ML19296C551
| ML19296C551 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/10/1980 |
| From: | Moody D Maine Yankee |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19296C536 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 WMY-80-9, NUDOCS 8002260635 | |
| Download: ML19296C551 (2) | |
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ENGINEERING OFFICE WESTBoRo. MASSACHUSETTS QiSE' 617-366-9; 7
.Q B.3.2.1 WMY 80-9 January 10, 1980 United States Nuclear Regulatory Commission
' Washington, D.C. 20555 Attention:
Office of Nuclear Reactor Regulation Mr. Harold Denton, Director License No. DPR-36 (Docket No. 50-309)
References:
(a)
NUREG-0578 - Lessons Learned Task Force Status Report and (b)
Short Term Reco=mendations dated July 1979 (c) USNRC Letter to All Operating Nuclear Power Plants dated September 13, 1979 (d) USNRC Letter to All Operating Nuclear Power Plants dated October 30, 1979 USNRC Letter to All Operating Nuclear Power Plants' dated (e)
October 17, 1979
Dear Sir:
Subject:
NUREG-0578, Item 2.1.6a Systems Integrity For Containing Radioactive daterials Outside of Containment The following information is provided in response to the recommendations and requests made in References (b), (c), and (d) relative to Item 2.1.6a.
To ensure the integrity of systems which are located outside the containment and which are likely to contain high level radioactive materials subsequent to an incident, Maine Yankee intends to use the program already in effect as a result of upgrading its inservice inspection program to full compliance with The Safety Class 2 and 3 systems located outside the ASME Code Section XI.
containment which may cortain radioactive materials will be leak tested in accordance with Articles IWA-5000, IWC-5000, and IWD-5000 of the referenced Those systems that are not Safety Class but could contain high Code Section.
level material will also be inspected to Articles _lA-5000 and IWD-5000.
Repairs to systems will be made in accordance with Articles IWA 4000, IWC-4000, and IWD-4000 and the Maine Yankee Quality Assurance Program.
It is felt that the above program as a supplement to presently installed gas and radiation monitors, daily' inspections by plant operators, plant surveillance testing, continuous monitoring of tank levels, and plant preventative maintenance practices will pW vide the necessary measures to assure tha integrity of the systems in question.
B 0022 60
U.S. Nuclear Regulatory Commission January 10, 1980 Mr. Harold Denton, Director Page 2 The systems which are likely to contain highly radioactive contaminants following an incident are listed below. The leakages have not yet been quantified but are considered negligible.
An inspection program has been developed and will be carried out prior to plant startup following the 1980 refueling outage. The program will identify and quantify leakages from the below listed systems, and the results will be forwarded to the NRC as soon as they are available.
The following systems have been determined to have the potential for containing high level radioactive materials:
Waste Gas System Letdown System Purification System Charging System Emergency Core Cooling System Residual Heat Removal System Primary Sampling System Post Accident Purge System Containment Air Particulate and Gas Monitcring System Seal Water Supply and Return Maine Yankee has additionally, in accordance with the requirements of Reference (e), completed a review of the North Anna Unit 1 incident as it' applies to our facility. We have determined that no design deficiencies exist and, therefore, no modifications are deemed necessary as a result of this review.
We trust that this information is satisfactory; however, if you have additional questions, please contact us.
Very truly yours, MAINE YANKEE ATOMIC POWER COMPANY dE24) au /
D.
Moody Manager of Operations EWJ/sec
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- m-M O B.3.2.1 January 22, 1980 WY 80-14 1
United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Office of Nuclear Reactor Regul; tion Mr. Harold Denton, Director
References:
(a) License No. DPR-36 (Docket No. 50-309)
(b) NUREG-0578 - Lessons Learned Task Force Status Report and Short Term Recommendations dated July, 1979 (c) USNRC Letter to All. Operating Nuclear Power Plants dated September 13, 1979
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(d) USNRC Letter to All Operating Nuclear Power Plants dated October 30, 1979
Dear Sir:
Subject:
NUREG-0578, Item 2.1 3.b Instrumentation for Inadequate Core Cooling ift accordance with the requirements of References (c) and (d), the N 1owing information relative to ':nstrumentation for the detection of
'triacrequate core cooling is hereby submitted.
