ML19326D941
| ML19326D941 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 07/07/1980 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Spellman G HOUSE OF REP. |
| Shared Package | |
| ML19326D942 | List: |
| References | |
| NUDOCS 8007250251 | |
| Download: ML19326D941 (2) | |
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,8 UNITED STATES 7
NUCLEAR REGULATORY COMNilSSION
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E WASHINGTO N. D. C. 20555
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JyL 7 1980
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Docket No. 50-318 i
The Honcr :le Gladys Spellman United S:;es House of Representatives Washington, D.C.
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Dear Congresswoman Spellman:
Your letter of May 15, 1980 to Mr. Carlton Kamerer of the NRC Congressional Affairs Office has been forwarded to me for response.
The following infor-mation is provided for your use in answering the May 1,1980 questions of your constituent, Mr. John A. May. The subject of this request is the radioactive gas leakage at the Calvert Cliffs Nuclear Pcwer Plant.
Mr. May's questions referred to the time approximately one week before his letter of May 1,1980.
A. radioactive gas release occurred close to this time on March 27, 1980. The cause of that release was a poor gasket seal on the Unit 2 degassifier relief valve. This release activated local area radiation alarms in the Auxiliary Building.
ConsequentQy, the builcing was evacuated in accordance with the Plant Emergency Procedures.
The maximum instantaneous release rate was calculated to be about 4% of our Technical Specification Limit.
It is estimated that the maximum integrated dose would be approximately 0.0007 mrem at a point south of the site.
Further detailed information on this event can be found in Enclosure 1, Nuclear Regulatory Commission Inspection Report 50-317/80-03 and 50-318/80-03.
Activities of this agency and our licensees following the accident at Three Mile Island (TMI) have been directed tcward further improve-ments in comercial nuclear pcwer plant design, construction, and operation.
These i@rovements are directed tcward both preventing accidents of this severity and toward mitigating the consequences of such an accident in the event it should occur. Recently Baltimore
, Gas & Electric issued a press release (Enclosure 2) describi.ng changes they have made following the TMI accident. Our safety evaluatio.;
of a number of these actions taken at Calvert Cliffs Unit Mos. 1 and 2 is provided for your information in Enclosure 3.
n IHIS DOCUMENT CONTAINS POOR QUALITY PAGES u
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t The Honorable Gladys Spellman We trust that this letter will provide you with the information needed to satisfactorily respond to questions from your constituents.
Sipcerely,
? Signed) T. A. Reher
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William J. Dircks
('" Acting Executive Director for Operations
Enclosures:
1.
NRC IE Inspection Report 50-317/80-03, 50-318/80-03, April 25, 1980.
2.
BG&E News Release, March 28, 1980.
3.
NRC Staff's Evaluation of Category "A" Lessons Learned Implementation, April 7, 1980.
s 4.
Letter dated May 1,1980 from John A. May.
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g, y UNITED STATES
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4 HUCLEAR REGULATOnY Commission
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suSa[AvcNur KING QF PRUSSIA. PENNSYLVANIA 19406 APR 2 5 1980 Occket Nos. 50-317 and 50-318 Baltimore Gas and Electric Company
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ATTN: Mr. A. E. Lundvall, Jr.
Vice President, Supply P. O. Box 1475 Baltimore, Maryland 21203 2 '.
Gentlemen:
Subject:
Inspection Nos. 50-317/80-03 and 50-318/S0-03 This refers to the inspection conducted by Mr R. Architzel of this office en March 10-23, 1980 of activities authorized by NRC License Mos. OPR 53 and DPR 69 at the Calvert Cliffs Nuclear Power Plant, Lusby, Maryland and to the discussions of our findings held by Mr 1
Architzel with Mr L. Russell of your staff at the conclusion of the inspection.
Areas examined during this inspection are described in thk Office of Inspecticn and Enforcement Inspection Report which is enclosed with this letter.
Within these areas, the inspection censisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
Within the scope of this inspection, no items of ncncompliance were obse rved.
In accordance with Secticn 2.790 of the NRC's " Rules of Practice,
Parc 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Occument Rocm.
