ML19257A274
| ML19257A274 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1979 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 WYR-79-163, NUDOCS 8001030513 | |
| Download: ML19257A274 (24) | |
Text
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Telephone 617 366-9011 Twx FIO-3 90-0 739 YANKEE ATOMIC ELECTRIC COMPANY n.3.2.1 WYR 79-163
(.T3h' 20 Turnpske Road Westborough, Massachusetts 01581
.Yauxes December 31, 1979 United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation Mr. Harold Denton, Director
References:
(a) License No. DPR-3 (Docket No. 50-29)
(b) USNRC Letter to YAEC, dated October 30, 1979 (c) YAEC Letter to USNRC, dated November 19, 1979 (79-141)
(d) YAEC Letter to USNRC, dated November 7, 1979 (79-131)
(e) USNRC Letter to YAEC, dated October 17, 1979
Dear Sir:
Subject:
Lessons Learned Short-Term Requirements This letter forwards information requested by your letter, Reference (b),
and/or committed to by us in our letter, Reference (c). This information is attached as follows:
Item Description 2.1.2 Relief and Safety Valve Test Program and Schedule 2.1.3b Instrumentation for Detection of Inadequate Core Cooling 2.1.4 Identification of Essential and Non-Essential Systems 2.1.5a Dedicated H2 Control Penetrations 2.1.6a Systems Integrity for Containing Radioactive Materials Outside of Containment 2.1.6b Design Review of Plant Shielding 2.1.7a Automatic Initiation of Auxiliary Feedwater 1667 040 0305/3 p 800:
b' U.S. Nuclear Regulatory Commission Decembe r 31, 1979 Attn:
Mr. Harold Denton, Director Page 2 Item Description 2.1.8a Post Accident Sampling 2.1.9a Reactor Coolant System Ventine 2.2.2b On-Site Technical Support Center (TSC)
We trust you will find this information satisfactory; however if you have any questions please contact us.
Very truly yours, YANKEE AT0MIC ELECTRIC COMPANY
<5'.
J. A. Kay Senior Engineer - Licensing JAK/kaf 1667 041
Page I of 23 F
ATTACHMENT Section 2.1.2 Relief and Safety Valve Test Program and Schedule By letter dated December 17, 1979, Mr. William J. Cahill, Jr., Chairman of the EPRI Safety and Analysis Task Force submitted " Program Plan for the Performance Verification of PRR Safety / Relief Valves and Systems,"
December 13, 1979.
Yankee considers the program to be responsive to the requirements presented in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" dated July, 1979, Item 2.1.2, which recommended in part, "c ommit to provide performance verification by full scale prototypical testing or all relief and safety valves. Test conditions shall include two phase slug flow and subcooled liquid flow calculated to occur for design basis transients and accidents."
The EPRI Program Plan provides for a completion of the essential portions of the test program by July, 1981. Yankee will be participating in the EPRI program to provide program review and to supply plant specific data as required.
Yankee is also evaluating alternate programs to accomplish the intent of NUREG-0578, Section 2.1.2, as clarified by your letter of October 30, 1979.
1667 042
y Page 2 of 23 ATTACHMENT Section 2.1.3.a Direct Indication of Power-Operated Relief Valve and Safety Valve Position Direct indication of power-operated relief valve and safety valve position will be provided by the installation of an acoustic accelerometer system on the PORV pipe and on each of the two safety valve pipes. Ecch valve position indicator system will be powered from the vital bus and will be indicated and alarmed in the main control room.
The accelerometers will be installed during the January, 1980 plant shutdown.
The system equipment within the vapor container is being qualified for the appropriate environment by the supplier, Babcock & Wilcox.
Environmentally qualified cables will be used. Seismic qualification will be equal to or better than the system to which the equipment is attached.
1667 043
Dage 3 of 23 ATTACHMENT Section 2.1.3b Instrumentation for Detection of Inadequate Core Cooling The analyses and emergency procedure guidelines to meet the requirements of this section have been submitted to NRC by the Westinghouse Owners Group with letter OG-18, C. Reed to D. Ross, dated October 30, 1979.
Yankee has used these guidelines to prepare plant-specific procedures.
The information required on the subcooling meter is attached.
The reactor vessel at Yankee Rowe has no taps or connections at the base of the vessel which would allow the connection of differential pressure level measurement devices. Therefore, the measurement of reactor water level to the bottom of the core by conventional methods is very difficult, if not impossible.
