ML19309D754
| ML19309D754 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/09/1980 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 WYR-80-43, NUDOCS 8004110419 | |
| Download: ML19309D754 (13) | |
Text
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Telephone 617 366-9011 Twx 760-360-0739 YANKEE ATOMIC ELECTRIC COMPANY z.3.2.1 WR 80-43
%L%
20 Turnpike Road Westborough, Massachusetts 01581
,M 5iKEE April 9, 1980 United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Office of Nuclear Reactor Regulation Division of Operating Reactors Mr. Dennis L. Ziemann, Chief Operating Reactors Branch #2
Reference:
a) License No. DPR-3 (Docket No. 50-29) b) YAEC Letter to USm C dated December 31, 1979, WYR 79-163 c) YAEC Letter to USNRC dated December 21, 1979, WYR-79-160 d) YAEC Letter to USNRC dated November 27, 1978, WYR-78-103
Subject:
Resolution of TMI " Category A" Implementation Audit Outstanding Items.
Dear Sir:
During our Category "A" implementation audit on April 2, 1980, your staff identified several areas that required additional information. This letter is in response to those concerns.
Item 2.1.1 Document the heater capacity required and the time within which that capacity must be available to maintain natural circulation at hot standby conditions.
Operating experience at Yankee Rowe has demonstrated that the operation of one group of pressurizer heaters (37.5 Kw) is required to meet the heat losses from the pressurizer with normal spray flow through the pressurizer at hot standby conditions. The ability to maintain natural circulation under emergency conditions would require less capacity than for normal operations or less than 37.5 Kw.
Westinghouse has performed a study for the Westinghouse Owners Group in which pressurizer heat loss calculations were performed to determine minimum heater requirements without offsite power for three different 8004110 N
t Mr. Dennis L. Ziemann, Chief Operating Reactors Branch 2 April 9, 1980 Page 2 pressurizer volumes. An extrapolation of the results of these calculations to the volume of the Yankee Rowe pressurizer verifies the required heater capacity. The Westinghouse study included a transient analysis of the loss of offsite power event to establish the time frame when heaters would be required to maintain RCS subcooling.
The results show that the ability to supply emergency power to the heaters within four hours will prevent loss of subcooling in the primary system following a loss of offsite power.
Verify that the necessary heater capacity will be available within the required time period.
Yankee Rowe has four groups of pressurizer heaters connected to 480V Bus 6-3 and four groups connected to 480V Bus 5-2.
Each of these 480V buses is connected to an emergency bus via two circuit breakers in series, and these tie breakers are operated from the main control room. To supply the required heater capacity from the emergency bus, Bus 6-3 and Bus 5-2 are cleared and the buses are re-energized by closing the tie breakers to the emergency buses. The time required to accomplish this, giving due consideration to all requirements of plant operating procedures, is 15 minutes from the occurrence of the loss of offsite power.
2.1.3a Provide the date by which valve position indicator circuits will be qualified.
A program to qualify the position indicators is currently underway at Babcock & Wilcox. The documentation of the results, in report form, is expected in the third quarter of 1981.
2.1.3b Provide your plans, including schedule, for increasing the number of core exit thermocouple inputs to the subcooling margin monitor.
In accordance with your present requirements, we will increase the number of temperature inputs to the saturation monitor from 4 core exit thermocouples to 8 thermocouples.
This upgrading will be completed prior to restart.
When will additional details of your proposed reactor vessel level meter, based on the results of evaluations referenced in your Dacember 31, 1979 letter, be provided for tEC review?
The details of our proposed reactor vessel level meter, based on the results of our evaluation referenced in our 12/31/79 letter, will be provided by June 1, 1980.
2.1.4 Provide an addendum to your 12/31/79 submittal so that all essential and non-essential systems are identified. Include the bases for the classification of all essential systems not provided in your 12/31/79 submittal.