If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you
'make a written application within 20 days to this office tp withhold such ir. formation from public disclosure. Any such application must be accompanied by an affidavit executed by the cwner of the information, which identifies the document or part sought to be withheld,"and which contairs a statement of reasons which addresses with specific'ity the items which will be considered by the Ccmmission as listed in subparagraph (b) (4) of Section 2.790. The information sought to be withheld shall be incorporated as far as possible into a separate part of the affidavit.
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APR 2 4 1980 Saltimore Gas and Electric Company 2
If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
No reply to this letter is required; howeyer, if you should have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
.i,
El on. Brunner, Chief Reactor Operations and Nuclear Support Branch Office of Ins. ection and Enforcement Inspection Report
Enclosure:
c Nur.bers 50-317/80-03 and 50-318/80-03 cc w/ encl:
R. M. Douglass, Manager, Quality Assurance L. B. Russell, Chief Engineer
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- f. Sydnor, General Supervisor, Operations QA R. C. L. Olson, Senior Engineer K. H. Sebra, Principal Engineer bec w/ enc 1:
IE Mail & Files (For Apprcpriate Ofstribution)
Central Files Public Document Recm (PDR) local Public Occurent Room (LPDR)
Technical Information Center (TIC)(NSIC),
Nuclear Safety Information Center REG: I Reading Room State of Maryland (2)
Dr. Steven Long, Administrator.for Nuclear Evaluations R. Architzel, RRI
- 0. Beckman, RRI n.
C. Ccwgill, RRI R. Cente/D. Haverkar-RRI J. Higoins, RRI L. Ncrrholm, RRI T. Rebelewski, RRI R. Gallo, RRI J. Shadlosky, RRI
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.s U.S. NUCLEAR REGULATORY COMMISS10N Q..
^W.4 OFFICE OF INSPECTI0fl AND ENFORCEMENT 4,T/p
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50-317/80-03
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Report Nc. 50-318/80-03 50-31/
Occket No.
50-318 OPR-53 C
License No.
npp-69 Priority Category C
Licensee:
Balti-cre Gas and Electric Cencanv P.O. Box 1475 Baltimore. Marvland 21203 Facility Name:
Calvert Cliffs Nuclear Power Station. Units 1 and 2 Inspectica at:
Lusby, Maryland Inspectica conducted:
March 10-28,1980 Inspectors:
8_ O b kh, R 4 [2V/7 o R. Archit::el, Resident Reactor Inspector date signed date signed date signed Approved by:
0Ek yI24[fo E. C. McCabe, Jr., Chief, Reactor Projects date signed Section No. 2, RO&NS Branch Ins:ection Sc==ary:
Inscection on March 10-28,1980 (Combined Recort Nos. 50-317/80-03 and 50-318/80-03)
Areas ins:ected:
Routine, onsite regular and cacksnif t inspection by the resident inspector (16 nours, Unit 1; 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, Unit 2). Areas inspected included the control room and the accessible portions of the auxiliary, turbine, service, and intake buildings; radiation protection; physical security; fire protection; plant operating records; and reporting to the t!RC.
Results: No items of noncompliance were identified.
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I jt UNITED STATES Et! CLOSURE 3 g 7, c.,f ;g NUCLEAR REGULATORY COMMISSION 4' ; y
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WASHING TON, D. C. 20555 QS"'
April 7, 1980 Occkets Nos. 50-317 and 50-318
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Mr. A. E. Lundvall, Jr.
Vice President - Supply Baltimore Gas & Electric Company P. O. Box 1475 Baltimore, Maryland 21203
Dear Mr. Lundvall:
Enclosed is the staff's evaluation of the implementation of Category "A" Lessons Learned requirements (excluding 2.1.7a) at Calvert Cliffs Nuclear Pcwer Plant, Units Nos.1 and.2.
Tnis evaluation is based on your submitted documentation and the discussions between our staffs at a site visit on February 19, 1980.
Based on our evaluation, we conclude that the implementation of the Category "A" requirements at Calvert Cliffs, Units Nos. I and 2 is acceptable. Certain items, identified in the evaluation, will be verified by the Office of Inspection and Enforcement.
