ML20049A506
| ML20049A506 | |
| Person / Time | |
|---|---|
| Issue date: | 02/17/1981 |
| From: | Stone J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Bryan S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19264A269 | List: |
| References | |
| FOIA-82-93 NUDOCS 8104060352 | |
| Download: ML20049A506 (1) | |
Text
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TI 2515/50 Issue Date:
2/17/81 MINIMUM INSPECTION EFFORT THAT MUST BE PERFORMED BY RESIDENT INSPECTORS AT LICENSED POWER REACTORS A.
Objective To ensure that systems and functions important to safety are routinely inspected during periods of severe manpower limitations or high reactive inspection effort at all licensed facilities.
B.
Background
We have experienced occasions when the reactive inspection effort has required a preponderant amount of both resident and region based inspector time.
The most recent effort was the crash Inspection Program associated with the review and verification of the Three Mile Island action items.
The purpose of this Temporary Instruction is to identify the minimum inspection effort that must be routinely completed by the Resident Inspection staff at each operating power reactor facility regardless of the amount and significance of any nonroutine inspection effort that may be identified in IE Headquarters or Regional Office communications.
Preliminary information relating to this matter was discussed in a memorandum, V. Stello, Jr., Director, Office of Inspection and Enforcement to the Regional Office Directors, dated November 25, 1980, and has been incor-porated in this Temporary Instruction.
/ identifies those specific daily and weekly activities / functions that must be routinely inspected by Resident Inspectors at operating power s
reactors regardless of other ongoing activities.
The only exception is when there is an ongoing operational event at the plant.
Then the resident inspection staff must follow the event until redirected by the Regional Office.
C.
Schedule This TI must be implemented whenever the requirements of the following modules cannot be completed; 1.
61726 - Monthiy Surveillance Observation 2.
62703 - Monthly Maintenance Observation 3.
71709 - Operational Safety Verification However, such implementation requires Regional Office approval by the Director, Division of Resident and Project Inspection.
This approval has an effective life of up to two mo,,tl.s.
Any further implementation of this TI requires the approval of the Regional Director.
Section I, and II of Enclosure 1 must be completed on a daily basis, i.e., at least 4 days out of a normal 5 day work week, whenever the plant is operating or temperature is above 2000F.
Section III of Enclosure 1 must be completed at least every 10 calendar days.
Section IV of enclosure 1 must be completed on a daily basis, i.e., 4 days out of a normal 5 day work week and Section V must be completed every 10 days whenever the plant is shutdown and temperature is below 200 F.
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Page 2 of 2 Issue date:
2/17/81 D.
Reporting Results of this inspection effort should be documented in routine inspection reports.
E.
Expiration, This-Temporary Instruction will remain in effect until the provisions are incorporated in MC 2515.
F.
IE Headquarters Contact Regional Office questions regarding this TI should be addressed to F. J. Nolan (492-8019).
G.
Module Tracking System Input (766 Data)
For module tracking system input, record actual inspection effort against module No. 25550B
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TI 2515/50 Page 1 of 5 Issue Date:
2/17/81 g
ENCLOSURE I MINIMUM ROUTINE INSPECTION EFFORT THAT MUST BE COMPLETED BY THE RESIDENT INSPECTION STAFF AT EACH OPERATING POWER REACTOR SITE REGARDLESS OF OTHER IE HEADQUARTERS OR REGIONAL 0FFICE DEMANDS I.
During Daily Entry and Egress from the Restricted Area:
Verify that the personnel radiological monitoring equipment (Portal monitors, etc.) is operable and that equipment and materials (exclu-sive of personal items) are monitored or evaluated for radioactive materials prior to release for unrestricted use.
II.
Daily Control Room Activities - Operating Facilities General:
Each inspector shall use the unit specific checklist (developed pursuant to Module-71709) for Ahe various operating modes to be used in verifying the licensees adherence to LC0 Requirements for those items observable at control room panels, in the control room and during plant tours.
A.
Verify proper control room manning and that access is controlled.
B.
Review control room and shift supervisor log books and operating orders to obtain information concerning operating trends and activities, and to note any out-of-service safety systems.
Apparent anomolies may require followup to assure compliance with regulatory requirements.
C.
Verify the status of selected control room annunciators and assure that the control room operator (s) understand the reasons why important annunciators are lit.
In addition, if an off-normal condition or false annunciation signal exists, the inspector snould assure that appropriate actions to return the situation to normal have been initiated. The inspector should verify that the corrective action has been initiated and completed in a timely manner.
D.
Review panel boards for the engineered safety feature systems and verify that the redundant train is available for service whenever one train is tagged out of service.
In addition, confirm that selected valves are positioned appropriately for operation.
E.
For PWR Facilities:
1.
Verify that the Panel Board display of control switches, indicating lights, etc., reflect the proper lineup for the auxiliary feedwater system trains.
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.9 Page 2 of 5 Issue Date:
2/17/81 ENCLOSURE I 2.
Verify that containment temperature and pressure are in conformance with LC0 requirements.
3.
Verify that control rod insertion limits (to assure shutdown margin) and power distribution limits are in conformance with Technical Specification (TS) requirements.
F.
For BWR Facilities:
1.
Verify that power distribution limits are in conformance with TS Requirements.
2.
Verify by examining panel indication that-the Torus Cooling System is aligned for service.
3.
Verify that primary containment temperature and internal pressure are in conformance with TS.
In addition, verify that the differential pressure between the drywell and suppression chamber is in conformance with TS.
G.
Review stack monitor recorder traces for the period since the previous review and followup on any indication of an apparent release.
H.
Verify by examining the panel indications that required onsite and offsite emergency power sources are available and aligned for auto-matic operation.
I.
Review the reactor protection system panel indications and verify conformance with TS requirements (LC0's).
III.
Weekly Inspection Requirements - Facility in an Operating Mode:
A.
Observe portions of a selected ongoing ESF maintenance activity and verify that the redundant train is available for service.
In addition, assure that approved procedures are used for tasks that appear to be beyond the normal skills of the worker and that the work is being per-formed by qualified personnel.
B.
