ML20039D546
| ML20039D546 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1981 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.B.2, TASK-2.B.3, TASK-2.F.1, TASK-2.K.2.13, TASK-2.K.2.19, TASK-2.K.3.30, TASK-TM FYR-81-165, NUDOCS 8201050176 | |
| Download: ML20039D546 (6) | |
Text
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YANKEE AT001C ELECTRIC COMPANY 2.C.2.1 FYR 81-165
&f 1671 Worcester Road, Framingham, Massachusetts 01701 sYauxes e~
December 31, 19 CP Y
RECEIVED United States Nuclear Regulatory Commission
. q j
Washington, D. C.
20555 h
4' N> j-E WM EJtafflangs Attention:
Mr. Denids M. Crutchfield, Chief 9
7 Operating Reactors Branch #5 4
Division of Licensing p
Re f erence s :
(a)
License No. DPR-3 (Docket No. 50-29)
(b)
YAEC Letter to USNRC, dated December 15, 1980 (WYR 80-136)
(c)
USNRC Letter to YAEC, dated July 10, 1981 (81-07-020)
(d)
USNRC Letter to YAEC, dated July 10, 1981 (81-07-017)
(e)
YAEC Letter to USNRC, dated July 14, 1981 (FYR 81-108)
(f)
YAEC Letter to USNRC, dated November 30, 1981 (FYR 81-157)
(g)
YAEC Letter to USNRC, dated September 8,1981 (FYR 81-132)
(h)
YAEC Letter to USNRC, dated December 15, 1981 (FYR 81-160)
(i)
YAEC Letter to USNRC, dated January 1,1981 (FYR 81-6)
(j)
YAEC Letter to USNRC, dated November 24, 1981 (F7R 81-155)
(k)
YAEC Letter to USNRC, dated November 21, 1980 (WYR 80-125)
(1)
YAEC Letter to USNRC, dated December 16, 1981 (FYR 81-160)
Subject:
Status of TMI Action Plan Item Implementation
Dear Sir :
This letter discusses those TMI items described in Reference (b) scheduled for implementation on January 1, 1982.
Ihe status of the proposed modification is described; and where schedular problems are encountered, a basis for the delay is provided.
In addition, the status of modifications required as a result of References (c) and (d) are provided.
II.B.1 Reactor Coolant System Vents Information describing our proposed modification was submitted in Reference (e) as requested.
The system has been installed and is awaiting NRC approval before declaring the system operational.
Procedures will be modified as appropriate upon staf f approval of the installed modification.
II.B.2 Design Review of Plant Shielding and Equipment Qualification Information describing our proposed modification and schedule for installation of the Po s t-Incide nt Cooling System is provided in Reference ( f).
Information providing the documentation and implementation schedules pertaining to environmental qualification of safety-related electrical equipment is contained in Reference (g).
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United States Wuclear Regulatory Commission December 31, 1981 Attention:
Mr. Dennis M. Crutchfield, Chief Page 2 II.B.3 Po s t-Accide nt Sampling Capability l
As required, a post-accident sampling station will be installed.
This l
action is the result of the staf f position contained in Reference (c).
This position disagreed with our implementation schedule from Reference (b) which proposed that the installation of the sample system depend on the SEP integrated assessment.
Since the issuance of Reference (c),
dated July 10, 1981, every ef fort has been taken to design and procure I
equipment necessary to install the sample station by January 1,1982.
We expect thac the building modification necessary to accomodate this equipment will now be completed by mid-January,1982. Equipment installation and testing is not expected to be complete until February 28, 1982.
The sample station has been manufactured and once installed will have the capability to obtain reactor coolant samples (liquid and g.nes).
The capability to obtain containment atmosphere samples (i.e., no5le gases, iodines and cesiums, and nonvolatile isotopes) is provided via a gas sample in the switchgear room at the hydrogen monitor 1c ation.
The capability to perform chloride sampling and analysis is described in Reference (1). The capability to perform boron sampling and analysis is described below.
Boron Analysis Capability Our installed post-accident sampling station withdraws a primary coolant sample under system pressure from any one of the four primary loops. The liquid sample is cooled, depressurized, degassed with argon to remove hydrogen and fission gases and is then diluted. Our panel was designed to make an 800-fold and 20-fold sample dilution for radionuclide and boron assay respectively. However, any dilution f rom zero to 1000-fold may be done.