This evaluation does not address the Technical Specifications necessary to ensure the limiting conditions for operation and the icng-term operability surveillance requirements for the systems modified during the Category "A" revi ew. You should be considering the proposal of such Technical Specifica-tions. We will be in comunication with ycu on this item in the near future.
Sincerely, l
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Robert W. Reid, Chief Operating Reactors Branch !4 Division of Operating Reactors
Enclosure:
Evaluation cc w/ enclosure:
See next page
Balticore Gas and Electric Company r"
cc w/ enclosure (s)-
James A. B'ddison, Jr.
Mr. Bernard Fowl er General Counsel President, Board of Ccunty MX G and E Suilding Commi s sioners Charles Center Prince Frederick, Maryland 20768 EE Baltimore, Maryland 21203 Director, Technical Assessment George F. Trewbridge, Esquire Division Shaw, Pittman, Potts and Office of Radiation Programs. _.
rr Trewbridge (AW-459) 1500 M Street, N.W.
U. S. Environmental Protection Agency 5
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Vashington, D. C.
20036 Crystal Mall !2 Arlington, Virginia 20450 Mr. R. C. L. Olson Saltimore Gas and Electric Company U. S. Environnental Protection Agency P__-i Roca 922 - G and E Building Regica III Omce
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Pos: Office Box 1475 ATTN:
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Baltimore, Maryland 21203 Curtis Builcing (Sixth Flecr)
M Sixth and Wainu: Streets EE-fir. Leon 5. Russell, Chief Engineer Philadel:hia, ?ennsylvania 19106 M
EEF Calver-Cliffs Muclear ?ower Plan 3altimore Gas ano Electric Company Raiph E. Architzel M,7 Lesby, l'aryland 20657 Resident Reac;cr Inspector -
EEE NRC !nspectica anc Infcrcement
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5ech el Power Corecration P. O. Box 437 s=
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Mr. J. C. Judd Lusby, Maryland 20557
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157a0 Shady Greve Road Gaithersburg, Maryland 20760 ME
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EE Cc::custion Engineering, Inc.
Acministrator, Power Plant Sitine Progra-
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Mr. P. W. Kruse,ilanager Energy and Coastal Zone Adminisdatien EE Engineering Ser vices Depart 9ent of Natural Resources 5E Pes: Office Box 500 Tawes State Office Building EE Windsor, Connecticut 06095 Annapolis, Maryland 21204
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Mr. R. M. Douglass, fianager
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s., v j CALVERT CLIFFS UNITS 1 & 2 EVALUATION OF CATEGORY "A" LESSONS LEARNED 4
IMPLEMENTATION Intr: duction Ey letters dated October 17, November 20, December 5 and 14,1979; Januzry 4, 22, and 30, February 1 and 29, and March 12 and 17,1980, 3aitinore Gas and Electric Company (BG&E or the licensee) submitted docu entatien of.the actions taken at Calvert Cliffs Units Nos. I and 2 (the plant). to implement the requirements resulting from TMI-2 Lessens i. earned. To facilitate our review of the licensee's actions, members of the staff visited the plant on February 19, 1980.
Evaluation The SRC's Category "A" Lessons Learned Recuirements and acceptance criteria tre focumented in NL' REG-0573 and NRC letters dated September 13 and Octcber 30, 1979. The number designatica of es:h item is censis en with the identifi-cati:ns used in NUREG-0E78.
2.1.1 Emer:enev power Sucolies Pressurizer Heaters The ;ressurizer propc rtional Featers are supplied from a Class lE bus.
Two
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banks of pressurizer backup heaters are fed from a a80V engineered safety fe nure load center. The redundant capacity of 450kw provided by the backup and proportional pressurizer heaters is sufficient to maintain natural circu-iation.
The load center breakers supplying the proportional heaters are trip;ed on an under<oltage signal after a loss of offsite power. These breakers can be closed manually when required after diesel generator load secuencing is complete.
Power Ocerated Relief Valves (PORVs) and Block Valves The :otive components of the FORVs are supplied frca safety related 480V
- otor control centers which have a diesel backup.
The control components of the 70RVs are supplied froc safety related 125 VDC battary buses.