Observe portions of a selected ongoing ESF surveillance and verify:
1.
Test instrumentation is calibrated 2.
Redundant system or component available for service 3.
Approved procedures used 4.
Work performed by qualified personnel C.
Review the results of a sample of at least one liquid and one gaseous release made during the week and verify conformance with regulatory requirements prior to release.
F.
TI 2515/50 Page 3 of 5 Issue Date:
2/17/81 ENCLOSURE I D.
Review the results of the most recent sample of the boric acid tank and the standby liquid control tank and verify that the con-centration was in conformance with TS.
In addition, confirm that tank levels are as specified.
E.
Conduct a visual inspection of selected piping between containment and the isolation valves for leakage or. leakage paths.
This should also include verification that manual valves are shut, capped and locked when required and that motor operated or air operated valves are not mechanically blocked.
F. 'During the performance of the above inspection activities and during entry and egress of plant and during tours examine the following on a sampling basis:
1.
Proper completion and use of selected radiation work permits.
2.
Proper use of protective clothing and respirators.
3.
Proper personnel monitoring practices (wearing of badges and j,'
use monitoring equipment).
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4.
Ignition sources and flammable material are being controlled in accordance with licensee procedures.
5.
Equipment tag outs in conformance with controls for removal of equipment frcm service.
6.
Plant housekeeping and cleanliness practices are in conformance with approved licensee programs.
7.
Normal security practices are being followed.
IV.
Daily Control Room Activities at Facilities in a Shutdown Mode - Cold Shutdown or Refueling A.
Verify proper control room and shift manning and that access is controlled.
B.
Review control room and shift supervisor log books and operating orders to obtain information concerning operating trends and activities.
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Issue Date:
2/17/81 ENCLOSURE I C.
Observe a sample of important control room annunciators ' expected to be in an operating stata for the specific shutdown mode and assure that the operator understands the reason for any abnormal indication.
D.
For PWR Facilities:
1.
Verify that the decay heat removal system is operable with a redundant heat removal path available.
2.
Boric acid concentration in conformance with T.S.
E.
For BWR Facilities:
1.
Verify that the mode switch is in the proper position 2.
Verify that one shutdown cooling mode loop of the residual heat removal system is operable.
F.
Verify by examining panel indications that the required emergency power sources are available.
G.
Verify by examining panel indications and interviewing operators that the required containment configuration is established.
H.
Review the jumper log to identify recently installed jumpers and then-verify that two are located as described.
J V.
Weekly Inspection Requirements at Facilities in a Shutdown Mode:
A.
Tour accessible areas of the facility to make an independent assessment of plant conditions and equipment conditions.
The following items should be observed or verified on a sampling basis during the tour:
1.
Selected vital area barriers are not degraded 2.
Proper completion and use of selected radiation work permits 3.
Proper use of protective clothing and respirators 4.
Proper personnel monitoring practices (wearing of badges and use monitoring equipment) 5.
Operational status of selected personnel monitors, area radiation monitors and air monitors.
6.
Proper posting, barricading, and control of selected radiological areas, i.e., radiation areas, high radiation areas, airborne activity areas and surface contamination areas 7.
Plant areas (including cabinet interiors) for fire hazards 8.
Fire alarms, extinguishing equipment, actuating controls, fire fighting equipment and emergency equipment for operability.
9.
Ignition sources and flamnable material are being controlled in accordance with licensee procedures
- 10. Review a sample of equipment tag outs to verify that the licensee has complied with TS LC0's regarding removal of equipment from service
- 11. Plant housekeeping and cleanliness practices are in conformance with approved licensee programs
TI 2515/50 Page 5 of 5 Issue Date:
2/17/81 ENCLOSURE I B.
Observe portions of a selected ongoing ESF maintenance activity and verify that the redundant train is available (as required) for service.
In addition, assure that approved procedures are used for tasks that appear to be beyond the normal skills of the worker and that the work is being performed by qualified personnel.
C.
Observe portions of a selected ongoing ESF surveillance and verify:
1.
Test instrumentation is calibrated 2.
Redundant system or component available for service 3.
Approved procedures used 4.
Work performed by qualified personnel D.
Verify by review of records and discussions with personnel that alarm set points for selected process and effluent radiation monitors are in conformance with requirements.
VI.
Other Requirements A.
The Resident Inspector must review major unscheduled transients and any unplanned reactor or turbine trips as described in inspection.
module 93702.
B.
The Resident Inspector must review details concerning any significant liquid or gaseous release reported to the Commission in conformance with 10 CFR 50.72.
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s October 7, 1981 SECY-81-582 POLICY ISSUE For:
The Comisspp6trsation Vote)
From:
William J. Dircks, Executive Director for Operations
Subject:
TMI Action Plan II.F.2 (NUREG-0737); Additional Instrumentation for Detection of Inadequate Core Cooling
Purpose:
To inform the Comission of the staff's current position on the placement of thermocouples to monitor core cooling conditions for BWRs, and to provide a status report and recommendations concerning the implementation schedule for reactor vessel level instrumentation for PWRs.
Background:
BWR Core Thermocouples Regulatory Guide 1.97, revised in accord with the TMI Action Plan, states that thermocouples should be installed in BWRs to monitor the cooling conditions in the reactor core. The ACRS, in August 1981 letters reporting on the review of the Fenni and Susquehanna reactors, recommended that the location of such thermo-couples be reevaluated. The staff agrees and describes its approach for doing that, below.
PWR Reactor Vessel Water Level System Item II.F.2 of NUREG-0737 and the TMI Action Plan require implementation of reactor vessel level instru-l inentation in PWRs. The ACRS has expressed concern I
about the schedule required for implementation of this requirement being too tight. We have assessed the progress of industry in development and installation of these systems and recommend that the existing imple-mentation date of January 1982 be replaced. The earliest practical date by which these systems could be installed, calibrated, and operable at all operating PWRs is the first refueling after January 1,1983, for reasons discussed below.
Discussion:
Status of BWR Core Thermocouples The ACRS supported use of core thermocouples in BWRs in its letter of November 10, 1980 to the NRC Executive Director for Operations, but called attention to the need for further study to determine the appropriate vertical location of such thennocouples.
