.- A 12 hour deinerted period may be sufficient. Based on discussions with the license during a September 11, 1986 sii.e visit, the i staff understands that it takes about 8 hoars to deinert the containment. l
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Therefore, it appears feasible to reduce the deinerting period in the Technical i Specifications. ' l l 1 l
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In the Vermont Yankee containment study the licensee has concluded based on operating experience that deinerted power operation is about 1.1 percent. of-the total operating time, and concludes that such a low percentage will not significantly impact safety. Since the licensee was satisfied that its design objectives have been met, any impacts asscciated with shorter deinerting times were not identified. This issue should be explored further in order to address the staff's goal relating to increased assurance that the presence of a combustible mixture inside primary containment is reduced to as low as practicable levels. To this end, the study should include an evaluation of past practice to determine the impacts to plant operation due to a shorter allowable deinerting period. 2.2.2 containment Sprays The licensee has identified and implemented means to improve the reliability / availability of drywell sprays through the use of the following installed features: a) redundant sources of water to the containment sprays; namely, the torus, the ultimate heat sink, and the cooling tower basin. ' b) redundant pumping capability for the containment sprays; namely, the RHR pumps and the SWS pumps. c) the capability to operate the containment sprays during loss of a.c. power by using d.c. operated valves and a diesel driven fire pump.
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. u - :aa:. w:.ca , ..:ai.aa.au d) the capability to use the diesel driven fire pump to pump water i through the RPV and into the containment during severe accidents that violate the RPV integrity. This mode could be used to cool the core debris and avoid over depressurizing the containment.
Additionally the licensee indiceted the possibility of automating
'the valves in the Reactor Building that are required to line up -
the diesel fire pump with the containment and RPV sprays. The licensee has concluded that it is practical to use the containment
-sprays to control containment pressure and temperature and cool core debris-during a severe accident. -
The staf.f concurs with the licensee that using containment sprays after a severe accident can assist in the control of containment pressure and temperature, scrub the containment atmosphere and cool the core debris. However, in order to accomplish these objectives sufficient spray flow must be maintained. The diesel driven fire pump may have insufficient flow capacity to oroduce a spray that is effective for pressure and temperature Control, scrubbing, and debris cooling. Insufficient flow to the spr?.ys can result in water droplets that are too large to scrub or control pressure and temperature, and spray coverage may be insufficient to cover the core debris. Additional alternate water supplies (and/or spray nozzles) to the containment sprays should be considered since the diesel driven fire pump cannot supply sufficient water to ensure fission product attenuation and heat removal. The IDCOR/BWROG study (Evaluation of proposed BWR Accident Mitigation Capacity Relative to Proposed NRC Change) ef the use of sprays is in conformance with the staff objectives. Further, IDCOR/BWROG estimated that only 250 gpm is necessary
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... .~ l to cool the core debris to prevent it from damaging the containment. Although 250 gpm may be sufficient to remove decay heat from the core debris when the spray is ideally distributed, a larger flow of water from the sprays may be required to establish a good spray distribution for fission product removal and to remove the latent heat from the melted core and debris.
The licensee's proposed usage of the containment sprays is a reliable and cost effective means for post severe accident control of containment pressure and temperature, scrubbing of the containment atmosphere, and cooling the core debris, and is consistent with the staff's views. However, in order to be assured of sufficient flow for severe accidents further study is recommended to better understand spray performance at derated flow conditions. 2.2.3 Venting With regard to the venting issue, the licensee discussed two of the most significant accident scenarios; station blackout and ATWS. The report concluded that containment venting should only be considered as a last resort. The staff agrees that venting the noble gas activity is a procedure of last resort. The licensee also concluded that venting may be of little value in preventing core damage for an ATWS. In the case of station blackout, the licensee sta au that opening a wetwell vent is difficult and the potential benefits are limited. Therefore, the licensee concluded that containment j venting is not considered to be practical. Even though the report concluded unfavorably with regard to venting, it I identified six possible vent paths from the existing piping arrangements. After evaluating each path, a "12 inch vent path" was proposed as being the j l _ ________________-___-___-------A
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m,w -: . x o. :. & y , ,. .- ' i j l most desirable. This proposed path bypasses'the Standby Gas Treatment System i filters. The modification would require that approximately 15 feet of 12-inch 1 f pipe be added and that one new motor-operated isolation value he added to the l line. The proposed three-inch Atmospheric Control System line includes the Standby Gas Treatment System. For this pathway, additional vendor analysis and testing would be required for the. reliability of the valves. Both paths would utilize the vent stack. The final selection was left open pending further
' study, which will include consideration of competing safety requirements.
None of the potential vent paths identified by the licensee has been shown capable of opening and closing at the pressures, temperatures, radiat' ion, and steam environments associated with severe accidents. The issues of a.c. independent power sources for the needed valves have not been evaluated by the licensee. The staff agrees that further study should be perfomed. In the study, consideration should be given to more recent investigations such as the Evaluation of proposed BWR Accident Mitigation Capacity Relative to Proposed NRC Change, submitted by IDCOR/BWROG, dated August 1986. , 1 In particular, the availability of adequate ' independent power under accident condition should be considered (station blackout). The Vermont Yankee study did not consider independent power sources because an earlier study (Harrington, R.M. and Hodge, S.A.; " Containment Venting as e Severe Accident Mitigation Technique for BWR Plants with Mark I Containment"; June 26, 1986) concluded that wetwell ventiag during Station Blackout has limited beneficial potential for BWR Mark I containment plants. This limited benefit may be significantly understated because of key assumptions in their analysis
j si; m y A . ~ :. A ip.w..:.n u2,w.4 w a n.n , y .g;;; m yyp p g.w w f.n % ;u Q 1 and, in addition, it should be weighed against the IDCOR finding that station.- y blackout is a significant contributor to core melt probability. .The study should also include an engineering evaluation that compares venting requirements
-with venting capabilities. The report stated that the need for substantial-venting cap 6bility may arise during an ATWS scenario in which'the energy in the fonn of steam that must be removed from the containment would be in the range of approximately 10% to 40% of rated reactor power. However, it is not clear if the vent paths noted earlier have this capability. .The IDCOR study indicated that vent sizes for power levels of 20 to 30 percent may require vents of 26 to 34 inches in diameter, whereas the sizes required to remove decay. heat after 10 minutes decay for other scenarios would range from 4 to 6 inches.
Another aspect of' venting that can have a 'significant effect on accident i consequences (onsite as well as offsite), and that needs further consideration, is the consequence of elevated releases. If the vent path includes the plant stack, the elevated release during venting can prevent the contamination of onsite buildings and equipment. As noted in the licensee's study, some vent paths have the potential for making portions of 3 the plant inaccessible for accident control and mitigation functions. Specifi-cally, some of the vent paths have the potential for releases into the Reactor and Control Buildings. One advantage of an elevated release through the plant stack is' that it maintains operator accessibility to vital plant areas. Elevated releases can also reduce substantially offsite dose consequences in the vicinity of the plant. Reduction factors of 10 or more are conceivable within two or ] i three miles. Hence, the licensee should evaluate and quantify the specific d 1 i
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i benefits that can be achieved with respect to offsite dose consequences by venting through the plant stack. Specifically, the licensee should consider , site specific meteorology and topography in the ev'aluation of, venting through the plant stack. The offsite atmosphere transport should extend to distances where the effects of stack height become negligible. 2.2.4 Core Debris Barriers The use of core debris barriers to prevent molten core debris from penetrating the steel containment shell or. torus vent pipes has been investigated by the licensee. During initial construction the bottom of the steel drywell at Vermont Yankee was. backfilled with concrete to E1. 238'. Should core debris melt through the reactor it will fall on this concrete in a 17' -2" diameter enclosed area (231.5 ft 2) known as the Subpile Room. The only paths available for core debris to exit the Subpile Room are through its only doorway, or through either the three-inch steel floor drain or the three-inch steel equip-ment drain. The Subpile Room floor and equipment drain lines are imbedded in the drywell floor concrete and are routed to sumps inside the drywell. There is a minimum of approximately 1.5 feet of concrete underneath the sumps and piping before reaching the steel drywell shell. The Subpile Room doorway is a curbless opening in the portion of the biological shield wall which supports the reactor skirt. 1 The licensee states that should core debris melt through the reactor onto the circular Subpile Room floor and stay molten with enough vertical head available to Cause it to flow, it Could pass out of the Subpile Room though l - _ - - _ - - - - _ _ _ _ _ _ - - - - _ - - - - - _ - _ _ _ - - _ _ _ _ - _ _ - - - - - _ _ - _ _ _ - - - _ _ _ _ - _ _ . - . _ - -. -
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r . 4 . L its only doorway, or through either the three-inch steel floor drain or the three-inch steel equipment drain. The Subpile Room floor and equipment drain i linis are imbedded in the drywell floor concrete and are routed to sumps inside the crywell. If the entire volume core debris is available for spreading, a maximum layer thickness of approximately 1.1 inches would result. Since the vent line is approximately 1 foot above the drywell floor, it is not expected that tie molten core debris could enter the torus through the vent pipes. The licensee concludes that Vermont Yankee analysis indicates that a scenario in which sufficient core debris melts t.nd propagates to the contain-ment boundary does not appear feasible. Vermont Yankee has a similar size drywell and reactor vessel as the Peach Bottom plant analyzed in WASH-1400, yet has less than half the number of fuel assemblies and control rods. Significant uncertainty exists in core and steel debris volume and transport analyses. The Vermont Yankee analysis indicates that core debris will not reach the drywell shell. Vermont Yankee already has diverse capability to spray water into the containment. The licensee states that additional physical barriers may be counter-productive as they may prevent containment spray from cooling the debris while it is confined in the Subpile Room. The Vermont Yankee analysis shows core debris control is closely coupled to contain-ment spray capability. Our preliminary assessment is that full core debris barriers as considered by the licensee may not be cost effective for Vermont Yankee. However, the l licensee should examine localized barriers protecting the downcomer opposite 0
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j Such a modification may not be costly, and will provide added assurance that f I core debris travel will net cause suppression pool failure or bypass. ) 1 2.2.5 Severe Accident Procedures i i Vermont Yankee's present symptom-oriented Emergency Operating Procedures (EPGs) are based upon the latest approved version of the BWROG EPG's (Revision 3). However, containment venting and reactor power control using water level for ATWS conditions were not included in the Vermont Yankee Emergency Procedures. The licensee has indicated a willingness to consider updating the emergency operating procedures to include the EPG Rev. 4 guidance on venting and ATWS power control. The licensee should implement the operating procedures currently being considered by industry in EPG Rev. 4, following NRC staff reviews. Any j future addenda which may result from the current program of enhancing the containments capability to mitigate the consequences of severe accidents. 2.2.6 Other Improvements Other potential containment performance improvements should also be considered. For example, the licensee should consider supplemental power supplies for improved automatic depressurization system operation for station blackout response. This is particularly important since loss of makeup at high pressure appears to be the largest contributor to core melt. I
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3.0 CONCLUSION
S The staff concludes that the Vermont Yankee containment safety study.has
'provided evidence.that the containment is mere capable of performing its function during severe accidents than previous assessments of Mark I type containments would indicate. That is, given a core melt accidents, a prelimi-nary evaluation indicates a less than 50 percent likelihood of containment failure. The staff also concludes that the likelihood and consequences of such a failure may be reduced substantially with modest improvements at discussed above.. The staff, therefore, recommends that the licensee perform the feasibility studies indicated by the study, and those proposed in this evaluation.