As an alternative approach to determining reactor water level, an attempt is being made to use analytical techniques for this determination. The approach will utilize the change in count rate at the source range detectors as water level changes. The change in count rate will be calculated using a monte carlo code.
A shipping cask library containing 22 neutron groups and 18 gamma groups is used with the monte carlo code to track both gammas and neutrons at the source range detector. To date, a three-dimensional mockup of the reactor has been set up, the sources of neutrons has been determined and the library has been converted from IBM equipment to CDC equipment.
Initial check-out calculations of the code are currently in progress. To date, no actual Yankee Rowe cases have been run.
If this approach is successful, it is anticipated that the work can be completed by March 31, 1980.
If the approach does not work, we will have that information much sooner.
By the use of neutron count rate for determining var:ations in reactor level, existing equipment can be used.
Procedures will be developed from the results of the analytical process to provide the plant operators with the nethods for interpreting the change in neutron count rate as it is affected by reactor water level variations. Therefore, this indication will be implemented by January 1, 1981.
The Westinghouse Owners Group approach uses conventional level detection methods on the assumption that a connection can be made available at the bottom of the vesset.
Since this appro ch cannot be used at Yankee, we have requested that Westinghouse evaluate the problem of full range vessel level measurement at Yankee Rowe and propose an alternate approach which could be considered in event that the neutron count rate analytical technique is not workable. We anticipate a response at about the same time we determine the feasibility of the analytical method.
6 044
t Page 4 of 23 ATTACHMENT Section 2.1.3.b Information Required on the Subcooling Meter Display Information Displayed (T-Tsat, Tsat, Press, etc.)
Tsat & Psat Margins Display Type (Analog, Digital, CRT)
Digital Continuous or on Demand Continuous Single or Redundant Display Single Location of Display Main Control Board Alarms (include setpoints)
Low Margin Alarm 25 F T Margin Loss of Steady State 50 F T Margin Overall uncertainty (OF, PSI) 3.60F, 40.8 psig Range of Display 4 Digit Display Qualifications (seismic, environmental, IEEE 323)
IEEE 323 - 1974 IEEE 344 - 1975 Calculator Type (process computer, dedicated digital or analog calc.) Microprocessor If process computer is sed specify availability.
(% of time)
Single or redundant calculators Single Selection Logic (highest T., lowest press)
Thigh, Plow Qualifiations (seismic, environmental, IEEE 323)
IEEE 323 - 1974 IEEE 344 - 1975 Calculational Technique (Steam Tables, Functional Fit, ranges)
Steam Tables Input Temperature (RTD's or T/C's)
T/C's Temperature (number of sensors and locations) 4 core exit Range of temperature sensors 0 - 7000r 1667 045
Pcge 5 of 23 Uncertainty
- of temperature sensors (OF at 1)
See Note 1 Qualifications (seismic, environmental, IEEE 323)
See Note 2 Pressure (specify instrumant used)
Rosemount 1153 Rosemount 1152 Pressure (number of sensors and locations) 2 Main Coolant Sys.
Range of Pressure sensors 0 - 3000 psig Uncertainty
- of pressure sensors (PSI at 1)
.25%
Qualifications (seismic, environmental IEEE 323)
See Note 3 Backup Capability Yes Availability of Temp & Press Availability of Steam Tables, etc.
Yes Training of Operators Yes Procedures Yes Notes:
1.
0 - 530 F
+2F 530 - 700 F 13/8%
2.
Yankee is currently verifying the qualification of the core exit T/C's.
3.
The existing pressure sensor is a Rosemount 1152 and is qualified to IEEE 323-1971 and IEEE 344-1975. The new pressure sensor is a Rosemount 1153 and is qualified to IEEE 323-1974 and IEEE 344-1975.
i667 046
Page 6 of 23 ATTACHMENT Section 2.1.4 Identification of Essential and Non-Essential Systems Containment isolation valves which have been identified as non-essential are as follows:
V. C. Fan Cooling Water Returns TV-408 V. C. Heating Returns TV-409 No. 4 Steam Generator Blowdown TV-401D No. 3 Steam Generator Blowdown TV-401C No. 2 Steam Generator Blowdown TV-401B No. 1 Steam Generator Blowdown TV-401A V. C. Drain Header TV-209 M. C. Vent Header TV-203 Sample Cooler TV-206 M. C. Drain Header TV-202 Valve Stem Leakoff TV-204 Neutron Shield Tank Sample TV-207 Auxiliary Steam to #1 F. W. Heater TV-410 Steam Drains to Condenser TV-404 Main Steam Line Drains to Condenser TV-406 Purification Pumps Suction from LPST PV-MOV-541 Low Pressure Bleed Sample TV-213 Atmospheric Steam Dump TV-411 Containment isolation valves which have been identified as essential and the basis for selection of each are as follows:
Component Cooling Return Header - TV-205 It is essential that this valve remain open to ensure long-term operability of the four main coolant pumps.
