ML20212C429

Summary of 861117 Meeting W/Util,Delian Corp,State of VT Public Svc Dept & Nuclear Info Resource Svc in Bethesda,Md Re Util Response to NRC Questions on Containment Safety Study.List of Attendees & Util Draft Response Encl

ML20212C429
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	12/23/1986
From:	Rooney V Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8612310002
Download: ML20212C429 (88)

Text

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f.i December 23, 1986 Docket No. 50-271 LICENSEE: Vermont Yankee Nuclear Power Corporation FACILITY: Vermont Yankee Nuclear Power Station

SUBJECT:

NOVEMBER 17, 1986 MEETING WITH THE VERMONT YANKEE NUCLEAR POWER CORPORATION (VYNPC)

RE:

Containment Safety Study On November 17, 1986, a meeting was held at the NRC headquarters in Bethesda, Maryland to discuss VYNPC'S response to the NRC questions pertaining to the Vermont Yankee Mark I Containment Safety Study (VYCSS). The 38 questions had been transmitted to VYNPC together with a pre.liminary assessment of the VYCSS by NRC letter dated October 24, 1986. is a list of individuals that attended the meeting. is the draft VYNPC response to the questions which was presented at the meeting.

The response to each question was discussed separately, and for certain questions clarification or amplification was requested. VYNPC agreed to revise the response accordingly, and submit formal answers soon after the meeting.

crid=1 ti;ad Ly Vernon L. Rooney, Project Manager BWR Project Directorate #2 Division of BWR Licensing

Enclosures:

As stated cc w/ enclosures:

See next page DISTRIBUTION

• Docket: File: :=

NRC PDR Local PDR PD#2 Reading VRooney ACRS(10)

RBernero DMuller 0GC-Bethesda EJordan BGrimes 8612310002 861223 DBL:PD#2 PDR ADOCK 05000271

.(

VRooney:cd P

PDR p,/4,) /8(p 0FFICIAL RECORD COPY

Mr. R. W. 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 Vern.ont Yankee Nuclear Power Corp.

Vermont Department of Public Service 1671 Worcester Road 120 State Street Framingham, Massachusetts 01701 Montpelier, Vermont 0560E 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 SeTectman 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 05602 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 John A. Ritscher Esquire Ropes & Gray 225 Franklin Street Boston, Massachusetts 02110

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ENCLOSURE 1 LIST OF MEETING ATTENDEES VERMONT YANKEE CONTAINMENT SAFETY STUDY 4

November 17, 1986 Name Organization R. Bernero NRR/ DBL V. Rooney

- NRR/ DBL /PD#2 4

W. Murphy Vermont Yankee R. J. Lodwick Vermont Yankee J. K. Thayer Yankee Atomic D. E. Yasi Yankee Atomic E. T. Burns Delian Corp.

J. R. Chapman Yankee Atomic-H. N. Jow Yankee Atomic Farouk Eltanila NRC/DSR0/ RIB D. B. Vassallo NRC/ DBL /F08 Jack Kudrick NRC/ DBL /PSB Phillip L. Paull Vermont Public Service Department Jerry Hulman NRC/ DBL /PSB Stephanie Murphy Nuclear Information &

Resource Service Wayne Hodges NRC/ DBL /RSB Eric Weiss NRC/IE/DEPER Jack Fulton Boston Edison Co.

Terry Pickens Northern States Power Earl Page Detroit Edison Gordon Bristol Vermont Yankee Robert G. Brown Delian Corp.

Jim VanHerrmann Delian Corp.

D. R. Muller NRR/ DBL /PD#2 A. C. Kadak Yankeee Atomic P. Davis VT. Consultant 4

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. r-ENCLOSURE 4 VERMONT YANKEE CONTAINMENT STUDY RESPONSE TO NRC

~ REQUEST FOR ADDITIONAL INFORMATION V

4 QUESTION 1 What is your. 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 estimate" 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 published BWR PRA.

Current on-going work by the NRC in the NUREG-1150 Program and other i

programs (e.g., MELCOR) may provide greater insights into this area.

The NRC statement from Page 3 that "the licensee's estimates appear optimistic considering the uncertainties..." is not appropriate. The

^

containment conditional failure probability point estimate was derived using a realist'ic "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 assumptions to provide a balanced assessment. The outcome of that assessment is presented in the Vermont Yankee Containment Safety Study as the "best estimate."

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y QUESTION 2 5

What is the effect on core damage frequsney when accident seque,nces TPUV, TPQUV, T PQX, T,PQW, T,yW M are included in the dominant E

accident sequences based on reduced battery life, the number and type of SRVs compared to Peach Bottom,'and on the CCFP7 (Note: The question 4

, originally inc1'uded T,QUV and not 7,PQUV; however, since the question appears to be directed at. stuck open valves, we believe it should have been T,PQUV.)