The motive and control comocnents of the PORY block valves are supplied from safety related 480V motor control centers which have diesel backup.
In each case the motivt ar.d control power for the block valve is supplied frcm a pcwer supply train different from that which supplies the associated PORV.
Fressurizer Level Indicator
,c cf tne pressurizer level instruments for each unit are ?cwered frca the vital DC buses and the third is powered frco offsite AC pcwer with diesel backup.
Tha licensee meets the recuirements of Category "A" for Item 2.1.1.
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2.1.3.a Direct Valve Position I,dication The direct valve position indicaticn of the two pcwar operated relie.f valve- (2 indicators) and two safety valves (1 indicator) have been accomnlished wit..
an acoustic monitoring system manufactured by Technology for Energy Cor-poration (TEC). The TEC 914. Valve Flow Menitor uses an acceleremete n the sensor ahead of a charge converter.
The accelerometer converts acceleration to charge, which is then converted to voltage. The processor unit.then.
indicates relative valve flow.
The flo'w indicator located in the control room is a bar indicator made up of 10 light emitting. diodes (LEDs) which iight up relative to the flow.
The system,is safety grade and has an alarm in the control room.
Based on the above we find the licensee has met requirement 2.1.3.a.
2.1.3.6 Instrumentation for Inadecuate Core Cooline Procedures have been ucgraded to aid operators in detection of inadecuate core ccoling and to assure apprcpriate actions are taken.
Additional procedures to be used by the operator to recogni e inadequate core cool.ng are being developed based on analysis and cuidelines required by Item 2.1.9, Transient and Accident
. Analysis, s
The CE Owners Group provided report CEN-il7 in response to Item 2.1.9 and Item 2.1.3.b which addresses the instrumentation for detection of inadequate core ccoling.
CEN-li7 concludes that present instrument response during various postulated means of approaching inadequate core cooling yield several sienificant patterns of indication available to the operator.
From these pat erns he can detect the approach to inadervate core coolirg.
In addition, the licensee has submitted a desic.i description of a reactor vessel water level measurement system.
This systrm, as well as CEN-il7, will ce reviewed at a later date.
Subcooli_nc Meter.
SG&E has installed a subcooled margin monitor (SMM).
The SMli is a micro-cceputer based instrument which continuously displays the subcooled margin to saturation.
It is designed for use as a pcst-accidect monitoring instru-ment.
The SMM is a Sarety Class 1, Seismic Class 1, Quality Class 1 instru-ment and is designed to meet IEEE Stds. 344-1975 and 323-1974.
The SMM provides the operatcr with continuous digital display of either the pressure or temperature margin to saturation. An alarm is provided as part of the SMM. Tempenture inputs are from two hot legs and two cold legs per
.r.e t e r.
The range is 212*-705'F.
The RTD input sensors are s ismic qualified 1
and are part of the existing Reactor Protection System (RPS)..There are two pressure inputs per metar with the range of 15-2300 psia.
pressure sensors are seismic cualified and are part of the existing RPS.
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i Installatien of the new dual transmitters has had compatibility orablems with the existing equipment.
One channel per meter is now operaticnal.
C;eration of the second channel will be delayed four weeks until receipt of compatible equipment for the remaining channel.
At present the plant computer ~ calculates the margin to saturation.
The com-4 puter uses incere thermocouples.
.L The licensee meets.the requirements of. Category "A" for item 2.1.1. h.
Operation of the second channel of the 'subcoolino meter will satisf.y
_the redundancy recuirement of Category "A" Item 2.1.3.b.
The redundancy at the present is provided by the plant computer. Our Office of Inspection and Enforcement (IE), will verify modifications to the SPMs have been completed.
2.1. 4 Centainment Isolation The NRC requirements are that the licensee is to:
(a) carefully reconsider their determination of which systems should be considered essential or non-
=ssential for safety; (b) modify systems as necessary to isolate all non-essential systems by automatic, diverse, safety grade isolation signals; and (c', modify systems as necessary to assure that the resetting of the contain-man-isolation signal.does not cause the inadvertent recpening of containment isclation valves.
