ML20207T416

Discusses Util PRA Info & Impact PRA Has on EPZ Size. Suggests Formulation of Two Concerns,Including PRA Being Used at Least to & Perhaps Past Point of Applicability. Related Info Encl

ML20207T416
Person / Time
Site:	Seabrook
Issue date:	01/13/1987
From:	Lyon W Office of Nuclear Reactor Regulation
To:	Newberry S Office of Nuclear Reactor Regulation
Shared Package
ML19306D588	List: ML20197D222 ML20205L245 ML20207A140 ML20207F465 ML20207Q051 ML20207Q052 ML20207S785 ML20207T164 ML20207T271 ML20207T335 ML20207T387 ML20207T396 ML20207T407 ML20207T416 NUREG/CR-4540, Forwards BNL 861117 Request to Amend Page 60 of NUREG/CR-4540, Review of Seabrook Psa:Containment Modes & Radiological Source Terms, to Clarify Wording Re Conditional Failure Probability
References
NUDOCS 8703230552
Download: ML20207T416 (57)

Text

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• .o January 13, 198,/ )2 NOTE FOR: Scott Newberry, Section Leader Facilities Operations Branch, DPLA

. FROM: Warren Lyon, Sr. Nuclear Eng'ineer Facilities Operations Branch, DPLA

SUBJECT:

SEABROOK STATION RISK AND EMERGENCY PLANNING PERSPECTIVE As you know, I have been working with Seabrook Station review for some time, and I spent several days last week at Brookhaven Nationar Laboratory (BNL) discussing BNL work pertinent to the Public Service of New Hampshire (PSNH)

Probabilistic Risk Assessment (PRA) information and its impact upon Emergency Planning Zone (EPZ) size. I forsee some difficulties with some of this PRA work. Consequently, it may be fruitful to postulate conclusions and see where thi's leads. I do so with the understanding that such postulates are preliminary and perhaps premature, but I believe we s,hould be critically examining the technical aspects as others ser.them as well as our own perceptions.

Suppose we formulate two major concerns:

1. PRA is being used at least to and perhaps past the point of applicability, given the present state of knowledge and what many people believe to be the uncertainty which must be assigned to the phencaena, and,

2. If correct, failure to appreciate this could lead both ourselves and PSNH into a non productive exercise which will take several months.

There are a number of implications. To understand this, first we need to examine part of the risk message provided by NUREG-0;96 (" Planning Basis for

, the Development of State and Local Government Radiological Emergency Response l

Plans in Support of Light Water Nuclear Power Plants", December 1979):

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usi i se .oo imo DISTANCE mal 111 PSNH has stated that Seabrook Station risk may be described by:

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I which they conclude means that Seabrook Station risk with emergency planning to a one alle radius is no greater than what the authors of NUREG-0396 considered to be an acceptable risk with emergency planning to a 10 mile  ;

radius. / i PNL has investigated several issues pertinent to the above. When they consider individual selected items which were not consideret in either the .

Seabrook FRA or NUREG-0396 and apply " pessimistic assumptions", they find:

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,, _ e ,, .e see imo DETAMCE WILSSR where i e cur *e labeled "BNL"' represents the upper values (most pessimistic) which they chasidered in their sensitivity work, and the cross-hatching indicates a range is involved, the lower values of which are of f the scale of the plot. An important point is that the perturbations to the PSNH curve have little to no impact at a 10 mile radius, but there is an impact at the one alle radius.

In each of the cases we have identified, f or practical purposes, the SNL perturbation can be reduced by using more realistic modeling, and probably can be shown to again fall within the one aile radius (for a conditional probability greater than 0.001) as concluded by PSNH. This modeling approach will be controversial because many people will disagree with what I believe to j be realistic representations, and I will not recommend certain of the modeling

! because I cannot establish sufficiently tight uncertainties at this time. .

(Steam generator tube rupture is a good example of this.)