The below listed instrumentation presently exists for Maine Yankee operators to utilize in making an assessment as to whether or not a condition of inadequate core cooling exists. Although not all instrumentation provides an exact quantitative measurement, the use of combinations of available data provides the operators'the necessary information to determine that sufficient core cooling is being provided.
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Core Exit Therinocouples 20 0-9000F Computer 1.
(to be extended to 23000F) 2.
Hot Leg RTD's ll/ loop NR 515-6650F MCB/M 1/ loop NR 515-6150F MCB/M, 3
Cold Leg RTD's DUPLICATE DOCUMENT Entire document previously entered into system under:
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February 1, 1980 United States Nuclear' Regulatory Commissien Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation Mr Darrell G. Eisenhut, Acting Director Division of Operating Reacters Referencas:
(a) License No. DPR-36 (Docket No. 50-309)
(b) USNRC letter to MYAPC dated October 18. 1979 (c) USNRC letter to MYAPC dated September 11, 1979 (d) MYPAC letter to USNRC dated November 20, 1979 (WMY 79-138)
(e) MYPAC letter to USNRC dated November 15, 1979 (WMY 79-133)
Subject:
Automatic Initiation of Auxiliary Feedwater System.
Dear Sir:
This letter is written to provide supplementary design information of the automatic initiation of the auxiliary feedwater system control grade installation at Maine Yankee (Attachment I); hereupon addressing your ihquirements identified during the course of the NRR Bulletins and Orders Task Force review, Reference (b).
Your letter, Reference (c) required information regarding a possible steam generator water hammer under the influencing automatic initiation of auxiliary feedvater flow. Experience gained at Yankee Rowe was utilized in the design of the Maine Yankee plant. As a result, the only water hammer event ever to occur in the Maine Yankee feedvater system was caused by rapid cycling of the feedvater regulating valve due to a loss of control air. There is no reason to expect additional water hammer events as a result of automatic initiation of auxiliary feedwater.
We trust this information is satisfactory; however, if you have any questions, please feel free to contact us at your convenience.
Very truly yours, f
MAINE YANKEE ATOMIC POWER COMPANY h
tF pA a x [0 Robert H. Groce 9
Senior Engineer - Licensing 9
3, RHG/ dis b
Ob Attachment
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1 ATTACHMENT 1 AUTOMATIC INITIATION OF THE AUXILIARY FEEDWATER SYSTEM - CONTROL GRADE REFERENCES (a) NUREG 0578, Itsued July 1979 (b) NRC Letter, dated September 13, 1979, " Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident."
(c) Letter, Maine Yankee-to the NRC, WMY 79-113, dated October 18, 1979 (d) Letter, Maine Yankee to the NRC, WMY 80-4, dated January 9,1980 REASON FOR CHANGE During the incident at the Three Mile Island (TMI) plant, the auxiliary feedwater cystem (AFWS) was unavailable for use due to operator error. The NRC has determined, as a result of their evaluation of the incident, that
" consistent with preventing the steam generators from drying out following loss of main feedwater and minimizing operator errors that could delay the timely initiation of the auxiliary feedwater system in a PWR plant the AFWS should be automatically initiated" (Reference (a)). Therefore, recent Yankee staff meetings on the TMI incident have resulted in a commitment (Ref. (c)) to install control grade equipment to automate the AFWS at Maine Yankee. A tafety grade system will be installed in 1981 when the design and procurement of safety grade equipment has been completed.
DESCRIPTION OF CHANGE This change will automatically start the two electric driven auxiliary feedwater pumps on a delayed one-out-of-three logic signal from a low water level in any one steam generator. This delay will be set for 5 minutes as per Reference (d). The low S.G. water level signals will be from the reactor regulating system (RRS) cabinets (UT1211X, Y, LT-1221X, Y, and 1231X, Y).
The setpoint for this low water level signal will be the same as the present Reactor Protective System (RPS) low S.G. water level trip.
The installed system will be single failure proof with on-line testing capability, which will be accomplished by the use of a test switch as shown on 11550-ESK-11AD. This system will also be designed to allow the operator to bypass a failed or inoperative train of the actuator logic provided that the other train is not in test. This will be done using a key lock switch with one key fitting the bypass switch of one train and the test logic of the other train.
