ML20205G535
| ML20205G535 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 10/25/1985 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20205G540 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.4.2, TASK-TM NUDOCS 8511130278 | |
| Download: ML20205G535 (14) | |
Text
i s BALTIMORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL. JR.
Vict PRESIDENT SuPPLv October 25,1985 Office of Nuclear Rt. actor Regulation U. S. Nuclear Regulatory Commission Washindton, DC 20555 ATTENTION: Mr. A. Thadani, Project Director PWR Project Directorate #8 Division of PWR Licensing B
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Amendment
REFERENCES:
(a) BG&E letter from Mr. A. E. Lundvall, Jr., to Mr. 3. R. Miller, dated February 22,1985 (b) BG&E letter from Mr. A. E. Lundvall, Jr., to Mr. R. A. Clark, dated October 16,1981 (c) BG&E letter from Mr. A. E. Lundvall, Jr., to Mr. R. A. Clark, dated October 26,1981 (d) Letter from Mr. R. A. Clark, NRC, to Mr. A. E. Lundvall, Jr.,
BG&E, dated November 6,1981, Safety Evaluation (e) Letter from Mr. R. A. Clark, NRC, to Mr. A. E. Lendvall, Jr.,
BG&E, dated July 30,1982, Safety Evaluation (f) Letter from Mr. R. A. Clark, NRC, to Mr. A. E. Lundvall, Jr.,
May 5,1983, Safety Evaluation (g) Letter from Mr. D. H. Jaffe, NRC, to Mr. A. E. Lundvall, Jr.,
BG&E, dated April 19,1984, Safety Evaluation Gentlemen:
The Baltimore Gas and Electric Company hereby requests an Amendment to its I Operatiag License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos.1 & 2, respectively, with the submittal of the proposed changes to the Technical Specifications.
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i Mr. A.Thadani October 25,1985 Page 2 CHANGE NO.1 (BG&E FCR 82-36)
Change page B 2-5 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
DISCUSSION The Technical Specification bases, Section 2.0, identify the design bases for providing a containment pressure signal input to the Reactor Protective System. The presently worded containment pressure-high bases are ambiguous in referring to the reactor trip setpoint initiation as being concurrent with actuation of safety injection.
NUREG-0737,Section II.E.4.2, required reducing containment pressure setpoints to minimum - values compatible with normal operation. In compliance with that requirement, the Reactor Protective System containment pressure trip was adjusted to a value of less than 2.6 psig and the Containment Isolation Signal (CIS)/ Safety Injection Actuation Signal (SIAS) trip was adjusted to a value of less than 3.0 psig. This proposed change clarifies the Technical Specification Bases to reflect actual system operation. A future revision to the Updated FSAR, Section 7.2.3.9, will reflect this clarification.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed charige has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of any accident previously evaluated; or No previously analyzed accident in the Updated FSAR is affected by this proposed change, since it only provides a clarification of the existing Technical Specification Bases.
(ii) create the possiblity of a new or different type of accident from any accident previously evaluated; or No new accidents previously unanalyzed will be created by this proposed change, since no modifications to the intent of the Technical Specifications are being made.
(iii) involve a significant reduction in a margin of safety.
This change documents a conservative, post-TMI reduction in containment pressure setpoints. The margin of safety in the bases is not reduced since the setpoints are now maintained closer to their trip setting.
s Mr. A.Thadani October 25,1985
- Page 3 CHANGE NO. 2 (BG&E FCR 85-64)
Change page 3/41-6 of the Unit 2 Technical Specifications as shown on the marl:ed-up copies attached to this transmittal.
DISCUSSION This proposal would change the Technical Specification 4.1.1.4.2.c which requires calculation of the Moduator Temperature. Coefficient (MTC) within seven days after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm. The change would allow the MTC to be calculated within seven days before to seven days after reaching RATED THERMAL POWER equilibrium boron concentration of 300 ppm.
This change would provide greater flexibility in the performance of the surveillance and has previously been requested for Unit I as a part of reference (a).
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposal has been reviewed against the standards set forth in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or (ii) create the possibility of a new or different kind of accident from any accident previously evaluated; or (iii) involve a significant reduction in a margin of safety.
This proposal would not affect the validity of the MTC calculation with regard to its function as a confirmation of the assumptions used in the accident and transient analyses for the fuel cycle. Therefore, the probability or consequences of an accident previously evaluated would not be increased and no margin of safety would be reduced.
This proposal would not modify equipment or change system alignments. The procedural change involved would not create the possibility of a new or different accident.