._ ... . . _ . ~ . . _ _ _ . . ~ - _ _ _

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PSNH can reduce each SNL perturbation, probably to the point of each being negligible, via suitable and acceptable actions regarding plant operation, perhaps in conjunction with einer plant changes. Although this may be accept'able to most people with respect to the analyses and plctted results,

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there will be some residual uncertainty which causes some difficulty. More importantly, such an action fails to address the question of undiscovered events and phenomena which, if recognized and included in the analyses, eight again result in exceeding the NURE6-0396 bound when applied to a ens mile planning radius. ,

The difficulty is in part due to:

1. PRA techniques are inadequate to provide a complete picture.

2. The " defense-in-depth" philosophy is perceived as being weakened.

These items will be examined in the following paragraphs.

1. Pgg,g(([,gg[,gtgv[gg g,qggg[g[g,g[gigtgg Consider events as divided into three types:

.. a. Those which have a high probability of occurrence or which may be reasonably investigated because a body of information is available which describes the potential occurrence,

b. Those which have a " negligible" probability of occurrence, and,

c. Those which-are bounded by the above two extremes.

Type a events are those which one may reasonably expect to occur over some< period of time. The key parameters are that they are recogni:ed and something is known, either directly or indirectly, of their frequency. A severe earthquake along a well studied, major f ult can fit in this category. Sooner or later, the event will occur and, if sufficient historical data are availatte, one may calculate roughly how often the event will occur. Reactor trip is another example of an event which can logically be expected to occur. The important characteristic of type a events is that they are amenable to analysis by application of PRA sethodology.

As defined, type b events essentially cannot occur since their probabilities are "negli'gible". The difficul'ty is that they do occur and this aust be recogni:ed and considered in a realistic nuclear power plant evaluation. Some type b events may be classified as "bi:arre'. (I have borrowed freely from a conversation with Pete Davis in development of this topic.) These individual events are enes which an analyst wi!!

calculate "cannot* occur due to the convoluted path which must be followed to achieve occurrence. PRA's will calculate probabilities of these individual bi:arre events as so low that reasonable people will conclude they may be neglected. An additional aspect of some beetsr h ,ea events is that no-one has recogni:ed their possibility. The Rancho SMo light bulb accident and the Brown's Ferry fire are type b events if the initiator is considered.as part of the path to be evaluated. Steae

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F generator tube rupture as a consequence of core melt associated phenomena is another example. .( This has not been considered until recently because the mechanisa leading to the potential failure was overlobked by the analysts.) The difficulty with type b events is that they can happen!

Enough of these events ha've occurred that the message is clear - they cannot be ignored. Equa!!y clear is the failure of PRA techniques to

" catch" them, and although one may argue that we are getting smarter and PRAs are getting better, the "bi:arre" events continua to occur.

Type e events are those events which occupy the grey region between types e and b. Type c events are more or less amenable to analysis via PRA technology, with the degree of success being a function of information and understanding which may be applied to the event or class of events.

We have witnessed development of a number of nuclear power plant PRAs. A general trend i's that the major PRAs are becoming more complete. (In this respect, the Seabrook PRA appears to be excellent.) The trend further appears to be the identification of more and acre possible accidents, and although the likelihood of each individual accident appears to become less~(perhaps due to its more carefully defined bounds and a multiplicity of events which replace one event), the total core  ;

nelt probability appears to be drifting' upward. Uncertainty remains

.high, and analysis techniques. remain "cohservative" such that the results are distorted to the point of being misleading. When one encounters event frequencies in the range of 10-6 to 10-7 per reactor year with

. large uncertainties, one' should begin to question whether these values 4

are meaningful. We are in this situation when applying PRA methodology to the Seabrook Station EPZ size consideration. (Note that this difficulty is implicitly recogni:ed in-our regulations, which do not rely on PRA alone.)

2. Iht_ftirn11:in:stath ahileinahYs - ~

This philosophy is based upon a number of separators between release of radioactive material and public exposure to that material. These includes

a. Fuel composition and structure

b. Fuel clad

c. Reactor coolant'systes pressure boundary

d. Containment

e. An " exclusion" zone

f. An area outside the exclusion :ene fence in which preplanned public response planning exists.

o This philosophy has served us well. When one considers the failure of PRA to fully identify events in nuclear plants, one should carefully examine any change which is based on PRA arguments. My perception is ,

that we have not seen sufficient justification that a blanket reduction  ;

in EPZ si:e will be widely accepted.

