ML20024D589
| ML20024D589 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/01/1983 |
| From: | Hiatt S OHIO CITIZENS FOR RESPONSIBLE ENERGY |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8308050291 | |
| Download: ML20024D589 (16) | |
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NNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I$f Q-Before the Atomic Safety and Licensing Board M.- AUG 4 1983 :-
) q C% cfnene.
In the Matter of )
g'dQ' & ter. .
Docket Nos. 50-% 4 CLEVELAND ELECTRIC-ILLUMINATING )
COMPANY, Et Al, < ) 50-W1 'E
) (Operating License)
(Perry Nuclear Power Plan'c, ) '
1 Units 1 and 2)
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OCRE BRIEF IN SUPPORT OF THE LICENSING BOARD'S 7 INTENT
- TO DECLARE A SUA SPONTE ; ISSUE- -
During the May 9, 1983 conference call the Licensing Board A
posed two questions to Staff and Applicants fo$ the purpose of 4
obtaining information which would help it " decide whether'or not to declare a sua sp5nte issue." Tr. 809.. The questions were:
6-(1)- To what' extent does.the simulator training for operators.of-the Perry reactors include training in
' differentiating different kinds of instrument failures fr'mo transient or. accident conditions; (2) To the extent tl$at some kinds of instrument failures are not simulated during training, please explain whethe'r the omission is detrimental to the safe opera-l tions of the reactor. ,
NRC Staff responded to these questions on June 13 by filing Applicants responded on July 15 the affidavit'of David H. Shum.
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- l. Because inter-' ' " .
! by filing the affidavit of Anthony F. Silaicos'ki.
venor Ohio Citizens for Responsible Energy.("OCRE")
finds these f
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~ affidavits -to bk so vague, conclusory,. and lacking .in factual and because OCRE possesses
-detail as to be totally deficient, I
information concerning this issue, OCRE is filing this brief f 1 ,
l supporting the Licgnsing Board's intent to declare a sua sponte issue. OCRE elieves that the matters raised constitute a serious
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8308050291 830801 -
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safety issue which should be addressed in this proceeding before PNPP can be allowed._to operate.
I. Staff's Affidavit On June 13, 1983 the Staff filed its response to the two questions. posed.by the Licensing Board in the form of the affidavit of David H. Shum. .
The response to the first question is so vague as to be practically worthless ~.. The Staff has essentially deferred the question to Applicants, in that the Staff assumes that simulator training will include dif'ferentiation of instrument failures from
- transient or accident conditions, that the plant operating pro-cedures to be developed by Applicants.should provide sufficient guidance to operators, and that other instrumentati'on will be available in the event of instrument failure.
~ The Staff's response to the second question is similarly vague. The Staff states that the ondssion of. "some kinds of instrument failure in simulator training will not be' detrimental, but does not identify-the failures so classified. The Staff then states that the worst case loss of instrumentation will be simulated during training (emphasis added) . The Staff again defers to the emergency operating procedures (discussed below)
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as providing. guidance with respect to this type of failure.
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OCRE finds the term " loss of instrumentation" to be crucial.
If loss means "not available", as if power is lost to an instru-ment channel, then it is likely that this failure would be. detected as such. Much more ominous is an instrument failure which pro-duces the display of erroneous but plausible data.which might be a ~ - - - --_ -----
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indicative of a transient cv accident condition. Such a failure puts the operator in the position of having to decide which instrument is correc . Reliance on other indicators may or may not be helpful, depending on.the' degree of independence (likelihood of common-mode failures) of the monitors and the adequacy of training'and diagnostic ability of the operator. Since OCRE believes that this.is'.the whole point of the Board s inquiry, the ' Staff's response has provided no enlightenment on the matter .
1 II. Applicants' Affidavit Applicants' response to the two questions posed by the Board, Q in the form:of the affidavit of Anthony F. Silakoski, was filed I July 15, 1983.
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This response did provide somewhat greater specificity as to types of instrument failures simulated. .
The affidavit lists three types of instrument-failures which are simulated:
failure of electrical bus powering an instrument channel, failure of the process computer, and simulated break in i the drywell which causes erroneous pressure readings in the control rod drive hydraulic control system.
It is not clear whether any of these failures represents the " worst case loss of instrumentation" which the Staff claims will be simulated during operator training.
Nor is it clear that these are the only types of instrument failures simulated.