CHANGE NO. 3 (BG&E FCR 85-3000, Supplement 2)
Change page 3/41-11 of the Unit 1 Technical Specifications as shown on the marked-up copy attached to this transmittal.
. Mr. A.Thadani October 25,1985 Page 4 DISCUSSION The NRC recently issued Unit 2 High Pressure Safety Injection (HPSI) flow balance and associated boration systems Technical Specification revisions as part of Unit 2 License Amendment No. 89. NRC review of the proposed changes required an additional revision to Limiting Condition for Operation 3.1.2.4. Specifically, it required that the two operable charging pumps have independent power supplies above 80% power. This change ensures that at least one charging pump would be available to deliver flow following a Safety Injection Actuation Signal as currently assumed in the Small Break LOCA analysis.' Charging flow is not essential for acceptable Small Break LOCA results below 80% power; hence, the Technical Specification imposes the requirement when operating above this power. Power supply independence was not part of the similar changes made
-for Unit 1, but is standard operating practice. The proposed change formalizes this practice and provides consistency between the requirements for the two Calvert Cliffs units.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change to 'the Technical Specifications has been reviewed against the standards in 10 CFR 50.92 regarding significant hazard considerations and has been determined to involve no significant hazards considerations, in that operation in accordance with the proposed change would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated, or No modification to plant equipment is made by this proposed change. The change formalizes current practice to ensure minimum assumed equipment performance for safety analy*is.
(ii) create the possibility of a new or different kind of accident from any accident previously evaluated, or No modification of plant equipment is made by this proposed change. Therefore, no new accident scenario is created.
(iii) involve a significant reduction in a margin of safety.
No accidents are adversely affected by the proposed change.
The proposed change formalizes current operating practices and imposes an additional control not presently included in the Technical Specifications. Therefore, the proposed Technical Specification change has been determined to involve no significant hazards.
- . Mr. A.Thadanil October 25,1985
.Page5 r
CHANGE NO. 4 (BG&E FCR 83-007)
Change pages 3/4' 3-40-and 3/4 3-41 and add page 3/4 3-41a of Unit I and 2 Technical
. Specifications as shown on the marked up copies attached to this transmittal.
DESCUSSION The NRC published guidance Technical Specifications in Generic Letter 83-37. Upon
' detailed review' of the Calvert Cliffs' installation of Containment Water Level Instruments, these guidance Technical Specifications were deemed to be too restrictive.
They require two channels as the required number of channels and one channel as the
' minimum channels OPERABLE. 'If BG&E adopted these Technical Specifications, a single
- transmitter failure would require a full unit shutdown to reduce dose' rates to levels acceptable for obtaining access to the transmitter for trouble shooting and repair. This is due to the neutron streaming and high gamma radiation at the transmitter's location on the 20' 9" elevation in the containment.' Discussions with NRC staff personnel have concluded
- that unrestricted operation with one channel INOPERABLE would be ;
perrnissible provided the INOPERABLE channel is restored to OPERABLE status at the next outage of sufficient duration to facilitate repair.
Other instrumentation, such as subcooled margin monitor, pressurizer water level, and refueling water tank water level, prevides an alternate means of' monitoring free liquid inventory in the containment building ouring an accident. The Containment Water Level Instrumentation provides corroborative information to the operator to assess the current status of an accident which is progressing. It does not provide any automatic function, nor does its operation contribute to assumptions in any accident analyzed in the Updated FSAR for Calvert Cliffs.
DETERMINATION OF SIGNIFICANT HAZAADS CONSIDERATIONS
- The proposed Technical Specification change has been reviewed against the standards provided in 10. CFR 50.92 regarding significant hazards considerations. This change constitutes an additional restriction, limitation, and condition not currently included in the Technical Specifications. -Also, the, Containment Wide Range Water Level Instruments provide only corroborative information to assist the operator in assessing the current status of an accident in progress. As such, the Technical Specification would not create the possibility of a new or different type of accident from any accident previously evaluated. Because a new. restriction, limitation, and condition is being added to the 1 Technical Specifications, no reduction in the margin of safety would result. Therefore, the proposed Technical -Specification change has been determined to involve no significant hazards considerations.
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. Mr. A.Thadani October 25,1985 Page 6 ir CHANGE NO. 5 (BG&E FCR 82-138, Supplement 1)
Change pages 3/4 4-5 and 3/4 6-1 of the Unit I and 2 Technical Specifications as shown on the marked-up pages attached to this transmittal.
~ DISCUSSION The proposed change would correct one typographical and one syntax error. The proposed corrections are administrative in nature and do not affect plant operation or safety.