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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 dominant sequences in a plant-specific analysis. However, many of these considerations were examined in the Vermont Yankee evaluation. A discussion of each item follows.

i o

Type of SRVs 4

Both Vermont Yankee and Peach Bottom have Target Rock three-stage relief valves. This is identified on Page A-4 of the Containment Safety Study. Because the designs are the same no change was made to i

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-4 of the Containment Safety l

Study as follows:

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.....=-

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:

4 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 QUV y

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 Peach Bottom than at Vermont Yankee.

t 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 Vermont Yankee quantification.

o Sequence Considerations The IPE methodology treats similar sequences together. Specifically, the IPE methodology bins together accident sequences caused by an IORV 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 QUV in the Vermont y

Yankee study and appear in Class ID.

The T QUV sequences in the y

Vermont Yankee Study include:

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T QuV 3

T PQUV C

T PQUV g

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T PQUV T

T PQUV p

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 Station Blackout Sequences T QUX. T POUV E

E The BWR IPE methodology addresses these specific issues on Pages 4-42 through 4-44.

The conclusion from the BWR IPE is that sequences of this nature have frequencies in the range of 3E-7/yr. The 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 that reported in the Vermont Yankee Containment Study for the total core melt frequency and for Class IB.

This increase represents an increase of:

0.16% for total frequency 0.81% for Class IB This would not alter the qualitative or quantitative conclusions of the evaluation.

o Reduced Battery Life i

The battery life used for Vermont Yankee is judged to be equal to or greater than that used for Peach Bottom and not to be a factor in the assessment of any of the sequences identified by the NRC.

o Recovery of RHR Pumps (TPW)

TPW sequences are long-term loss of containment heat removal i

sequences. These sequences can be effectively mitigated by any I

containment heat removal pathway, i.e.'

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l' RHR Main condenser Containment venting It can be shown that whether en SORV exists or not would have little or no impact on 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.

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QUESTION 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 estimate that the frequency of loss of off-site power at Vermont Yankee is 0.07/yr.

i

RESPONSE

The Containment Safety Study conservatively estimated the frequency of loss of off-site power as 0.07 events / year based on Vermont Yankee's record of having no total loss of off-site power events in over 14 years of operation.

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 NRC'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.

The NRC states that industry average value for loss of off-site power is j

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 characteristics, 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 of f-site power." Thus, site-specific analysis and site experience provides a more accurate estimate of frequency of loss 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 frequency as calculated using NUREG-1032 and does not deviate significantly from industry experience of about 0.1 events / year. 5104R

t.

QUESTION 4 Verify that the total battery capacity available at Vermont Yankee is greater tha'n 2,175 ampere-hours, and that it could be maintained at a voltage greater than 1.75 volts / cell in high ambient temperature during the accident for six to eight hours.

9

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 1E Alternate i

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 1.75 volts / cell.

The am'bient temperature for the batteries is not expected to increase at all during the station blackout scenario; however, if the temperature did increase, the effect would be to increase battery capacity not reduce it.

T 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 required 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. 5104R

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QUESTION 5 How often are the RHR/RHRSW interconnecting valves actuated to assure that the valves work properly?

RESPONSE

The RHR/RHRSW interconnecting valvas are stroked monthly ("RHR Valve Operability Surveillance," OP-4124) to assure that the valves work properly.

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j' QUESTION 6 How often are the interconnecting valves between the RHRSW and the Fire Protection System (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 System and the RHRSW/RHR crosstie valves are normally open.

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QUF.STION 7 How readily can the MSIVs be reopened following closure at oper,ating

. 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?)?

RESPONSE

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:

1.

Rx Low-Low Water Level - 82.5 Inches 2.

Main Steam Line High Radiation - 3x Normal Full Power Background (NFPB) 4 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 Tannel High Temperature - 212 F 5.

Main Steam Line Low Pressure - <800 psi (Rx Mode Switch in RUN) 6.

Low Main Condenser Pressure - <12" HgA b.

Under emergency conditions, the operator is permitted by the Emergency Operating Procedures (EOPs) to bypass all valid isolation signals except 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.

i Conditions 5 and 6 may be cleared by the operator by placing the t

reactor mode switch out of RUN and by use of a keylock bypass switch on a Control '

5104R

Room back panel, respectively. All other isolation conditions can be bypassed by initiating lifted lead and jumper precedures (Appendix, OE-3100). The following five-step procedure must then be carried out:

1.

Place valve control switches (total of 11) in the CLOSED or SHUT position to satisfy the inadvertent opening / reset perm'issive 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.

5.

Open the inboard MSIVs (4) via control switches.

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QUESTION 8 i

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It is not clear! 'aow the CCFPs A ven on Page 74 of the report were i

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. i obtained. Please' explain.