In the
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August 11, 1981 ACRS letters (reporting on nd
Contact:
L. E. Phillips, NRR:DSI:CPB g
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. Susquehanna reviews) from ACRS Chairman J. C. Mark to NRC Chairman N. J. Palladino, the ACRS recommended that further study be given to placement of a small number i
of thermocouples in a more accessible location.
Regu-latory Guide 1.97 Revision 2, provides for inclusion of a minimum of four thermocouples in each BWR core quadrant but does not further specify the location of such thermocouples.
The staff did study the placement of thermocouples in i
BWRs in the process of formulating the requirements in Regulatory Guide 1.97.
Our study included our own calculations and those of consultants in addition to a review of General Electric Company studies and recom-i mendations on the subject.
We concluded that inclusion of core thermocouples, one in each of several Local Power Range Monitor (LPRM) assemblies, would provide operationally useful data about core cooling when the ECCS core spray is degraded.
We also concluded that the vertical location of the thermocouples in the LPRM assemblies was not critical and that the location in LPRM assemblies used by GE in its analyses was the most practical.
Our evaluation is reported in the LaSalle SSER (NUREG-0519) and was called to the attention of ACRS in a memorandum from H. Denton to Chairman Mark (Enclosure 1) and in staff presentations during the ACRS review of Fermi 2.
However, we have no reason to disagree with the ACRS conclusion that placement of the thermocouples at another location may prove to be more practical and just as useful for accident diagnosis.
Therefore, in response to the ACRS recommendation for a reevaluation, we have prepared a letter for BWR appli-cants and licensees (Enclosure 2) requesting that they perform a study to confirm the most suitable location for thermocouples for their facility.
We will en-courage and expect to receive generic responses, perhaps for the various classes of BWRS. We do not expect this to affect the June 1983 implementation date i
of Regulatory Guide 1.97, i
l Although this approach will be responsive to the stated concern of the ACRS, it will not completely resolve all controversy surrounding these thermocouples.
The BWR l
owners are expected to again appeal the need for the thermocouples to the ACRS in the near future.
Status of Reactor Vessel Water Level Measurement Systems for WRs The purpose of a reactor water level measurement system is to provide the reactor operator with instrumentation to directly monitor water level in the reactor vessel and thereby the adequacy of core cooling so that he
I may implement actions to correct or avoid conditions of inadequate core cooling.
Instrumentation presently
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available in PWRs does not provide an unambiguous indication of availability of cooling capacity for the core.
At this time, additional instrumentation proposed by applicants and licensees includes three types of water level measurement (Enclosure 3):
(a) Westinghouse Reactor Vessel Level Instrumentation System (RVLIS),
(b) Combustion Engineering Heated Junction Thermocouple (HJTC) system, and (c) National Nuclear Corporation (NNC)/EPRI Neutron Detector system.
In addition, a modified differential pressure (dp) system which includes hot leg dp measurement to the top of the candy cane is-under study by some B&W reactor applicants and licensees.
It is expected that the details of such a system will be submitted for review soon by B&W or one of the B&W reactor owners.
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Westinghouse dp System The Westinghouse RVLIS utilizes two sets of three dp cells.
These cells measure the pressure drop from the bottom of the reactor vessel to the top of the vessel, and from the hot legs to the top of the vessel.
This dp measuring system utilizes cells of differing ranges to i
monitor different flow behaviors with and without pump operation. The dp signal is processed and converted i
into the reactor water level.
When the reactor coolant pumps are not operating, the RVLIS. reading will be indicated on the narrow range scale ranging from zero to the height of the vessel.
This reading represents the equivalent collapsed ~ liquid level in the vessel.
When the reactor coolant pumps are operating, the RVLIS reading will be indicated on the wide range scale.
With the pumps running the RVLIS reading is an indication of the void fraction of the vessel water and steam mixture.
Upper range RVLIS is used for head venting 4
operations during long term recovery.
The Westinghouse summary report, " Westinghouse RVLIS System),-(UHI Plant)quate Core Cooling (ICC) (7300, and (Microp for Monitoring Inade L
submitted in December 1980 and is under staff review.
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-The review includes results from a test program being perfonned at the Semiscale facility in Idaho.
Our review and~the test program are consistent with the i
original plan (Enclosure 4) but are approximately one l
month behind at this time.
We expect to complete our i
generic evaluation, except for the analysis of the I
final performance test, by the end of 1981.
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. The Westinghouse dp system has been selected by licen-sees for use in eighteen reactors.
We are told that twelve of these reactors will have the system installed prior to the current January 1, 1982 deadline in NUREG-0737.
The systems for Zion Units 1 and 2 are already installed.
Most of the operating license applicants who have selected the Westinghouse system expect to complete installation prior to power operation (Summer 1, McGuire 1 and Diablo Canyon).
Licensees have requested that installation be delayed until the first refueling outage for Sequoyah Units 1 and 2, North Anna 1 and 2, and Salem 2.
The staff concludes that those plants which have coninitted to install the Westinghouse dp system are substantially in compliance with the NUREG-0737 pro-vided that major technical problems are not encountered in proof tests of the system that affect its generic approval.
One problem attributed to an atypicality in the Semiscale test model has been encountered and is expected to be resolved by a final performance test to be performed in mid-November.
Evaluation of this additional test will delay the generic SER until January 1982.
In addition, a LOCA survivability test to evaluate the instrument performance after it has been subjected to large break blowdown conditions will be perfonned in February 1982.
However, this test is confinnatory in nature since the objective is to assure a correct interpretation of the instrument readout after such an event, and not necessarily to require continued operability as a design basis.
The results of the large break LOCA survivability tests and con-clusions on the post-LOCA interpretation of level instrumentation will be reported in conjunction with our review of calibration test data obtained from plant specific reviews.
In summary, installation of Westinghouse dp systems are expected to be completed on schedules ranging from the present to the first refueling after January 1,1982.
We believe that our generic review of the system will be substantially complete by the end of 1981 and that approval of plant specific installations for incorpo-ration into plant operating procedures can be completed on schedules ranging from mid-1982 to mid-1983, depending on the submittal schedules for final system descriptions and calibration data from hiividual plants.