The proposed changes are described below:
Page- Change 3/44-5 . . 25 gpm maximum pressurizer level . . ." to ". . 25 gpm pressurizer level . . ."
3/46-1 "OEPRATION" to " OPERATION" DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change has been reviewed against the standards set forth in 10 CFR 50.92
- and has'been_ determined. to invole no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or (ii) create _ the possibility of a new or different kind of accident from any accident previously analyzed; or (iii) involve a significant reduction in a margin of safety.
This proposed change is administrative in nature and does not involve a change to equipment, procedures, or limits.
CHANGE NO. 6 (BG&E FCR 85-63)
Change pages 3/4 4-9,3/4 4-10, and 6-18a of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
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.. Mr. A.Thadani-3 October 25,1985 Page 7 DISCUSSION This administrative change updates and clarifies the listing of Special Reports of I Technical Specification 6.9.2. Two items already listed have been clarified; and two new items have been added. Both of these newly listed reporting requirements already exist in Section 3/4 of the Technical Specifications, and are being added to Section 6 for completeness.
~ The two items to be clarified are:
- i. . Steam Generator Tube Inspection Results, and
- m. Radioactive Effluents Item '(i) covers three separate, possible reports. Changes have been made to this specification to more clearly delineate the reporting requirements. The change to item (m) corrects an obvious typographical error. The Technical Specification for Liquid
- Radwaste Radioactive Effluents is 3.11.1.3.
Items (o) and (p), as indicated on the attached marked-up pages, are to be added to the list of Special Reports. Previous license amendment submittals added reporting requirements to Section 3/4 Technical Specifications, but inadvertently failed to update
- the Administrative Controls listing, in section 6.
, DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) - involve'a significant increase in the probability or consequences of an accident previously evaluated; or The proposed change is administrative in nature, and does not affect plant operation or design in any way.
(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or No new or different kind of accidents from those previously evaluated in the Updated FSAR are created by this proposed change.
(iii) involve a significant reduction in a margin of safety.
This proposed change does not involve any reduction in the margin of safety; it only entails an administrative clarification and update of reporting requirements.
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. Mr. A.Thadani October 25,' 1985 Page8 CHANGE NO. 7 (BG&E FCR 85-65)
Change pages 3/4 5-5, 3/4 5-5a, and B 3/4 5-2 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
DISCUSSION This proposed change would delete the surveillance which verifies the correct position of the electrical position stop for isolation valves in the High Pressure Safety Injection (HPSI) System. These isolation valves fully open on a Safety Injection Actuation Signal
-(SIAS) to provide flow to the reactor via the normal and auxiliary injection headers to each leg.
This change is proposed since the function of these valves has changed as a result of License Amendments 104 (U-1) and 89 (U-2). Prior to these amendments, the valves would mid-position on a SIAS to throttle the high pressure injection flow at 170 + 5 gpm to each leg. The purpose of Technical Specification 4.5.2.g was to verify the throttle position of these valves. This surveillance is no longer necessary since the valves now fully open on a SIAS. These valves would continue to be cycled as required by Technical Specification 4.0.5 and full stroke of the valves would be verified by observing the valve position on the control boards. The position transmitters and indicators for these valves are all safety grade instruments. After maintenance which could affect the performance of these valves, the valves would be satisfactorily cycled prior to being considered OPERABLE.
This proposal would also change the basis for Technical Specification 3/4 5.2 to be consistent with deleting the surveillance which verifies the throttle valve position stops.
In addition, this proposal cancels and supersedes the previous request for amendment concerning this surveillance included in reference (a).
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposal has been reviewed against the standards set forth in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
.(i). involve a significant increase in the probability or consequences of an accident previously evaluated; or The HPSI isolation valves would continue to be cycled as required by Technical Specification 4.0.5. Valve position, however, would be verified by observing the valve position on the control boards in lieu of measuring stem travel during post-maintenance valve cycling. Valve position indication on the control boards is an appropriate indicator of full stroke for these valves and is safety grade. Therefore, the operation of the HPSI System during previously evaluated accidents would not be affected.
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, Mr. A.Thadani:
October 25,;1985.
Page 9 :
i
'(ii) create the possibility of a new or different kind of accident from 6 any accident previously evaluated; or - -
This proposed change does not involve a change to any component, system
- g. alignment, or operating procedure and would not create the possibility of a
, new or different accident.
-(iii) involve a significant reduction in a margin of safety.