Itishurthe'runclearwhethercombinationsofsuch instrument failures or of instrument failures with transient or accident conditions are ever simulated.
! Of the three. types of instrument failures specified, the first
! two appear to belof a' type that is easily diagnosed. The extensive
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failure of electrical power (e . g. , the failure of an entire electrical
division) would result in the failure of all instruments, monitoring a number of different variables, which are powered by that system.
The loss of a large $ $mber o'f instrument would logically alert the operator to the likely cause, loss of electrical power.--1/ Less extensive electrical failures might also be readily detected, depending on its effect on the display; e.g.,. loss of power would cause indicator lamps and readouts to fail to light, or -
cause meters'or recorders'tc display an obviously erroneous value.
The second type of failure, that of the process computer, would similarly be easily detected due to the large number of g variables monitored and displayed by the computer. Observation.
- - of other displays would lead the operator to' discover the computer failure. -
-6 That Applicants mention the third type of failure is interesting for two reasons. First, the significance of this parameter is no't apparent. Scram system status would be more appropriately monitored by other variables, such as reactor power and control rod
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position. Secondly, Applicants failed to mention a most crucial variable which is affected by the condition postulated: reactor
~~ 1/ Applicants. claim that station blackout is part of simulator training, but'do not explain how an operator is to control the plant with the extensive loss of instrumentation inherent
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in that sitsation. The BWR emergency procedure guidelings do not. address loss of electrical power or station blackout.
_2/ CRD hydraulic system pressure is not a RPS setpoint (FSAR S 7.2.1.1). Grand Gulf (a BWR/6 like Perry) does have a limiting condition for operation that all scram accumulators be operable, and an associated surveillance requirement that the pressureaof each control rod scram accumulator be verified at least once every 7 days (Grand Gulf Technical Specifications, NUREG-0926, pp. 3/4 1-8-9). An alternate source of hydraulic (continued next page)
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b vessel water level.
The heat from a.small break LOCA in the dry-well could cause erroneously high vessel level readings, at a time when monitorIn'g' of that indication is vital .
(More about vessel level indication below.)
Applicants respond to the Board's second question by stating that-operators are trained to compare multiple indications of plant variables and "to respond to the worst case parameter
when instrument failures create ambiguity. This is supposedly, reinforced by the emergency instructions This statement, (which are discussed below).
as well as the emergency instructions, aives little indication of how operators are trained to react when instrument
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failures create ambiguous readings that diverge toward opposite s_ extremes, each of which is undesirable.
Operator confusion may be compounded if false signals actuate or.otherwise' affect plant equipment, the operation of which may. lend crddibility to the faise reading.
. (A specific example of this situation is described below.)
III.
Adequacy of Emergency Procedure Guidelines Since both Applicants and Staff appear to be relying extensively on t,he plant energency procedure guidelines as a panacea to the problem of possible operator confusion due to instrume'nt failures, it is appropriate .to examine these guidelines. The Perry-specific instructions will be prepared from the generic emergency procedure guidelines described in the General Electric Topical Report 1
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continued.'d is provide by the-control rod dri' pressure necessary for contro ve pumps. Each CRD also
_has an internal ballcheck valve'that will allow reactor pressure (if greater than ,600 psi):to-insert the ' control rod.
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NEDO-24934, " Emergency Procedure' Guidelines, BWR l-6.F- These generic procedures consist of two guidelines (one for RPV level control'and one for containment control) and seven contigencies (level restoration, rapid RP,V. depressurization, core cooling without injection, core cooling.without level restoration, alternate shutdown cooling, RPV flooding, and level / power control).
Two additional guidelines, on. combustible gas control and secondary containment. control, are to be added in the future. These pro-cedures are said 'to be superior to previous BWR emerger.cy. instructions in that they are symptom-based rather than event-based. The Staff
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in its safety evaluation of NEDO-24934 stated that "the use of symptoms rather than events as bases for actions, eliminates c errors resulting from incorrect diagnosis of events, and addresses multiple. failures and operator errors." (Generic Letter 83-05, dated February 8, 1983).
However, symptoms are indicated to the operator by instrumentation.
The inability to diagnose instrument failures when using symptom-based procedures may prove as damaging to plant safety as the inability to diagnose events when using event-based procedures, especially since the new procedures may neven be entered if the operator incorrectly interprets the instru'ent m readings.