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. Combustion Engineering HJTC System The HJTC system measures reactor coolant liquid inven-tory with discrete HJTC sensors located at different levels within a separator tube ranging from the top of the core to the reactor vessel head.
The basic princi-ple of system operation is the detection of a temperature difference between adjacent heated and unheated thermo-couples.
In a fluid with relatively good heat transfer properties (e.g., water), the temperature difference between the adjacent thermocouples is very small.
In a fluid with relatively poor heat transfer properties (e.g., steam), the temperature difference between the thermocouples is large.
The separator tube provides for a steam-water interface at the collapsed liquid level and thermocouples at discrete axial levels within the tube indicate the presence of water at the measure-ment level.
The description of the Combustion Engineering (CE) HJTC system has been documented by individual users and is currently under staff review. We have been in connunica-tion with CE and with the CE Owners' Group in an effo,t to conduct the review generically.
The review of the initial submittal on the docket for San Onofre Units 2 and 3 (Amendment 23, February 1981, Docket Numbers 50-361/362) is one to two months behind the review schedule originally projected (Enclosure 4).
However, functional performance test data on the system have not yet been submitted even though the tests have been run and the vendor claims good results.
Performance tests on the production probe should be complete by the end of the year.
On this basis, we estimate that generic approval of the system can be completed near the end of the first quarter of 1982.
The installation of a HJTC system requires modification of incore instrument flange assemblies, containment penetration assemblies, and mineral insulated cabling.
For San Onofre, which is the lead plant using the HJTC system, the required plant modifications can not be completed prior to mid-1982, even though the HJTC probe assembly can be delivered during the first quarter, 1982 (Enclosure 5).
Southern California Edison has requested that the installation of the system be delayed until the first quarter,1983, during its first refueling shutdown.
. Those plants that have selected the HJTC system (there are eight in total and several ordered the system as early as one year ago) cannot physically complete installation before the latter part of 1982, even though the probe assemblies may be available early in the year.
Therefore, plants comitted to the CE system are not able to meet the current NUREG-0737 imole-mentation schedule. We believe that all plants comitted to this system will be able to complete installation on a schedule consistent with San Onofre, i.e., first refueling after January 1,1983.
Staff review of individual installations and calibration data and approval for incorporation into emergency procedures will probably extend to mid-1984, depending on individual submittal schedules.
NNC/EPRI Neutron Detection System The principle used is the detection of photoneutrons from the reactor coolant system.
The ratio of the count rates from two sets of detectors (one set above and the other below the reactor vessel) provides an indication of reactor water level.
This system was proposed by Alabama Power Company for interim use and developmental testing on Farley Units 1 and 2.
Documentation of early experience with the system was provided in July 1981 and is currently under staff review.
We have requested that additional information be provided by October 1.
The system is still in the research and development stage and we are not optimistic that it will gain our approval. The licensee projects that the system could become operable by mid-1983.
If the prospects for this system are not more favorable b; early 1982, we will urge the licensee to select another system.
Vessel level Systems to be Proposed A vessel level system based on differential pressure measurements is currently being evaluated by B&W and B&W Owners.
We estimate that I to 2 years would be required to complete the design evaluation, procure-ment, and installation of such a system after completion of preliminary design.
Progress by licensees and OL applicants with B&W reactors has been uniformly poor.
The staff will not license a B&W reactor that does not show reasonable progress in this area.
We have taken a consistent position with TMI-1 for approval of restart.
In view of the good progress being made by others, we believe it is justified to require that B&W owners have their systems installed and operable before restart following the first refueling outage after January 1, 1983.
However, somewhat later dates may be acceptable, as described in the Recommendation, below.
Recommendation:
It is recommended that the Commission allow the staff to delay the implementation schedule for PWR applicants and licensees for installation of the vessel level measurement system.
In the next two weeks, NRR will describe an approach for renegotiating practical implementation schedules for NUREG-0737 requirements for all operating reactors.
If the Commission endorses that approach, then this requirement for vessel level indicators in PWRs would be decided on a case by case basis.
The decision for each plant would be made in conjunction with decisions on the other TMI work remaining to be done on that plant and in full view of the equipment procurement and installation constraints described above.
As a general matter, we would expect the result to be that most of the level measurement systems would be installed before startup following the first refueling after January 1,1983.
William J. Dircks Executive Director for Operations
Enclosures:
1.
Memo for Chairman Mark fm H. Denton, " Installation of Core Thermocouples in LaSalle."
2.
Draft ltr to BWR Applicants & Licensees.
3.
Information Report for Commissioners from W. J. Dircks, SECY-80-529.
4.
Staff Review Schedule for NUREG-0737 Section II.F.2.
5.
Figure 8-1 Earliest Possible ICC Installation Schedule - Songs 2.
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., to the Office of the Comissioners' coments should be provided direi' Secretary by c.o.b. Thursday, October 22, 1981, Comission Staff Office coments, if'any, should I.a submitted to the ition copy to the Comissioners NLT October 15, 1981, with an infid Office of the Secretary.
If the paper is of sui 'l.i nature that it requires additional time for analytical review M'; coment, the Comissioners and the Secretariat should be apprised of when ci;.aents may be expected.
DISTRIBUTION Comissioners Comission Staff Offices Exec'Dir for Operations Exec Legal Director ACRS ASLBP ASLAP Secretariat i
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JUL 141931 MEMOPJdDUM FOR:
J. Carson Mark, Chairman, ACRS j
FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation t
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SUBJECT:
INSTALLATION OF C0F.E THEF.':0 COUPLES IN LaSALLE.
In yeu-ietter cf April 16 to Chairri.. Herdrie reprdinc LaSalle County P
Station Units 1 and 2, the Cccittee recc..en:ie:: Inat a study of tne feasibility of the use of core outlet or core subassembly thermoccuples be completed prior to reaching a decision on the recuirement for the LaSalle plant.
J This is to call to your attention Section 18, item (3) of the LaSalle l
SER Supplement No.1, NUREG-0519, where we described the staff's study.