As stated In (i) above, the operation of the HPSI System during accidents would not be affected.
m CHANGE NO.' S (BG&E FCR 85-60)
Change pages 6-18a and 3/4 6-26 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
s DISCUSSION This proposed change allows for the use of containment atmosphere grab sampling as a-
' back-up means of measuring containment-hydrogen concentration. Current Technical Specifications require plant shutdown -following a 30 day ACTION statement on one Inoperable hydrogen analyzer. This change will allow for continued plant operation ,
provided a grab sampling capability is demonstrated.
The prime function of the hydrogen analyzers is to monitor the containment atmosphere
~ for hydrogen following a Loss Of Coolant Accident (LOCA). The hydrogen analyzing system is common to both units. The system consists of two hydrogen analyzing subsystems, each consisting of a' hydrogen analyzer cabinet, sample select cabinet,
- hydrogen sequencer panel, and remote hydrogen recorder. Each hydrogen analyzing subsystem can monitor the containment hydrogen concentration at six points, three in each containment . structure. These locations have been selected to provide representative samples of the containment atmosphere. The sampling lines are run in groups of three through two separate containment penetrations.
' The original Calvert Cliffs hydrogen analyzers were replaced in 1982 per NUREG-0737 Section II.F.1.6. References (b) and (c) requested a temporary relaxation of the 30 day ACTION statement to provide for installation and testing of the upgraded analyzers. The NRC approved this request per reference (d), permitting one hydrogea analyzer to be inoperable for a period of approximately eight months. During this period, the provisions of Technical Specification 3.0.4 (which would prevent either unit from changing l operational modes with a hydrogen analyzer inoperable) were suspended for Technical Specification 3.6.5.1.
. Mr. A.Thadani October 25,1985 Page' 10 Due to unforeseen complications, we were forced to request an extension of the above stated relief in order to allow for the completion of the hydrogen analyzer modification. The NRC, per reference (e), granted the additional two-month relief, stating, ". . . A single hydrogen analyzer is sufficient to perform post-LOCA hydrogen sampling for Calvert Cliffs Units 1 and 2. . .."
In 1983, we again requested relief of Technical Specification 3.6.5.1. This proposal involved a two month period in which a single hydrogen analyzer could be inoperable for the purpose of replacin NRC, per reference (f)g , granted the solenoid valves this request, with environmentally reiterating qualified their position. as stated invalves.
the The aforementioned Safety Evaluation Reports.
During any period when one analyzer subsystem is out of service, the second subsystem is available to perform all necessary sampling evolutions. In addition, the availability of the hydrogen " grab sampling" capability provides added redundancy. This sample bomb equipped with a septum plug is located on the 45' level of the Auxiliary Building. The Laboratory Analyst can use a syringe to withdraw and subsequently analyze a containment atmosphere sample with the analyzer pump operating, and the sequencer lined-up to the appropriate 135' containment elevation sample point. This capability has been demonstrated on several occasions.
It is important to ncte that Calvert Cliffs' design incorporates two independent hydrogen recombiners as the primary means of hydrogen control after a LOCA. Each recombiner is designed to handle twice the flow rate necessary for maintaining the hydrogen concentration at three volume percent (Safety Guide 7 requires less than four volume percent). Emergency Operating Procedures require placing both recombiners in service if hydrogen concentration exceeds one percent.
In summary, we are seeking relief of Technical Specification 3.6.5.1 because of the unnecessary consequences of a dual plant shutdown due to one hydrogen analyzer exceeding its 30 day ACTION statement. We feel a shutdown is not warranted with one analyzer operable and a second analyzer pump capable of obtaining a containment grab sample as a back-up. Repairs will be adequately tracked pursuant to Technical Specification 6.9.2. (Hydrogen Analyzers, Specification 3.6.5.1, will be added as item (q.)
to Technical Specification 6.9.2.) Additionally, with the explicit requirements as stated in this proposed Technical Specification, we feel that Technical Specification 3.0.4 is not applicable for the same reasons.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or
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, Mr. A.Thadani October 25,1985 Page 11 A single hydrogen analyzer, coupled with containment atmosphere grab sampling capability, is sufficient to perform post-LOCA hydrogen sampling for Calvert Cliffs Units 1 and 2.
(ii) create.the possibility of a new or different type of accident from any accident previously evaluated; or No new or-different kind of accidents from those previously evaluated in the Updated FSAR are created by this proposed change.
(iii) involve a significant reduction in a margin of safety.
The bases for Technical Specification 3.6.5.1. state that equipment and systems required for the detection and control of hydrogen gas will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. This margin of safety is not significantly reduced. A single hydrogen analyzer and recombiner are sufficient to perform post-LOCA nydrogen sampling and control.