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- g ENCLOSURE 2 VERMONT YANKEE CONTAINMENT STUDY REQUEST FOR' ADDITIONAL INFORMATION p '
-1. What is your estimate of the overall uncertainty of conditional containment failure probability and its basis?
- 2. . What is the affect on core damage frequency when accident sequences TPUV, TPQUV, ET PQX, ET PQUV, Tg QUV, TPW are included in the dominant accident sequences based on reduced battery life, and the number and type of SRVs compared to Peach Bottom, and on the CCFP7
- 3. Given that the national average'value for frequency of loss of offsite power is on the order of 0 22/yr justify, on the basis of the Bayesian estimate that the frequency of loss of offsite power at Vermont Yankee is 0.07/yr. -
- 4. Verify that the total battery capacity available at Vemont Yankee is greater than 2175 ampere-hours, and that it could be maintained at a voltage greater than 1.75 volts / cell in hich ambient temperature during the accident for 6-8 hours.
.5. How often are the RHR/RHRSW interconnecting valves actuated to assure that the valves work' properly?
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- 6. How often are the interconnecting valves between the RHRSW and the fire protection system (fire pumps) actuated to assure that the valve works
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- 7. How readily can the MSIVs be reopened following closure at operating conditions?,jWhat interlocks must be bypassed and how complicated are the procedures (e.g., must the differential pressure across the MSIVs be reduced:for the valves to be re-opened)? i l
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- 8. It is not clear how the CCFPs given or page 74 of the report were obtained.- Please expir.in.
- 9. Waar'SLCS modifications are proposed for V.Y.? Page 86 discusses the p
aavantages 'of two different possible modifications, but gives no commitment i te +4ther
- 10. Identify the testing and maintenance requirements you use for the diesel driven fire pump. Do these requirements conform to those contained in the National Fire Codes? Also identify any reliability infomation fot* the system such as outages and failures to start on demand. What outage time limitations do you use for the system while at power?
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- 11. Identify the scope of modifications required to the spray system, or increases in the pumping capacity, to assure a uniformly distributed spray with proper droplet size (as cpposed to a dribble) if the diesel fire pump were used in a core melt event. Approximately what would !
the costs be of such modifications? i i
- 12. Can portable AC generatort be used effectively to power vital valves and/or small pumps for station blackout accidents? If so, what modifications would be required, and what would be their approximate costs?
- 13. In section 2.2.1 you conclude that the containment can be " expected to withstand pressures approximately two times design prior to failure."
Provide the bases for your conclusion.
- 14. In section 2.2.10.3/ isn't the water supply i' rom at least a portion of the cooling towers also available?
- 15. The use of the Vernon Hydroelectric Station is referenced in Section 2.2.11.1, and discussed in more detail in Section 4.4.2.3. Reliability estimates are presented on page 62. Please provide the basis for the reliability estimates with reference to both the historical operation of Vernon Hydro, and the transmission line and substations to Vermont Ytnkee.
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- 16. TheNitrogenContainmentAtmosphereDilution(NCAD)systesisreferenced s
on pages 25 and 115. What maintenance and surveillance procedures are used to ensure operability?
- i. i
- 17. In Section 4.1.4 MARCH /RMA and MAAP code pachage for Vermont.Ydhl:ee is referenced. Were calculations made for Vermont Yankee, or were'ihe results of computation for other reactors evaluated for the Vermont Yankee design? What calculations were made?
- 10. Invessel,and exvessel steam explosions were not considered credible (pg.
t 55, 1st para.) based upon research. Identify the research that forms the basis for this conclusion.
- 19. As we understand tables 4.7 through 4.10, the designations E, L, and NCL refer to early (E) or late (L) containment failure estimates, and NCL refers to no containment failure. The second designators H. N and L refer to high, medium and low releases, respectively. Towhatsrteht ,
can mitigation through manual actions in the time avasicble, ar.d Ir. the temperature and radiation environments associated with such accident l types be expected to be. successful for early failures; for late failures?
+
' , Spect fica 11y), for the combustible gas control, spray, and venting evalua-tions, what do you judge the effectiveness of the existing plant and procedures to be versus the possible improvements for early and late sequences?
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.20. NPSH during spraying is identified as'a concern on page 117. To what extent will- further investigation be undertaken to determine whether NPSH is an issue? Yerify that procedures exist for the operator to lineup ECCS water sources outside the containment in the event NPSH requirements are not met. If analysis indicates it is an issue, what do you propore be done to eliminate or reduce the level of concern?
- 21. Venting is considered for the station blackout sequences only. Please discuss your rationale for not considering other everts when venting may be beneficial.
- 22. Since there is a substantial difference between the heights of the VYNPS
, plant stack (318 feet) and the reference plant stack (500 feet), indicate how this was taken into account in the comparison of the two plants.
- 23. Evaluate the differences in offsite dose consequences due to venting at I ground level versus through the stack. ' sing J site specific meteorology and topography, provide an estimate of the offsite dose differences l between the two types of reler,se as a function of distance from the site.
-)
- 24. On page 125, rapid containment depressurization which could fail the drywell is offered as an uncertainty relative to containment venting.
What analysis and/or tests are being conducted to reduce this uncertainty? If no analysis or tests are contemplated, what actions are proposed to minimize the uncertainty? L l
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- 25. - Remote manual.value operation is discussed in Section 5.3.5.1.1.,
primarily with respect to station blackout. To what extent can the remote vent valve and any spray valve alignment be counted on for the other classes of sequences you assessed? That is, if remote manual operation is not available, would the local environment the operators would encounter allow s>ccessful local operation?
- 26. Severe accident venting discussed in Section 5.4 does not include an eveluation of the reliability of the ADS system. Given the types of
-severe accident challenges you have described, provide your estimates of the reliability of ADS valves; i.e., their potential as a suppression pool bypass path. Can battery packs or portable generators b9 used to assure high reliability? If so, at what approximate cost?
- 27. What is the approximate cost for improving the valving for the diesel fire pump?
- 28. For the improvement options you have evaluated, what maintenance and surveillance guidelines would you propose to use?
t
- 29. To what extent do you coasider the option of drywell flooding to be 1
effective? If effective, would you include the option in future revisions to your emergency operating procedures.
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l 30. It is estimated that the maximum debris layer thickness on the drywell floor would be approximately 1.1 inches (page 136). Provide the bases for such a conclusion.
- 31. What is the thickness of the vent dur/c betwet:n torus and drywell? (Fig 1, 1
Page F-6).
- 32. It is stated that " additional physical barriers are believed to be counter-productive as they may prevent containment spray from cooling the debris while it is confined in the Subpile Room." Please elaborate how such a barrier would prevent the spray from effectively cooling the debris. (p137).
- 33. It is indicated that, with the exception of the one-inch nitrogen CAD line and the six-inch nitrogen purge line, the pipe lines associated with four potential vent paths'are likely to fail. Provide background information which led to this conclusion (p 131).
- 34. It is implied that a layer of debris (1.1 inch thick) would not penetrate the drywell steel shell and enter the torus (p 136).' If such is the conclusion, please discuss why such a burn through is unlikely while core debris is attacking the dry well floor. There is a gap between drywell steel shell and concrete shield. If the molten core were to burn through i
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the steel shell at the indicated corium elevation,'what would prevent the fission gas.from entering the reactor building since the concrete shield outside the drywell shell is not designed as a pressure boundary?
- 35. It is stated that 135 psia is a reasonable value for the VY Cor.tainment failure pressure (page F-5) a) What is the uncertainty range associated with this value?
b) What would be a change in core melt and conditional containment failure probabilities associated with'the uncertainty? c) Provide references for the Ames and Sandia calculations mentioned in the Appendix F. (page F-1)
- 36. Your evaluation of deinerting indicates a relatively few hours of power operationwhilethecontainmentisdeinerted(i.e.,about1%inthe RunMode). For each such instance, please identify the following:
a) the number of hours deinerted; b) the purpose for the deinerted condition, and whether it was successful; before shutdown was required by your technical specification, and c) the' power level and its corresponding reactor pressure at which containment entry and exit were made. Please indicate the impacts you would expect for a deinerting technical specification to either 16 hours, or 12 hours. i m k
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- 37. Please provide your estimattes of pressure and temperature as a function of time for the accident sequences you analyzed for CCFP estimates.