What of Seabrook Station in light of the above? Seabrook Station, in my opinioni,is at least as safe as most of the other power plants in operation in I this country. There is nothing unique in the nuclear steam supply and

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4 associated systems which would indicate a higher than normal core melt likelihood. The Seabrook containment is large, which has the advantage of providing volume for expansion of material in the event of a core melt, and it is the strongest containment I have encountered. The concrete is basaltic, which will minimi:e generation of gases if there should be a core sett (in contrast to those containment basemats which are based upon limestone concrete). Finally, ths containment configuration is such that the likeli. hood of reaction of significant quantities of core melt with the containment atmosphere is minimized. This realistically makes the chances of an early containment failure exceedingly small under core seit conditions, as well as making it unlikely that the containment will fail in the long ters. . The only weak points.I see involve possible containment bypass situations or anything involving an accident outside containment; and a effort has been initiated to deal with these. Given these aspects of Seabrook Station, there is a good basis f or assessing changes in the EPZ configuration. I suggest the following technical approach (based upon suggestions which Steve Long has made):

1. Assess the risk due to Seabrook Station assuming a 10 si'le EPZ si:e and further assuming the full cooperatien of state and local officials.

2. Assess the risk due to Seabrook Station assuming the EPZ situation as it presently exists, with full cooperation of the State of New Hampshire and

local New Hampshire officials, and with PSNH personnel and planning substituting for Massachusettes officials and planning' insofar as is realistic. The si

e should remain at 10 miles.

3. Compare the above.

My suspicion is that there will be little difference.

I will add one more observation to this communication. Energency planning for Seabrook Station, or any other_ plant for that matter, should be based upon the premise that a core melt accident will occur. I say this not in the belief that it will, but simply fras the viewpoint that prudent and intelligent people should f ollow such an approach (just as prudent and intelligent people will have developed catastrophe plans to deal with serious accidents whether or not there is a nuclear-power plant in the vicinity). This planning should be based upon a realistic appraisal of plant features, not one in which so such conservatism has been added that the accident response becomes unrealistic. At the same time, one can make allowances for the unforseen.

For example, the Seabrook Station containment will, with a high likelihood, prevent release of radioactive sai: rial during the early hours of a core selt accident. This should be recogni:ed in emergency planning. At the same time, prudent planning should be based on protective action initiation prior to actual core melt.

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Warren C. Lyon

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PERSONAL NOTES 8/27/86 R. Barrett's notes on public NRC f f. PNL) meeting with PSNH 11/12/86 S. Long's notes on public meetinc l (undated) Topics for Pratt prepared for telephone conversation with BNL by S. Long (undated) Discussion Topics prepared by S. Long for meeting in T. Novak's office on 12/12/86 12/12/86 S. Long's notes on meeting in T. Novak's office.

1?/18/86 Notes of S. Newberry on riview of RNL draft report.

I?/22/86 Handwritten note from S. Long to V. Noonan concerning review of YAEC 1502 1/6/87- Outline for Summary (of BNL report) prepared by S. Long 1/7/87 S. Long's notes from meeting at RNL to discuss draft report.

1/8/87 L. Soffer's notes from meeting to discuss how to handle PSNH waiver request.

1/12/87 L. Soffer's notes on meeting in F. Kantor's office.

1/12/87 R. Barrett's notes on meeting with BNL, 1/13/87 Handwritten note from S. Newberry to E. Rossi through V. Benaroya on

" Review schedule of PSNH package in accordance with Rationale 1,2,3."

, . 1/14/87 S. Long's notes from public meeting.

1/14/87 S. Newberry's notes from public meeting.

1/14/87 E. Rossi's notes from public meeting.

1/14/87 S. Davis's notes from public meeting.