The licensee's submittals of November 20, 1979 and February 29, 1980 identi-fied the essential and non-essential systems and orovided the bases for the essential system classification.
The containment isolation system has been modified so that containment pene-trations associated with non-essential systems are either locked closed or are automatically isolated on diverse caraceters including the safety injec-tien actuation signal (SIAS) and the containment isolation signal.
SIAS is initiated on either high containment pressure or icw pressurizer pressure.
The isolation valve control circuits have been modified to prevent inadvertent cpening after resetting the isolation signal. This has been acccmplished by wiring the valves' control switches so as to fann a reset permissive, i.e.,
4 resetting will only be accomplished if all the isolation valves' handswitches in a given circuit are in the isolation position. With this design, deliberate operator action is required to reopen each isolation valve after the isolation signal is reset.
In addition, each circuit can be bypassed by a bypass switch located in a locked closed cabinet in the cable spreading area.
This bypass provides a backup reset capability in the event of'a failure in tne reset circuitry.
The above mcdifications have been made for all the isolation valves, except the oxygen and reactor coolant system sample valves.
Physical constraints require that this reset design objective be accomplished using additional lock-in relays that will interruct the power to the valves and fail them l
closed once an isol.ation signal is received; the circuit will keep the valves closed until the operator repositions the valve control switch to the open
- csition.
These acdifications will be made within 30 days of receipt of the required equipment and prier to June 1, 1950.
Our conclusicn is that with the modification of the sample valves the licensee's containment isolat:en design i
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1 meets the NUREG-0578 Section 2.1.4 containment isolatien requirements and is therefore acceptable.
Our Office of IE will verify that modifications to the sample valves centrol circuits have been completed.
2.1.5.a Dedicated penetrations for External Recombiners or post-Accident External
- purce Systam The NRC's position is that dedicated containment isolatien systems should be used for the external reccmbiners or purge systems that meet redundancy and single fat'ure requirements. This requirement dces not apply to the licensee since recombiners located wholly within the containment ar6 T2 sed.
.l.5.c Recembiner procedures The NRC's position is that the procedures for use of the recombiners be reviewed considering shielding requirements and personnel exposure limitations.
The plant utili:es recombiners located inside the centainment. Controls for operating the recombiners are located inside the control rocm.
During the site visit we discussed the licensee's review of-the recombiner oper-ating procedures and agreed that no modifications are receired.
e have concluded that the licensee has me-the NUREG-0575 recuirements for
..review of the recombiner procedures, Section 2.1.5.c.
2.1.o.a 5ystems Intecrity
- s The licensee has provided a list of those systams which he has determined may contain radicactivity folicwing an accident.
These systems are the safety injection, containment spray, shutdown cooling, containment sump recirculation, and reacter coolant and containment atmosphere sampling system.
He has also provided a description of the immediate leak reduction program which included walk down inspections to identify leakage, cleanup and repair of these systems.
The licensee has also measured and reported the final system leak rates to the NRC.
The licensee has established a surveillance test program for the systems which may centa n activity following in accident which includes testing once per refueling cycle.
In order to assure that radioactivity will be restrained to those systems.
specified and to allcw for operation of the reactor coolant pumps, the licensee
- has ccanitted to incorporate a procedure for ccmoing the reactor coolant drain tank to the cont;'nment sump.
IE will assure that this procedure is in place.
Cur October 30, 1979 clarification letter requests the 'lic$nsee to include a review of potential release paths due to design and coerator deficiencies as discussed in the October 17, 1979 letter regarding North Anna. The licensee has analyted their piant witn regard to the Ncrth Anna Incident and concluded no corrective ac-ion is necessary.
5 sed on the above information, we conclude that the licensee has met the Categcry "A" requirements for this item.
2.1.5.5 7'.! - Shieldine Revies The licensee's Januarv 4,1980 submittal includes a design review of plant s.ieldino and enviroimental qualification of equipment. The licensee has
- erformed the design review assuming the systems identified in Item 2.1.6.a contain radioactivity. The licensee has used the source term as specified in the October 20 letter for his review. The licensee has deter-mined hich radiation areas and identified components which may be attected.
Tha licerisee also discussed possible modifications in the affected areas.