NEDO-24934 does provide some guidance.to operators regarding instrument readings in the following general precautions:
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S(x letter dated November 16, 1982 from D. Davidson of CEI to A. Schwencer, NRC, re response to request for additional information on 'SER Outstanding Issue 21.
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CAUTION #1 Monitor the general state of the plant. If an entry condition for either (prodedure developed from the level control guideline]
or (procedure developed from the containment control guideline]
occurs, enter that procedure. When it is determined that an emergency no longer exists, enter [ normal operating procedure].
, CAUTION #2 ,
Monitor RPV water level and pressure and primary containment temperatures and pressures from multiple indications.
CAUTION 93 If a. safety function initiates automatically, assume a true initiating event has. occurred unless otherwise confirmed by at least two independent indications.
CAUTION #6 f '
Whenever [ temperature near the instrument reference leg vertica; runs] exceeds the temperature in the table, the actual RPV water level may be anywhere below the elevation of the lower instrument s.; tap when the instrument reads below the indicated level in the table. ,
Indicated
' Temperature
- Level Instrument any 537 in. Shutdown Range Level (500 to 900 105 F -109 in. Wide Range Level (-15 0 to +6 0 i 3100F 19 in. Narrow Range Level (0 to +60 in.
3790F 239 in. Fuel Zone Level (200 to 500 in.)
- [ List in order of increasing temperature.]
CAUTION #7
[ Heated reference leg instrument] indicated levels are not reliable during rapid RPV depressurization below 500 psig. For these conditions, utilize [ cold reference leg instruments] to monitor RPV water level.
The guidelined further caution the operator thht:
If [ temperature near the cold reference leg instrument vertical runs] reaches the RPV saturation limit, enter [ procedures developed from CONTINGENCY #2]. (Contigency #2 is Rapid RPV Depressurization<
A plot of RPV saturation limit vs. RPV pressure is provided in step CC-2.2.2). [ Note: PNPP uses only cold reference leg level instrumentation. See' January 14, 19 83 letter from M. Edelman of
Also, if the RPV water level cannot be determined, the
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.' -o-operator is instructed to enter CONTINGENCY #6, RPV Flooding.
While these steps are an improvement in operator awareness of the problems existing -with RPV level instrumentation, they do not totally solve the problem. The operator must rely on instrument readings before even enuering'the emergency procedures.
The procedures can also be exited at any time, depending again on the operator's interpretation of instrument readings. The danger of an operator relying on false instruments is amply indicated by the following passages:
The EPG provides " cautions" End instructions to the operator which are intended to allow the operator to cope with the vessel level. inaccuracies or erratic indications prdduced by the above q mentioned conditions. These instructions do not' permit accurate
(" determination of vessel level but only state that under certain conditions actual. vessel level could be below the lower instrument tap when indicated vessel level is on scale. This leaves the s operator without an accurate indication of vessel level under emergency conditions when he' is following a procedure whose primary objective is restoration.and/or maintenance of acceptable vessel level. (Report by*S. Levy, Inc., SLI-8118, October 1981, p. C-4.)
(I)f the operator fails to recognize that he has lost level indication and has a false high reading of water level, he might take action to throttle or stop ECCS systems in order to avoid filling steam lines or to reduce load on emergency power systems.
In this case, the flashing or boiling in the reference legs could lead to operator actions prejudicial to plant safety . . . . We believe that operator recognition of loss of accurate level information as addrened in the emergency procedure guidelines is cumbersome at best. The operator is to relate indicated water level and drywell temperature using a table contained in a caution statement of the emerge' hey procedures. Indicated water level values beyond the ranges shown in the table are to be mistrusted. Automation of these actions and decisions seems
. in order. (Memo for' Roger J. Mattson from Themis P. Speis dated January 15, 1982, re Errors in BWR Vessel Water Level Indication, pp. 11 and 13.)
RPV water level is a crucial variable, with a sensing system vulnerable to recognized problems-4/ Other plant variables are also For further discussion on the problems with BWR level in-
--4/ strumentation, see " Safety Concern Associated with Reactor (CONTINUED NEXT PAGE)
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depended upon for entry into. and monitcring throughout the EPGs.
These include: RPV pressure, drywell pressure, suppression pool temperature, drywell temperature, containment temperature, suppression pool level, APRM level' (neutron flux) , and control rod position.