CHANGE NO. 9 (BG&E FCR 84-3002, Supplement 1)
Change page B 3/4 7-1 of the Unit 1 and 2 Technical Specification as shown on the marked-up copies attached to this transmittal.
DISCUSSION This change deletes the words "1971 edition" from the basis for Technical Specification 3/47.1.1. The reference to the 1971 edition of the ASME Boiler and Pressure Vessel Code is incorrect. Calvert Cliffs Unit 1 is subject to the requirement of ASME Code Section XI,1974 edition, with Addenda through Summer 1975, while Unit 2 is subject to the 1977 Edition, with Addenda through Summer 1978 for pump and valve testing only.
ASME Code requirements are met by Surveillance Requirement 4.0.5, which sets requirements for testing and inspection of ASME Code Class 1, 2, and 3 components.
. Specifications 3/4.7.1.1 do' not supercede the Code requirements controlled by Specification 4.0.5. Therefore, it is not necessary to defir.e the specific code edition in the basis -for Specification 3/4.7.1.1. Such a definition could be misleading and contradict more formal commitments.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change to the Technical Specifications has been reviewed against the standards in 10 CFR 50.92 regarding significant hazard considerations and has been determined to involve no significant hazards considerations, in that operation in accordance with the proposed change would not:
. Mr. A.Thadani October 25,1985 Page 12 (i) involve a significant increase in the probability or consequences of an accident previously evaluated, or (ii) create. the possibility of a new or different kind of accident from any accident previously evaluated, or (iii) involve a significant reduction in a margin of safety.
The proposed change is editorial, correcting an error in the Technical Specification bases. Therefore, the proposed change has been determined to involve no significant hazards.
l CHANGE NO.10 (BG&E FCR 85-1044)
Change page 3/4 7-61a of Unit 1 Technical Specifications as shown on the marked-up copy attached to this submittal.
DISCUSSION The proposed change deletes one snubber from Table 3.7-4 of the Unit 1 Technical Specifications which was replaced by a sway strut. A provision in Table 3.7-4 allows snubbers to be removed from safety-related systems for the purpose of replacement by a sway strut provided that Table 3.7-4 is updated at the time of the next License Amendment request. This request fulfills that requirement. Based on a reevaluation of the piping load calculations, it has been determined that this snubber is suitable for replacement by a rigid sway strut due 'to low thermal expansion attributable to the supported equipment. The sway strut was installed with the same geometry as the snubber, therefore, no decrease in the seismic resistance of the associated system will ,
occur.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS The proposed change has been reviewed against the standards provided in 10 CFR 50.92 regarding significant hazard considerations. Reference (g) has evaluated the replacement of snubbers with rigid sway struts for similar replacements and found no significant hazards. Since the sway strut can carry all analyzed accident loadings and thermal movement is insignificant, the proposed change has been determined to involve no significant hazards.
CHANGE NO.11 (BG&E FCR 84-110, Supplement 1)
Change page B 3/4 3-2 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
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.- Mr. A.Thadani October 25,1985 Page 13 DISCUSSION This proposal would change the basis for Technical Specification 3/4 3.3.4, Meteorological Instrumentation, to make it consistent with the appropriate guidance for meteorological instruments. This change is administrative in nature and has been requested by the NRC. The meteorological instrumentation is consistent with the recommendations of Regulatory Guide 1.23, ."Onsite Meteorological Programs," dated February 1972, as supplemented by Supplement I to NUREG 0737.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposal has been reviewed against the standards set forth in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or (ii) create the possibility of a new or different kind of accident from any accident previously evaluated; or (iii) involve a significant reduction in a margin of safety.
This proposed change is administrative in nature and does not involve a change to any component, system alignment, or operating or test procedure.
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.
, Mr. A.Thadani October 25,1985 Page 14 FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. (1900630) in the amaunt of
$150.00 to the NRC to cover the application fee for this request.
Very truly yours, (JA%2A 1 --
hrthur-E.Lundvall,Jr.
Vice President - Supply
- STATE OF MARYLAND :
- TO WIT:
CITY OF BALTIMORE :
Joseph A. Tiernan, 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 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 Seal: chod ,el JW Notary Public My Commission Expires: 7 [e 6 D' ate AEL/RMS/BEH/MTF/ dim /sjb e
Attachments cc: D. A. Brune, Esquire G. F. Trowbridge, Esquire D. H. Jaffe, NRC T. Foley, NRC T. Magette, DNR
. _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ . _