ML20083K463

Application to Amend Licenses DPR-53 & DPR-69.Amends Revise Tech Specs to Reflect Minor Title & Responsibility Changes & Incorporate Containment high-range Radiation Monitors,Per Generic Ltr 83-37

ML20083K463
Person / Time
Site:	Calvert Cliffs
Issue date:	04/09/1984
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20083K466	List: ML20083K463 ML20083K484
References
GL-83-37, NUDOCS 8404160182
Download: ML20083K463 (11)

Text

..

r BALTIMORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 Antwun E. LUNDVALL,Jn.

April 9,1984 vice ParslOENT Su ppty U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 ATTENTION:

Mr. James R. Miller, Chief Operating Reactors Branch #3

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No.1 & 2, Docket Nos. 50-317 & 50-318 Hequest for Amendment R EFERENCES: (a)

BG&E letter from A. E. Lundvall, Jr., to 3. R. Miller, dated September 16,1983 (b)

BG&E letter from A. E. Lundvall, Jr., to 3. R. Miller dated November 10,1983 (c)

NRC letter from D. H. Jaffe to A. E. Lundvall, Jr., dated November 17, 1983, Amendment No. 88 to Facility Operating License No. DPR-53 Gentlemen:

The Baltimore Gas & Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos.1 & 2, respectively, with the submittal of the enclosed proposed changes to the Technical Specifications.

CHANGE NO. l_

(BG&E FCR 84-33)

Remove existing pages 6-2 and 6-3 of the Unit I and 2 Technical Specifications, and page 2.3-8,5.1-1,5.3-1,5.5-1 and Figure 5.2-1 of the Appendix B Technical Specifications and replace with attached, marked up pages, Attachment (1) to this transmittal.

DETERMINATION OF SIGNIFICANT HAZARDS CONSIDER ATIONS The proposcd changes are needed to reflect minor title and responsibility changes for certain supervisors within the Baltimore Gas and Electric Company organization. The minor title changes do not affect the responsibilities as described in Chapter 12 of the Updated FSAR or the Technical Specifications.

The Fire Prevention Unit, which originally reported to the General Supervisor-Finance, now reports to the Manager-Real Estate and Office Services.

Because of this organizational responsibility shif t, the corporate responsibility for the Fire Protection Program is now assigned to the Vice President-General Services.

g K W m!r W (lAa M l

0404160102 040409 g.y

~~

PDR ADOCK 05000317 P

PDR ggg

\\1r. James R. \\1 iller

~

April 9,1984 Page 2 I

E E

Since the administrative changes to the Environmental Tecnnical specifications do not authorize any release of e f fluents not currently authorized and do not affect responsibilities described in the Environmental Technical Specifications, they will have no adverse af tects on the Environmental Impact Appraisal. A cost-benefit study also is not necessary since the proposed changes represent no cost other than the f ee f or the cense Amendment.

O The proposed cnanges to Tables 6.2-1 and 6.2-2 are adininistrative in nature, since they are needed to more accurately reflect the current organizational structure within the r

ikitimore Gas and Electric Company.

As sucn, the proposed changes have been 4-

?

determined to involve no signiticant hazards considerations, in that operation of the E

f acility in accordance with the proposed amendment would not:

(i)

Involve any increase in the probability or consequences of an accident previously evaluated; or e

I (ii) create the possibthty of a new or dif ferent kind of accident f rom any accident previously evaluated; or (iii) involve any reduction in the margin of safety.

k CHANGE NO. 2 (BG&E FCR 82-06)

Remove existing pages 3/4 3-16 through 3/4 3-28 of the linits Nos. I and 2 Technkal Specifications, and replace with attached, iaarked up pages, Attachment (2) to this transmittal.

u x,

DETEllMINATION OF SIGNIFICANT HAZARDS CONSIDER ATIONS

/

y h

This proposed change to the Technical Specifications is being processed as requested by NRC Generic Letter 33-37. These proposed changes will incorporate the containment high range radiation monitors and the main vent wide range noble gas ef fluent monitors 6

{

into the Technical Specifications. Although every attempt has been made to follow the i

guidance Technical Specifications in the generic letter, some minor changes were l

necessary to make the surveillince requirements consistent with the f acility design.