The licensee has stated that components which may be adversely afrected-will be identified and corrective acticns completed by January 1,1981, if equipment is available. They have also identified areas where access may be required.
For these areas, corrective. actions will be taken to assure that the necessary functions can be perforfned. A detailed evaluation of the submittal will be performed at a later date.
We conclude that the licensee has met the Category "A" requirements for this item.
2.1.7.b Auxiliarv Feed :lew Indication Indication of auxiliary feed ficw provided to each steam generator is safety grade and is powered from vital power supplies.
As a backup to the safety grade auxiliary feed flow indication each steam generator has four safs:y grade level char..els with control recm readout.
'*a find this satisfies the Category "A" requirements for this item.
2.1.3.a post-Accident Samolino_
The licensee's January ?,1980 submittal contains a design review of the plart sampling capability for primary coolant and containment air samples assuming ? scutce as specified in NUREG-0578.
The licensee has incorperated interim procedures for cbtaining and analyzing a reactor coolant sample folicwing an accident. They also incorporated in erim procedures for obtaining and analyzing a containment air sample with the existing system.'
IE will assure that the procedure is in place.
The licensee has providad a preifminary design of the prcposed plant modi-ficaticns necessary to teet the Category "3" requirements for reactor coolant and containment atmosphere sampling.
i Based on the above, we conclude the licensee has met the Catecory "A" require-ments for this item.
1 2.1.8.b Hich Rance Radiation Snitors The licensee has provided ecuipment and implementing precedures' to quantify noble gas release rated from the plant vent, condensor air ejector and steam safety release and dump valves if the existing instrumentation goes offscale.
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IE will assure that the procedures are in place.
The licensee has provided a description of his syste : to be used to da: ermine radioicdine and particulate effluents.
They have also modified existing procedures for obtaining effluent samples to allow for potential high dose rate levels following an accident.
IE will assure that the appropriate procedures have been modified.
l Eased on the above information, we conclude that the licensee has met the
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Category "A" requirements for this item.
2.1.8.c Imoroved Iodine Instrur.entation The licensee has ' committed to use charcoal cartridges to collect air samples in occupied areas. The sample cartridges will be counted using one of the plant Geli systems, which has been dedicated for analyzing air samoles.
The samples can be counted in a short enough time to allow for operator protection.
This system meets the requirements of NUREG-0573.
The licensee has also pro-vided assurance that all areas occupied by essential pe:;onnel will be moni-torsd. Therefore, we conclude that the licensee meets the requirements of N'JREG-0578, Item 2.1.8.c.
2.2.1.a. Shift Suce-visor Resconsibilities The NRC requirement for this item is to revise, as necesitary, the responsi-bilities of the Shift Supervisor such that he can provide comnand oversight i
cf operations and perform management review of ongoing operations that are important to safety.
During the staff's site visit we reviewed the licensee's management directives and revisions to their administrative procedures, QAp-25, plant Operations.
k's have determined that these directives and procedures satisfy the require-j ments of NUREG-0578, Item 2.2.1.a. for delineation of Shift Supervisor responsibilities.
2.2.1.b Shift Technical Advisor (STA)
The NRC requirement is for the licensee to provide an on. shift advisor tn the Shift Supervisor to serve the two functions of acciden:. assessment and coerating experience assessment. As a supplement to the ocerating staff, the STA must be available to the control room to assist in diagnosing an oTf-normal event.
To satisfy the r s requirements, the licensee.has implemented a program, described in the-Jecember 14, 1979 submittal, wherein the staff for the two plants is in: ased to include one additional (SOL) holder or degreed engineer.
This i :ividual would satisfy the recuired STA accident assessment function.
The o:a ating experience assessment function of the STA is satis-fled by a standir. co mittee staffed by onsite engineers and augmented as r.ecessary by engineers from the Enginee-ing Department.
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'rle have reviewed the ifcensee's submittal describing their STA programs.
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additien, during the site visit we discussed the ?r: gram with the licensee and detarmined that a satisfactory STA program is in opera-ion.
We find tha.
s-their STA program is in agreement with the staff's requirements describad in Section 2.2.1.c of N" REG-0578 and is therefore acceptable.