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RESPONSE

- L The basis of conditional probaM11tv hf early containment failure (CI) for each class of accident is as 'Ic11ows :

e Accident Class, CI

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Failure mode was estimated tc.he-byarogen burn. The probability is based on 'a Shoreham PRA study.

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Failure probability was estimat!ed as the, fraction of accident sequedces in Class IC that +;ould' result in* elevated containment pressure-(140 psia) before reactor vescal failure.

i Given the large probability of dryvell failure by overpressure (see c.

containment event tree far Classes II and IV).

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o QUESTION 9 T,

What SLCS modifications are proposed for Vermont Yankee? Page 86 discusses the advantages of two different possible modifications, but gives no commitment to either.

3, 1;

RESPONSE

At the present time the preliminary design proposed for Vermont Yankee is to utilize two 43 gpm pumps for injection. However, a final design has not been approved 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 (10CFR50.62)." The letter states that Vermont Yankee will

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implement a design or operational modification of its Standby Liquid Control System during its second refueling outage after July 26, 1984 (summer 1987 outage).

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OUESTI N 10

. Identify tha testing'and stintenance requirements you use for the diesel-driven fire pump. Do these requirements conform to thos'e contained i v in'the National Fire Codes? Also, identify any reliability information I

for,the system such as outages and failures to start on demand. What JO houtagetime!1imitationsdoyouuse-forthesystemwhileatpower?

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RESPCNS2' i.

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The testing and maintenance requirements for the diesel-driven fire 4

pump are l'

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 alarm circuitry.

4.

DFP starting battery - once per cycle - cleaning and inspection.

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5.

DFP - A0 rounds (each shift) - battery charger and fuel oil level.

b.

Although the above testing and maintenance requirements do not conform 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.

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c.

The FP System is considered operable with:

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1. 'Two fire pumps operable and lined up to the fire suppression loop (Note: one. electric, one diesel-driven).

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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 sectionalizing control or isolation.

The following table indicates those periods when the diesel fire Pump has not been available.

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Data Duratien Reason 6/23/80 6.0 hrs Preventative Maintenance (PMs) 5/14/80 2.0 Calibration 10/06/81 1.5 Check operation of generator 4/21/82 8.0' PMs 5/21/82 7.0 Battery chhrger 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 6/28/85 6.5 Replace battery terminals

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8/09/85 1.0 DFP battery 10/29/85 2.0 Replace exhaust muffler 2/10/86 2.0

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6/11/86 1.5

(?)

TOTAL DURATION:

71.5 hrs NOTES 1.

This information pertains to the period 6/28/80 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. 5104R t

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The following information indicates the reliability of the diesel fire pump:

A.

FAILURE TO START Date Reason 8/10/76 Starting solenoid relay 9/30/85 Poor battery connections B.

STARTING BATTERIES DEGRADED Date Reason 5/13/80 Bad cells on one battery; battery charger timer

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9/30/85 Bad connection to one of two starting batteries 1/07/86 Low voltage both batteries; blown battery charger fuse d.

With the Fire Suppression Water Supply System inoperable, a backup Fire Suppression Water System must be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the reactor must be placed in hot standby within the next six hours and cold shutdown within the following thirty hours.

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9 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 l

such modifications?

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 are considering further evaluation of Vermont Yankee's spray design.

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' 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

The most " vital" valves under station blackout conditions will be a.

those which are required by steam-driven pumps (HPCI, RCIC) supplying makeup to the RPV, and those valves (SRVs) which provide a means for removal of the heat 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 heat sink throughout this accident scenario.

Primary containment vent valves are air-operated with ac solenoid valve contr'o1 (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.

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 /conteinment injection.

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Modifications and inventory requirements to support these options will be developed if design change options are pursued.

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QUESTION 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.

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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 (SAS16, 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 1.94 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 failure."

In addition, the BWROG on Mark I Containment Ultimate Strength Analysis is presently involved in an effort that will attempt to determine the actual vessel failure pressure for Mark I containments. The group will also attempt to show that early overpressure containment failure under severe accident conditions is unlikely.

Reference A: Vermont Yankee Project - Containment, Contract 7-6202, General Design Criteria, Revision 1, Chicago Bridge and Iron Company, February 25, 1969. 5104R

QUESTION 14 In Section 2.2.10.3, is the water supply from at least a portion of the cooling towers also available?

RESPONSE

I The deep basin under the west cooling tower is available to parts of the Service Water (SW) 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

RBCCW:

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)

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j QUESTION 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 Yankee.

RESPONSE

In addition to Vermont Yankee's diesci 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

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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 i

directly to either Emergency Bus 3 or 4 via remote manual breaker 4

operations from t'he 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 I

buses can be written as:

1 U=H+0+V+C I

I where f

represents the unavailability of the hydroelectric stations and H

l any required active breaker transfers at the hydroelectric plant.