Mr. Dennis L. Ziemann, Chief Operating Reactors Brcnch 2 April 9, 1980 Page 3 In addition to those essential systems identified in our 12/31/79 submittal (Reference (b)), the following systems are also classified essential.
High Pressure Safety Injection / Low Pressure Safety Injection - SI-V-14, CS-V-621 These systems are required to mitigate the consequences of an accident.
Main Feedwater BF-CV-lOOO, 1100, 1200, 1300 This penetration is classified as essential because the emergency boiler feed system (EBFS) piping ties into the main feedwater system piping.
The EBFS is used to mitigate the consequences of an accident.
MC Feed to Loop #4 - CH-V-611 This system is part of the safety injection system and is required for hot leg injection.
Component Cooling to MCP - CC-V-667, 663, 671, 675 i
This system is classified essential to ensure long-term operability of the reactor coolant pumps.
Containment H7 Vent System - HV-V-5, 6 This penetration is required to vent hydrogen gas from the containment.
ECCS Recirculation - SI-MOV-516, 517 These valves are normally closed and are remotely opened from the MCB to provide long term ECCS recirculation in the event of an accident.
Table 3.6-1 of the Technical Specifications identifies the containment penetrations. Those penetrations that have not been identified as essential or non-essential in Reference (b) or above are classified as non-essential penetrations and are isolated with either check valves, locked-closed manual valves or other passive means. A summary of the essential and non-essential containment isolation valves is provided in the attached list.
Document your modifications to the containment isolation control system to include the following NUREG-0578 requirements: 1) Diverse containment isolation signals for non-essential system; 2) Resetting of containment isolation signals shall not result in the automatic loss of contaitynent isolation and 3) Reopening of containment isolation valves shall require deliberate valve-by-valve or penetration-by-penetration actions (i.e. group reopening is precluded).
l Diverse containment isolation signals are provided to isolate l
non-essential systems with automatic trip valves. The diversity in j
signal generation is provided in the SIAS logic which is actuated by low
Mr. Dennis L. Ziemann, Chief Operating R rctors Branch 2 April 9, 1980 Page 4 I
reactor system pressure (sensed by a pressure transmitter and a pressure switch in the containment) or high containment pressure (sensed by two pressure switches outside containment sensing containment pressure).
Resetting of containment isolation signals is not an automatic process.
For each individual valve, the operator must physically operate two separate switches in two independent operations in order to reset the containment isolation signal.
This process is in accordance with your requirements for valve-by-valve deliberate operator action to reopen containment isolation valves.
2.1.5 Provide a description of the use of the H2 purge system-for hydrogen control.
Show that the hydrogen purge system can withstand a single failure without jeopardizing containment integrity or the hydrogen control function.
Discuss proposed category B modifications to the hydrogen purge system to meet the NUREG-0578, Section 2.1.5 requirements. Provide your schedule for these modifications.
The air supply portion of the hydrogen purge system is shown on attached drawings FM-26A and M-10.
The existing system consists of a service air compressor with receiver backed up by three instrument air compressors (one recently installed and not shown on drawing) with two receivers.
Air flow is from the receivers via series valves CA-V-825 and HV-V-5 and 26 through either parallel solenoid valves HV-SOV-1 or 2.
These valves are either normally open, remotely operable, or accessible following an accident.
A backup air supply route is from the receiver via valves CA-V-687 and 688 and through locked open valve HV-V-26 through either HV-SOV-1 or 2.
CA-V-687 and 688 are not accessible during the initial phases of an accident but would be accessible approximately one week after the accident. Presently, purge system operation is not required until approximately three months after an accident.
The system is M S except for the portion which penetrates the containment and the portion which is common to the exhaust portion which is safety class 2.
Based on the above, YAEC concludes that the air supply portion of the hydrogen purge system meets the single failure criteria.
The exhaust portion of the purge system is also shown on drawings FM-264 and M-10.
The system is placed in operation by aligning the required valves in accordance with Procedure No. 2658.