Containment Pressure Sensing Lines - TV-211 and TV-212 These are instrument lines used to monitor pressure in the vapor container.
They provide the operators with information to determine if there is a pipe break ins ide containment. The system is dead ended just beyond the valves. A plant design change request is in progress which will physically remove these valves.
Turbine Steam Bypass - PCV--402 Closure of this valve with a turbine trip will result in the loss of the normal heat sink. This will result in lifting the secondary safeties resulting in a potentially unmonitored ground release verses a normal diluted monitored stack release. The secondary safeties have never been challenged in the almost 20 years of plant operation.
T.he existing loss of load accident analysis shows that the steam dump can handle this transient without lifting secondary safeties. To trip this valve on a SIAS places an unnecessary transient on the plant with a potential for an unmonitored ground release.
i667 047
Page 7 of 23
~ Main Steam Turbine Throttle Valves (2),
These valves close on a turbine trip or reactor scram.
If these valves are tripped on a SIAS, an inadvertent SIAS will cause a turbine trip and reactor scram which places an unnecessary transient on the plant.
If a valid SIAS is present, the reactor protection system will trip the plant which causes the valves to close.
Auxiliary Steam to Air Ejector Gland Steam Seal and Auxiliary Feedpumo - TV-405 Closure of this valve on a SIAS will isolate the steam supply to the emergency boiler feed pump.
Post H2 Control System Vent Valves - HV-SOV-1&2 These valves are needed in the extreme long term.
They are normally open, fail open energize to close valves.
They provide H2 sampling capability and post accident hydrogen vent capability. Manual valves downstream of these valves are normally closed per the technical specifications.
Air Particulate Monitor-In VD-SOV-301 Air Particulate Monitor-Out VD-SOV-302 These valves isolate the Containment Air Particulate Monitor. This instrum'ent provides the operators with early indications of a main coolant system leak and indication of the amount of fuel failure. For this reason, Yankee has classified these valves as essential.
t i667 048
Page 8 of 23 ATTACKHENT Control Penetrations Section 2.1.5.a Dedicated H2 The hydrogen vent system utilizes an existing 2" penetration used to supply service air to the containment. Two manual valves outside containment isolate the service air header when the hydrogen vent system is in operation.
Following a loss of coolant incident, the hydrogen venting system provides the capability to sample and periodically vent the vapor container. Two open-ended pipes, either of which can be isolated by closing a solenoid valve located inside the vapor container, allow the vapor container atmosphere at the midplane or near the top of the vapor container to be directed to the sample and control station. The solenoid valves shut automatically on a Containment Isolation Signal (CIS), and can be reopened from the switchgear room.
The penetration qualifies as a combined design in accordance with your position clarification of October 30, 1970 The penetrat".on is single-failure proof for containment isolation purposes. The single fsilure of an automatic valve will not prevent the system from functioning properly.
The penetration meets the requirements of General Design Criterion 54 and 56..
The system piping arrangement is shown on the attached sketch, M-10.
Based on the above, YAEC believes that the existing system meets the requirements of this topic and therefore, no modifications are required.
1667 049
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Page 10 of 23 ATTAC M NT Section 2.1.6.a Systems Integrity For Containing Radioactive Materials Outside Of Containment Yankee has implemented a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels.
In general, the program consists of periodic leak tests on specified systems to verify system integrity and identify component leakage. During the leak test the system will be visually inspected to identify leaking components. This identified leakage will be properly recorded and reported to the Program Administrator for corrective action.
The Leak Reduction Program consists of two parts:
A - Continuing leakage Identification Program normally performed quarterly on operating systets and B - Integrated Leakage Identification Program normally performed at refueling intervals.
Continuing Leakage Identification Program Specified systems in the program will be visually inspected while in service on a periodic basis, usually quarterly. The inspection will be conducted in accordance with the specific system surveillance or inspection procedures.
Any leakage identified will be recorded in the specific procedure and reported to the Program Administrator. The Program Administrator will determine the action required and ensure that action is taken.