I 1

represents operator errors defeating the successful connection.

0 represents active and passive electrical system hardware failures V

I at Vermont Yankee.

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C represents those common events, such as extreme wind conditions, that cause loss of power at Vermont Yankee and defeat the Vernon tie line.

Determination of H iThe best available information indicates that the Vernon Hydroelectric Station was unavailable for a total of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 minutes in a 21-year

-5 period making the average unavailability 1.3 x 10 In response to a grid collapse, the hydroelectric station would have to separate from the grid to allow the tie 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

~0 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:

H = 1.3 x 10

+ 6.5 x 10 '

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= 6.6 x 10 Determination of 0 The estimates for operator inappropriate action are as follows:

0 **

Phase I 0-2 hours

.1 Phase II 2-4 hours

.05 Phase III 4-10 hours

.01 Phase IV 10-24 hours

.01 It should be noted that only two events were recorded one of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 20 minutes and one of 4 minutes. The station recovered quickly from the latter event which was initiated by a lightning strike.

• • The values above are taken from the following documents:

(1) BWR Individual Plant Evaluation Methodology.

(2) A. D. Swain and H. E. Guttman, Handbook of Human Reliability Analysis With Emphasis on Nuclear Power Plant Applicatiot.s, NUREG/CR-1278. 5104R

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 V Two breakers-have to close to feed either bus. In addition, it is assumed

' hat the diesel breaker must open (this is conservative since failure of t

i this breaker to close may have been the cause of " diesel failure to supply

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emergency bus").

Therefore. V can be approximated as 1.7 x 10 This value was supplied by one of our consultants whose expertise is in the area of PRA analysis.

Determination of C 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 l

assessed not be be applicable to the Vermont Yankee site.

-3 C could range from 4 x 10 to.1.

These values were also supplied by our PRA consultant. This information had to be taken from the referenced material because no Vermont Yankee-specific data was available.

i f

The upper estimate (0.1)

.'s used for the point estimate quantification in i

j this analysis.

Summary i

1 The unavailability of Vernon Hydro as an effective AC power source to the emergency buses given a station blackout is:

4 U=H+0+V+C U = 6.6E-4 + 0 + 1.7E-3 + IE-1 I

J i j 5104R i

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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

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QUESTION 16 The Nitrogen Containment Atmosphere Dilution (N CAD) System is 2

referenced on Pages 25 and 115. What maintenance and surveillance procedures are used to ensure operability?

RESPONSE

The Nitrogen Containment Atmosphere Dilution (N CAD) at Vermont Yankee 2

provides the capability to inject nitrogen gas into the primary containment and to vent the containment gas mixture at a controlled rate through the Standby Cas Treatment System. This method of reducing the hydrogen and oxygen gas concentrations in the containment provides a

" defense in-depth" design approach to combustible gas control as a backup to the inert containment.

The N CAD System, consists of nitrogen supply piping and valves, as well 2

as vent gas piping and valves. Nitrogen gas from the inerting system's nitrogen supply tank or other on-site or off-site bulk storage sources can be manually cross-connected to the N CAD nitrogen supply piping from 2

outside the Reactor Building.

The only vital active components of the N CAD System located in the 2

Reactor Building are nitrogen supply valves and vent gas valves. The valves must operate remotely to align N CAD flow paths and to provide 2

containment isolation.

The following waintenance and surveillance procedures apply to the Nitrogen CAD Systemt o

Remote valve operability - monthly (OP-4125).

1 Local / remote valve indication test - each refueling outage (OP-4102).

o Type A - Primary Containment Integrated Leak Rate Testing (OP-4029).

o Inservice Testing Program - Type A and B Valve Operability (AP-4024).

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QUESTION 17 In Section 4.1.4 MARCH /RMA and MAAP code package for Vermont Yankee is referenced. Were calculations made for Vermont Yankee, or were'the results of computation for other reactors evaluated for the Vermont Yankee '

design? What calculations were made?

RESPONSE

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 following 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 sequence analyses were performed specifically for Vermont Yankee using the MAAP code:

1.

Station blackout for 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. 5104R

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QUESTION 18 In-vessel and ex-vessel steam explosions were not considered credible (Page 55, first paragraph) based upon research.

Identify the research that forms the basis for this conclusion.

RESRONSE IDCOR has studied in-vessel and ex-vessel steam explosions [1], [2] and found that the energy transfer mechanisms sufficient to threaten the reactor vessel or the containment of current LWRs do not exist. Also, information p'ublished in NUREG-1116 [3] supports 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 I

neglected in source term determination.

t a) In-Vessel Steam Explosion The energy release from large scale steam explosions has been hypothesized to be sufficient to cause containment failure.