Laboratory studies have been perf ormed that clearly demonstrate that as little as 1 ml of a solution containing 50 ppm baron can be analyzed by the standard mannitol potentiometric method with an uncertainty of about 5%.
Guidance from NUREG-0737 suggests a factor of 2 for enalytical measurements on post-accident samples is all that is required.
Furthermore, our routine boron analysis procedure has been modified to physically separate the buret f rom the sample beaker and electrodes by about three feet.
The titrant is delivered to the sample vessel by syringe tubing. This modification allows lead brick shielding to be placed around the sample vessel, if necessary.
Calculations show that the diluted reactor coolant sample can be handled and that the resulting personnel exposure will not prohibit completion of any analysis for boron which may be requested by the plant emergency director.
Additionally, our procedure for post-accident sampling and analysis c1carly directs the chemist to have all laboratory equipment operable before sampling, to pre plan and rehearse the sequence of steps for analysis and to collect _only enough sample as may be required for the
~
United States Nuclear Regulatory Commission December 31, 1981 At tention :
Mr. Dennis M. Crutchficid, Chief Page 3 analysis.
These prudent steps are essential to assure that any radiation exposure incurred daring sampling is dinimized.
II.F.1.1 Noble Gas Ef fluent Monitor The high range noble gas ef fluent monitors, as described in Reference (1), will be operational by the end of January,1982.
The necessary conduit, cabling and supports for this equipment have been installed; however, manufacturer testing on the monitor has not been completed.
This delay has resulted f rom manufacturer dif ficulties in equipment calibration.
These problems have only been recently resolved and the equipment will be installed after completion of the testing.
I I. F.1. 2 Sampling and Analysis of Plant Effluents De si gn, procu remen t, and implementation of equipment necessary for monitoring of radioiodine and particulate effluents was initiated upon receipt and evaluation of Reference (d), dated July 10, 1981.
This item had originally been deferred to SEP as indicated in Reference (b).
Due to the redirection provided in Reference (d), this equipment will not be installed and operational until March 1,1982.
The installation of a system to provide stack iodine and particulate sampling is based on the exposure criteria stated in Reference (d).
Sampling in radiation fields exceeding this criteria will be performed when necessry by of f site radiation survey teams as personnel report to the Emergency Operations Facility.
II.F.1.3 Containment High Range Radiation Ifonitor Modifications necessary for the containment high range radiation monitor will be completed upon receipt and installation of environmentally qualified cable and connectors f rom the manufacturer.
Qualification testing is currently in progress and is expected to be completed during Ja nua ry, 1982.
The equipment is expected to be operational by Februa ry 28, 1982.
II.F.1.4 Containment Pressure Monitor This equipment will be installed and operational by January 1,1982.
II.F.1.5 Containment Water Level Monitor This equipment will be installed and operational by January 1,1982.
II.F.1.6 Containment H9 Monitor This equinment is expected to be installed and operational by March 31, 1982.
This delay is due to equipment delivery delays for system completion concurrent with power supply equipment delays (described below).
Furthermore, a week long test (i.e., system heatup and stablization) must be successfully performed before acceptable operation is achieved.
Uni,ted States Nuclear Regulatory Commission December 31, 1981 Attention:
Mr. Dennis M. Crutchfield, Chief Page 4 II.K.2.13 Thermal Mechanical Report This item requires a detailed analysis of the thetual-mechanical conditions in the reactor vessel during recovery f rom small breaks with an extended loss of all feedwater. Westinghouse (in support of the Westinghouse Owners Group) is performing an ana2ysis for generic Westinghouse plant groupings to address this issue which will be submitted to the NRC by the end of 1981.
This report is currently under review to determine its applicability to the Yankee Plant.
This review will be conducted during the next month and the results presented at that time.
II.K.2.19 Reactor Coolant System Voiding During Transients The possibility is remote that upper head voiding could occur during natural circulation cooling at the Yankee Plant. Recent analysis, specific for tha Yankee Plant and submitted to the NRC via Reference (j),
confirms that design features of the reactor vessel allow for ef ficient cooling of the upper head region.
Operators are trained to recognize symptoms of the St. Lucie event and are prepared to implement proven methods for safe rec ove ry.
The St. Lucie event was addressed by Westinghouse Owners Group Letter OG-57, dated April 20, 1981.