If the action required involves maintenance, the system / component will be reinspected following the maintenance to establish the leakage rate for the system / component.
The leakage rate determined will be recorded on a Master List.
Integrated Leakage Identification Program Specified systems in the program will be subjected to an Integrated Leak Rate Test on at least a refueling interval to establish their Integrated Leakage Rates. During the performance of the leak test an integrated leak rate will be determined, where system configuration permits.
Where an integrated leak rate cannot be determined, an attempt will be made to quantify the leakage by component where leakage is found.
The results of the Integrated Leakage Rate Test (ILRT) will be evaluated by the Program Administrator. Corrective action will be taken where necessary to reduce the leakage to as-low-as-practical and repeat the ILRT where necessary. The results of the ILRT will be recorded on the Master List.
Systems which are exempted from the Integrated Leak Rate Test are delineated in Table "A".
The reason for this exemption is also stated in Table "A".
1667 051 i
Page 11 of 23 Systems Excluded Free The Leak Reduction Prngram Systems listed in Table "B" are excluded from the Leak Reduction Program.
The reason for this exclusion is also stated in Table "B".
In most cases the exclusion is based on the requirement that two pressure boundaries must fail, i.e. a double failure, before leakage to the environment would ensue.
Leakage Rate Measurements Leakage rate measurements for all systems in the program are delineated in Table "C".
In many cases no leakage was found.
Those systems which were not able to be placed in operation due to the piant's operating status are so designated. Table "C" represents an initial attempt at quantifying system leak rates. As more experience is acquired, the leak rate monitoring program may be modified to provide a more workable tool for evaluation of system performances. As leak rate measuring procedures are improved through experience, the leak rates delineated in table "C" may change.
It is Yankee's intention, through this Leak Rate Reduction Program, to keep leak rates as-lcw-as practical.
~
We have reviewed the North Anna Unit 1 incident, as it applies to our facility. To date no design or operator deficiencies nave been identified.
No modifications are deemed necessary as a result of Unis review.
I 1667 052
Page 12 of 23 Table "A" Systems exempted from Integrated Leak Test 1.
Shutdown Cooling Reason: Closed System -
Impossible to perform ILT 2.
Main Coolant Bleed Reason:
Undergoes con-stant LRT's 3
Charging & Volume Control Reason: Undergoes con-stant LRT's 4.
Waste Gas Reason: Cannot Isolate Table "B" Systems Exempted from Program 1.
Component Cooling Reason: Double Failure 2.
VC Heating / Cooling Reason: Double Failure 3
Feedwater/ Steam / Blowdown Reason: Previously estab-lished Technical Specifi-cation leak rate limit 4.
Cavity Fill Reason: Double Failure 5.
VC Air Charging Reason:
Double Failure 6.
Cavity Purification Reason: Double Failure 7
NST Sample Reason: Double Failure 8.
VC Ventilation & Purge Reason: Double Failure 9
Demineralized Water to Vapor Container Reason: Double Failure 10.
Low Pressure Vent Header Reason: Double Failure 1667 0153
Page 13 of 23 Table "C" Leakage Rate Measurements SYSTEM LEAK RATE (GAL / DAY) 1.
Safety Injection (ECCS) System 10 2.
Vapor container Recirculation System 0
3 Shutdown Cooling System e
4.
Main Coolant Bleed 0
5.
Charging and Volume Control System 400am 6.
Safety valve Discharge a
7.
Low Pressure Surge Tank (LPSI) Cooling System 0
8.
Purification System 0
9 Low Pressure Surge Tank and Appurtenances 0
10.
Jaste Gas System 0
- 11. Waste Liquid System 0
12.
Vapor container Drain Tank and Line 0
13 Post Accident H2 Vent System 0
- 14. Neutron Shield Tank Tell Tales 0
- 15. Main Coolant /Heise Pressure Line 0
- 16. Main Coolant Drain Line 17 Main Coolant Sample System 0
- 18. Vapor Container Air Particulate Detector 0
19 Fuel Handling System 0
20.
V.C. Pressure Monitoring System 0
21.
Valve Stem Leak Off System 0
aNo data available due to inability to place system in operation because of plant
- status, anThis system leakage varies primarily due to leakage from the positive displacement. charging pumps. This leakage is controlled cnd directed to the Gravity Drain Tank.