Generally, this has been conceived to result from in-vessel steam explosions which fail the primary system and subsequently the I

containment. The issue to be addressed is whether large scale steam explosions of this magnitude could occur during degraded core accidents.

I A review of the literature regarding in-vessel steam explosion was performed to allow quantitative characterization of this potential containment failure mode [1-10). The principal sources of information I

used in the Vermont Yankee assessment and the conclusions drawn from those sources are as follows:

o IDCOR Reports [1, 11, 12]

Steam Explosion Review Group (SERG) [3]

o 5104R i

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l o

American Physical Society Study Group [6]

o Mark I Containment Event Tree for Application in NUREG-1150 [9]

Based upon the information presented in these sources, the conclusion

was reached that the probability of a steam explosion sufficient to cause RPV and containment failure is small. In addition, the failure e

probability at elevated RPV pressure may be zero and at low pressure may be 1E-3 to zero depending upon the modeling assumptions. Most of the sequences in the Vermont Yankee analysis would be at high pressure and therefore a value of IE-4 is used consistent with Steam Explosion Review Group (SERG), a flow type melt model (MELPRI), and the lack of sufficient energy released in a very short time.

b) Ex-Vessel Steam Explosion It has been postulated that energetic steam explosions caused by molten material dropping into shallow water pocls 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]

I o

American Physical Society [6]

o Sandia Preparation of CET for NUREG-1150 [9]

There appears to be less general agreement regarding the possibility and impact of ex-vessel steam explosions as opposed to in-vessel explosions.

It does appear, however, that the likelihood of a sufficiently severe ex-vessel steam explosion to threaten containment is impossible if the core melt process is a flow type.

l l 5104R i

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References

[

1.

Technical Support for Issue Resolution, Fauske and Associates, FAI/85-27, dated July'1985.

l 2.

IDCOR Task 14.1.

3.

Steam Explosion Review Group (SERG), "A Review of the Current Understanding of the Potential for. Containment Failure Arising From In-Vessel Steam, Explosions," NUREC-1116, U.S. Nuclear Regulatory Commission, Jurie 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.

1 5.

J. B. Rivard, et,al., Identification of Severe Accident Uncertainties, NUREG/CR-3440, dated September 1984.

6.

Richard Wilson, et al., Radionuclide 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 Accidents at the Peach Bottom Atomic Power Station (Preliminary Draft for i

Review) SAND 86-1135, dated May 12, 1986.

10. Reactor Safety Study, WASH-1400, October 1975.

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11.

IDCOR Technical Summary Report, November 1984.

12.

IDCOR Technical Report 14.18. " Key Phenomenalized Models for Assessing Nonexplosive Steam Generation Rates," June 1983.

9 8

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, 5104R

. ~. -

w QUESTION 19 As we understand Tables 4.7 through 4.10, the 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

. refer to high, medium, and low releases, respectively. To what extent can

' mitigation through manual actions in the time available, and in the temperature and radiation environments associated.with such accident types 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 effectiveness of the existing plant and

~

procedures to be versus the possible improvements for early and late sequences?

RESPONSE

The probabilistic analysis results are provided in Tables 4.7 through 4.10 and take into account the existing plant design and procedural guidance.

Mitigation through manual actions as presently prescribed in emergency 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 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 ensuccessful 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 System) can all be 2 5104R i

Ch 0

accomplished by manual actions outside the Reactor Building provided ac power is available. For the station blackout scenario, the existing plant design does not support remote operation. Local operation of containment sprayvalvesintheReactorBuildingisconsideredthemostfeSsible containment 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 carrently remotely operable under station blackout events.

However, the'N CAD System would not be expected to be needed due to our 2

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 are to be reconsnended. At this time, it is difficult to judge the relative effectiveness of potential design improvements.

l 5104R

.-.__.2.

QUESTION 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? Verify that procedures exist for the operator to line up 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?

RESPONSE

The ECCS systems were analyzed at full-rated flow conditions to determine what the available NPSH (ANPSH) for each would be with suppression pool 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 ruction strainers. Torus airspace pressure was assumed to be 14.7 psia.

t 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 175 F.

Efforts to quantify this effect and to provide additional guidance to operators will be based on industry and BWROG studies. 5104R

QUESTION 21 Venting is considered for the s*.ation blackout sequences only. )Please discuss your rationale for not considering other events when venting may be beneficial.

RESPONSE

Venting is discussed for both the st'ation 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,summarises 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-element proposal relative to the station bisckout event. The conclusion was that wetwell venting could be utilized for containment p'ressure control in lieu of containment spray, but further study was required to determine the most appropriate procedural / design J

changes to pursue. Section 5.7.4 provides the recommendationc 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 advisable during an ATWS event unless no other method of pressure control is available. In addition, further studies were recommended to determine l

the uncertainties / risks associated with venting as compared to other containment pressure control methods 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 i

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.

l Vermont Yankee should perform additional plant-specific analysis to insure l

any decisions on venting are based on a sound engineering foundation."

i i 5104R

QUESTION 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.