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T.: Wy DISCUSSICN TOPI'.,

I..BNL Draft Report Deficiencies II. List NRC Reviewers for-BNL Dr, aft Report

a. cover technical 'ssues i

b' . 62velop staff position on BNL report III. Adequacy of Information Base to Evaluate Seabrook Request

a. Materials Available

1. BNL Review of EPSS and RMEPS E. BNL Review of PSA back end

3. LLL Review of PSA front end

4. PSNH Responses to LLL Questions

b. Accuracy of PRA IV. Criteria for Comparisons

a. Absolute Risk or Conditional on Core Melt

b. PAGs or 200 Rem dose only

c. Ad Hoc Expandability Basis V. BNL Report Review Schedule

- a. PSNH Request Expected Next Week

b. ACRS Schedule

1. Subcommittee Meeting on 1/29/$7

2. Full Committee Meeting on 2/5/86 I

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SSPSA KEY EVENTS e

j e ORIGINAL SSPSA SUBMITTED 2/84 f,_j[

{ e LLNL REVIEW OF SSPSA PLANT MODEL 4/85 g he BNL REVIEW OF CDNTAINMENT MODEL 2/86 -

! e SSPSA UPDATE SUBMITTED 7/21/86 j e MEETINGS - IN BETHESDA '

8/6/86;8/27/86;9/23/86;11/12/86 j - AT BNL l- 8/14/86;10/16/86;10/17/86

{ - AT.SEABROOK 9/8/86;9/9/86;10/15/86 4

e ACRS - SUBCOMMITTEE 9/26/86 FULL COMMITTEE 10/10/86 l- e DRAFT BNL REPORT ISSUED 12/8/86 e APPLICANTS PETITION FILED 12/18/86

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GENERAL COMMENT

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SUMMARY

DOES NOT REFLECT FULL SCOPE OF "5 i 4EVIEW ,

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v'# /#

ROLE OF RMEPS AND SENSITIVITY STUDY NOT EXPLAINED i #

i CONSERVATISMS SHOULD BE IDENTIFIED i

e EXPLAIN FOCUS IS EARLY RISK PROFILE i e SSPSA MORE COMPLETE THAN WASH-1400

! e BNL REVIEW HAS NOT DIMINISHED TECHNICAL JUSTIFICATidN FOR A 1 MILE EPZ r

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e RHR SYSTEM INTEGRITY

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EXPERIMENTS VS MODELS e CONSERVATIVE PROBABILISTIC ASSESSMENT

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GENERIC VS PLANT SPECIFlC TREATMENT

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REANALYSIS USING BNL ASSESSMENT OF CONTAINMENT PRESSURE CAPACITY -

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s ENCLOSURE TO ANSWER NO.12

Additional EP Enforcement Actions Date Plant Nature Of Violation 11/8/83 Cooper Failure to conduct required -

annual audits for a period of 3 years.

5/25/8a San Onofre 2, 3 Failure to implement the emergency plan as reouired, and failure to recognize and declare an unusual event in a timely fashion.

6/21/84 Ginna Failure to train / retrain all personnel assipned to the emergency organization.

9/22/84 Turkey Point 3, 4 Failure to include appropriate protective action schemes into the emergency plan implementing procedures, inability of Shift Supervisors to make appropriate protective action reconnenda-tions during walk-throughs and inability of responsible emer-gency response personnel to perform dose calculations.

9/22/84 Turkey Point 4 Failure to include appropriate protective action schemes into the emergency plan implementing procedures, inability of Shift

Supervisors to make protective action reconnendations during walk-throughs and inability of responsible emergency personnel to perfom dose calculations.

l i

12/11/84 Big Rock Point Failure to follow procedures for implementation of emergency plans upon exceeding limiting Conditions for Operation.

2/6/85 Peach Bottom 2, 3 Procedural deficiency - Inade-quate initiating conditions relative to EALs.

Markey/IE 02/03/87 , _ _ _ _ _

l l

Da te Plant Nature of Violation i 2/11/85 D.C. Cook 1, 2 Failure to notify offsite authorities in a timely manner after declaration of Unusual Event.

2/13/85 Big Rock Point Failure to notify the NRC of exceeding an LCO as reouired by the emergency plan.