2.2.1.c Shi#t and Relief'Turnever Procedures The NRC requirement is for the licensee to assure that procedures are adecuate to provide guidance for a complete and systematic turnover between the offgoing and encoming shift to assure that critical plant parameters are within limits and that the availability and alicnment of safety systems are made kncwn to the oncoming shift.
The licensee's submittal indicated thatIchecklists and locs have been developed
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which satisfy our acceptance criteria. Turther, he has established a system to evaluate the effectiveness of the shift turnover procedure.
During the site visit our check of the revised shift turnover procedure checklists and logs confirmed that the licensee has addressed this position.
We conclude that the licensee has satisfied the recuirements of Item 2.2.1.c related to shift turnover procedures.
Adecuacy of the checklists
.and locs will be performed by the Office of IE and will be documented by approcriate Inspection Reports.
2.2.2.a Control Rocm Access
' Procedure CA?-25 establishes the authority of the persen in charge of the con rol recm to limit access to the control room.
The ' licensee has imple-mented procedures in the Site Emergency Plan defining lines of communication and authority between the control room and the different emergency centers.
Precedures also exist which establish the line of succession for personnel in charge of the centrol reem and limit the personnel in charge to those holding a SROL'.
We find the licensee has met the Catecory "A" requirements for this item.
2.2.2.b Technical Succort Center (TSC)
A TSC has been established en the 55' Elevation of the Log and Test Instru-ment Room. This area is adjacent to the control room and is included in the Centrei Rocm Ventilation System.
This reem is directly accessible from the turbine building c-the control recm.
The dedicated communications to the cen rol recm and Emergency Operations Center will be prdvided by dedicated channels of the ;'.r.: paging system.
The licensee has installed an additional head set in the ~;: to permit simultaneous communication between the TSC and the control roce
-f the TSC and the Emergency Operations Center.
Piens and procedures fer e' 1eering/ management support of the TSC have been established and are containe:
1 the Site Emergency Plan. The licensee ha's located the " Record Retri..al Center" in the TSC. This will contain a.ll clant physical data, drt ines and parameters.
Selected plant parameters can be read out 'n the TI: via tem orary recorders connecced to a portion of the star up/pnysics es: ;anei. The licensee has provided details' for the icng term TSC. We conclude that this satisfies the Category "A" require. ents for m
this item.
s 2.2.2.c Cnsite Ocerational Suecort Center (OSC),
The licensee has established an OSC from which there is the capability to communicate with the control room.
The OSC is located in the service building, across the turbine deck from the control rocm. The Site Emer-gency Plan has been modified to establish lines of communication and management of the OSC.
'a'e consider this to meet the requirements for this item.
NRR Reactor Coolant System Ventino The licenses has preposed a desi~gn for venting of the reactor coolant system in fulfillmenc of the Short-Term Lessons Learned Requirement.
Conclusion Sised on the above, subject to our Office of IE verification as noted, we find that imolementation of the Category "A" Lessons Learned Requirements at Calvert Cliffs Units Ncs. I and 2, is acceptable.
Dated: April 7, 1980
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EtjCLOSURE 4 a
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}983
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6707 Gateway Blvd.
District *ieights, Md. 20028
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May 1, 1960 y
T rs. Gladys Spellman v
F.9. Veeresentative
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c.annon Sida - Dist. Office 6CD< u lcrest 9d.
e ysttsville, vd.
cuellman:
near vrs.
I am writing this letter in reference to the nuclear ucwer ulant which is located i~n Calvert Cliffs,
\\
aryland.
v A week azo, I read in the papers that there vs.a a malfunction in the plant.
The paper said that a sms11 amount of radioactive zas leaked out of the nuclear tower plant because of a malfunction inside of -
the building.
Why was this caused?
I would like to know what exactly hacoened in the plant and what measures are being taken to protect the nearby community from a possible incident that happened at Three vile Island.
s The paper never really explained what happened at the Calvert Cliffs plant, so I would like to have an answer to the accident at the plant.
Very trul,J Tours,
{Nt 0 ="/7.aa pm~
o John A. May 4
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