A channel check is not useful f or the containment high range radiation monitors due to their locations in the containment being dif ferent. Although post accident radiation i

levels would be expected to be approximately equal on the radiation monitors, the channels indicate dif ferent levels during normal operation. The main vent wide range ef fluent monitors only have one channel of installed instrumentation per unit. Therefore, a channel check is not possible because no equivalent instrument is available for such a surveillance check.

y With the exception of the channei check we have followed the guidance Technical Specifications forwarded in the Generic Letter.

v L

IE

Mr. James R. Miller April 9,1984 Page 3 This proposed change constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications.

As such, the proposed change conforms to an example of an amendment considered not likely to involve significant hazards considerations, item (i) as provided on page 14870 of the Federal Register Notice dated April 6,1983.

The proposed change will not:

(i) involve any increase in the probability or consequences of an accident previously evaluated; or (ll) create the possibility of a new or different kind of accident from any accident previously evaluated; or (111) since the changes provide substantial assurance that equipment which provides useful information to the operators in post accident situations as to the extent of the accident, it may actually result in an increase in the margin of safety.

Based upon the above, this proposed change has been determined to involve no significant hazards considerations.

CHANGE NO. 3 (BG&E FCR No. 83-1042, Unit 2 Only)

Delete the following pages of the Unit Two Technical Specifications and add new pages as indicated (proposed replacement pages are attached):

(a) delete page 7-5 and add new page 7-5 (b) delete page 7-Sa and add new page 7-5a (c) delete page B7-2 and add new page B7-2 The purpose of this change is to reflect final modifications to the Unit Two Auxillary Feedwater System as they pertain to completion of the cross-connect between the Unit One and Unit Two motor-driven pump trains. This change would bring the Unit Two Auxiliary Feedwater System Technical Specifications (TS 3/4.7.1.2)into agreement with the existing Unit One Technical Specifications.

BACKGROUND in Reference (a) we proposed changes to the Unit One Technical Specification, (TS.3/4.7.1.2) to reflect the completion of Auxiliary Feedwater System (AFWS) modifications including the addition of a motor-driven pump train and a cross-connect between the discharge piping of the Unit One and Unit Two motor-driven pumps.

Reference (a) indicated our intent to request similar Technical Speci!! cation changes for Unit Two at a later date.

4_.

Mr. James R. Miller April 9,1984 Page 4 Reference (b) modified our initial license amendment request with regard to the surveillance requirements for flow verification (TS 4.7.1.2.c) and the discussion of flow setpoints and flow setting error bands in the Bases (B 3/4.7.1.2).

In accordance with the above, Technical Specifications are now requested for Unit Two to achieve consistency with the existing Unit One Technical Specifications and to credit the motor-driven train cross-connect. The proposed Unit Two Technical Specifications are identical to those for Unit One as approved by Reference (c) with the exception that the LCO footnote pertaining to the initial 30-day test period is no longer applicable and has been deleted. No other changes to the LCO definition are proposed.

A two-part remedial action statement (TS 3.7.1.2.a.1) is proposed for an inoperable motor-driven pump. The existing specification requires that an inoperable motor-driven pump be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed specification would extend this parlod to a maximum of 14 days by taking credit for the continued ability to align two steam-driven AFWS pumps to automatic initiating status.

In the event that any two of the three AFWS pumps were to become inoperable at the same time, a new remedial action statement (TS 3.7.1.2.b) is being added that would require the operators to verify within one hour that the remaining pump is aligned for

automatic initiation and that the cross-connect between the Unit One and Unit Two motor-driven trains is operable and capable of delivering AFW flow to the affected unit upon manual initiation. In addition, the operators would be required to restore a second pump to automatic initiating status within 72 ' hours. If these actions could not be succesfully completed,'the Operators would be required to place the unit in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

To ' ensure that plant. operational flexibility!!s not unnecessarily restricted in the event

.that a motor-driven or a single steam-driven pump becomes inoperable, we propose a

" statement (TS 3.7.1.2.d) that deletes applicability of TS 3.0.4 as. long as any two Auxiliary Feedwater trains can be aligned for automatic initiation. This provision wi!!

ensure that the intent of the LCO will continue to be satisfied.