RESRONSE Stack height was not a consideration in our study. The focus of ou'r study was on the containment failure probability. Only plant characteristics that influence 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.

. 5104R

t QUESTION 23 4

Evaluate the differences in off-site dose consequences due to v,enting at ground level versus through the stack. Using site-specific meteorology and topography, provide an estimate of the off-site dose differences j

between the two types of-releases as a function of distance from the site.

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.'

I.

The differences in off-site dose consequences due to ground level venting i

versus a primary vent stack release can be estimated by comparing the t

?

atmospheric diffusion factors (CHI /Q values) from these two release pathways. Cumulative probability distributions of hourly CHI /Q values at i

l the site area boundary (0.16 to 0.44 alles) and at distances 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 Stack Ratio (Gnd/Stk)

Site Boundary 1.4x10 7.2x10~IO 1.9x10 9

-5 1 Mile 2.5x10 2.2x10-6 11,4

-6 2 Miles 9.9x10 6.3x10-6 1.6 3 Miles 5.8x10-6 4.8x10-6 1.2 4 Miles 3.9x10-6 3.5x10 1,1

-6 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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t QUESTION 24 s

On Page 125, rapid containment depressurization which could fall the drywell is offered as an uncertainty relative to containment ver$ ting.

What analysis and/or tests are being conducted to reduce this uncertainty? If no analysis or tests are contemplated, what actions are

' proposed to minimise the uncertainty?

i

RESPONSE

The uncertainty referred to results from a concern addressed by the BWROG on Emergency Procedure Guidelines (EPGs) in their draft report,

" Development of BWR Containment Venting Procedures." The failure

{

postulated would originate from a suppression pool swell which would result from the clearing of the drywell vents or downcomers and the flashing of the suppression pool liquid. This concern is primarily with the opening of large lines in MARK I or MARK II containments where the rate of depressurization is largest.

There are presently no analyses and/or tests being conducted to reduce this uncertainty. However, as mentioned above, the issue is being I

addressed by the BWR Owners Group and our future actions will consider their final response.

1 1

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Remote manual valve operation is discussed in Section 5.3.5.1.1, primarily with respbt to station bla(.kout. To what extent c.an the remote vent valve and any spray valve alignment be counted on for the other classes of 4

s sequences you psses dd?'. That is, if etmete manual opezation is not f

3

.available,woul'dthelocalenviror.medttheoperatorniwouldencounterallow successful local-operation?

3 j

RESPONSE

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)

(1 L1 The remote manual valve operation discussed in Section 5.3.5.1.1 deals i

with RHR valves 183,,1844 16h, and 31A. These valves are necessary to line up the 41}se? fire yump for spraying the diywell af ter the initial five hours ofca st'ationiolackout. The valves are located in the Reactor Building and may not b,4 ccesa%1e if se.ere fbel failure occurred prior n ;

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to taking action 'to opes them due to ares radiation dose rates. AC power

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independent reaou manual control for these valves is; being studied as j

'describedinour[responsetoQuestion19'.

I For the other classes of sequences assessed injthe report, the diesel fire pump would not be necessary for spraying the dryvell. The RHR System with the torus water inventory as the primary source of water would be utilized i

intil the torus water temoerature exceeded the.specifled maximum value.

At this point, the Service Water System could be crosatied into the "A" loop of the RHR System to provide an ultimate backup capability to inject l

water into the reactor vessel and/or containment. from the Connecticut l

River. The valves required to utilise both apray patts are operable from 4

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 j

of these valves from the Control Room would not be possible in the event 2

of a station blackout since they are all ac powered.

(Operation would be 1

1

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i 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 cot be accessible since they are located in the Reactor Building and 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 manual actions and temporary

' connections. It is Vermont Yankee's position that the primary method of containment pressure control in severe accident sequences should be the use of containment sprays.

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OUESTION 26 1 -

,e Severe accident venting discussed ir. Section 5.4 does not include an evaluation 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 3

bypass path. Can battery packs or portable generators be used to assure high reliability? If so, at what approximate cost?

RESPONSE

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 1

environmental conditions, including high pressure, temperature and

{

possibly radiation during the postulated severe accidents. DC control power to the solenoids, as well as sufficient instrument nitrogen gas pressure, is needed to insure proper operation of the SRV valves.

Te The reliability of our ADS valves for severe accident service is primarily influenced by three factors:

(1) valve failure rate history at Vermont Yanke', (2) de control power availability, and (3) instrument nitrogen gas e

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"piecsure.