3/8/85 WPN-2 Failure to notify the NRC Operations Center in a timely inanner as required when the high pressure core spray system had actuated and dis-charged into the WNP-2 reactor vessel.

5/3/85 Sequoyah 1, 2 Failure to test sires.s as required. Failure to provide protective action recommenda-tions as required.

7/8/85 Palo Verde 1 Failure to trein/ retrain all personnel assigned to the emergency organization as recuired.

7/25/85 Quad Cities 1, 2 Failure to train / retrain all personnel assigned to the emergency organization as required.

7/30/85 Davis-Besse Failure to notify offsite authorities until 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after classification of an event.

8/26/85 Hatch 1, 2 Failure to include evaluation of interfaces with state and local governments in the annual audit.

I l

Markey/IE L 02/03/87

Da te Plant Nature of Violation 1/10/86 Fort St. Vrain Emergency preparedness training for emergency re-sponse personnel inadequate, no training program for STAS or corporate support personnel, inadequate audits of interfaces with state and local govern-ments for a 2 year. period and failure to correct a previous exercise deficiency.

1/29/86 Fort Calhoun Inadequate emergency prepared-ness training for key members of the emergency response organization.

Failure to adequately audit the emergency preparedness program as required.

2/28/86 Diablo Canyon 1 Failure to train / retrain all personnel assigned to the emer-gency organization as recuired, and failure to maintain emer-gency kits as required.

6/9/86 Ra'ncho Seco Failure to train / retrain all personnel assigned to the emergency organization as re-quired.

Failure to maintain procedures and provide adequate guidance for computer based dose cal-culations at the TSC.

Failure to maintain procedures to calculate dose rates and in-tegrated doses for unmonitored releare pathways; failure to maintain dose assessment cap-ability in the EOF.

Markey/IE 07/03/87

~.

4-Da te Plant Nature of Violation 6/9/86 Rancho Seco Failure to maintain procedures for dose proiection based on inplant monitoring.

Failure to maintain procedures as required relative to pro-viding protective action recommendations based on plant conditions.

Failure to maintain procedures as reouired relative to re-inoval of energency kits and fa.ilure to provide guidance relative to requirements for protective equipment and .

supplies for offsite assembly areas.

Failure to maintain procedures as required relative to the description and location of the first-aid facility, TSC and equipment at the California State Forestry Fire Fighting Academy and the Herald Fire Department.

Failure to retrain selected emergency response personnel in advanced first-aid, standard multi-media and CPR as required for a period of 3 to 5 years.

Failure to review and discuss Emergency Action Levels with State and local authorities on an annual basis.

1 i/24/86 Fort St. Vrain Failure to train / retrain personnel assigned to the emergency organization, and failure to correct exercise deficiencies.

Markey/IE 02/03/87 _ _ _ , _

Da te Plant Nature of Violation 8/14/86 sRiver Rend 1 Failure to review and maintain call lists for emergency re-sponse personnel as required.

9/3/86 Turkey Point 3, 4 Failure to maintain the facility smergency plan as required by 10 CFR 50.54(q).

9/17/86 H. B. Robinson Failure to train / retrain all personnel assigned to the emer-gency orgerization and failure to designate accident assess-ment or damage control team members as required.

Rancho Seco Failure to adequately implement 10/22/86 the emergency plan during an Unusual Event 12/26/85.

- EP Coordinator did not sound the plant alarm or announce -

appropriate messages over the public address system as required.

- Initial notification to State and county emergency response agencies did not contain all required information.

- The counties were not provided with hourly updates on plant status after the initial notification as required.

- Actual alann setpoints for the Auxiliary Puilding stack radiation monitor was higher than stated in the applicable emergency plan implementing procedure, itarkey/IE

' 02/03/87

L Da te Plant Nature of Violation Brunswick 1, 2 Inadequate emergency prepared-10/28/86 ness training and qualification l program and failure to train /

retrain all personnel assigned to the emergency organization as required. l 10/29/86 Catawba 1, 2 Failure to conduct all re-quired drills, e

Markey/IE

- 02/03/87 [