- 38. It is not clear from your evaluaticn why the probability estimates of early failures with higher releases are lower than for Class IV events.
Please explain if verding of ATUS sequences before core melting was assumed? _.___ -. _ _ __ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - - - _ - _ - _ - _ - _ . _ _ = - - _ - _ - - --- - - - - - - - - - - - - --. --
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I h ACh(y\ ANN NAEMA
.. October 3, 1986 ' J i I
/o Docket No. 50-271 LICENSEES: Vermont Yankee Nuclear Power Corporation FACILIlY: Vermont Yankee Nuclear I'ower Station _
SUBJECI: SEPTEMBER 11, 1986 MEETING WITH THE VERMONI YANKEE NUCLEAR POWER CORPORATION (VYNPC) RE: Containment Safety Study On Sept. amber 11, 1986, a meeting was held at the NRC headquarters in Bethesda, Maryland to discuss the completed Vermont Yankee Mark I Containment Safety Study. Enclosure 1 is a list of individuals that attended the meeting. Enclosure 2 is a handout of the slides presented by VPNPC at the meeting. The study had been submitted on the Vermont Yankee docket by letter dated September 2,1986, prior to the meeting. VYNPC described the contents of the study and answered various specific technical questions. the staff acknowledged the receipt of the report and comented that plant specific information was provided which was expected to be useful in the generic BWR Containment Requirements effort. Furthermore, the staff noted that, considering the timing of the study, Vernont Yankee dovetails neatly as a sample case for generic containment requirements activities. Consistent with this purpose, the staff intends to review and comment on the study by late October to the extent of identifying significant areas of agreement an! disagreement and providing questions. VYNPC commented that in ordar te provide timely response to questions, the continuation of this level of effort might conflict with certain other schedule commitments to NRR. The staff advised VYNPC to identify such cases to Project Management for discussion. 0 Wo! c'Id by Vernon L. Rooney, Project Manager BWR Project Directorate #2 Division of BWR Licensing
Enclosures:
DISTRIBUTION As stated Docket File RBernero cc w/ enclosures: Lo a DR See next page PD#2 Reading EJordan VRooney BGrimes ACRS (10) 0FFICIAL RECORD COPY DBL:PD#2 VRooney:cb l'/oJ /86
Mr. R. H. Capstick Vermont Yankee Nuclear Power Vermont Yankee Nuclear Power Corporation Station cc: - Mr. J. G. Weigand Mr. W. P. Murphy, Vice President & President & Chief Executive Officer Manager of Operations Vermont Yankee Nuclear Power Corp. Vermont Yankee Nuclear Power Corp. R. D. 5, Box 169 R. D. 5 Box 169 Ferry Road Ferry Road Brattleboro, Vermont 05301 Brattleboro, Vermont 05301 Mr. Donald Hunter, Vice President Mr. Gerald Tarrant, Commissioner verment Yankee Nuclear Power Corp. Vermont Department of Public Service 1071 Worcester Road 120 State Street Framinghane, Massachusetts 01701 Montpelier, Vermont C5002 New England Coalition on Public Service Board Nuclear Pollution State of Vermont Hill and Dale Farm 120 State Street R. D. 2. Box 223 Montpelier Vermont 05602 Putney, Vermont 05346 Vermont Yankee Decommissioning Mr. Walter Zaluzny Alliance Chairman, Board of Selectman Box 53 Post Office Box 116 Montpelier, Vermont 05602-0053 Vernon, Vermont 05345 Resident Inspector Mr. J. P. Pelletier,-Plant Manager U. S. Nuclear Regulatory Commission Vermont Yankee Nuclear Power Corp. Post Office Box 176 Post Office Box 157 Vernon, Vermont 05354 Vernon, Vermont 05354 Vermont Public Interest Mr. Raymond N. McCandless Research Group, Inc. Vermont Division of Occupational 43 State Street
& Radiological Health Montpelier, Vermont 05602 Administration Building 10 Baldwin Street Regional Administrator, Region I Montpelier, Vermont Q5602 U. S. Nuclear Regulatory Commission 631 Park Avenue
. - Honorable John J. Easton King of Prussia, Pennsylvania 19406 Attorney General State of Vermont 109 State Street Montpelier, Vermont 05602 i John A. Ritscher, Esquire l Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110
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[, Enclosure 1 j i i LIST OF MEETING ATTENDEES ~ VEP.NONT YANKEE CONTAINMENT SAFETY STUDY September 11, 1986 ; Name Organization. D. Muller NRR/ DBL /PD#2 V. Rooney NRR/ DBL /PDf2 E. Wenzinger, Sr. NRC Region I J. Hulman NRR/ DEL /PSB D. Vassallo NRR/ DBL /F0B R.' Houston NRR/ DEL .j M. Thadani NRR/ DBL /PD#2 , D. Shum NRR/ DBL /F0B / G. Lainas NRR/ DBL F. Eltawila NRR/DSR0/ RIB W. Hodges NRR/ DBL /RSB i F. Lodwick VYNPC i J. Thayer Yankee Atomic Electric Co. S. Schultz Yankee Atomic Electric Cc. l W. Murphy VYNPC J. Goetz Washington L. Marinos Reporting Servi:e T. Landers New York Power Authority J. Graf, Jr. New York Power Authority L. Clifford GE . S. Floyd Carolina Power & Light Co. K. Holtzclau GE N. Edwards NUTECH
~ -
T. Pickens Northern States Power M. Jmay - TVA-BFN R. Bachmann" NRC/0GC
^
S. Markiewicz State of Massachusetts K. Campe NRR/ DBL /PSB J. Coats
- Vermont Public Interest Research Group E. Weiss Harman & Weiss Union of Concerned Scientists S. Murphy Nud.-Information & Resource Service E. Fotoponlos Sechtel S. Bokim MR/ DBL /PSB F. Coffman LM/ SORI A. Notafrancesco NRR/ DBL /PSB
- C. Reid Bechtel
- 0. Scott Southern Co. Services J. Kudrick NRR/ DBL /PSB l
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l i l I VERMONT YANKEE CONTAINMENT SAFETY STUDY O 9 I
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I VERMONT YANKEE CONTAINMENT SAFETY STUDY a
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i INTRODUCTION PERSPECTIVE SCOPE RESULTS
SUMMARY
11 MARK I CONTAINMENT DESIGN REVIEW III SEVERE ACCIDENT ANALYSIS AND QUANTIFICATION BWR DOMINANT SEQUENCES CONTAINMENT CONDITIONAL FAILURE PROBABILITY , o METHODOLOGY o ACCIDENT SEQUENCE QUANTIFICATION o INITIATING EVENT FREQUENCIES o CORE MELT FREQUENCIES o CONTAINMENT EVENT TREES
,- o CONTAINMENT PERFORMANCE RESULTS IV MARK,1 CONTAINMENT POLICY ISSUES o l HYDR 0 GEN CONTROL o DRYWELL SPRAY CAPABILITY o CONTAINMENT VENTING }
o CORE DEBRIS CONTROL e - o SEVERE ACCIDENT TRAINING PROCEDURE V DECISION ANALYSIS FOR RECOMMENDATIONS l o RECOMMENDATIONS AND INSIGHTS I l
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PERSPECTIVE e INITIATED - TO RESPOND TO QUOTED STATEMENT THAT MARK I CONTAINMENTS HAVE A 90% CONTAINMENT CONDITIONAL FAILURE PROBABILITY e -AGREED UPON BY VERMONT YANKEE, THE NRC AND THE GOVERNOR OF THE STATE OF VERMONT ON JUNE 30,1986 e PROGRESS INITIATED JULY 1, 1986 INTERIM STATUS MEETING IN BETHESDA ON AUGUST 6, 1986 FINAL REPORT COMPLETE AND TRANSMITTED
- TO NRC AND SIATE OF VERMONT ON
,- SEPTEMBER 2, .1.986 b
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l l - STUDY SCOPE e MARK i CONTAINMENT DESIGN REVIEW o - QUANTIFY A VERMONT YANKEE SPECIFIC CONTAINMENT , CONDITIONAL FAILURE PROBABILITY GIVEN A l SCENARIO LEADING TO CORE MELT e ADDRESS MARK I CONTAINMENT POLICY ISSUES HYDR 0 GEN CONTROL DRYWELL SPRAY CAPABILITY CONTAINMENT VENTING CORE' DEBRIS CONTROL h - SEVERE ACCIDENT TRAINING AND PROCEDURES O 9
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ta STUDYRESU111 . . . e VERMONT YANKEE CONDITIONAL CONTAINMENT FAILURE PROBABILITY OF 7% e CONTAINMENT CAPABILITY IMPROVEMENTS IDENTIFIED PROCEDURE CHANGES / ENHANCEMENTS ISSUES FOR FURTHER EVALUATION INSIGHTS BASED ON SEVERE ACCIDENT SEQUENCES a e O i { l
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MARK ! CONTAINMENT DESIGN REVIEW e- SCOPE: COMPARE VERMONT YANKEE WITH THE MARK I , REFERENCE PLANT 10 IDENTIFY DESIGN AND OPERATIONAL FEATURE DIFFERENCES SIGNIFICANT TO CONTAINMENT PERFORMANCE DURING SEVERE ACCIDENT SEQUENCES e' RESULTS: 4 SIGNIFICANT DESIGN FEATURES WERE IDENTIFIED
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RATIO 0F CONTAINMENT "lZE TO REACTOR POWER ELECTRICAL PLANT DIVERSITY. DIESEL GENERATORS TIE LINE TO ADJACENT HYDROELECTRIC STATIONS CONNECTION FROM DIESEL FIRE PUMP TO LPC1/ CONTAINMENT S? RAY O e O n l
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4 . l CONTAINMENT FAILURE PROBABILITY EVALUATION q i e BACKGROUND WASH 1400 (90%) DEFINITION OF CCFP e VERMUNT YANKEE ANALYSES APPROACH ACCIDENT SEQUENCE IDENTIFICATION ACCIDENT SEQUENCE QUANTIFICATION CONTAINMENT EVENT TREE DEVELOPMENT CET QUANTIFICATION e
SUMMARY
OF RESULTS O 9 6
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WhSH-JLIDOASSUMPTIONS e ALL SEQUENCES (MEll/NON-MELT) ARE ASSOCIATED WITH CONTAIN e CORE MELT RELEASE CATEGORIES
- 1. STEAM EXPLOSION !