.Two changes to the' Surve111ance. Requirements of TS 4.7.1.2 are proposed.. First is Jdeletion of the requ!rement toi verify a-160 'gpmisetpoint on the controller.for the automatic flow control valves (TS 4.7.1.2.c.4). Licensing grade' transient analyses have'

'. Indicated that the quantity of flow! automatically' delivered,to the. steam generators -

a

during the

first 10. minutes, a maximum flow; of 1300 gpm' (runout),,during the worst

overcooling transient-or a minimum flow of 0 gpm (zero), during the worst undercoollng -

. transient is. acceptable.- Nevertheless, the automatic flow setmint is being identified in c

1 e the TS Bases in recognition of the AFW system's functional design basis.

~

f 7p-

x 8

. Y

- ~

t z

I*

'I.^-

6 r

t 5 -:

m m

m.

m

.m.

m-..

m

r.

Mr. James R. Miller April 9,1984 Page 5 The second proposed change is a clarification of TS 4.7.1.2.c for testing the automatic functions of the AFWS. The new specification would require that all automatic valves in the flow path actuate to their correct positions and each Auxiliary Feedwater pump automatically starts upon receipt of each AFAS test signal. Consistent with the above

- discussion concerning the quantity of flow automatically delivered by the AFWS, the requirement to verify a modulated flow of 160 gpm 10 gpm is deleted. To ensure that the system can deliver a minimum specified flowrate a new requirement is added (TS 4.7.1.2.c.2) that provides for a demonstration that 200 gpm nominal flow can be delivered to each flow leg. The purpose of this test is to ensure that no flow path degradation has occurred (e.g., obstructions in the line) during the surveillance interval.

The test will be performed by manually aligning each of the flow legs and individually verifying their capability to pass 200 gpm flow.

Finally, the Bases for TS 3/4 7.1.2 are revised to include a new nominal flow setpoint and instrument loop error band in accordance with Reference (b).

The transients affected by AFWS performance, and the assumptions used the analysis of these transients, are being listed in the basis for TS 3/4.7.1.2. Licensing grade analyses have demonstrated that ~ for the first ten minutes no flow is needed for undercooling transients, and that the maximum Auxiliary Feedwater suction leg flow is acceptable for

- overcooling transients. Although the operational nominal flow setpoint of.200 gpm is discussed, it is made clear that flow fluctuations beyond the discussed band are allowable.

The new interim nominal flow setpoint does not represent a limiting safety system setting required to maintain the assumptions of the FSAR. This setting is defined only to support enhanced equipment reliability and prudent operations to ensure that the effects -

- of the following abnormal operating events are mitigated:

1..

AFAS actuation concurrent with a low steam generator backpressure condition will not result in motor-driven AFW pump runout.

'2.

AFAS actuation concurrent with a plant trip that does not involve an excessive cooldown transient, will not result in excessive cooldown (with ~ subsequent safety injection actuation) -within' the - first ten minutes of assumed no operator action.

The ' nominal' flow ;setpoint and ~ associated error. band ' will be finalized ' during' the..

Unit Two refueling outage scheduled to begin in April _'1984. Modifications are i

~

upcom ng planned :ior the automatic flow control features of. the motor-driven pump train to improve ! system performancel during lowL steam generator backpressure conditions.)

Following completion _ of these modifications, testing will be performed to identify the '

final setpoint. The value of this setpoint will then be submitted as a supplement to this

- request.-

J d

.d.'

a.