In addition, there have been three instances when the air operatorr associated with the valves failed inspection tes ting. The last occurrence was in July of 1976. These three failures i

were; attributed to the diaphragms on the operators. The diaphragm material was changed in 1976 and their replacement frequency was also increased. Since that time, there have been no additional failures.

DC control power is provided from reliable redundant battery banks. The dc. solenoid valves require very little power to cperate.

It is expected

' 5104R A

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.

.The instrument nitrogen gas is supplied from an on-site liquid nitrogen storage tank. No electric power is needed to maintain this supply system

' operable. The Nitrogen Gas Supply System is comprised of very few active components, and operating experience has been very good.

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 Significance of Generic Safety Issue 61."

The best estimated core melt frequency of an SRV line-break accident

-13 reported in that study is less than 1.0 x 10

/ reactor year, which is small compared to the other accident types reported in the Vermont Yankee Containment Safety Study.

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QUESTION 27 What is the approximate cost for improving the valving for the' diesel fire Pump?

RESPONSE

A conceptual design study is underway that will address the cost of improving certain valves to aid the operators in aligning the diesel fire I

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

QUESTION 28 For the improvement options you have evaluated, what maintenanc and surveillance guidelines would you propose to use?

RESPONSE

9 A number of design improvements are being considered by Vermont Yankee.

As our engineering studies progress, we will determine which improvements are to be reconsnended. Maintenance and surveillance guidelines will be addressed as part of the detailed design change process to implement the reconsnended improvement. Maintenance and surveillance requirements are generally determined with consideration of a number of factors including importance to safety, equipment reliability, vendor reconsnendations, etc.

t,

5104R

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. 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 TAF if RPV flooding / injection (core cooling) capability is lost at some subsequent point in time. To be effective the following conditions 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:

With primary containment water level above that corresponding a.

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).

i 4.

Capability to vent the RPV outside the primary containment to ensure that the RPV will be flooded via containment flooding by reducing the pressure within the RPV to as close to atmospheric pressure as can be achieved. 5104R

All of these conditions must. exist simultaneously for an appreciable amount of time (approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) before core cooling can be effectively achieved by this means.

If all other means of pore coolingarelostpriortothis,itstillprovidesameansfErcore debris control and protection of the primary containment structure.

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.

8 5104R

QUESTION 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.

~

RESPONSE

At the time of vessel melt-through, it was assumed in an analysis of MARCH /RMA that the debris was separated into two layers - the upper layer being the lighter melted metallic materials and the lower layer being the denser oxide.' The debris temperature was calculated to be higher than the liquidus temperature of the metallic melt and lower than the liquidus temperature of the oxide 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 3

58,000 pounds of zirconium (145 ft ) and 56,000 pounds of steel (115 2

ft ).

The total area of the drywell floor is approximately 1,240 ft (see Figure 5-13).

In addition, there are two sumps whose total volume is 156 ft. Therefore, assuming the sumps fill completely, there will 3

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. 5104R

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.o QUESTION 31 i

What is the thickness of the vent duct between torus and drywell I

(Figure 1, Page F-6)?

l

RESPONSE

e The vent duct la 1/4 inch thick with a 6-foot 9-inch inside diameter.

8 5104R

's QUESTION 32 It is stated that " additional physical barriers are believed to'$ e b

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 vould prevent the spray from effectively cooling the debris

'(Page 137).

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 exit the Subpile Room and spread over the drywell ficor, more water will be available to cool the debris and there will be more debris surface area to be cooled.

r 5104R

QUESTION 33 It is indicated that, with the exception of the one-inch nitrogen CAD line andthesix-inchnitrogenpurgeline,thepipelinesassociatedlwithfour potential vent paths are likely to fail.. Provide background information i

which led to this conclusion (Page 131).

RESE0NSE 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:

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."

4 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 capable 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 into duct work prior to exiting the Reactor Building.

The discussion in (c) above applies here as well. '

5104R m

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, please 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 Given 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 5104R

'n o

Each of these 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 of the 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 possi,ble in view of the radiative 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.

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 MAAP 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. 5104R

\\

i Once the debris causes local seal penetration failure, the debris is anticipated to fall or be ejected (deper. ding upon the RPV pressure).

}

. o Mass Transport to the Drywell Floor i

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 RPV 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 transport process, it is anticipated that the debris would cool more than 300 C.

I 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 i

debris contact with the drywell shell wall.

l o

Debris Spreading i

1 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 l

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 I

with no drywell injection (e.g. drywell sprays) then the drywell wall temperature could be high, i.e., on the order of 1500 F.

4 If the drywell shell is not adequately restrained (and the 1

. Judgment is that the shell in this area is adequately restrained) then a high temperature creep-rupture, failure mechanism could result.

-54.