1667 054
Page 14 of 23 ATTACHMENT Section 2.1.6.b Design Review of Plant Shielding As a result of the preliminary shielding review at Yankee Rowe, it was obvious that primary emphasis for shielding design changes should not be direc'ted toward individual areas of the plant, rather toward the shielding of the source. However improbable the postulated event may be when considering the inherent stability and thermal inertia of the Yankee plant, to meet the shielding requirements with the specified source term, it will be necessary to construct a concrete shield around the elevated vapor container.
The vapor container vessel will continue to function as the primary containment structure. The concrete enclosure will provide shielding, and will not function as a secondary containment.
Based on the shielding provided by the new structure, local area shielding will be upgraded to meet access requirements.
Yankee has had a study performed to estimate the schedule and manpower requirements for design and construction of a containment shield enclosure.
Yankee's review of the scoping study has not been completed, but it indicates a project of at least three years duration with a minimum of 18 months for design engineering and licensing, assuming design input information is available when required. However, there are areas of design input which depend on the results of SEP evaluations, including seismic design and consideration of missiles. This dependence could impact the schedule. Yankee has already informally requested that the missile review be advanced in the SEP schedule.
Based on the impact of the SEP review schedule upon the design of a major project of this nature, Yankee requests that NRC include its design review of plant shielding in the SEP program.
Yankee has identified the systems carrying radioactive fluids under pos t-acc ident conditions and has reviewed the shielding design of the areas and equipment requiring access. The control room, portions of the primary auxiliary building, and the area adjacent to the safety injection pumps are the areas which have to be addressed.
The review indicated that the shielding provided by the rear wall of the control room, the side which is furthest away from the containment, should be increased to reduce the shine dose. Yankee is proceeding with the design of increased shielding for this wall and will have it installed prior to January 1, 1981.
To allow access to and limited stays at the sampling sink in the primary auxiliary building, it will be necessary to move or rebuild the sampling station. Modifications would include routing recirculation lines to closed drains, bringing valve stems and nipples through shielding, and if necessary, using lead glass curtains. The requirements for these changes are contingent on the resolution of the changes required by the review of post-accident sampling capability (Section 2.1.8.a).
1667 055
Page 15 of 23 Access to the area adjacent to the safety injection pumps is not required for short-term operation. However, if long-term operation is required, there is the possibility that access may be required for maintenance purposes.
In this event, it will probably be necessary to shield the three trains from one another and it may be necessary to reroute the cross-connecting piping.
Yankee believes any reronting of safety injection piping, if required, should be deferred until the SEP review is completed in order that all contingencies may be considered (seismic, missiles, etc.).
1667 056
Page 16 of 23 ATTACHMENT Section 2.1.7.a Automatic Initiation of Auxiliary Feedwater In our response to Item 2.1.7.a in Refcrence (c) and in Response 4 of the enclosure to Reference (d), we indicated that we would address the probability of inducing steam generator water hammer in the design for automatic initiation of the auxiliary feedwater system. Your letter of November 9,1979 does not require automatic initiation of the auxiliary feedwater system; therefore, evaluation of steam generator water hammer is not required for this mode of operation.
1667 057
Page 17 of 23 ATTACHMENT Section 2.1.8.a Post-Accident Sampling A design and operational review of the capability to obtain reactor coolant and containment atmosphere samples has been completed. Yankee presently has the capability to obtain a containment air sample and perform on-line hydrogen analysis with a single analyzer and to perform isotopic analysis of a grab sample.
With regard to reactor coolant sampling, Yankee has the capability to handle an unpressurized reactor coolant sample with a radioactivity concentration up to several mci /ml, which is three orders of magnitude higher than normal samples, but three orders of magnitude lower than the NRC source term on reactor coolant.
Sampling and analytical procedures have been revised according to the above concentration limit.
To meet NRC requirements for reactor coolant sampling and analyses, Yankee is evaluating several concepts which may separately, or in combination, allow the required sampling and analyses to be performed. One conceptual design includes the provision of a sampling dilution capability which will allow conventional analysis techniques to be used. Another involves an investigation into the current availability of instrumentation which can provide the required analysis by on-line monitoring. To meet the total requirement, a combination of schemes may be necessary.
In the evaluation of these concepts there are a number of items which require resolution before a decision can be made as to which direction should be pursued.
In addition, if on-line analysis proves capable of meeting the total requirement and is available, this could remove the need for the sampling station and impact the modifications required as a result of the shielding
. review. Yankee expects to resolve these items in time to implement the necessary modifications by January 1, 1981.