This results in flow of hydrogen through either parallel valves HV-SOV-1 or 2, through locked open valve HV-V-26 and locked closed valve HV-V-6 through a hydrogen analyzer, flowmeter, EPA filter, charcoal filter, valve HV-V-13 which is locked open, a radiation monitor to the primary vent stack via valve AR-V-649 which is required to be open for power operation.
Mr. Dennis L. Ziemann, Chief I
Operating R:::ctors Br nch 2 April 9, 1980 Page 5 The system design meets the single failure criteria except for valve HV-V-6 which is locked closed and required to be open for the system to function.
All other active valves have either a parallel redundant valve, are locked open, or are normally open.
Containment isolation redundancy is provided by having HV-SOV-1 and 2 as inside containment isolation valves and CA-V-688, and HV-V-5 and 6 as locked closed, outside containment isolation valves.
Thus, the containment isolation feature meets the single failure criteria.
The following modifications are being evaluated to insure that the failure of HV-V-6 will not preclude the ability to purge hydrogen.
- 1) Addition of a parallel valve
- 2) Installation of a parallel purge pipe with required valves.
The required modification will be completed by January 1, 1981.
The exhaust and supply portions of the system use common equipment.
This may require that system operation be alternated between the exhaust and supply modes. The system was designed and sized for this mode of operation. Procedure OP-2658 will be reviewed and revised as necessary by January 1, 1981 to reflect any equipment changes or operational changes resulting from the proposed modification described above.
2.1.6b Provide more details of the design review for plant shielding.
Include the identification of vital areas and operator actions which will be necessary following an accident and the associated radiation fields.
Include a discussion of the extent of potential plant modifications and a justification for deferring them to the SEP program.
Include a discussion of the equipment qualification review.
A shielding review of the Yankee Rowe Plant has been completed. Primary emphasis was given to the direct radiation from the unshielded containment structure. A dose analysis was performed for on-site locations assuming a TID 14844 source term in the containment atmosphere. The results of this analysis demonstrated that most of the buildings on-site would not be habitable for several days following an event of this magnitude. Further calculations show that a concrete shield approximately 3 feet thick surrounding the containment, would be required to reduce the radiation levels to acceptable levels in the early stages of the accident.
i A review of plant emergency procedures define five areas in the plant that may require access within a short time following an accident.
Control Room /Onsite Technical Support Center The shielding of this complex was found to be adequate with the exception of the north wall. Skyshine radiation penetrating the north wall was I
Mr. Dennis L. Ziemann, Chief Operating Reactors Branch 2 April 9, 1980 Page 6
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found to exceed present day criteria.
An additional 12 inches of concrete was recommended for this wall and design of this additional shielding has been initiated.
Hydrogen Monitoring / Containment Air Sampling Station The hydrogen monitoring / containment air sampling station is located in the switchgear room and was found to be adequately shielded.
Auxiliary Feedwater Station The auxiliary feedwater station is located in the boiler room which is not shielded from the containment. The present auxiliary feedwater system is manually operated; however, this system will be supplemented by a new auxiliary feedwater system which can be operated from the main control room.
(See Reference (c), page 5, for details)
Liquid Sampling Station The liquid sampling station is located in the primary auxiliary building which contains no radiation shielding from the containment. The design of a new post accident liquid sampling station has been initiated. The location of the new station will depend on the final decisions regarding containment shielding and source term.
CIS Relay Station The containment isolation actuation relays are also located in the primary auxiliary building.
In some cases, it may be necessary for an operator to manually reset these relays to allow individual valve operation from the control room. A design change is in progress to allow these relays to be reset from the control room.
As a result of the shielding review discussed above, it appears that consideration be given to shielding the containment, or to justifying a more realistic source term based on a plant systems analysis. Yankee believes these considerations should be deferred into the SEP review.