In addition, o' r report in response to the Committee's recocmendation i:,
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u included in Section 22, Item II.F.2, of the same document.
The staff j
concluded that the use of incore thermocouples in BWRs to provide diverse indication of water level and to monitor core cooling is feasi-ble.
We are requiring LaSalle and all other BWR's to incorporate thermocouples into the ICC monitoring system prior to June 1983 in accordance with Regulatory Guide 1.97.
In a related matter, your letter of June 9 to Mr. Dircks requested an early heeting of the staff with ACRS to discuss issues relating to Section II.F.2 of NUREG-0737.
We expect to have that meeting in September.
If the conmittee has questions concerning our LaSalle report or our intent with other BWRs in this area, the staff will be prepared i
to discuss them at that meeting.
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g, Harold R. Denton, Director 4
Office of Nuclear Reactor Regulation
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DRAFT ENCLOSURE 2 LETTER TO BWR APPLICANTS AND LICENSEES As-you know, the staff has stated its intention to condition operating licenses for BWRs to require inclusion of thermocouples in the core.
Our evaluation of the placement of the core thernoccuples is discussed in the LaSalle SSER (NUREG-0519) which concluded that location of one thermocouple in each of several LPRM assemblies (four per core quadrant) was a feasible and acceptable approach.
Based on the GE BWR Owners' Group appeal of this requirement in the August 6 ACRS review meeting on Fermi 2, the ACRS recommended that further study be given to placement of a small number of thermocouples in a more accessible location.
We request that you perform such a study and confirm that the LPRM assemblies are the most suitable location or propose an alternate location and/or number of thennocouples to monitor inadequate core cooling.
If an alternative is proposed, provide an evaluation of the advantages and disadvantages of the recommended installation in terms of relative effectiveness to monitor the approach and existence of inadequate core cooling and in terms of the installation and maintenance costs in dollars and occupational doses, including how these costs may be a function of the required installation date now specified as June 1983 in Regulatory Guide 1.97.
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i ENCLOSURE 3
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NUCLEAR REGULATORY COMMISSION WASHINGTON, c. c. :osss 4
'INFORMATION REPORT FCR:
The Ccm issioners FRCM:
William J. Dircks, Executive Director for Operations
SUBJECT:
TMI ACTION PLAN - II.F.2; ADDITIONAL INSTRUMENTATION FOR MEASUREMENT OF COOLANT LEVEL IN REACTOR YESSEL PURPOSE:
To provide the Comission with an infomation paper on the
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status of the technology for measuring reactor vessel water level.
BAC5: GROUND:
During the Octooer 28, 1980 Ccmission meeting regarding clarification of TMI Action Plan requirements, the staff informed the Comission that it would provide an information paper that would discuss the status of the technology for measuring reactor vessel water level and provide a basis for its reccmended impler,entation schedule. The attached document is provided in response to that ccmitmmt.
DISCUSSION:
The status of Instrumentation for Detection of Inadequate Core Cooling (ICC) was discussed at the LOFT / UTILITY Technology Transfer Meeting (LOFT Meeting) on October 16 and 17,1980 at Idaho Falls, Idaho. Presentations were made by NRC (NRR and RES), INEL, ORNL, Westinghouse, CE, EPRI and DAVCD covering the NRC requirements and research program, the state of the art and current industry programs. A discussion following the session was conducted among the utilities, instrument suppliers, reactor vendors, and NRC. The main concerns discussed at this meeting were:
Requirements of the TMI Action Plan Task II.F.2 Status of Current Industry Planning PRC research coordination NRC plan and action The enclosure to this paper provides details. The principal concern expressed is that there has not been sufficient systems-level testing to demonstrate the acceptability of any device proposed to measure vessel level, and that the required imple-mentation date of January 1,1982 is thus in some jeopardy.
Centact:
T. Huang, NRR, 49-29420 L. Phillips, IRR, 49-27140 l
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The staff believes that further delay of the required implemen-tation of level measurement systems beyond January 1,1982 would result in slower progress and may or may not result in better level measurement systems which would further enhance plant safety.
We therefore believe that the current schedule requirements (January 1,1981 for selection and documentation of the measure-ment system and January 1,1982 for installation) should be maintained.
However, we also believe that scrne flexibility should be maintained to permit development of improved systems where the specific documentation submittals include or subscribe to a viable development and testing plan to prove the concept selected on reasonable schedule.
/
William Dircks Executive Director for Operations
Enclosure:
As stated s
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_El.CLOSukE Th! LUf;CEEN 0F ThI ACTICN PLAf4 II.F.2 TC LLTLRMINE A LOGICAL IMPLEMEhTATICh DATE FOR REACTOR VESSEL LEVEL INS'TRuhEhTAT10N l
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NRC Re:uirements Before THI, hRC had no requirements for instrumentation for detection of inadequate core cooling.
Following the TMI accident, the Lessons Learned Task Force (NUREG-0578) and the Office of Nuclear Reactor Reg-ulation (the 1979 clarification letters) have established requirements for the instrur.entation for detection of inadequate core cooling. As stated in NUREG-0576, Category A items shall be implemented prior to January 1,1980, and Category S items prior to January 1,1981.
(Cate-gory A: develop procedures and describe existing instrumentation; new instrurent design, subcooling meter. installation, and implementation schedule.
Category B: complete new instrument installation.) This implementation schedule was described in the October'.1979 lette~r from H. R. Denton (NRC) to All Operating Nuclear Power Plants and confirmed in NUREG-0660, "NRC Action Plan." Consistent implementation schedules for NT0Ls are required by NUREG-0604, "Thl-Related Requirements for New Operating Licenses."
However, based upon the staff's review of state-cf-the-art, design progress, and the equipment procurement situation, the staff has recently proposed that the implementation schedule for reactor vessel level Category B requirements be rcscheduled to January 1,1982 and the documentation required for Category B should be reported by the licensee by January 1,1981. This latest implementation schedule was described in a clarification letter dated October 31, 1980 to all Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits. As indicated by the following discussion, even the revised schedule is in jeopardy.
s 2.