The capability for manual override of this signal for individual AFWS pumps will be retained so that the operators may override when parameters indicate the AFWS is no longer needed.
Also, an annunciator will be used to indicate an automatic AFWS start signal has been generated when the pumpp have started, when the system is in test, or when the system has been bypassed.
Page 2 DESIGN CRITERIA By design, this control grade change uses sound engineering judgement and practice to assure no interaction with protective channels. Also, the criteria stated in Reference (a) and listed below were employed:
1.
The design shall provide for the automatic initiation of the auxiliary feedwater system.
2.
The automatic initistion signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.
3.
Testability of the initiating signals and circuits shall be a feature of the design.
4.
The initiating signals and circuits shall be powered from the emergency buses.
5.
Manual capability to initiate the auxiliary ~feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of syst em function.
6.
The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.
7" The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.
Also, the capability for manual override for individual AFWS pumps will be retained so that the operators may override when parameters indicate the AFWS is no longer needed.
The auxiliary feedwater system is a safety system.
This change vill add a control grade automatic start logic circuit to the A
present manual initiation of the motor driven auxiliary feedwater pumps.
failure of this control grade system will not prevent the AFWS from operating manually.
In the light of the above discussion and the analysis by NSD (Reference (d)),
the proposed modification does not increase the probability of occurrence of a previously evaluated accident, create the possibility of a new type of accident, or reduce the margin of safety as defined in the basis of any Technical Specifications.
The proposed modification has been analyzed to assure that it does not create any unreviewed safety questions as defined in 10CFR59.59(a)(2)..
ATTACISIENT C This attachment to be submitted immediately upon management concurrence.
ATTACIBIENT D ONSITE TECHNICAL SUPPORT CENTER I.
REFERENCES (a) NRC letter dated September 13, 1979, D G Eisenhut to All Operating Nuclear Power Plants (b) NCREG 0578 "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" (c) NRC letter dated October 30, 1979, H R Denton to All Operating Nuclear Power Plants (d) NRC letter dated October 10, 1979, D G Eisenhut to All Power Reactor Licensees II.
DEFINITION A.
Reference (a), in implementing the recommendations of Reference (b),
required in Section 2.2.2b that:
"Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident.
The center shall be habitable to the same degree as the control room for postulated accident conditions.
The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.
A complete set of as-built drawings and other records, as described in ANSI N45.2.9-1974, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions.
These documents shall include, but not be limited to, general arrangement drawings, P6ID's, piping system isometrics, electrical schematics, and photographs of components installed without layout specifications (e. g., field-run piping and instrument tubing)."
Reference (c), Section 2.2.2b provided additional details of this requirement in its "clari fication".
B.
In order to provide technical and management support to the command and control personnel during abnormal occurrences, Maine Yankee has established an Onsite Technical Support center as described in Section III capable of the following functions:
1.
Monitor the status and trends of the reactor core during accident conditions to predict, evaluate, and limit core damage.
This evaluation shall include teactivity and thermal-hydraulic parameters.
2.
Monitor the status and trends in containment during accident conditions to evaluate, predict, and limit releases of radioactive material. This evaluation shall include paraneters which would assist in the determination of core damage as well as control the release of radioactive material.
3.
Monitor the status and trends of those systems which provide reactivity control, heat removal, and contain-ment integrity capabilities to p.rovide the operator with alternate.nethods for performing these vital functions.
4.
Monitor the status of plant information systems to ensure reliability of the parameters used for Iters 1 through 3 above and provide the operator with a reliability assessment of control parameters when in doubt.
5.
Provide for operational assessment of the above four functions and translate their results into recommendations for shift operating personnel.
6.
Provide for management decisions to determine the overall course of action to be taken during accident situations, and provice liaison with senior NRC management personnel.
7.
Provide information as required to the emergency coordination center and control room.
III.
DESCRIPTION A.
Location and physical size The technical support center consists of the plant computer center and the second floor of the technical support building. As shown in Figure 1, the computer center is adjacent to the control rocm and connected to the second floor of the technical support building by a stairway. This arrangenent provides 3252 ft2 of space divided into of fices, file space, computer space and a sanitary f acility.