According to FSAR Table 7.5.1, some of these j
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Vessel Level Instrumentatiob in Boiling Water Water Reactors" by the NRC's Office'for Analysis and Evaluationi of Operational Data, January 1982. This report. discusses -
possible false high level indications due to leaks in the
'~ reference leg -(a number of such events have occurred; see Appendix A of the AEOD Report) .which can affect control (feedwater flow).and safety systems (high-pressuresECCS) that actuate on various level signals. . BWR vessel water level.is measured by sensing the differential pressure across
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two instrument lines. One line, the. reference leg,'is con-nected to a condensing chamber connected to the reactor steam space and maintains a constant level. The other line, the variable. leg, is a tap connected lower on the RPV. Typically several such taps for different level ranges share-a common reference leg. Decreased level in the reference leg, due to leaks or boil-off*results.in false high RPV level-signals.
False high level indication would reduce feedwater flow into
. the RPV. Vessel water -level would rapidly decrease; when the level reaches the low level setpoint, the operator will be confronted with both high and low level alarms.
This problem'could affect safety systems as well. Aside from the obvious : operator errors likely in this situation, the AEOD report indicates that GDC.24 and IEEE-279 may be violated, such that a false high level indication, coupled with a second random failure, may prevent automatic or timely actuation of safety functions, such as scram, containment isolation, and HPCS and RCIC.
Applicants (January 14', 1983 letter to B. Youngblood) claim that such a scenario. (reference leg break in one division and second
" consequences . random failure in another) would not cause
. of immediate concern." This: conclusion
- of course. assumes correct op'erator action ~s. Applicants also
' ~ have committed.to making design modifications to vessel instrument lines which purportedly reduce the effect of reference leg boil-off and flashing due to high drywell temperature. This case probably represents a worst-case
" common-mode failure" of level instrumentation, since all level indicators,.and signals to safety and control systems, would be falsely high. Even assuming that these mechanical fixes i
I have eliminated.this problem, operator action in the event of a reference leg break which presents conflicting information
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to the operator is still at issue.
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e variables have only two channels of safety-related display instrumentation available in the control room (notably RPV pressure, RPV vessel level, and containment pressure) . This configuration leaves the operator with the task of having to decide which of two instruments is indicating correctly. Some indications, e.g., ECCS, diesel generators, emergency service water system, and combustible gas control, have no red'undant instrumentation for each component or subsystem; the rationale for this is that the systems as a whole have redundancy (e.g.,
2 ESWS loops, 2 diesel generators). This situation could lead an operator to declare an' entire ECCS subsystem, electrical division, or ESWS loop inoperable on the basis of a faulty 6- display, or conversely, to disable.a system by operating it beyond its limits because a hazardou,s or faulty condition was not correctly displayed.
_ There may be non-safety instrumentation monitoring some of these variables, but it is questionable whether these should be relied upon during accident conditions since non-safety equipment is.not environmentally qualified.
The process computer.likewise may not aid the operator.in the diagnosis of instrument failures. FSAR S 7.7.1.8 indicates that the process computer accepts analog signals from plant instrumentation. ktappearsthataninstrumentfailurethat results in an erroneous but plausible indication will merely.
be confirmed by the process computer as a credible reading, thereby adding to operator confusion.
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IV. Human Tendencies Even.the best.e.mergency instructions are rendered useless if they are not followed.by control room operators. Swain and Guttman, in their comprehensive Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications (NUREG/C R-12 7 8, October 1980) point out that precedures are not always followed:
A common error in human reliability analysis is td assume that availhble written procedures will-be used and used properly.
Even in work with severe penalties for nonuse or!. incorrect use of written procedures, nonuse and misuse have been' observed.
In the absense of such penalties, it is reasonable to assume g
a greater frequency of such practices. At every plant Brune g
and Weinstein visited (NUREG/CR-1368, " Development of a Checklist
- for Evaluating Maintenance, Test, and Calibration Procedures" (1980,
at least one manager or supervisor expressed. concern that personnel
- c. might not' use of follow written procedures as intended. The nonuse or misuse of written procedures is obviously related to'the quality of administrative control in a plant . . . .these pr'oblems are also related to the quality of the written materials:themselves.
(p. 14-13) _5/
. Swain and Guttman offer other insights on human reliability in the area of diagnosis of instrument failures. Instrument l
l failures, especially those in which a meter or recorder fails with the pointer reading in the normal operating band of the scale, I are recognized as a contibutor to human error (p.11-4) . Pens of chart recorders are more likely to stick than pointers of panel meters (p. 11-7).