Procedu ral restrictions at the Yankee Plant are imposed to ensure safe natural circulation cooling without upper head voiding.
In addition, the Ref erence (j) submittal confirmed that adequate supplies of condensate grade feedwater are required by the Technical Specifications for the plant.
Additionally, Westinghouse (in support of the Westinghouse Owners Group) has performed a generic study which addresses the potential for void formation in Westinghouse designed nuclear steam supply systems during natural circulatin cooldown/ depressurization transients.
This generic study was also submitted to the NRC by the Westinghouse Owners Group Letter OG-37, dated April 20, 1981.
In addition, the Westinghouse Owners Group has developed a natural circulation cooldown guidelin 2 that takes the results of the study into account so as to preclude void formation in the upper head region during natural circalation cooldown/depressurization transients, and specifies
.those conditions under which upper head voiding may occur.
These generic guidelines have been procedurally implemented at the Yankee Plant and were submitted to the NRC in Westinghouse Owners Group Letter OG-64, dated November 30, 1981.
II.K.3.30 Revised Small Break LOCA Methods The scopa and the schedule of the program to be pursued in resolving the ECCS model concerns, as identified by this Action Plan, was transmitted in Reference (k). We had planned on pursuing two parallel approaches.
One approach -involved revision of the currently licensed RELAP4-EM Code.
The other involved the development and qualification of an Evaluation Model (EM) package based on the RELAPS computer code.
The final a
United States Nuclear Regulatory Commission December 31, 1981
. Attention:
Mr. Dennis M. Crutchfield, Chief Page 5 selection was to be based on various modeling requirements, accura te prediction of experiments, and user oriented features. We anticipated that the selection effort as well as the code modifications effort would be completed by January 1,1982.
Since the issuance of our letter, RELAPS was selected. Various versions of RELAPS were obtained from Idaho National Engineering Laboratory (INEL) as soon as they were made available to outside users.
Areas needing improvements were identified, in addition to the development work needed to incorporate EM requirements.
Integral sa well as separate ef fect tests were selected for the qualification of the code, and preliminary results have been obtained.
Our experience exercising RELAP5 against various integral and separate effect tests indicates that several of the constitutive equations in the code as released by INEL (e.g., interphase drag, critical heat flux, heat transf er logic) could be further improved.
In addition, code stability (e.g., boil of f under low flow conditions) is being given further attention. We have revised the work scope beyond that originally planned to address these areas; as a result, the schedule for response to Item II.K.3.30 has been revised to mid-April,1982.
Emergency Power Supply Upgrade Additional emergency power supplies to accomodate some of the equipment modifications required by the implementation of the TMI Action Plan (NUREG-0737), were required. A study was undertaken to evaluate the short-term and future ac and de emergency electric power needs for the Yankee Plant. Based on the results of this evaluation, we plan to make the following modifications to the on-site emergency power suppites:
1.
Station batteries #1 and #2 are approaching end of life and will be replaced with ides.tical or larger batteries.
2.
New Class lE inverters will be installed to provide a two train ac vital instrumentation power supply to feed safety class instrumentation.
This modification will be made in conjunction with the replacement of the existing station batteries #1 and #2 during the 1982 refueling outage.
3 A non-Class lE Uninterruptible Power Supply (UPS) will be installed to temporarily supply power to some of the instrumentation required by NUREG-0737 until the Class lE station batteries and inverters are installed.
These modifications will result in a significant overall improvement in the plant ac and de distribution systems and provide the necessary capacity to connect and power all of the required loads.
Installation of the temporary UPS was scheduled for January 1, 1982; however, due to delays in shipment of 2
the equipment, we expect that completion will be delayed.
The building necessary to accommodate the UPS has been completed and installation of associated equipment (i.e., cable, conduit, instrumentation) is in progress.
Baring no further delays, installation of the temporary UPS, connecting of
,, Uni,ted States Nuclear Regulatory Commission December 31, 1981 At t ention :
Mr. Dennis M. Crutchfield, Chief Page 6 power leads, and testing should be completed by March 31, 1982. Meanwhile, TMI equipment installed by 1/1/82, with the exc'ption of the containment hydrogen monitor, will be powered f rom emergency sources.
We trust this information is satisfactory; however, if you have any questions, please contact us.
Very truly yours, YANKEE AT0!!IC ELECTRIC COMPANY 57f e<__. A J. A. Kay Senior Engineer - Licensing JAK: dad
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