ML20008E766
| ML20008E766 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/27/1981 |
| From: | Merritt W BOSTON EDISON CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.6, TASK-2.B.2, TASK-2.F.1, TASK-2.K.3, TASK-2.K.3.27, TASK-TM 81-44, NUDOCS 8103090586 | |
| Download: ML20008E766 (14) | |
Text
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B O STO N EDISDN COMPANY OENERAL Orrects 500 SovLevow Srstar?
SomToN. Massachusetts 02199 W. J. M ER RITT MANAGER NUCLEAR ENGIN EERING DEPARTM ENT February 27, 1981
@ BECo. Ltr. a81-44 Dk c'0 7
Mr. Darrell G. Eisenhut, Director H
v Division of Licensing 2,
p' Office of Nuclear Reactor Regulation
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U.S. Nuclear Regulatory Commission Ycp/'g,g Washirigton, D. C.
20555 License No. DPR-35 Docket No. 50-293 TMI NUREG 0737 Commitments
Dear Sir:
As requested by Mr. Mark Williams of your staff, attached please find additional information and commitments pertaining to item numbers I.C.6, II.B.2, II.F.1 and II.K.3.27 of NUREG 0737.
In addition, as a result of our review of all previous correspondence in this regard, other items have been re-assessed and are included in this attachment for your review.
Should you have any additional questions or concerns, please do n]t hesitate to contact us.
Very truly yours, b6 Lo Y6 Attachments y
//
8103090586
l.C.6 Guidance on Procedures for Verifying Correct Performance of Operating Activities Boston Edison will develop a station policy incorporating the requirements of TAP Item't.C.6.
This policy will be in place by March 31, 1981.
Procedures effected by this policy will be revised by June 1, 1981.
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11.B.2 Design Review of Plant Shielding Per telecon on February 25, 1981, between Messrs. Dave Verelli and Mark Williams of the NRC and Jim Keyes and Rod Cavalieri of Bost6n Edison Company, Mr. Verelli concurred that our January 5, 1981 responce, to TAP Item ll.B.2 was adequate.
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ll.E.4.2 (7) Containment Purge and Vent Isolation Valves Must Close on a High l
Radiation Signal Although this requirement was formally issued by the October 31, 1980 letter, t
design criteria had not been clearly established providing guidance for utilities
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to implement this modification.
Boston Edison participated in the Owner's Group
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efforts to establish design criteria which states the radiation instrumentation does not need to be safety grade. We are not in agreement with this position.
We are presently developing design criteria based on safety grade equipment and f
l are investigating material availability to implement this requirement. We will i
L provide a design description and implementation schedule by April 15, 1981.
In the interim, we believe that timely protection is provided by other i
signals (Low low reactor water level and High Drywell pressure) which minimize of'f-site doses during a loss of coolant' accident.
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j II.F.1 Noble Gas Effluent Monitors i
The high range effluent monitoring system consists of three on-line radiation I
monitors located at the Main Stack, the Reactor Building Vent, and in the Turbine Hall (turbine building operating floor). All three monitors have been sized and located to allow the measurement of the release rates and concentrations indicated on Table 1.
The calculated release rates are based on the estimated response of the monitors to the actual expected mixture of isotopes available for release in an accident involving fuel damage.
(xe-133 response is based en the xe-133 decay energy and intensity only).
Although the main stack high range monitor is capable of measuring the range of xe-133 concentrations specified in Position ll.F.1 of NUREG 0737 (ie:
10 to 5
10 u Ci/cc), it is more meaningful to consider the capability of the monitors to measure gross release rates of an isotopic mixture which is characterf.stic of.
an a<tual fission product mix.
In the case of an accident involving the release of a realistic' mixture of fission products, the range of the main stack and reactor building vent monitors is adequate to measure release rates which correspond to a drywell leak rate in excess of 100 times the technical. specification limit of 1%/ day.
The range of the turbine building monitor is adequate to measure release rates equivalent to a drywell leak rate in excess of 10 times the technical specifications.
Clearly, these monitors are adequate to quantify releases under any abnormal condition.
TABLE I flain Stack Rx Bldg. Vent Turbine Bldg.
Range of Monitor 4
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4 10 1 to 10 10 to 10 10 3 to 10 readout (R/hr) r Cross Activity 6
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10 Release Rates 10 - 10 10 - 10 10 - 10 at.. time 0 i
(u Ci/sec)
Flow Rates from Release Point 2000 - 10000 25000 - 200000 35000 - 210000 (SCR1) l Equivalent xe-133 6
12 5
11 5 - 1x10 11 Release Rates 9x10 - 5x10 9x10 - 8x10 2x10 (u Ci/sec) i Equivalent xe-133 1
6 8x10 2 - 8x10 1.5x10 2 - 1.5x10 3
3 Concentration 10 - 20 I
(u C1/cc) 4 1
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11.F.1 Attachment 2. Iodine and Particulate Sampline Continous sampling of iodine and particulate is provided for the main stack and reactor building vent release points.
In the unlikely event of an accident at PNPS which results in the release of radioactive effluents containing halogens, the normal iodine collection devices would Le retrieved and analyzed provided access to the sampling location is available and the sampling device itself does not result in a dose rate to operators which is prohibitive In the event the sampling device is unavailable, it is possible t'o make conservative estimates of iodine release rates by using the noble gas gross activity monitors described in section 11.F.1.