2

..... i.

m.m-M

)

J

.Mr. James R. Miller April 9,1984

. Page 6 DETERMINATION OF SIGNIFICANT HAZARDS CONSIDER ATIONS Our review of the changes requested by this application indicates that, with the following exception, no significant hazards considerations are involved. By strict application of

'the criteria contained in '10 CFR 50.92, the proposed change to TS 3.7.1.2.a must be treated as a significant hazards consideration in that it represents a relaxation in the action statement for restoring an inoperable motor-driven pump to service (14 days versus 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />)in the absence of compensating system improvements.

~

The subject relaxation is justified because the proposed Technical Specification requires that the standby steam-driven pump be aligned to automatic initiating status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, thereby satisfying AFWS design basis requirements by ensuring the availability of two pump trains.

This specification was shown to be acceptable for Unit One in Reference (c).

In all' other respecti the proposed. changes 'do not involve a significant hazards consideration in that they more accurately reflect the manner in which the Unit Two AFWS, with the motor-driven train and Unit One cross-connect, will be operated and tested in the future. These changes do nots (1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or i

(ll)

Create the possibility.ci~a new.or> different kind of accident from any accident previously evaluated; or

(111)

Involve a significant reduction in the margin of safety currently.

provided by the Technical Specifications.

QHANGE NO. 4 (BG&E FCR 84-1028) '

. Remove old page 5.6-3 and replace with new page 5.6-3 (see attached mark-up). This.

change modifies, the reporting criteria for measured level of radioactivity in offsite -

! environmental media.

~

This request for amendment is made pursuant to the requirements of 10 CFR 50.91 and 50.92 and the provisions of TS.5.6.2.6 as it relates to changes in the. Environmental-

[

Technical Specifications. ;

DOCUMEIN

~ ^

c-Environmental Technical Specification 5.6.2.b, Non-Routine Radiological Environmental Operating Reports, currently states in part l

r

~, '

'"...'If a ' confirmed level 'of radioactivity at any off-site location In any environmental medium. exceeds' ten times the " background" value, a written report shall be submitted to the Director of the NRC Regional _ Office..." ~

u s

g

~

s f:

4 g

c.%

y s

4

,..y

.e-,,

y y.g m,e,.,

.y 9-yy e

,q.e

, g,,, +.

Mr. James R. Miller April 9,1984 Page 7 At the time this Technical Specification was originally included in the operating license, the~ criteria of " ten times the background value" was considered appropriate given the then existing state-of-the-art in radiation detection and radioisotope measurement.

Since the issuance of the operating license, radiation measurement technology has improved substantially, permitting the resolution, and the measurement of individual radioisotopes with far greater accuracy and sensitivity than has been possible in the

^

past.

To some extent, improvements in energy resolution have been manifested as apparent increases in the measured levels of individual isotopes.

In addition, improvements in instrumentation have resulted in the elimination of noise which in the past was a rather large component of the measured " background" level.

The combined effects on increased energy resolution for individual isotopes and the reduction of background noise as afforded by modern instrumentation has had the net effect of significantly increasing the difference between naturally occurring background

,c levels of radioisotopes and the measurable levels or radioisotopes in environmental media attributable to the operation of Calvert Cliffs.

For example, Ag-110m levels measured in oysters collected at the Camp Canoy sampling location during the past three years have been consistently higher than the " ten times

background" criteria.. This has occurred even though the concentration of Ag-110m in liquid effluents has always been _ a negligible fraction of that allowed.by 10 CFR 20,

' Appendix B.

As a result, we have had to submit Non-Routine Radiological Reports pursuant to

' TS 5.6.2.6 on all but 'a routine basis. We do not believe that these reports, or 'the

~

associated Licensee Event Reports required by TS 6.9.1, reflect the original intent of the Environmental Technical Specifications.' As discussed in each of the said reports,' the measured levels of-Ag-110m have been well.below those which could result in a whole

body oriorgan dose. commitment approaching the design objectives of 10 CFR '50,

. Appendix I, or limits specified in 40 CFR 190.- Thus, these reports are unnecessary since -

1 they identify neither an unsafe nor potentially unsafe condition in the environs.

The proposed change would require a report to be' submitted 'only when the projected annual ! dose commitment, calculated _ in accordance' with ' Regulatory Guide 1.109,'

Revision 1,-was equal to or. greater than one mrem for a maximally exposed individual..