- 2. OVERPRESSURE FAILURE DIRECT 10 ATMOSPHERE
- 3. OVERPRESSURE FAILURE THROUGH REACTOR BUILDING
- 4. ISOLATION FAILURE e RELEASE CATEGORIES 2 & 3 DOMINATE RISK e SEQUENCES WHICH CONTRIBUTE COMPRISE 90%
ATWS LOS.S.0F RESIDUAL CORE HEAT REMOVAL
. CURRENT ' STUDY e DIFFERENT DOMINANT SEQUENCES IDENTIFIED e KNOWLEDGE OF CONTAINMENT FAILURE PHENOMENA AND CRITERIA HAS EVO , , _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ _ - - - - - . _ - - - - - - - - - - - - - - - - - - - - - - - ^ - - - - - - - * - - - - - - - - - - - ~ ~ - - - - - - - - ' - - - ' ' - ~~
i i l . . I 1 i CONTAINMENT CONDITIONAL FAILLIRE PROBABILITY I L 1 l l . , o CALCULATE THE CONTRIBUTION OF EACH ACCIDENT CLASS TO TOTAL LIKEllH00D OF CORE MELT l e DETERMINE THE CONDITIONAL FAILURE PROBABILITY. 1 0F CONTAINMENT FOR EACH ACCIDENT CLASS e CALCULATE THE WEIGHTED AVERAGE OF PRIMARY ) CONTAINMENT FAILURE PROBABILITY I l
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INITIATING EVENT FREQUENCIES 1 Number of Verimont Yankee Category Events in Best Estimate Types of Events Frequency Data Base ~ (yr-1) 1 Translants Resulting in 20 Reactor scram with Bypass 1.62 Potentially Available 2 Events Bounded by Loss of
- 2
.Feedwater 0.16 3
MSIV Closure 4 0.32 4 Events with Bypass Not 8 Available 0.65 5
. Inadvertent opening of One 0 or More SRVs 0.062 6
Loss of Off-site Power 1 - 0.07' e e 4
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l l SLTKKARY OF CORE MELT FREQUENCY TYPES AT VERMONT YANKEE i l l . . 1 l Class Description Frequency Fraction 1 l . (Yr-1) IA Loss of Makeup: RPV 1.33E-5 .43 at high pressure IB Loss of Makeup: 6.2E-6 .20 station Blackout IC Loss of Makeup: 2.6E-6 .08 ATWS ID Loss of Makeup: 3.9E-6 .12 RPV at low pressure IE Loss of Makeup: 7 4E-7 .01 Loss of DC buses II Loss of Containment 2.1E-6 .07 Ma,at Removal III LOCAs 7.3E-7 .02 TV ATW3 2.2E-6 .07 Y . Interfacing LOCA Netlitible g litible ! 1 TOTAL 3.1E-5 100% [ 1
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- HAND CALCULATIONS e IDCOR TECHNICAL REPORT INFORMATION e_ ENGINEERING JUDGMENT e
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. CATEGORIZATION t
EH/M EL_ LM LL ..ECf_ l 1 1 BASE 7 12 4 9 68 S1 10 12 3 9 66 S2 3 9 4 10 74 S3 1 3 6 16 74' WHERE: S1 -
.EARLY CONTAINMENT FAILURE INCREASE FOR BOTH HIGH AND LOW PRESSURE MELT DOWNS S2 -
INCREASED OPERATOR ACTION CREDIT FOR ATWS EVENTS S3 - S2 PLUS CLASSIFICATION OF TW CLASS AS LATE VERSUS EARLY RELEASE
[ w .G acy xsoa . L ,, . , A .y a m i - ,. ' i: . . . ., e , . . . s , 4. . . . _ _ . .a n . MRK I CONTAINMENT POLICY ISSUES e FYDR0 GEN CONTROL-e, DRYWELL SPRAY CAPABILITY e ' CONTAINMENT VENTING e CORE DEBRIS CONTROL e
- SEVERE ACCIDENT TRAINihG AND PROCEDURES 9
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HYDROGEN CONTROL l l e OBJECTIVE: PREVENT HYDR 0 GEN COMBUSTION FOLLOWING SEVERE l l ACCIDENTS e VY STATUS
- CONTAINMENT INERTED DE-INERT TIME CONTROLLED BY TECHNICAL SPECIFICATIONS OPERATING HISTORY INDICATES DE-INERT TIME IS L l% OF TIME PLANT IS IN RUN MODE VY RECENTLY CONVERTED.T0 N 2 CAD e EVALUATION: CURRENT VY PERFORMANCE AND DESIGN MEETS OBJECTIVE e RECOMMENDATIONS: NONE
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_ _ _ _ _ _ _ ..________.__________..._______________._______._________________.____.__m_ . _ _ _ _ _ _
j DRYWELL SPRAY CAPABILITY e OBJECTIVE: PROVIDE AC INDEPENDENT CAPABILITY TO SPEAY DRYWELL 10 LOWER PRESSURE AND TEMPERATURE l l e VY STATUS:
- NORMAL CAPABILITY FROM RHR, BACKED UP BY SERVICE WATER (DEPENDENT ON AC POWER)
DIESEL DRIVEN FIRE PUMP CAPABLE OF BEING AllGNED TO LPCI/ CONTAINMENT SPRAY (AC INDEPENDENT)
- WATER SUPPLY FROM COOLING TOWER DEEP BASIN (AC DEPENDENT) e EVALUATION: CURRENT CAPABILITY ACCEPTABLE, ENHANCEMENTS POSSIBLE TO INCREASE RELIABILITY OF SUPPLY FROM DIESEL FIRE PUMP e fiEC0 EMENDATIONS:
- ENHANCE THE OPERATION OF THE DIESEL FIRE PUMP TO DRYWELL SPRAYFLOWPATH
- IMPROVE ABillTY TO RESTORE AC POWER
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e OBECTIVE: PREVENT UNCONTROLLED LOSS OF CONTAINMENT INTEGRITY e VY STATUS:
- ONLY VERY SMALL VENTING PATHWAY PROCEDURALIZED
- STUDY IDENTIFIES 6 VENTING PATHWAYS
- RISKS OF VENTING DISCUSSED e EVALUATION: CAPABillTY FOR WETWELL VENTING, TO ACHIEVE FISSION PRODUCT SCRUBBING, SHOULD BE ENHANCED
~
c RECOMMENDAfT0NS: ENGINEERING EVALUATION OF UPGRADING WETWELL VENTING TO BE INITIATED
, VENTING GUIDELINE DOCUMENT. PREPARED OUTLINING 32 POTENTIAL
. VENT PATHS, REQUIREMENTS FOR USE, ACCESS, ETC.
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CORE DEBRIS C0flTROL e OBJECTIVE: PRECLUDE CONTAINMENT FAILURE DUE TO ' SHELL CONTACT ' o VY STATUS:
- SMALL CORE VOLUME IN LARGE DRYWELL P WILL NOT MIGRATE TO DRYWELL SHELL
- DRYWELL SPRAYS WILL " REFREEZE" CORIUM .
e- EVALUATION: CURRENT VY CONFIGURATION MEETS OBJECTIVE e RECOMMENDATIONS: NONE 9 0 9
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, SEVEERE ACCIDENT' TRAINING AND PROCEDURES i
e OPACTIVE: ENSURE OPERATORS READY TO USE PLANT BEST ADVANTAGE IN SEVERE ACCIDENTS e VY STATUS:
- VY HAS IMPLEMENTED E0P's BASED ON REV. 3 0F e
EVALUATION: VYMEETSCURRENTGUIDANCE e RECOMMENDATION:
- IMPLEMENT REV. 4 0F EPG's WHEN AVAILABLE
- CONSIDER REVISION TO EXISTING E0P's FOR:
o STATION BLACX0VT o CONTAINMENT VENTING o DRYWELL SPRAY USING DIESEL FIRE PUMP
. . o GUIDANCE TO SUPPLY CST VIA FIRE SYSTEM o SAMPLE SUPPRESSION P0OL BEFORE TRAN'SFER T o PROCEDURALIZE 22H /0 CONTROL o COMPLETE DEVELOPMENT OF VY SPECIFIC LEV l'
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DECISION ANALYSIS FOR RECOMMENDATIONS e TWO PART TEST
- WHICH SEVERE ACCIDENTS WARRANT CONSIDERATION?
- WHAT IS ACCEPTABLE CONTAINMENT PERFORMANCE?