Status of Current Industry Planning Four types of systems for detection of Inadequate Core Coolink (ICC) were described in the LOFT / UTILITY Technology Transfer Meeting.
The instru-mentation described included the Westinghouse Differential Pressure
' Measurement of Reactor Vessel Water Level (Figure 1), the CE Heated Junction Thermocouple System (HJTC) (Figure 2), the EPRI sponsored Noninvasive Water Level heasurement Technique (Neutron Detectors) (Figure 3),
and the DAVC0 hicrowave Liquid Level Gauge for A High Radiation Environment (Figure 4). Among those four systems three (excluding the DAVC0 system which is under development) have been' currently proposed by utilites for use to meet NRC requirements. The DAVC0 Microwave Liquid Level Gauge was demonstrated in the presentation and aroused considerable interest from the utilities, the research community and the NRC. However, it still needs an extensive research and development effort before it could be installed in an operating, reactor.
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Baseo en the presentation in the n.ecting, it appears that each of the incustry proposed systems has inherent covantages and disacvantages.
6:ever, none of the proposec systems has been deranstratec to be reliable uncer various simulated accident concitions which could challenge the valicity of the c4asurements.
It should be noted that some utilities, niostly B&W owners, are taking the position j
that no additional ICC instrumentation is neeced based on their analyses. They I
rely primarily on the subcooling n,onitor to detect the advent of ICC and on core exit thermocouples and hot leg RTDs to cetect the near advent and exist-ence of ICC.
They contend that existing level ueasurenant n.ethods do not meet all of the NRC design requirements. The staff has indicated in corres-pondence to BLK owner utilities (letter fron. D. Eisenhut dated September 29, 1980) that this position is unacceptabic to the staff. The early warning of the subcooling meter is ambiguous (not necessarily loss of coolant); a reliable level measurement system would enhance detection of the advent of ICC anc pro-vide prompt indication of the effectiveness of ceasures to prevent or recover from the condition.
3.
T<RC Research Coordination Ecth INEL and ORfit presented the status of developnIent of instrumentation for detection of ICC during the LOFT Eeeting. They have investigated the use of absolute temperature type thermocouples, heated junction type thermo-t couples, a torsional ultrasonic level probe, etc. (see Table 1) and have perforned an analysis of the c.ethods consicered achievable for detection of-ICC. The data obtained from these RES research programs are promising; however, the instrumentation systems need further development and testing to support design for installation in com.cecial reactors. fiRC/ DOE research facilities which can be made available for such testing include THTF, SEMISCALE, and LOFT. LOFT is an especially appropriate test bed for final integral
.I testing since it is a PWR with most of the test relevant characteristics of a comercial PWR, but includes extensive instrumentation systems not available on a concercial PWR.
The instrumentation relevant to ICC monitoring f.ncludes thermocouples at discrete levels above and within the core (including fuel i
clad thermocouples), a conductivity probe for level monitoring, and gama densitometers to indicate coolant quality conditions in the coolant piping.
The objectives.of power plant instrumentation research sponsored by RES are (1) to provide proof of principle for selected conceptual designs, (2) to provide confirmatory test data on specific conniercial designs, (3) to extend the instrumentation technology where appropriate, and (4) to pro-vide a technology transfer from research programs. The research is to pro-vide developmental data relating to functional capability; the qualifica-t1on testing of systems and components must be performed by the industry.
In response to the TMI Task Action Plans I.D.5 and II.F.2, RES has been con-ducting work on possible liquid level measuring concepts.
In order to advance NRC i
licensing requirements, the facilities for reactor safety research such as LOFT, SEMISCALE, TNTF and advanced instrucentation test facility, etc. will be used, under hRC research coordination, to test industry supplied prototype systems which are being considered for installation in operating reactors.
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The vendors and NS will neet and establish tests nee:ed fer the industry pr:; posed systems at urious test facilities.
In scre insW;ts wl.ere tne testing schedule at MC/00E facilities does not sucport rroduct develo: cent on a schedule consistent with the January 1,19?? installa-tion requirenent, %c supplier is planning to perform essential ceveles-ment testing at inoestry facilities.
Later testing at NRC/DCE facilities would then be confirmtory in nature.
The status of propusec test plans foilows:
(1) The Westinghouse Reactor vessel Level Instrumentation System (RYLIS) is being installed in the SEMISCALE facility.
Data will be taken during the upcoming test program planned for the last cuarter of CY 1980, and the testing is expected to be continued until the end of fiscal year 1981.
(2) The CE Reactor Vessel Level Monitoring System (RVLMS) is in the final design stage; however, CE's prototype has been tested in THTF at ORNL.
An additional test program at CE facilities is in progress.
Installa-tion of the CE system for testing in either the SEMISCALE or in the LOFT facility is being negotiated between RES and CE. The first test may occur early in 1981 and the program possibly concluded by the end of fiscal year 1981.
(3) The EPRI sponsored NNC Neutron Detector (which is being installed for testing in Tarley 1 & 2) is being installed in the LOFT facility.
Installation is exrected to be completed in tirae for the L3-6 test in mid-December loSO, which includes partial core uncovery. Testing is expected to be completed by the end of FY 81.
(4) The DAVC0 Microwave Liquid Level Gauge is still in the development stage.
It has to be subjected to proof of principle testing in separate effects facilities, such as autoclaves, prior to an integral test in the LOFT or other facility.
To go through a prerequisitt series of tests will take approximately six months to one year.
DAVC0 has provided window materials to NRC for materials testing which is now in progress. The prototype could then be installed in integral test facilities for tests which may be concluded by the end of the fiscal year 1982.
Potential industry sponsors of this system indicate that an effort will be made to complete the development at an earlier date.
(5) The torsional ultrasonic probe (Figure 5) is still. in the research and development state and has no comercial sponsors. However, the probe is available and can be tested and improved. The test may be perfor1r.ed in the early part of 1951 and could be concluded by the end of the fiscal year 1981.
4 NRC Plan and Action None of the instrurentation systems proposed by the industry to date have been tested under simulated reactor operating conditions to substantiate l
functional reliability and accuracy, i
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The utility representative' who attended the LOFT meeting have requested that the II.F.2 implementation dates be extended to allow completion of instrument development programs before a system must be selected for pro-curement. They have expressed the concern that systems procured prior to the completion of development and testing may prove to be unacceptable to NRC, and would result in unnecessary expense.