The response of these monitors to a mixture of noble gases alone and a mixture of noble gases and halogens in the relative proportions of the reactor core has been calculated. Based on this information and the assumption that the ratio of iodine to noble gases in the core (with the proper decay time correction) it is possible to estimate conser-vative halogen release rates.
Based on the above calculation it would be pos-0 4 and 10 uCi/see from'the sible to. estimate halogen release rates of between 10 turbine building. Of course the presence of halogens does not necessarily mean the release composition is similar to the core composition but the method des-cribed above would certainly result in a higher estimate of halogen release rates than are actually ocqurring.
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L 11.F.1.4 Containment Pressure Monitor l
11.F.1.5 Containment Water Level Monitor 11.F.1.6 Containment H2+02 Moni,or 11.B.3 Post-Accident Sample System i
l BECo letter #80-310 dated December 15, 1981 provided the following imple-7 j
mentation dates for the referenced TAP Items:
Containment Pressure 2/1/81 Containment Water Level 2/1/81 H02 2 Monitors 6/1/81 Post-Accident Sampling 6/1/81 As a result of numerous problems encountered while developing the detailed design for this modification, the implementation schedules have been revised
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Containment Pressure 1/1/82 e
Containment Water Level 1/1/82 i
H +02 Monitors 1/14/82 2
Post-Accident Sampling 1/14/82 l
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r 11.K.3.3 Reporting Safety and Relief Valve Challenges Below is our revised Relief Valve Challenge Report which contains some minor revisions to our previous submittal of January 5,1981.
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Challenge Date S/RV #
Reason for Challenge Remarks 5/17/80 RV203-3A Performed tech spec surveillance All valves tested OK 3B Testing 3C 3D L
7/25/80 RV203-3D Perform oper test per temp.
Valve failed to open Procedure TP 80-65 7/25/80 RV203-3D Test after inspection Valve would not open t
7/25/80 RV203-3D Test after inspection Valve would not open 7/25/80 RV203-3C Perform oper test per temp. procedure Tested OK j
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TP 80-65 7/25/80 RV203-3B Perform oper test pe r temp, procedure Tested OK j
TP 80-65 e
7 26/80 RV203-3A Test after maintenance related to Tested OK problems with RV203-3D l
7/26/80 RV203-3D Test after maintenance Would not open i
7/26/80 RV203-3D Test after maintenance Tested OK i
l 7/26/80 RV203-3D Retest (reliability)
Tested OK j
8/1/80 RV203-3A Operability Test Tested OK i
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8/1/80 RV203-3B Opetability Test Tested OK 8/1/80 RV203-3C Operability Test Tested OK f
8/1/80 RV203-3D Operability Test challenged 4 times Valve did not open - 4 tim <
8/3/80 RV203-3D Operability Test after maintenance Tested OK 8/5/d0 Rv203-3D Operability test 8 times Test OK 8 times l
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Challenge
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Date S/RV#
Reason for Challence Remarks 8/30/80 RV203-3D Accelerated test program Tested OK 10/1/80 RV203-3D Rx scram / operator open & closed valve Valve opened - would not l
via control switch close 10/1/80 RV203-3C Rx scram / operator open & closed valve Operated satisfactorily l
via control switch j
10/5/80 RV203-3D Test arter maintenance Tested OK 10/,/80 RV203-3A Valve opened due to hi instrument nitrogen pressure to the RV solenoid J
10/6/80 RV203-3A Tert after maintenance Tested OK 10/31/80 RV203-3A Valve opened due to hi instrument nitrogen pressure to the RV solenoid i
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ll.K.3.15 Modify Break Detection Logic to Prevent Spurious Ieciation of HPCI and RCIC Cooling Boston Edison has developed a design to satisfy this requirement. -In light of our f
I scheduled refuel outage commencing in September 1981, this modification will be imple-mented during tha scheduled outage. We believe this to be an acceptable position since i
L the modification would require HPCI and RCIC system cutage alternately which reduces the safety margin of the plant, even though it is allowed by technical specifications.
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f 11.K.3.22 Automatic Switchover of RCIC System Suction i
Boston Edison has evaluated TAP Item II.K.3.22 Automatic Switchover of RCIC suction, and we believe it to be unnecessary. During events that require s,ignificant amounts of high pressure coolant injection, the 4250 gpm HPCI system is relied upon to provide the j
required coolant makeup. RCIC is not relied upon during this circumstance due to its relatively low flow capacity. Failure to operate hPCI would result in operation of the
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l ADS and LPCI system. Therefore, RCIC is not needed during events requiring large amounts i
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of High-pressure coolant injection. During events requiring the RCIC system to operate the large volume of water available to the RCIC suction in conjunction with the low flow 4
of RCIC gives the operator considerable time to check condensate storage tank level and perform manual switchover of RCIC suction from *.he control room, as per existing station l
procedures. Therefore, BECo feels the addition of an auto switchover of RCIC is unneces-sary and existing procedures are fully adequate.
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r 11.K.3.27 Provide Common Reference Level for Vessel Level Instrumentation Marker plates will be installed on all necessary level instruments referencing l
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instrument to the top of active fuel by July 1,1981.
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l ll.K.3.28 Verify Qualification of Accumulators on Automatic Depressurization System Valves BECo letter #80-310 dated December 15, 1981 states that theOwnerE'Groupevalua-
. tion will be submitted to the staff by January 1, 1981. This is a typographical error the correct date in accordance with NRC Schedule is January 1, 1982.
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