- Human tendencies may override the instructions in the EPGs l
to monitor multiple indications of a variable. Operators have 5
~~/ One would suspect that procedures which involve actions contrary to normal operator response might be less likely to be followed. The BWR emergency procedure guidelines CONTINUED NEXT PAGE
a tendency to fixate on.an anomalous display: In responding to a d'isplay, an operator may focus his attention
,on a particular, display.,to,the exclusion of all others. Often, the operator will initiate the indicated action and concentrate his attention on that display, waiting for a change-in the readout.
This " funneling" of attention is more likely to occur when the operator is under stress. An occasional display malfunction is " sticking": i.e., a
- pointer' on a panel meter. or a. pen on a chart recorder jams for some reason and:no longer yields useful information. Usually there are s'everal redundant displays for any significant parameter, and the operator can. refer to one of them for the required reading .
, .until the primary'displayris repaired. However, there is a strong
, tendency;to focus onijust.one display without cross checking.
'Because of the operator.'s involvement in the corrective' action to be taken, this is most likely to occur when the display sticks in a position indicating-the need for immediate corrective action.
It is less likely to occar when the sticking display does not indicate a need for immediate actionbbecause the.. operator will l be scanning the associated displays as well. When an operator uses an instrument.that has jammed without'any indication to h
% that effect, we estimate aprobability of .1 that -the operator will fail to cross-check until some other'ihdication, such as an alarm, alerts him that something is amis.s. (p. 11-11) c.
an'd : There are several LERs of operators relying on a chart recorder indication and performing inappropriate actions because of a
. stuck pen. In some incidents, operators apparently relied exclusively on one display when other. displays would have indicated
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the appropriate action required. (p. 3-26) l Still other human tendencies may lead to misdiagnosis of instrument failures. E.g.., an operator's first reaction to indications of an event considered unlikely to occur (such as a design-basis LOCA) would be one of sheer disbelief, or the incredulity response. An example of this response (in a different industry) is provided: contain several'such actions: lowering water
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level to top of the active fuel to reduce power in Contingency
#7, Level / Power Control; letting fuel clad temperature increase to 20009F and.then opening the safety / relief valves for steam cooling (more' ef fective .at. higher dlad temperatures)in Con-tingency #3; and terminating injection into the RPV, regardless of whether core cooling is adequate, to preserve containment integrity. Generic Mer . 8 3- 6, pp . 19, 2.2., so, 3 9 .
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I , i (I)n one refinery- the first indication the control room operator had of a serious fire.was that.many alarms occurred and many instruments behaved abnormally. This* operator's first response i . was to run, upstairs;4nd demand of an instrument technician, "What's wrong with my . instruments?" By..the time he returned to the control room, it.was-too late to take action that might have reduced the
- loss due - to the' fire., (p. 3-58) l The incredulity. response may be reinforced by'the perception that,many;actuations of safety-related alarms in nuclear power
- plants .are ' f alse ;~(p..:10-17) .
There also may be.a bias in the. u,. [ , . ~ operator's reaction to true alarms and instrument fail.ures. Swain and~; Guttman note: Operating personnel have two primary responsibilities: keeping the plant on-line.and ensuring".its safe operation. Because serious safety. problems rarely occur.at a plant, most f-of the operator's attention is directed to the-instruments and controls related to the first responsibility. Interviews with operators indicate that they do not expect. serious safety. problems. Furthermore, they. have confidence in -the ability of the plant's safety systems to cope with possible safety problems automatically. These attitudes, coupled with the usual lack of pra^ctice in
-d'ealing with the unexpected, may result. in reluctance on the part of an operator to take action that would interfere with keeping
- the plant on-line. (p . 10-16 )
This bias might lead an. operator to view alarms and instruments indicating serious safety problems as false alarms, while in-l
- terpreting false instrument readings that.would threaten con-i tinued plant operation (such as a false.high RPV water level l
indication.of incipient steamline flooding)'as true. Such bias
~would tend to decrease plant safety.