The intent of the SEP review is to determine the need for plant modifications and to defer potential major modifications of this nature until an integrated assessment can be performed. This integrated assessment would give consideration to all SEP topic reviews and determine the impact of the modification on existing plants systems to obtain the optimum design. Therefore, the need for a shield enclosure should be reviewed under SEP because, as noted in Reference (b),
construction of a shield enclosure is a major project of at least three years duration involving many design interfaces which depend on the results of SEP topic evaluation.
An environmental qualification review of all safety related electrical equipment at Yankee Rowe was performed, under the SEP program, in 1978.
The results of this review were forwarded to the EC in Reference (d).
l The EC has requested that a reanalysis of environmental qualification be performed under the SEP program using updated guidelines. This
Mr. Dennis L. Ziemann, Chief Operating R; actors Branch 2 April 9, 1980 Page 7 reanalysis is presently being performed and is expected to be completed by June 1980. At that time the results will be submitted to the tRC and will include all safety related equipment as required by NUREG-0578.
2.,.l. 8a Commit to provide the design details of the improved sampling system by May 1, 1980.
The design details of the improved sampling system will be submitted by May 1, 1980.
2.1.8b Provide a commitment to incorporate the procedures for monitoring of steam relief and dump valves for noble gases prior to plant startup.
Prior to plant startup, the procedures for monitoring of steam relief and dump valves for noble gases will be incorporated.
2.2.la Commit to provide an annual reinforcement of 1) company policy relating to the shift supervisor responsitilities (Item 2.2.la position 1) and 2) an annual review of the administrative duties of the shift supervisor.
A reinforcement of company policy relating to the shift supervisor responsibilities (Item 2.2.la position 1) and a review of the administrative duties of the shift supervisor will be accomplished on an annual basis.
2.2.1.c Prior to restart, modify your " Operations' Department Surveillance Schedule" AP-2005, Rev.14, Item 7 to include an evaluation of all the aspects of the shift and relief turnover procedures. This evaluation should not be limited to verification of system alignment.
Plant Procedure AP-2005, " Operations Department Surveillance Schedule",
Item 7 will be modified to include an evaluation of all aspects of the shift and relief turnover procedures. The evaluation will not be limited to verification of system alignment. This procedural change will be implemented prior to restart.
2.2.2.b Prior to restart, revise procedures to dedicate one channel of Gaitronics for communication between the TSC and control room.
Prior to restart provide an override capability at the TSC and ECC in-plant bell telephone stations to assure dedicated communication path.
Plant procedures will be revised to dedicate one channel of the Gaitronics for communication between the TSC and the control room when the TSC is activated. A capability to assure a dedicated communications
Mr. Demis L. Ziemann, ChiGf Operating Re ctors Branch 2 April 9, 1980 Page 8 path at the TSC and the ECC will be provided. Both of the above items
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will be implemented prior to restart.
We trust this information is satisfactory; however, if you have further qgestions, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY
. 6 tay J. A. Kay Senior Engineer - Licensing JAK/ dis l
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ESSENTIAL NON-ESSENTIAL A.
AUTOMATIC ISOLATION VALVE X
TV-401A No. 1 SG Blowdown X
X TV-401B No. 2 SG Blowdown X
TV-401C No. 3 SG Blowdown X
TV-401D No. 4 SG Blowdown X
TV-408 Containment Cooling Water Return X
TV-409 Containment eater Condensate Return VD-50V-301 Air Particulate Monitor-in X
VD-S0V-302 Air Particulate Monitor-out X
HV-SOV-1 Hydrogen Vent System X
HV-S0V-2 Hydrogen Vent System X
TV-202 Main Coolant Drain X
TV-203 Main Coolant Vent X
TV-204 Valve Stem Leakoff X
TV-205 Component Cooling Return X
Main Coolant Sample X'.
TV-206 TV-207 Neutren Shield Tank Sample X
TV-209 Containment Drain X
X
.TV-211 Containment Pressure Sensing X
TV-212 Containment Pressure Sensing TV-213 LP Sample X
l
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i
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CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION E_SSENTIAL NON-ESSENTIAL A.