4 SEVERE ACCIDENT SEQUENCES
- MOST CRiilCAL TO CONTAINMENT PERFORMANCE ARE:
STATION BLACK 0UT ATWS WITH MSIV CLOSURE o ACCEPTABLE CONTAINMENT PERFORMANCE
- CONTAINMENT SHOULD REMAIN INTACT, WITHOUT EXCESSIVE LEAKAGE FOR 2f1 HOURS FOLLOWING EVENT INITIATION e RECOMMENDATIONS REPRESENT THE CONCENSUS OF THE STUDY TEAM
. REGARDING PROCEDURAL AND DESIGN ENHANCEMENTS T0 IMPROVE CONTAINMENT (
3 PERFORMANCE. 1
- MAXIM 12E USE OF EXISTING PLANT FEATURES i
- AUGMENT EXISTJNG DESIGN WHERE CLEAR BENEFIT IS APPARENT l
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INSIGHTS BASED ON SEVERE ACCIDENT SEQUENCES 4 e a e i
__ <.-.2._ _. . _ _ - . . .a.... . . . . . s. . _ . . . . , .s PRQCEDUREADD1TIONS/ ENHANCEMENTS e PROVIDE PROCEDURAL GUIDANCE FOR RESTORATION OF AC P A STATION BLACK 0UT e PROVIDE A REPAIR PROCEDURE OUTLINING THE RESTORATI TIE LINE IF DAMAGE SHOULD OCCUR e PROVIDE PROCEDURAL GUIDANCE FOR OPTIMIZATION OF DC POW ! FOLLOWING A STATION BLACK 0UT e PROVIDE PROCEDURAL GUIDANCE FOR USE OF THE SRV's FOR MA DEPRESSURIZATION FOLLOWING A STATION BLACK 0UT e PROVIDE PROCEDURAL GUIDANCE TO ALIGN THE DIESEL FIRE PUMP FOLLOWIhG STATION BLACK 0UT e PROVIDE LEVEL / POWER GUIDANCE FOR ATWS
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4. INS!GHTS BASED ON SEVERE ACCIDENT SEQUENCES e t PROVIDE ADDITIONAL EVALUATION, GUIDANCE AND OPERATOR TRA { ON RESPONSE OF THE RHR PUMPS TO HIGH SUPPRESSIO ] TEMPERATURES l
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e IMPROVE RELIABIU TY OF THE NITROGEN SUPPLY TO e EVALUATE DESIGN AND POST-ACCIDENT RELIABILITY WATER SYSTEM e EVALUATE STANDBY LIQUID CONTROL SYSTEM DESIG 1 O O 4
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. o UN! TED STATES -
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' 8 g NUCLEAR REGULATORY COMMISSION g r WASHINGTON, D. C. 20655
%,..... September 19, 1986 -
NOTE FOR: R. Bernero, Director, Division af BWR Licensing l W. Houston, Deputy Director, Division of BWR Licensing G. Lainas, Assistant Director for BWR, Division of BWR Licensing
-D. Muller, Project Director, BWR Project Directorate #2 W. Hodges, Chief, BWR Reactor Systems Branch ulman, Chief, BWR Plant Systems Branch Project Manager, BWR Project Directorate #2 FROM: Mohan C. Thadani, Project Manager BWR Project Directorate #2 l Division of BWR Licensing
SUBJECT:
VERMONT YANKEE CONTAINMENT SAFETY STUDY REVIEW this note is to confirm my understanding of the activities and responsible individuals for review of the subject study. Arrange for a site visit, on September 29 and 30, 1986, to Vermont Yankee. Site visit participants will be Vern Rooney, Gus Lainas, derry Hulman, Wayne Hodges, and Mohan Thadani. Vern Rooney is responsbile for coordinating the site visit with the licensee. By October 10, 1986, prepare a draft of a letter to Vermont Yankoe which will enclose the staff's preliminary conclusion on the study and request additional information. Responsible individual is Mohan Thadani. By October 17, 1986, prepare (1) sumary of NRC evaluation of the subject study and (2) a set of questions requesting additional information.
- Responsible individuals are Jerry Hulman and Wayne Hodgcs.
By October 23, 1986, finalize the letter (and the two enclosures) to Vermont Yankee under R. Bernero's signature. Reponsible individuals for assuring timely dispatch of the letter are Mohan Thadani and Vernon Rooney. Arrange a meeting in Bethesda, Maryland with Vennont Yankee on November 13, 1986, to get response to NRC questions. Responsiblein theindividual afternoon,is Vernon Rooney. i On December 24, 1986 write a letter to Vermont Yankee under R. Bernero's signature giving tentative final review based on draft
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1 i e 2 8 Generic Letter. Responsibiccoordina.torsare'MohanThadanian5 ! Vernon Rooney. - Mohan C. Thadani, Project Manager BWR Project Directorate #2 Division of BWR Licensing O B
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o CONCEDRED h l STETISTS ,i s,..4. .m s. m . 6...... nc enm g u a. . ,o l 5 7-Septecher 15, 1956 b p Y The Hcr.orable Madeleire M. Kunin d Covernvr of Vermont rentpclict, Vercent 05S02 Dect Governor Kun!n: U: hr.e htd the opocrtunity to take a prelicir.ary review of Vercent Yankee's se-cr lied Centainnent Saf ety Study and to att2nd the testing en September 11, IMO, s t thich Yankee Atcolc presented its study to the NRC. We cocoend you fcr dscid!nc to get an independent review of this work. f or the state. Ve would alec like to offer cur thoughts for your censideraticn and to= place Yar.kee's paesentatien in proper centext. Yankee's report is a failing effort to ccme up with some alternative, lower, probability figure than NRC's estimate that 90'4 of severe accidents will result in cont !nsent failure for plants such as Vermont Yankee. The heart and starting peint of the report is the ASC's Reactor Saf ty Study, c.lled VASH-1400, dating. frc the mid-1970's. When the AEC first produced the report, it was trucpeted cc "preving" that nuclear plants are safe; you may remember the widely publiciced clairs elllion. at the time that the chance of a sericus nuclear accident was ene in a Fu.r years later, after belated peer review, the NRC issued a pclicy statement repudiating the validity of VASH-1400 as a measure of overall nuclear plant rafety. (We enclose a copy of the Union of Concerned Scientists' 1977 review of WASH-1400, Tha Risks of Nuclear Power Resetors, which played a significant role in the original pest review cf that document.) The NRO's 1979 policy statement discwning this application of VASH-1400 stated, inter alia: "In particular, in light of the Review Group conclusions cri accident prebabilitics, the Cocclesion dces not regard as reliable the Reactor Sefety Study's numerical estimrite of the overall risk of reactor accident." NRC Statement on Risk Assessoent and the Reactor Safety Stucy Report (UASH-1400) in 1.ight of the Risk Assesscent Review Group Report, January 18, 1979, p. 3. In brief, the thousands of arbitrary assumptions, dozens of unverified computer codes, and its inherent incompleteness (e.g., failure to consider whole c!rsses of accidents, such as those initiated by sabotage or external events) made WASH-1400's risk ectimates subject to such overwhelming uncertainty as to make them unusable for predicting the risk of serious accidents, according even to the NRC. Be:ause the Vermont Yunkee study takes WASH-1400 as its fundamental base, the Yankee study is subject.to precisely the same problems. Cambridge Office: 2o Church Street Cambridge, Maaachusetts 02238 - (617) 547 5552 ___--____, Eru " F M E E 5' F '" N N N N U ---
, . , " - - .. .; u _ . ,
o . w,,. f in fact, "ankee has taken a fur.dsmentally deficient study (UASH-1p00) a,n d , , vit5;ur cerrtetint anv of its flawe. compounded t,he uncertainties bytyiving l Vercent Yankee essentially arbitrary degrees of " credit" for differences betwesn thc Vereent Yankee plant and the Peach Bottom plant used in the study. No plant- - specific asnessment was ever done for Vermon,t Yankee. Perhaps most remarkably, l l Yankrc's st0dy'oakes no effort to dettroine the uncertainty bounds associated
- s. with its probability estimate, a failure which it attributes to the " limited time available.* Considering that WASH-1400's overall risk estimates were repudiated ;
by the NRC precisely because cf their unknown yet enormous uncertainties, Yankee'r presentation a single probability nucber for containment failure is unjustifiable and, in our viev, disingenuous, hention should also be made of Yankee's lengthy listing of " enhancements" to g plant design or operation made over the past decade. In the vast cajority of a these cases, rost of which Yankee calls " miner" (Yankee report, p. 24), the change wr,s necessary in order to bring the plant into linc with the rules or commitments which the AEC assumed to be met at the time of its initial licensing. [ That is, a modification was needed to achieve the original level of safety { 4 icplicit in the license (and, ironically, VASH-1400), not to go beyond. This
- port-licensing history is quite typical. As operating experience, tests, and ! research have disclosed safety problems, plants have had to make changes to hardware and precedures. Another irony, of course, is that the farther a plant actually was from the level of safety assuced at its licensing, the longer a list such as that provided by Yankee will be.
{ Even as to t here changes which Yankee asserts to constitute major improvements in safety, there is much reason to be skeptical of such claims. An illustrative exarple is Yankee's claim that it "has incorporated an electrical ecmpenent
' environmental qualification program." Yankee Report, p. 5. This issue, # equipment qualification, involves determining whether electrical and mechanical safety equipment is able to survive the harsh conditions (high temperature, pressure, radiation, etc.) created by various accidents and perform the functions necessary to protect the public (shutting the reactor down, cooling the core tr3 containment, and monitoring and mitigating the accident consequences).
However, the degree to which safety has been improved by equipment qualification is unknown. NRC is currently reviewing the utilities' equipment 3 qualificatfor. claims "using standards and criteria which are based on engineering judgement and have in many instances not been thoroughly validated." NRC has also observed that " utilities . . . have essentially no significant ongoing research efforts in equipment qualification to resolve questions and confirm the methods Iused for equipment qualification) . . . ." Impacts of Budget Cuts on NRC's Atility to Assure Safety (Overview), transmitted by V. Stello Executive Director of Operations, to Samuel J. Chilk, Secretary, April 30, 1986, p. 19, emphasis in original. Therefore, the actual degree of safety improvement at Vermont Yankee is open to serious question. Perhaps the issue of Yankee's versus NRC's estimate of the probability of containment failure can be put into perspective by understanding that, within the uncertainty inherent in the probabilistic risk assessment methodology, there is really not e very significant difference between a 7% and a 90% containment failure estimate. As NRC asserted in its September 11, 1986, meeting with the boiling water reactor owners' group, its current operttive assumption for General c ._.m__ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _u.- a a _~
} ...a...~. . w. . - . . . . , - : _....- . ... ._. . . , l . .. . F. . . . . ' f i I* L Electric Mark I plants is that "i out of 2 core melts gives fa3 fairly large I releasc." That is. within our ability to predict, it is just as likely as not ' f thr t a severe accident would result in less of containment integrityiYo'r Tctrar.t Yankec. There are many possible scenarios for such a releare and few would l dicpute that the consequences could be greater than those at Chernoby.. In ' our
' view, such oddr.are plainly unacceptable.