f , Swain and Gut $ man are pessimistic about nuclear power' plant operators' capabilities for coping with serious accidents, largely
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due to the stress and confusion.in such situations (p. 10-23). A contributor to this stress is difficulty in diagnosing the emergency (p . 17-10). The-operator's ability to cope with i emergencies is dependent on his skill level and. degree of famil-
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e iarity with the unusual condition, and.the " extent to which
, the displayed information directly supports the actions the operator should take to cope with the situation." (p. 17-9)
It would thus seem that. simulator training in diagnosing instrument failures would.be an extremely desirable and important requirement for operator training. Staff and Applicant affidavits do not clearly indicate the adequacy or extent of this type of training, nor.4.are: the.-BWR emergency procedure guidelines suf ficient to preclude operater misdiagnosis.of instrument readings. OCRE therefore would support the Board's declaration.of a sua sponte issue on this matter. V. Standards for.Sua Sponte Issues , ,6 A Licensing Board has.the power to raise sua sponte any significant environmental or safety issue in operating license hearings. Consolidated Edison of New York (Indian Point, Units 1, 2, and 3),'ALAB-319, 3 NRC 188, 190 (1976); consolidated l Edison of New York. (Indian . Point, Unit.3), CLI-74-28, 8 AEC 7 (1974). l 10 CFR 2.760a provides that in an OL proceeding-"(m)atters not put into controversy by the parties will be examined and decided by the presiding officer only wheye he or she determines that a serious safety, environmental, or common defense and security matter exists." The TMI-2 accident vividly demonstrated that human error (much of which resulted from misdiagnosis of instrument readings) can rapidly exacerbate plant transients, leading to core damage , l and potent'ially! devastating public health consequences. It should l be noted that in " Precursors to Potential Severe Core Damage
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Accidents: 1969-1979; A Status Report" (NUREG/CR-249 7, June 19 82) 36% of all the precursors, and 38% of the significant precursors, were found to involve human error (p. 5-18). Based on these facts and on:the above analysis, OCRE would. conclude that the training 6f operators in the diagnosis of instrument failures is indeed a serious safety matter. It furthermore cannet be concluded that the~ Commission's '
' initiatives in the human factors area since TMI will effectively reduce these problems. A General Accounting Office report,
" Problems and Delays Overshadow NRC's Initial Success in Improving p Reactor Operators' Capabilities" (GAO/RCED-83-4,. December 15, 1982) indicates that, although the NRC has made progress in short-term 6 actions to improve operators' abilities to cope with accidents, the NRC's efforts with respect to long-term actions have lost momentum and have been subject to delays.. Simulator training has been designated as a long-term issue. See NUREG-0933, "Prioritization^of Generic Safety Issues", wherein all the TMI Action Plan tasks relating to simulator training are considered
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covered by Item I.A.4.2, " Training Simulator Improvements - Long Term." Given the NRC's failure to effectively' address long-term actions, the Commission cannot be relied upon to resolve this important problem. Plant-specific resolution therefore becomes imperative to assure that the safe operation of~ Perry will not be compromised by human errors resulting from inadequate training. OCRE thus urges the Board to declare a sua sponte issue regarding this matter. 1 Respectfully submitted,
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CERTIFICATE OF SERVICE- h AUG This is to certify that copies of the foregoing kere s$h,ved 5 93 by deposit in the U.S. Mail, first class', postage prepaid, iq$'f,eerr 2nd. day of August, 1983 to those on the servic,e list below. Utoi:T 4 su y,,l e 1) g Susan L. Hiatt OCRE Repfesentative
'8275.Munson Rd.
Mentor, OH 44060 (216) 255-3158 SERVICE LIST D Terry Lodge, Esq. Peter B. Bloch, Chairman Atomic Safety & Licensing Board McCormack, Pommeranz, & U.S. Nuclear Regulatory Comm. . Lodge n 824 National Bank Bldg. Washington,'D.C. 20555 Toledo, OH 43604 Dr. Jerry R..Kline . Atomic Safety,& Licensing Board U.S.._ Nuclear Regulatory Commission Washington, D.C. 20555 Mr. Glenn O. Bright Atomic Safety & Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 James R. Cutchin, IV, Esq. Office of the~ Executive Legal Director U.S. Nuclear Regulatory Commission Washington,' D.C. 20555 Jay Silberg, Esq. l Shaw, Pittman, Potts, & Trowbridge 1800 M Street, NW Washington, D.C. 20036 l Docketing'& Service Branch Office of'the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety.& Licensing Appeal Board Panel , U.S. Nuclear Regulatory Commission Washington, D.C. 20555 . 1
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