The staff is of the opinion that existing technology provides reasonable confidence that a Op measurement system, comparable to the Westinghouse system, can enhance the quality of ICC monitoring systems. Additional analyses and testing of existing systems are required to identify any situations where erroneous measurements may be indicated and to provide for data processing thich identifies and negates or alarms such erroneous indications if they should occur. Also, the staff may need to reconsider the advisability of some requirements (e.g., level measurement with pumps running) based on the analyses and test results.
'Je believe that heated thermocouple systems and absolate temperature type thernocouple systems monitoring deviation from s3turation at discrete axial locations can be provided with only slightly less confidence in their ultimate acceptability. Other systems promise potential advantage in simplicity and possibly in reliability but have greater risks with respect to development schedule and ultimate acceptability.
The staff believes that further delay of the required implementation of level easurement systems beyond January 1,1982 would result in slower progress and may or may not result in better level measurement systems which would further enhance plant safety.
'Je therefore believe that the current schedule requirements (January 1,1981 for selection and documentation of the measurement system and January 1,1984 for installation) should be maintained. However, we also believe that some flexibility should be maintained to permit development of improved systems there the specific documentation submittals include or subscribe to a viable development and testing plan to prove the concept selected on reasonable schedule.
The burden to arrange for appropriate test programs and to provide the selected systems for testing will be on the industry.
NRC will cooperate by making NRC/
DOE facilities available for testing and by evaluating the acceptability of some instrumentation concepts by the end of CY 81.
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(Dif ferential Temp. Type) prototype units shield Possible LOFT to be scheduled Indicates wet or dry surface reflect-Heated T/C INEL New prototype; No comercial
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& development problems ll:utron Detector NNC Proof of prin-Reliabilityof LOFT late 1980 No leads in vessel (EPRI ciple needed signal inter-sponsored)
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SECY-81-582A December 29, 1981 (Notation Vote)
For:
The Comissioners From:
William J. Dircks, Executive Director for Operations
Subject:
Implementation of TMI Action Plan II.F.2 (NUREG-0737) for Babcock and Wilcox Reactors; Additional Instrumen-tation for Detection of Inadequate Core Cooling
References:
(1) SECY-81-582, "TMI Action Plan II.F.2 (NUREG-0737);
Additional Instrumentation for Detection of Inadequate Core Cooling," dated October 7, 1981.
(2) Memorandum from S. J. Chilk to W. J. Dircks, "SECY-81-582 - TMI Action Plan II.F.2 (NUREG-0737);
Additional Instrumentation for Detection of Inade-quate Core Cooling," dated November 16, 1981.
Purpose:
To: (1) inform the Comission concerning current plans of Babcock and Wilcox (B&W) reactor owners for compliance s
with TMI Action Plan II.F.2 and (2) provide, as directed by the Comission, an option for ordering B&W owners to incorporate a Westinghouse or Combustion Engineering level monitoring system.
Background:
In Reference (1), the staff provided a status report and recomendations concerning the implementation schedule for reactor vessel level instrumentation for PWRs.
In Reference (2), the staff was requested to develop an option for ordering B&W plants to incorpo-rate Westinghouse or Combustion Engineering level monitoring systems.
The staff has contacted B&W owners to learn of their current intentions with respect to level monitoring systems.
Our recomendations con-cerning the suggested option are provided herein.
Contact:
L. E. Phillips, NRR:DSI:CPB X-29472
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Discussion:
Status of Reactor Vessel Water Level Measurement System for B&W Reactors i
Based on a recent staff telephone survey, all of the B&W reactor licensees, with one exception, prefer or I
have selected some version of a hot leg differential pressure (dp) system based on a design development and evaluation being performed by B&W.
Davis-Besse 1, which is the exception, has under study conceptual designs for both the B&W dp system and the Westinghouse dp system.
The results of the study and a detailed,
engineering and installation schedule will be submitted by June 1, 1982.
In addition, the licensees for Oconee Units 1, 2, and 3 and Crystal River 3 are comitted to conclude the conceptual design of their systems and to submit a detailed engineering and installation schedule by mid-May 1982.
The licensee for TMI-1 has proposed a hot leg dp system monitoring the top 10 feet of the candy cane and plans to complete its design by February 15, 1982 and to install the system during the 1983 refueling outage.
The licensee for ANO-1 has comitted to install a B&W type dp system for the hot i
leg using existing taps, but has not provided details of the design or a schedule for implementation. The licensee for Rancho Seco has indicated an intent to install a hot leg level monitoring system using taps at the top and bottom of the hot leg; the bottom tap would be on the suction piping of the decay heat removal system.
The status of reactors discussed above is also sumarized in the enclosure Table 1.
Preliminary Staff Evaluation of preferred Concepts Several variations of the hot leg dp monitoring system have been discussed with the staff by licensees and applicants for B&W reactors. However, detailed engi-neering descriptions and evaluations of the concepts have not been provided for staff review.
Therefore, the discussion and preliminary evaluation of dp moni-toring concepts is predicated on the following as-sumptions:
(1)
Proposed dp concepts can be shown to function in an acceptable manner with pumps tripped by calcu-lations and testing.
(2) Concepts which do not include dp across the core (vessel bottom tap) will not provide a reliable indicator for trending a loss of coolant inventory with the pumps running.
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.. (3) Concepts which do not include dp from the vessel head to hot leg will not provide indication of voiding in the reactor vessel head until the bubble extends to the top of the hot leg nozzle.
(4) The detailed design of proposed systems will be accomplished in an acceptable manner with hardware which can be environmentally qualified.
A single dp measurement over the top 10 feet of the hot leg (as proposed for TMI-1) would detect voiding at the top of the' candy cane.
It will probably track hot leg level for a sufficient distance to distinguish between over cooling transients and a loss of coolaat inventory indicative of an approach to core uncovery.
It would also provide valuable information to support reactor coolant system venting operations and to confim that natural circulation operation is not interrupted by voiding in the candy cane.