I Furthermore, the situation today is not greatly improved over that in 1975 l w regarding the state of scientific knowledge of the phenomena occurring during severe accidents which would challenge containment integrity. For exacple, in a Dececher 1955, the NRC's Advisory Coosittee on Reactor Safeguards told the NEC tL . the computer codes used to analyze severe accidents, including continment l y perforennce, "should not be given much weight in making decision." We believe this is sound advice. 1 ollowing the ACRS's review, the NRC staff told the Commission that: Larre uncertainties exist in analyzine in-vessel b se vere accident behavior at core ciel t condit f or.c it.cluding clad oxidation heating, hydrogen !
! generation, core celt progression, and fission product and aerosci generation and transport.
**s These processes are currently modeled . . . by NRC codes ,_ . which are not verified against experimental data. ~These codes contain many_
arbitrarv assumptions and latre uncertafrties. Furthermore, alrost no data exist at the hieber temperatures near fuel meltine (3.100K) nor on the core-eelt procrescien process itself; so the analycis is of necessity cuite uncertain. , suera. p. 3, 1.. t pact of Budget Cuts on NRC's Ability to Assure Safety (Overview), emphasis added. In closing, we woul6 only add that UC5's experience with probabilistic risk assessment has convinced us that, in the manner used by Yankee, the technique is Basically, core a public relations effort than a credible scientific endeavor. it may be the analyses are manipulable to support predetermined conclusions, recalled that. prior to the Challenger disaster, the same technique was " sed All to claim that the probability of solid rocket booster failure was 1 in 60 s0. too often probabilistic risk assessments have been employed to stave off meaningful safety improvement. It is our hope that Chernobyl will begir to turn
}
that around, at least insofar as plants of the Vermont Yankee design are concerned. UCS would be happy to work with you toward that end. ll Very truly yours, , '
/
, h / '
i , i
'/
Ellyn R. Weiss Robert D. Pollard General Counse! Nuclear Safety Engineer 1
- - - - - - _ _ _ - - - - _ _ - - _ _ _ _ - - - - _ _ . - -_ _ . U
...-...~.--..----.-.....:-w . - - -. - - -
September 9, 1986 Docket No. 50-271 LICENSEE: Vermont Yankee Nuclear Power Corporation - FACILITY: Vermont Yankee Nuclear Power Station
SUBJECT:
AUGUST 6, 1986 MEETING WITH THE VERMONT YANKEE NUCLEAR POWER CORPORATION (VYNPC) Re: Containment safety study status On August 6, 1986, a meeting was held at the NRC headquarters in Bethesda, Maryland to discuss the status of the Vermont Yankee Mark I containment safety study. Enclosure 1 is a list of individuals that attended the meeting. i Enclosure 2 is a handout of the slides presented by VPNPC at the meeting. VYNPC reported that the Vermont Yankee plant had been compared to Mark I containment design assumed for generic studies with respect to: hydrogen control, drywell sprays, containment pressure control, core debris control, and training and procedures. These comparisons were almost complete, and at the time of the meeting, inputs by consultants were being condensed and integrated into a single report. The report should be complete at the beginning of Septer.ber, as scheduled. O.Tft: tic:vMby Vernon L. Rooney, Project Manager BWR Project Directorate #2 , Division of BWR Licensing
Enclosures:
As stated cc w/ enclosures: See next page BWR#2 VRooney
]/fi/86
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1; l l L Mr. R. W. Capstick E Vemont Yankee Nuclear Power Corporation-l, Vemont Yankee Nuclear Power Station . cc: ' Mr. J. G. Weigand W. P. Murphy, Vice President & President & Chief Executive Officer Manager of Operations Ve.rmont Yankee Nu'elear Power Corp. Vermont Yankee Nuclear Power Corp. R. D. 5. Fox 369 R. D. 5, Box 169 Ferry Road Ferry Road Brattleboro, Vermont 05301 Brattleboro, Yemont 05301 Mr. Donald Hunter, Vice President Mr. Gerald Tarrant . Commissioner Vermont Yankee Nuclear Power Corp. Vermont Department of Public, Service
- 1671 Worcester Road 120 State Street Framingham, Massachusetts 01701 _ Montpelier, Vermont 05602 -
New England Coalition en Nuclear Pollution Hill and Dale Fam Public Service Board R. D. 2. Box 223 State of Vermont Putney, Vermcnt 05346 120 State Street Montpelier, Vermont. 05602 Mr. Walter Zaluzny Chairman, Board of Selectman Vermont Yankee Decommissioning Post Office Box 116 Alliance
' Vernen Vemont 05345 Box 53 Montpelier, Vermont 05602-0053 J. P. Pelletier, Plant Meneger Ve mont Yankee Nuclear Power Corp.
Post Office-Box 157 Resident Inspector Vernon, Yemont 05354 U. S. Nuclear Regulatory Comission Post Office Box 176 Raymond N. McCandless Vernon, Yemnnt 05354 Vemont Division of Occupatient1 -
& Radiological. Health Yement Public Interest Administration Building Research Group, Inc.
10 Baldwin Street 43 State Street Montpelier. Vemont 05602 Montpelier, Vermont 05602 Honorable John J. Easton Thomas A. Murley Attorney General Regional Administrator State of Vermont Region I Office 109 State Street U. S. Nuclear Regulatory Comission Montpelier, Veraent 05602 631 Park Avenue King of Prussia, Pennsylvania 19406 John A. P.itscher Esquire Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110 "WW'P *iNW_ Titty WRW trtPMtKMthM 3M,' tNrN. M - 2 ffNYMJ!pt Myr eTyy;rzyte tw* fey C s****'?_ _qwN 31*Wy,p.gge ,RMAMpgggA
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Enclosure 1 LIST OF MEETING ATTENDEES
' ^
August 6, 1986 Name Organization H. Abelson NRC/ DBL /PDf? D. Muller NRC/ DBL /PD#2 J. Thayer Yankee Atomic Electric Co. R. Lodwick VYNPC P'W. Murphy VYNPC S. Shultz Yankee Atomic Electric Co. R. Auluck KRC/ DBL /PD#1 P. Paull State of Vermont G. Tarrant State of Vermont S. Murphy NIRS D. Shum NRC/ DBL / FOB T. Elasser NRC/RI P. Leech NRC/ DBL /PD#1 M. Thadani NRC/ DBL /PD#2 R. Bachmann NRC/0GC J. Gray, Jr. New York Power Authority J. Kopeck NRC/PA J. Coates Vermont Public Interest Research Group D. Hew Harmon & Weiss R. Houston NRC/ DBL M. Hodges NRC/ DBL /RSS G. Lainas NRC/ DBL /BWAD r'R. Bernero NRC/ DBL
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_ u ._c _ _ . . . _ . . . ._ .w,z..w.us.).! o - Enclosure.2. YERMONT YANKEE CONTAINMENT SAFETY STUDY . -
- 1. PURPOSE AND OBJECTIVE OF EACH TASK INCLUDING PRELIMIN RY RESULTS MARK 1 DESIGN REVIEW o WASH 1200 1 DESIGN COMPARISON o KEY DESIGN DIFFERENCES o VERMONT YANKEE DESIGN
SUMMARY
VERMONT YANKEE CONTAINMENT CAPABILITY o VERMONT YANKEE APPROACH o STUDIES AND RESOURCES USED i o ACCIDENTS AND TRANSIENTS STUDIED o RESULTS UTILIZING KEY DIFFERENCES AT VERMONT YANKEE o DEFINITION OF "90%" o VY CONTAINMENT CONDITIONAL FAILURE PROBABILITY CURRENT MARK I ISSUES i o DEFINITION AND TECHNICAL STATUS OF 5 ISSUES o APPLICABILITY AND NEED FOR FURTHER STUDY o ACTIVE INDUSTRY AND NRC EFFORTS
- 1. STATUS
- RESULTS TO DATE ,
l- - SCHEDULE FOR COMPLETION l l 1 l e n -n - -nn - -,,n ~-n,nn - ,,=-- -n- ~n- - - w
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. DESIGN AND OPERATIONAL FEATURES COMPARISON Design / Operational Data Versnont Yankee (WASH-1400) ,
General Plant Data - Plant Type CE BWR GE BWR Containment Type Mark 1 Pressure Suppression Mark I Pressure Suppression Rated Thermal Power, NWt 1.593 3.293 Rated Core Flow, Ib/hr 48.0 x 10E6 102.5 x 10E6 Rated Steam Flow, Ib/hr 6.43 x 10E6 13.381 x 10E6 Reactor Data Ins,ide Height, ft-in 63 - 1.5 72 - 11 Inside Diameter, in. 205 251 Containment Data Internal Design Pressure, psig 56 56 grywell Date Cylinder Diameter, ft. 33 38.5 ft. 62 67 t'pherical Diameter,3 Free Air Volume, ft 134.000 159.000 To:vs Data Major Diameter, ft. 98 111.5 Minor Diamater, ft. 27.66 31 Water Volume, stin/ max., ft3 68,000/70,000 123,000 Free Air Volume, ft3 114,200/112,200 119,000 Vent Pipes Data Number . 8 8 Internal Diameter, ft. 6.75 6.75 ggyr40mer Pipe Data Number 96 96 Internal Diameter, ft. 2 2 j Submergence, ft. (nominal) 4 4 l Number of WW-DW Vacuum IG 12 Breakers ) I I 1 l
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. Desian/ Operational Date Vermont Yankee (WASH-1400)
- i. x Secondary Containment Data r
. s .