AUTOMATIC ISOLATION VALVE (Continued)
PCV-402 Steam Dump to Condenser TV-404 Main Steam Drain to Condenser X
TV-405 Auxiliary Steam to Air Ejector X
TV-406 Main Steam Drain to Condenser X
TV-410 Auxiliary Steam to #1 Feedwater Heater X
l TV-411 Atmospheric Steam Dump x
Turbine Generator Throttle Valves X
B.
CHECK VALVES SI-V-14 Safety Injection (HP)
X CS-V-621 Safety Injection (LP)
X CH-V-611 MC Feed to Loop #4 X
CC-V-667 Component Cooling to MCP #1 X
CC-V-663 Component Cooling to MCP #2 X
CC-V-671 Component Cooling to MCP #3 X
CC-V-675 Component Cooling to MCP #4 X
CC-V-649-Component Cooling to Sample Cooler X
CC-V-653 Component Cooling to Neutron Shield Tank Coolers X
CC-V-660 Neutron Shield tank Fill X
1 9
CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ESSENTIAL NON-ESSENTIAL R.
CllECK VALVES (Continued)
SW-V-820 Service Water to Containment X
Cooler #1
'SW-V-821 Service Water to Containment Cooler #2 X
SW-V-822 Service Water to Containment Cooler #3 X
SW-V-823 Service Water to Containment
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Cooler #4 X
llc-V-1199 Steam Supply to Containment Heaters X
C.
Manual Valves SC-MOV-551+553 Shutdown Cooling - In x
SC-MOV-552+554 Shutdown Cooling - Out X
CH-M0V-522 MC Feed to Loop Fill Header X
X CS-V-601 Shield Tank Cavity Fill CA-V-746 Containment Air Charge X.
IIV-V-5 Containment H2 Vent System X
X,
. IIV-V-6 Containment H2 Vent System CA-V-688
$ontainmentH2VentSystemAirSup' ply x
CS-MOV-500 ' '
Fuel Chute' Lock Val.ve X
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CONTAINMENT ISOLATION VALVES a
VALVE NUMBER FUNCTION ESSENTIAL NON-ESSENTIAL C.
ManualValves(Cont'd)
CS-CV-215 Fuel Chute Equalizing X
CS-CV-216 Fuel Chute Dewatering X
Pump Discharge VD-V-752 Neutron Shield Tank-Outer Test X
VD-V-754 Neutron Shield Tank-Inner Test X
BF-V-4-1 Air Purge Inlet DF-V-4-2 Air Purge Outlet x
HC-V-602 Air Purge Bypass SI-MOV-516 ECCS Recirculation X
SI-MOV-517 ECCS Recirculation X
BF-CV-1000 SGJ1 Feedt:3ter Regulator X
BF-CV-1100 SG52 Feed. tater Regulator X
BF-CV-1200 SGi3 Feedwater Regulator X
BF-CV-1300 SGF4 Feedc ter Regulator X
PR-V-610 Main Coolant Heise Pressure Ga'uge X
PU-V-543 Purification Systcm Containrtent Srap Suction x
PU-V-544 Purification System Containment Sump Suction y
VD-Y-1093 SGil Emergency Feed (SI)
X VD-V-1094 SGf2 Emergency Feed (SI x
VD-V-1095 SG'f3 Emergency feed (SI y
VD-V-1096 SGf4 Emergency Feed (Si x
_r
s CONTAINMENT ISOLATION VALVES l
l VALVE NUMBER FUNCTION ESS M IAL N ESS M IAL D.
Other 18" Bolted Manway X
Demineralized Water Supply (Blank flanged)
X Cavity Purification (Blank flanged)
X LP Vent Header (Blank flanged)
X Personnel Airlock X
Electrical Penetrations X
Equipment Hatch X
Containment Leg Expansion Joints X
Fuel Chute Expansion Joints I
e 9
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