~'*

This level is consistent 'with the limits specified in 40 CFR 190 for annual doses received g'

by individuals from_ all nuclear power cycle pathways and is conservative with respect to _

the design objectives specified in 10 CFR.50, Appendix 1..

4 5

4 2

^

1

Mr. James R. Miller April 9,1984 Page'8 t'

- DETERMINATON OF SIGNIFICANT HAZARDS CONSIDER ATIONS The proposed change does not represent a significant hazards consideration inasmuch as:

(1) an increase in the probability or consequences of an accident previously evaluated is not involved, (11) the possibility of a new or different kind of accident is not created, j

and

' (iii) the margin of safety of plant operations is not reduced.

The existing requirement to evaluate the cause of any radioisotope which is detected at a concentration equal to ten times the background level would not be affected by this proposal.

CHANGE NO. 5 -(BG&E FCR 84-35)

Change surveillance requirement 4.7.11.1.1.b and 4.7.ll.l.2.a.2 as shown on the attached

- marked-up pages 3/4 7-67 and 7-68 for Unit I and pages 3/4 7-59 and 7-60 for Unit 2 Technical Specifications.

. DETERMINATION OF SIGNIFICANT HAZARDS CONSIDER ATIONS

~

Water for the fire protection system!!s supplied by two full capacity fire pumps. One pump is electrically-driven and the other is diesel engine-driven. ' The two fire pumps are designed to start automatically and stop manually. The electrically driven pump starts automatically on a low, header pressure of 95 psig with the diesel engine-driven pump being started at 85 psig. The diesel engine-driven pump is arranged to. provide backup for the~. electrically-driven pump in case the latter does not start or does not maintain adeqate pressure at the header.

The fire pumps are demonstrated operable in accordance with Surveillance Requirements 4.7.11.1.1 and '4.7.11.1.2.

The : electric and -diesel-driven ' fire pumps are tested on a-131. day. staggered test basis. : Section 4.7.ll.l.2.a.2 specifies that the diesel driven fire

- pump must operate for 30 minutes on recirculation flow. Section'4.7.ll.1.1.b specifies that the electric-driven fire pump must operate for 15 minutes on recirculation flow.

The requirement ' to test the _ pumps ~ in. the ' recirculktion mode,is an unnecessary

. restriction. The test can also be performed in the' normal standby line-up. Relief valves are provided at each fire pumps' discharge to meet or exceed the National Fire Codes.

t r

---n

Mr. James R. Miller April 9,1984 Page 9 i

Circulation Relief Valve. Each pump shall be provided with an automatic relief valve set below the shutoff pressure at minimum expected suction pressure. It shall provide circulation of sufficient w from overheating when operating with no discharge.pter to prevent the pump To supplement the Surveillance Test Procedures (STPs) performed to meet the Surveillance Requirements, Preventative Maintenance (PM) testing is performed to meet the plant's weekly test requirement. The PM testing is performed in the normal standby line-up. The maintenance history indicates that testing the fire pumps in the standby line-up has not adversely affected either fire pump.

The operation of the facility in accordance with the proposed amendment to the operating license would not-l (i) involve an increase in the probability or consequences of an accident previously evaluated; or (11) create the possibility of a new or different kind of accident from any accident previously evaluated; or (iii) involve a reduction in a margin of safety.

The proposed amendment does not change the intent or conclusions of the surveillance test. The ability to start and run the fire pumps is equally valid in the standby mode.

The performance of the fire pumps is tested in another section,4.7.11.1.1 f.

CHANGE NO. 6 (BG&E FCR 84-34)

Change Tables 3.3-11, Fire Detection Instruments, to include the additional instrumentation as shown on the attached marked-up pages 3/4 3-45 and 3-46 for the Units 1 and 2 Technical Specifications.

DETERMINATION OF SIGNIFICANT HAZARDS CONSIDER ATION The proposed change incorporates additional' fire detection instrumentation in the Technical Specifications. The rooms added to Table 3.3-11 are now equipped with heat, flame and/or smoke detectors, and an alarm system. The annunciators in the control room provide an audio-visual alarm which indicates the location of the affected area.