5104R i

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o-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 than.01 conditional on the fact that the slurry could even reach the drywell shell wall.

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 pmst 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

L 1

The pathway would be restricted and would promote refreezing of o

the interface debris that could seep through the containment drywell failure pathway. Therefore, it can be modeled as an intermittent gas flow path through a tortuous path and 1nto the

~

Reactor Building.

Based upon both the supporting MAAP and MARCH calculations, the radionuclide releases associated with such a leakage pathway into the lowest Reactor Building elevation are found to be low releases.

If the Reactor Building is effective, as it is judged to be at Vermont Yankee for this sequence, the radionuclide 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.

e e

j i 5104R

~---

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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?

Provide references for the Ames and Sandia calculations mentioned in c.

the Appen' dix 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. Structural characteristics such as actual yield strength, strain hardening, coupling of sections, 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 conservatisms, 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 f

to determine the actual vessel failure pressure including uncertainty calculations.

b) Civen 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. 5104R

t c) The source of data from the Ames and Sandia calculations is from NUREG/CR-3653:

1 NUREG/CR-3653, SAND 83-7463, " Final Report Containment Analydis Techniques - A State-of-the-Art Susunary," March 1984.

o Browns Ferry by Ames Laboratory, Section 16.1.

o Peach Bottom Containment by Sandia National Laboratory, Section 17.1.

6 5104R i

QUESTION 36 Your evaluation of deinerting indicates a relatively few hours of power operation while the containment 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 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

RESPONSE

See table on the following page.

. 5104R

Primary Contelament Deinertion Durina Power Operntion REACTOR POWER REACTOR PRESSURE DATE REASON DURATION (ENTRY / EXIT)

(ENTRY / EXIT) 6/8/82 RRU-3, -4 Repair

?-

S/D (14.7 psia 8/27/82 Recirc Pump Seal Replaced

?

S/D (14.7 1/7/83 Turbine Moisture Separator Leak and CU-19 Line Leak 14.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> S/D (14.7:

2/2/83 High Drywell Sump Leakage Indicated Valve Packing Leakage 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 70%

1020 3/4/83 Refueling 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

S/D (14.7 6/22/83 MSIV-80C and -86A Timing Out of Range 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> 231 1020 8/27/83 Inspection of RV-70B, High Tailpipe Temperature 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> S/D 800 1/5/84 HP Surveys 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> S/D

<14.7 1/20/84 RV-65B Packing Leak 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> IRM Range I 1020

(<1%)

4/16/84 MSIV-80C, Partial Closure Test i

Failure 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> S/D 800~

l 6/15/84 Refueling 23.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> S/D (14.7 8/9/84 MSIV-80A Indication Ground - EQUAL Connector 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> S/D

<14.7 9/18/84 Rx Shroud Lift 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> S/D (14.7

.,s 9/30/84 MSIV-80D Timing Failure 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> S/D

<14.7 9/20/85 Refueling (and Recirc Pipe Replacement).

18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> S/D

<14.7 8/20/86 B-Recirc Pump Vibration Monitor and Motor Oil Level Switch Problem 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 56%

1020

1. Duration is defined as the total time the containment oxygen concentration is above 4% (Tech. Spec. 3.7.A.7(b)).

5104R _-_ _____________ -.

. J...... - -.:....

t QUESTION 37 Please provide your estimates of pressure and temperatures as a) function of time for the accident sequences you analyzed for CCFP estimat'es.

t

' RESPONSE 37 Attached Figures 1 through 11 show some of the key thermohydraulic parameters calculated by MARCH /RMA as functions of time for T,C,C2 (A'NS) sequence. Attached Figures 12 through 21 show some of the key thermohydraulic parameters calculated by MARCH /RMA as functions of time for a station blackout sequence (station blackout for more than six hours, coupled with failure of HPCI, RCIC, and diesel fire pump after six hours).

. 5104R

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QUESTION 38 Itisnotclearfromyourevaluationwhytheprobabilityestimadesof 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 Frequencies Versus Hiah or Medium Release Frequencies There are many types of accident frequencies which could lead to a core melt accident. For the spectrum of accio, 4ts, 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 passive 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 successfully 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 minimize the radionuclide releases even after containrent 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 radionuclide 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 5104R

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Passive Wetwell airspace failures (i.e., the containment failure mode identified in WASH-1400)

Reactor Building decontam'ination o

Active

- Drywell sprays or coolant injection to drywell Each of the mitigation measures has the capability to reduce the radionuclide source terms to the low category (L). Calculations using both MAAP and MARCH /RMA indicate that radionuclide 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 ATWS 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. 5104R

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1 b) Venting During AIWS Venting' as a mitigation measure for ATWS was not used for tiro reasons:

The current Vermont Yankee procedures require that there be no o

unusual radionuclide 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., pe'ak 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).

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