However, it would not trend voiding with the pumps running and would not indicate void famation in the reactor vessel head until vessel water level reaches the hot leg nozzle.
It would also fail to provide a continuous indication of coolant inventory loss proceeding to core uncovery, and would not track the replenishing of coolant inventory.
The tracking range could be extended to about 5 feet above the core by adding a tap at the lower level of the hot leg.
Voiding in the vessel head could be detected by a dp transmitter between the vessel head and the hot leg.
Loss of coolant with the pumps running could be detected by a dp transmitter between the bottom of the vessal and the hot leg, but is perhaps not necessary if it can be shown that other instruments provide indication of voiding and that pump trip procedures will make inventory tracking with pumps running unnecessary. As presently proposed, the single dp measurement over the top 10 feet of the hot leg is unacceptable to the staff.
Other monitoring concepts discussed with the staff include dp transmitters between the top of the candy cane and the lower level of the hot leg or between the top of the reactor vessel and the hot leg.
By way of comparison, the Westinghouse concept is superior in that it includes dp between the bottom and top of the vessel and the hot leg.
Our preliminary conclusions are that an acceptable dp monitoring system for B&W reactors must include the following:
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. (1) a dp transmitter between the vessel head and the hot leg designed to indicate voiding in the vessel head and to track vessel level to within 5 feet of the top of the core (based on lower level of the hot leg nozzle in some reactors);
(2) a dp transmitter from the top of the candy cane to a level in the hot leg which is sufficiently low to distinguish between the most severe overcooling transient and a loss of coolant inventory; and (3) a dp transmitter sensing pressure change from a tap at the bottom of the vessel and designed to trend voiding with the pumps running, or acceptable justification for not providing for level trending with the pumps running.
Preliminary Evaluation of Combustion Engineering and Westinghouse Level Monitoring Systems for Application to B&W Reactors Based on discussions with vendors and infonnation provided by licensees, the staff continues to believe that application of the Combustion Engineering or Westinghouse systems to B&W reactors is feasible, though possibly expensive.
Information concerning the application of these systems follows:
Combustion Engineering Heated Junction Thennocouple System (1) Accessibility for installing the probe in the vessel head - Control Rod Drive Mechanism (CRDM) housing nozzles and thermocouple nozzles in the vessel head have been considered. Modifications to the access nozzles and vessel internals would be necessary in all cases.
Costs compared to alternate approaches may be unfavorable.
(2) Procurement - A lead time of about one year prior to a shutdown is needed to accomplish the modifications and installation.
(3) Acceptability - The HJTC system effectiveness on a B&W reactor should be compcrable to that on other reactor types and the concept is acceptable.
- Westinghouse dp System (1)
Feasibility - Locations for pressure taps at the bottom of the vessel, top of the vessel, and on the hot leg are feasible.
(2) Procurement - A lead time of up to 18 months is needed to supply the microprocessor equipment for the Westinghouse system.
Somewhat less time would be needed to provide and install the rest of the hardware.
(3) The effectiveness of the Westinghouse system on a B&W reactor,.with proper location of pressure taps, is expected to be acceptable.
Recomendation:
Recomend that the Comission:
(1) Direct staff that all B&W reactor licensees be ordered to conclude their conceptual designs and to submit detailed engineering, procurement, and installation schedules for an acceptable reactor collant system level monitoring concept not later than May 1,1982.-
(2) Note that the order will:
a)
Define " acceptable concepts" based on the preceding discussion, including the standard Combustion Engineering and Westinghouse concepts.
b) Direct that the procurement and installation schedule should be targeted for the first refueling after January 1, 1983. If the schedule submitted is not consistent with the target schedule, the submittal must include detailed justification for non-selection of the other acceptable concepts.
c) Direct licensees to be prepared to show evidence of a continuing "best effort" to assure that the procure-ment and installation proceed in accordance with the submitted schedule.
Scheduling:
Comission action should be deferred until briefings by vendors, but action is requested by January 21, 1982 in order that staff may process the order in time to be able to assure licensee implementationbynextrefuqingdate.
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/yt/C \\ A L%
', filiam J. Dircks t
W Executive Director for Operations
Enclosure:
Table 1
i
!..~ Commissioners' comments should be provided directly to the Office of the Secretary by c.o.b. January 15, 1982.
Commission staff office comments, if any, should be submitted to the Commissioners NLT January 8, 1982, with an information copy to the Office of the Secretary.
If the paper is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.
DISTRIBUTION:
Commissioners Commission Staff Offices EDO ELD ACRS ASLBP ASLAP
)
l TABLE 1
[
TMI ACTION PLAN II.F.2 STATUS
SUMMARY
(DECEMBER 1981)
FOR B&W OPERATING REACTORS Type of System Reactor Selected Schedule Milestones 1.
THI-1 B&W Hot Leg dp a) Order by 4/82 Coments: No hot leg tap is available at the (Top 10'belowcandycane) b)
Install 01st refueling nozzle level. Thermocouple nozzles after 2/82 (based on are available in the vessel head for assumed delivery by 2/83).
installation of the CE Heated Junction Thermocouple (HJTC) probe; but the estimated cost for rework is one million dollars.
2.
Rancho Seco Prefer the hot leg Comit to install B&W System, Coments :
Lower tap in the hot leg would be on dp system using however, installation schedule the suction piping of the decay heat taps at the top and is not provided.
removal system.
bottom of the hot leg.
3.
ANO-1 B&W Hot Leg Level Comit to install B&W System, Coments: The penetration at the top of the hot Monitor however, installation schedule leg already exists; the existence is not provided.
of the bottom tap is uncertain.
4.
Oconee 1, 2 Hot Leg dp Measure-Design and schedule will be and 3 ment provided by 5/14/82 (Top 10' below candy cane) 5.
Davis-Besse 1 No System Selected Detailed schedule available Coments:
Packages are out for conceptual 6/1/82.
Installation designs for the B&W system and probably in 1983 refueling Westinghouse system.
outage.
6.
Crystal River B&W Hot Leg dp A detailed engineering and installation schedule to be provided by 5/16/82.
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