Free Air Volume- 2,120,000 2,400,000 Mitimation Systems Desien Data MPCI System Number of Trains or Subsystems 1 1 Number of Pumps / Train 1 1-Design Flow / Train 4,250 spm at 1,120 to 150 psid 5,000 spm at 1,100 to 150 psis Electrical Power DC only (turbine-driven) DC only (turbine-driven) RCIC System i Number of Trains er ii,ubsystems 1 1 Number of Pumps / Train 1 1 Design Flow / Train 400 600 gym Electric.a1 Power DC only (turbine-driven) DC only (turbine-driven)- EMB Svstum
~
Number of Trains or Subsystems 2 2 Number of Pumps / Train 2 2 Number of HKs/ Train 1 2 Design Flow / Train 14,400 spm 20,000 spm Electrical Power Emergency ac and de Emergency ac and dc ; Source of Water Torus Torus ' service water (riv W service wate Ultimate backup (Diesel fire purap )- cross-tie capatility) Emergency Diesel Generator , Systems Number of Emergency Buses 2 4 per unit Numbar of Emergen:y Diesels T* 4* - Shared between two units
- Single-Unit Site - Requires *Two-Unit Site - Requires two one out of two amargency out of four emergency diesels diesels for safe shutdown. for safe shutdown.
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VERMONT YANKEE - WASH 1400 KEY DESIGN DIFFERENCES J
, RATIO DESIGN PARAMETER VERMONT YANKEE / WASH 1400 RATED THERMAL POWER .48 DRYWELL VOLUME / POWER 1.75 TORUS WATER VOLUME / POWER 1.19
-TORUS AIR VOLUME / POWER 1.96 HPCI PUMPING CAPACITY / POWER 1.77 RCIC PUMPING CAPAC11Y/ POWER 1.40
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SUMMARY
;% l Q l l 'o SM,ALL REACTOR PLANT (NSSS) IH LARGE MARK I CONTAINMENT a
ll l' o ENGINEERED SAFETY FEATURES CAPACITY e
- o, .i
+, RESIDUAL HEAT REMOVAL CAPABILITY i 1
I o ELECTRIC DRIVEN MAIN FEED PUMPS o DC SYETEM CAPACITY a DIVERSITY-
- 8 HOUR BATTERY RATING !
- APPENDIX R BATTERIES .
- SPECIAL PURPOSE BATTERIES ~ !
i
- UNINTERRUPTIBLE POWER SUPPLIES !
o AC SYSTEM RELIABILITY & DIVERSITY !
- VERNON HYDR 0 TIE LINE I
- NORTHEAST GRID RELIABILITY
- SEPARATE HIGH LINE RIGHT OF WAYS l
4 o DIESEL FIRE PUMP CROSSTIE TO RHR SYSTEM ! i l s l 4
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- f, l.1 STUDY APPROACH I
o IDENTIFY THE DOMINANT' ACCIDENT SEQUENCES WHICH CAN LEAD TO SEVERE-ACCIDENTS ,^,T VClMONT YANKEE.
.is o QUANTIFY THE DOMINANT' SEQUENCES USING A' REFERENCE MARK I MODIFIED FOR THE UNIQUE VERMONT YANKEE FEATURES..
i o DEVELOP THE CONTAINMENT EVENT TREE-TO DISPLAY THE PATHWAYS TO
' SAFE MITIGATION AND POTENTIAL RADIONUCLIDES RELEASE:10 THE.
ENVIRONMENT. o QUANTIFY THE CONTAINMENT EVENT TREE USING AVAILABLE ESTIMATES
'0F MITIGATION RELIABILITY AND STANDARD MODELS.
o CALCULATE THE CONDITIONAL FAILURE PROBABILITY OF CONTAINMENT. .I 1
s ,, QLu. ~ ~ . ,. ..x.. . .. a. * ~> w .<. a . : . . " .. , , 5,r . .u ., L. . .iL". i.;,;4:e %$Li A d:bc4 u 4" ., 'sCd%2sida iSbud'PLv L.. , STUDIES & RESOURCES STUDIES , l o WASH 1400 o IDCOR
- IPE METHODOLOGY
- TECHNICAL
SUMMARY
REPORT
- TASK REPORT DOCUMENTS o DRAFT BWR MARK 1 PSA's o SEVERE ACCIDENT SEQUENCE ANALYSIS PROGRAM i
RESOURCES o YANKEE ATOMIC ELECTRIC COMPANY o DELIAN CORPORATION o FAUSKE AND ASSOCIATES o RISK MANAGEMENT ASSOCIATES o GENERAL ELECTRIC COMPANY o PICKARD LOWE & GARRICK, INC. _ _9./L . A LTW TA9 "73. O EYd7 . . _ m
BWR DOMINANT ACCIDENT SEQUENCES CLASS 1 LOSS OF COOLANT MAKEUP IA HIGH PRESSURE IB STATION BLACK 0UT IC ATWS-ID LOW PRESSURE IE LOSS OF DC 1 1 I CLASS 11 LOSSOFCONTAINMENTHEATREM0hAL CLASS 111 LOCA
)
CLASS IV ATWS { l 1 1 I 1 i _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - - . _ _ _ _ - - . - - - - - - - . _ ___ )
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a4 VERMONT YANKEE RESULTS DOMINANT SEQUENCES
.STATIONBLACK0UT KEY DIFFERENCES:
o DIESEL FIRE PUMP CROSSTIE-o VERNON HYDR 0 TIE LINE o DC SYSTEM CAPABILITY
, o DIESEL GENERATOR RELIABILITY o NEW ENGLAND GRID STABILITY l
l l 235. KEY DIFFERENCES: i o CONTA,1NMENT SIZE VS. POWER l
! o EXISTING ATWS MODIFICATIONS AND COMMITMENTS o LARGE RHR SYSTEM HEAT REMOVAL CAPACITY l
i o MSIV REOPEN CAPAB!LITY o EMERGENCY OPERATING PROCEDURES 1 _._..m__- . . _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _
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DEFINITION OF "90%"_ WASH-1400 ASSUMPTIONS' L o ALL SEQUENCES (MELT /NON-MELT) ARE ASSOCIATED WITH CONTAIN!'.ENT FAILURE o CORE MELT RELEASE CATEGORIES
- 1. STEAM EXPLOSION
- 2. OVERPRESSURE FAILURE DIRECT TO ATMOSPHERE
- 3. OVERPRESSURE FAILUEE THROUGH REACTOR BUILDING 11 . ISOLATION FAILURE o RELEASE CATE50 RIES 2 & 3 DOMINATE RISK o SEQUENCES WHICH CONTRIBUTE COMPRISE 90%
ATWS LOSS OF RESIDUAL CORE HEAT REMOVAL CURRENT STUDY E o DIFFERENI DOMINANT SEQUENCES IDENTIFIED o KNOWLEDGE OF CONTAINMENT FAILURE PHENOMENA AND CRITERIA HAS EVOLVED _ _ _ _ _ _ _ _ _ _ b
F-'- en
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( . c l ' L l .: CONTAINMENT CONDITIONAL FAILURE PROBABILITY l L o CALCULATE THE CONTRIBUTION OF EACH ACCIDENI CLASS 10 10TAL LIKELlH00D OF CORE MELT o IEIERMINE THE CONDITION /L FAILURE PROBABILITY OF CONTAINMENT FOR EACH ACCIDENT CLASS o CALCULATE THE WEIGHTED AVERAGE OF PRIMARY CONTAINMENT FAILURE PROBABILI 1' 6 1
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,1 CONDITIONAL CONTAINMENT FAILURE PROBABILITY
' I QUANTIFICATION DOMINANT FRACTIONAL CONTAINMENT CONTRIBUTION ACCIDENT CONTRIBUTION FAILURE TO WEIGHTED SEQUENCE (CORE MELT) PROBABILITY AVERAGE LOSS OF COOLANT MAKEUP A A AxA LOSS 03 CONTAINMENT HEAT REMOVAL s B sxB LOCA -c C cxC ATWS o D oxD - - .
n 1.0 ) = CFFP E, ( , O WHERE CCFP = CONDITIONAL CONTAINMENT FAILURE PROBABILITY l. l 1 l I. l l t -- -- -- - - - - - - - -- - - - - - - _ - - - - - - - - _ -- - -- - - - - --_ - - ------ - --
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CURRENT TECHNICAL ISSUES . . .
- 1. ' HYDR 0 GEN CONTROL 2.. DRYWELL SPRAYS
- 3. CONTAINMENT PRESSURE CONTROL
- 4. CORE DEBRIS CONTROL' !
- 5. TRAINING AND PROCEDURES 9
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n HYDROGEN CONTROL OBJECTIVEt PREVENT HYDR 0SEN COMBUSTION CAUSED FAILURE SUGGESTIONS: 0XYGEN CONTROL INERT TO START CONTROL INGRESS OF OXYGEN ISSUES IDENT!FIED BY NRC: WHEN AND HOW LONG NOT INERfED PRESENT VY CAPABILITY RELATIVF TO PROPOSED REQUIREMENTS
- 1. CONTAINMENTINERTED
- 2. ELECTRIC POWER NOT REQUIRED TO MAINTAIN INERT
- 3. TECH. SPECS. CONTROL DEINERT TIME .