I

~

National Fire Codes, Volume 2, Chapter 20, Section 2-6,1983 i

M_r. James R. Miller April 9,1984 Page 10 In the Federal Register dated April 6,1983, the Nuclear Regulatory Commission provided guidance for license amendments which were likely or not likely to involve Significant Hazards Considerations. As an example of amendments that are considered not likely to

-involve Significant Hazards Considerations the Federal Register states:

...A change that constitutes an additional limitation, restriction or control not presently included in the Technical Specifications".

- The proposed change incorporating the recently installed fire detection instrumentation in the Technical Specifications constitutes additional control not previously required.

The installation of additional fire detectors improves the ability of Operators to recognize the existence and location of a fire. This information aids the Fire Brigade in containing and extinguishing a fire.

The operation of the facility in accordance with the proposed amendment to the operating license would not

' (i) involve an increase in the probability or consequences of an accident previously evaluated; or '

(11)-

create the possibility of a new or different kind of accident from any accident previously evaluated; or (111) involve a reduction in the margin of safety.

CHANGE NO.7 ~(BG&E FCR 82-169)

Remove existin6 Page 6-21 of the Unit I and 2 Technical Specifications and replace with the marked-up page 6-21.~

DETERMINATION OF SIGNIFICANT HAZARD 5 CONSIDER ATION Unnecessary exposure to high radiation areas is administratively controlled as required F

-by 10 CPR 'Part 20.203(c)(2) in lieu of " control devices" or an " audible alarm." High radiation areas are defined in 10 CPR Part 20.202(b)(3) as accessible areas where the radiation imparts a whole body dose rate greater than 100 mrem /hr. ' Positive controlis

'further enhanced by locking and barricading high radhtlon areas with a dose rate greater--

then 1,000 mrem /hr.

Presently, the Shift Supervisor on duty is authorized to administratively control access into the locked high radiation areas. The proposed change would provide the Radiation Control Supervisor, in addition to the Shift Supervisor, the authority to administratively -

control access into high radiation areas greater than 1,000 mrem /hr. This change is also consistent 'with the ALARA program, since the Radiation Control Supervisor and the Radiation Control Unit are most familiar with the appropriate radiological protection.

V

Mr. James R. Miller April 9,1984 Page 11 The proposed change constitutes an administrative change and does not involve a significant hazards consideration. Expanding the authority to control access to locked high radiation areas will not increase the probability or consequences of an accident previously evaluated. Typically, the few changes that affect dose rates (e.g., mode changes, fuel movement, ion exchanger line-ups, etc.) are internally communicated to the Radiation Control Unit. Similarly, the Radiation Control Unit informs Operations of non-routine entries into locked high radiation areas. A new or different kind of accident from any accident previously evaluated is not possible since the locks and radiation sources have not been altered. The margin of safety will not be reduced, since the Radiation Control Supervisor also can administer the keys to the locked high radiation areas and ensure control over unnecessary exposure.

SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.

FEE DETERMINATIOf(

We have determined, pursuant to 10 CFR 170.22, that this Amendment request consists of one Class IV and one Class 1 Amendment. Accordingly, we are including BG&E Check Number All7632 in the amount of $12,700 to cover the fee for this request. This check also forwards the fee required for the April 4,1984, submittal.

Very truly ours j.

MotW v STATE OF MARYLAND :

TO WIT:

CITY OF BALTIMORE Arthur E. Lundvall, Jr., being duly. sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland that he i

provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation.

WITNESS my Hand and Notarial Seals h/E#de b CW l

Notary Public My Commission Expires dw //,/fb

'AEL/LES/gla Attachments i

~

Mr. James R. Miller

' April 9,1984 Page 12 cc:

D. A. Brune, Esquire G. F. Trowbridge, Esquire D. H. Jaffe, NRC T. Foley, NRC T. Magette, DNR, State of MD i --

V

'Y

____--______________-.______________-______y