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Access to the control room is provided for key personnel. Normal access is provided through either the turbine hall or outside door at the east end of the technical support building.
B.
Habitability The technical support center provides two levels of protection.
That sector located in the computer room is at the same elevation as the control room and is within a steel framed concrete filled block wall structure which shares ventilation with the control room.
Its habitability ^would thus be challenged only in the most extreme circumstances.
The sector located on the second floor of the technical support building is monitored by portable radiation monitc.rs.
In the event that this sector became uninhabitable, the functions of this area would transfer to the computer and control rooms.
C.
Data Acquisition By encompassing the olant computer, the technical support center has available nearly all parameters necessary for assessment of plant status. Any parameter not on the computer may be readily obtained from the control room instrumentation.
Up-to-date plant design documents are routinely maintained in the drawing file section shown in Figure 1 and therefore are readily available to support technical support center functions.
D.
Communications Facilities The technical support center is provided with direct communications with the NRC, control room, emergency coordination center, emergency news element, the Yankee Nuclear Services in Westboro, state police, State Emergency Operations Center, Department of Energy, Corporate Headquarters and Plant Security as shown in the attached table.
IV.
OPERATION A.
Staffing Plans for staffing of the technical support center to perform the functions described in Section I are being incorporated in the site emergency plan as required by Section 2.2.2b of Reference (a). This effort is being carried out in accordance with the requirements of to Reference (a) and Reference (d).
Procedures for manning shall be included in the implementation procedures for the s'te emergency plan.
B.
Activation Guidance and criteria for activation of tite technical support center are being included in the site energency plan discussed in A. above.
This criteria shall provide recommendations for manning levels based upon the severity of the event in progress.
V.
UPCRADING A.
Evaluations are unde rway to determine the necessity for and extent of improvements to the technical support center in the following areas :
1.
TSC data requirements are being considered in an ongoing effort to upgrade the plant computer.
This effort will include the potential for offsite data transmission.
2.
An evaluation is underway to verify that the computer room ventilation is included with the control room ventilation HEPA filter system in accident situations.
3.
An evaluation is underway to determine appropriate permanent radiation monitoring equipment for the technical support center.
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Maine Yankee Com=unications Facilities Type Description Quantity Facility 1
Wiscasset Exchange Trunk 3
NET &T PBX Dial Access 2
Bath Exchange Trunk PBX Dial Access 2
NET &T 3
Wiscasset Exchange Line Private Dial Telephone 1
NET &T 4
Nuclear Regulatory Commission 1
AT&T 5
Civil Emergency Preparedness 1
NET &T Private Dial Telephone 6
Central Maine Power Company 3
CMP Dial Tielines Microwave 7
State Police Radio system 8
Maine Yankee security Radio 9
Plant Telephones PBX Dial 10 Plant Intercom System 4 Lines + Pegf r.g 11 CMP Dispatcher Augusta 1
CMP (For emergency use with patch Microwave to other locations) 12 Yankee Nuclear Services Division PABX Dial Tieline 1
Utility Microwave system TABLE 1 (cont)
Maine Yanke q
Telecommunications Communications Facilities From These I,ocatirns 7,o oca lons Listed on Le h Col un Telephone I
Intercom.
Technical Support Emergency Coord Control Room l'
Center center Two-way Radio 1/ 0/8.0
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Control Room 9,10
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Technical Support 9*10 9,10 Center l
s Emergency Coordination Center 7,9,10 9,10 l
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State Police 1,2,6,7,11 i
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State EOC 1,2,6,11.
1,2,6 1,2,5,6 (CEI,)
i NRC 1,2,4,11 1,2,4 1,2,3,4 DOE Department of Energy 1,2 1,2 1,2,3 i
i Corp. lleadquarters 1,2,6,11 1,2,6 I
1,2,3,6 (Augusta)
Yankee Nuclear Servi'e-c 1,2,6,12 1,2,6,12 l
1,2,3,6,12 Division (Westboro) l Emergency News l
Element
- 6 UJ 1,2,3,6 Plant Security 9,10' 9,10 3,9,10 M LE 1