- 4. PLANT SHUTDOWN IF TECH. SPEC. CANNOT BE MET e
. _____.___m- _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
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DRYWELL SPRAYS { OBJECTIVE: SPRAY WATER T0:
, 1. LOWER PRESSURE
- 2. COOL VULNERABLE EQUIPMENT
- 3. QUENCH DEBRIS
- 4. SCRUB AEROSOLS SUGGESTIONS:
- 1. SPRAY IN DRYWELL
- 2. INDEPENDENT BACKUP WATER SOURCES AND PUMPS H0SE CONNECTIONS USE OF FIREMAINS ISSUESIDENTIFIEDBYNRC:
- 1. RISK 0F IMPLOSION
'2. RISK OF HYDR 0 GEN COMBUSTION AFTER STEAM CONDENSATION
- 3. MANUAL' ACTIONS AND TIMING PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS: _
- 1. EXISTING FLOW PATH FROM DIESEL FIRE PUMPS TO VESSEL OR DRYWELL SPRAYS (2500 GPM)
- 2. AVAILABLE FOLLOWING STATION BLACK 0UT
- 3. IF EMERG DIESEL AVAILABLE, SYSTEM OPERABLE FROM OUTSIDE REACTOR BUILDING j
- 4. EXISTING OPERATING PROCEDURE DESCRIBES USE OF SYSTEM ,
a.,=l..L.. c c - . 2.x.x. . . , a:.. u.a;..:.:. .c.an . ... o , . m ....... ..c.... a .w..:. .~ ... .a - . a u :. 1 CONTAINMENT - PRESSURE CONTROL OBJECTIVES: 1. AVERT UNCONTROLLED OVERPRESSURE FAILURE 2.- CONTROL RELEASE PATH (SCRUBBING) SUGGESTIONS: <
~
- 1. SUBSTANTIAL CAPABILITY 10 VENT WETWELL
- 2. REMOTE / RELIABLE CONTROL OF VENT VALVE
- 3. ABILITY TO RECLOSE VENT ISSUES IDENTIFIED BY NRC:
- 1. DELIBERATE RELEASE OF RADI0 ACTIVITY
- 2. WHAT CONSTITUTES REMOTE / RELIABLE CONTROL?
- 3. IS VENTING TO SECONDARY J. CONTAINMENT ACCEPTAELE?
- 4. WHAT IS APPROPRIATE ACTI0N PRESSURE?
PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS: .
- 1. EIGHTEEN INCH ATMOSPHERIC CONTROL SYSTEM VENTS
- 2. THREE INCH ATMOSPHERIC CONTROL SYSTE.M YENT .
- 3. TWENTY AND EIGHTEEN INCH NITROGEN NRCE LINES i
)
- 4. SIX INCH NITR0 GEN PURGE LINES
- 5. ONE INCH NITROGEN CAD LINE - EXISTING PROCEDURE
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i CORE DEBRIS CONTROL
~ ~ -
OBJECTIVE: REDUCE LIKEllH00D OF FAILURE BY DIRECT CONTACT OF CORE DEBRIS WITH DRYWELL WALL. SUGGESTIONS:
- 1. USE PRACTICAL DEBRIS RETARDING BARRIERS
- 2. CONSERVE SUPPRESSION P00L WATER AS A QUENCHING P0OL ISSUE IDENTIFIED BY NRC: WHAT IS PRACTICAL?
PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS 1
- 1. SMALL CORE DEBRIS VOLUME COMPARED TO PREVIOUS STUDIES
- 2. DRYWELL SUMPS COMBINED VOLUME APPR0XIMATELY'200 FT3
- 3. >1200 FT DRYWELL FLOOR SURFACE AREA
- 4. DOWNCOMERS APPR0XIMATELY ONE FOOT AB0VE DRYWELL FLOOR i
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- s. . . .
OBJECTIVE: ENSURE OPERATORS ARE READY TO USE PLANT FEATURES TO BEST ADVANTAGE IN SEVERE ACCIDENTS SUGGESTIONS:
- 1. CLEAR SYMPT 0M BASED STRATEGIES (IllTEGRATED) ;
.2. REMOVAL OF UNNECESSARY INHIBITIONS
- 3. TRAINING ISSUES IDENTIFIED BY NRC:
- 1. COMPETING SAFETY REQUIREMENTS
- 2. DEGREE OF TRAINING PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:-
- 1. REV. 3 0F EMERGENCY PROCEDURE GUIDELINES (EPG's) IMPLEMENTED
- 2. ACTIVE PARTICIPATION IN REV. 4 DEVELOPMENT
- 3. VY PLANT SPECIFIC SIMULATOR COMPLETED IN 1985
- 4. EXTENSIVE OPERATOR TRAINING PROGRAM INCLUDES SEVERE ACCIDENT. PREVENTION AND MITIGATION I
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. 1 1
Review of '
- Varmont Yankee Containment Safety Study by
, Norman C. Rasmussen 4 Professor of Nuclear Engineering 4 Massachusetts Institute of Technology l Septenber 1,1986
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GENERAL COMMENT
S - . .
- 1. Extent of Review __,
The review consisted of reading the entire report in deta;.1 to develop p an understanding of the general approach taken. No attempt has been made tocheckanycalculationsorthecorrectnessofdatafromanyreferences.j _ Any place a statement raises a question it has been noted. this does not mean it is wrong'but it is something to be verified because it surprised me.
- 2. Comments on the Approach The general approach was to compare the Vermont Yankee Plant to the Peac$BottomplantusedinWASH1400toidentifymajordifferencesthat might effect the liklihood of significant accidental releases releases of radioactivity. The impact of these differences on either the probability of core damaging accidents or the containment performance was then evaluated. This information was used to estimate differences in the
. probability of large releases of radioactivity. -g Although it would ,have been better to have done a full scope PRA, this would have taken at least a year. Therefore, given the time constraints i the approach taken seems a good compromise and I feel the rather conservative way that it was carried out gives results that I believe somewhat overstate the probability of serious releases at Vermont Yankee.
It seems an honest effort has been made to look fer weaknesses in Vermont
. Yankee relative to the Peach Bottom plant. 3 There appear to be three particularly favorable points for Vermont Yankee in the comparison. They are: (a) the Vernon Hydro Station offers a source of off site AC independent of the grid, (b) the diesel driven fire
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emergency, 'and (c) the containment vclumt relative to the power level'is much bigger in Vermont Yankee than in Peach Bottom. The point is well made that these factors should 'all -improve Vermont Yankee performance under severe accident conditions over the Peach Bottom plant. These benefits are partially offset by some advantages that the' Peach Bottom plant has over the Vermont Yankee plant. I felt one important advantage that Vermont Yankee has over Peach ; Bottom was not mentioned. The risk curves for WASH 1400 were based on the failure probability estimates for Peach Bottom, but for fission product inventory a one thousand MWe (approximately 3500 MWth)' plant. Thus, the initial inventory in Vermont Yankee is less than one-half the WASH 1400 release even if everything else remains the same. This significant difference is not mentioned in the report and I think.it should be. The approach in estimating containment performance in Chapter 4 seems sound. I believe the containment event tree (CET) appr:ach used is consistent with current practice in PRA's. The approach seemed tnorough and somewhat conservative in scme of the assumptions. The insights 61atussed in Secticn 4.'5 seemed well supported by the analysis. I felt the discussion of venting and its possible problems was well done. My first. reaction has been that venting was a good idea but this analysis shows that there can be serious limitations to its usefulness. L I found Chapter 5 to be a good summary of the current status of.the plant. The suggestions for improvements or further analysis seemed to be reasonable and based on the previous analysis. The analysis of the NRC concerns relative to Vermont Yankee was well done. l _ _ _ - - _ _ _ _ _ _ - - - _ - _ - - - - _ - _ _ _ _ - - - - - _ - _ - - - - - _ - _ . - - - - -- _ - - _ _ _ _ _ _ - - - - - 1
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L I would complement the authors on the thoroughness's of the ' analysis and . in covering the NCR concerns in'a professional way. The willingness to identify weaknesses or possible weaknesses in the Vermont Yankee design and procedures in a frank and open way was excellent. I certainly got the
. impression that nothing was being covered up. I can find no major criticisms of the report and its approach to the problem.
4 4
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I SPECIFIC COMMENTS - I CHAPTER 3 page 3-5: Summation signs missing from the equation. page 3-6: Last sentence of first paragraph seems incomplete. ' page 3-7: Last sentence in the paragraph under 3, I think should say "This could substantially reduce some of the consequence of the event". CHAPTER 4 page 4-14: I don't see where the equation U = H + 0 + V + C ever gets ! i used. I don't find values for H, 0, V and C, and so I don't understand why it's there. page 4-15: Third line from the bottom, "Therefore HPCI was judged to be effective in the initial phase", "Was" seems to be missing. page 4-18: Last sentence, first paragraph. Last phrase is " Reactor wate.r level referenced by flashing"--something seems to be missing. page 4-29: The molten core on the floor would be a very strong radiative heat source. Is it possible that radiation heat transfer to the walls i could damage them by thermal stresses or overheating, without direct contact of the molten material with the steel? page 4-29 also: Second line from the bottom, "Not to be proceduralized" seems like bad grammar to me. page 4-31: Under " Reactor Building Effectiveness". I'm always worried when filters are said to be capable of capturing large amounts of radioactivity. If the amounts are large there'll be a significant heat source in the filter and one must be sure that the filter is cooled enough to prevent it from burning.
T X T .- _ . . _ . - . _ . - . . . . - . . . - - . . .. . . . . . . . ... e page 4-34: Fourth line from the top: " Assumptions withhold the largest", should read "...which hold the largest", I believe. page 4-37: End of last paragr6ph says, " Priorities should be given to dry well sprays." No further comments are made about dry well sprays here and leaves the reader puzzled. I think you should mention that discussion of these sprays will be made in the next section. page 4-38: "He" in the first line should be "The". page 4-46: Definition of D should be buses not bases. page 4-60: Second line from the bottom in the footnote: Bases should be buses. page 4-61: A footnote explaining EH, EM, EL would be useful. It's hard to remember what these letters stood for when you first read the table. . l CHAPTER 5 i page 5-2: Refers to figure 5.1 but it's not clear to me what percentage metal-water reaction is assumed to generate the curves in figure 5.1. page 5-4: Number 4. It's not clear to me how dry well sprays limits excessive negative pressure from developing. page 5-6: Fourth line from the top: " Vacuum breakers operate as designed". If they do, do they permit oxygen to come in and create a hydrogen combustion problem. page 5-6: Next to last paragraph says, " Sprays are designed as a safety grade system." I don't believe this was true on Peach Bottom. CHAPTER 6 page 6-1: Second line from the bottom: think "all" should be inserted to say " applicable to all the BWR Mark I containments". x . _ - _ - _ - _ _ _ _ _ _ _ _ - . .. - - . . . . - . _ . y}}