1667 058
Page 18 of 23 ATTACRMENT Section 2.1.9 Reactor Coolant System Venting Yankee will install equipment to vent both the reactor vessel head and the pressurizer by January 1, 1981. This equipment will meet the requirements of NUREG-0578 as clarified by your letter of October 30, 1979.
Sketch A, attached, provides a flow diagram of the proposed system. The reactor vessel head vent consists of a 1" motor operated valve, installed downstream of an existing manual vent valve. A flow restriction limits flow rates through the vent path to below the charging system makeup capability.
The vessel head vent is piped to the inlet of the post accident fans to insure adequate mixing.
The pressurizer vent consists of the existing solenoid operated relief valve, its block valve, and associated piping. Yankee is presently reviewing the qualifications of this equipment to determine if it meets, or can be upgraded to meet, the required design criteria.
If these criteria cannot be met, a motor operated valve with a suitable restriction orifice will be installed.
The vent paths will be safety class 1 for all portions where a piping failure would result in a LOCA.
In the reactor vessel vent portion the piping downsteam of the restriction orifice up to and including the motor operated valve is safety class 2.
The valves meet the single failure criteria when considered together, i.e.,
two vent valves, one on the reactor vessel head and a redundant vent on the pressurizer. These vents will meet the requirements of IEEE-279.
The following addresses specific NRC design considerations specified in the
' October 30, 1979 clarification letter.
A.1 The vent system has been designed to enhance the plant's ability to provide core cooling and maintain containment integrity.
A.2 Procedures addressing the use of the RCS vents will be provided by January 1, 1981 or before the system is placed into operation, whichever 18 sooner.
C.1-The hot legs are vented through the reactor vessel head vent.
A procedure to insure that sufficient decay heat removal is provided in the U-tube region will be provided by January 1, 1981. The vent system provides the capability to vent the pressurizer.
C.2 The nominal volume of the main coolant system is 3200 ft3, while the nominal capacity of the three positive displacement charging pumps is 100 gpm total. The failure of one charging pump reduces this capacity to approximately 66 gpm which is equivalent to approximately 533 3
ft /hr. Therefore, it will take 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to vent 1/2 of the main coolant system volume while still providing makeup capability with the charging system.
The reactor vessel venting portion of the system is designed such that inadvertent opening of the valve will not exceed the makeup capacity of 1667 059
Page 19 of 23 the charging system. This is accomplished by providing a flow restriction upstream of the motor operated reactor vessel vent valve.
3 The 533 ft /hr of charging pump capacity is equivalent to approximately 1200 SCFM of hydrogen displacement with the main coolant system at 2000 psi.
The vessel vent will discharge directly into the suction of the three post accident fans, each of which has a capacity of 6000 SCFM. The failure of one of the fans results in a 10% by volume hydrogen concentration in the fan duct work.
2 which The solenoid operated relief valve has a throat area of.601 in is equivalent to a nominal 7/8" orifice. Opening of this valve would exceed the makeup capacity of the charging system and would result in the venting of half of the main coolant system in less than one hour.
The displacement of 1600 ft3 of hydrogen, at 2000 PSIA into the vapor container (net volume = 860,000 ft3) would result in approximately 25%
by volume hydrogen concentration.
C.3 A flow restricion is provided at the reactor vessel vent location to limit RCS mass flow out the vent to less than the capacity of the charging system. The pressurizer solenoid operator relief valve is provided with a block valve which can be closed remotely.
C.4 The motor operated vent valve and motor operated block valve will have direct position indication in the control room. The solenoid operated relief valve has direct position indication provided by an acoustic accelerometer (see response to Section 2.1.3A).
C.5 Each vent will be remotely operable from the control room via operator action. However, power leads must be connected to the vent valve at the MCC which is located outside the control room and outside the vapor container.
C.6 Seismic design of each vent will be consistent with the present plant seismic design. When the SEP seismic review is completed, the vent will be seismically qualified to plant seismic values.
C.7 The system is designed to appropriate safety grade requirements.
In addition, the valves at each vent location are powered from different emergency buses.
C.8 The block valve for the solenoid operated relief valve will have the same qualifications as the solenoid operated relief valve.
C.9 During normal operation, power is removed from the motor operated vent valve.
In addition, the valve will be key locked closed in the control room under administrative control. The solenoid operated relief valve is normally closed while its block valve is normally open.
Both valves are under administrative control.
C.10 The reactor vessel vent portion of this system vents directly to the suction of the post accident fans. This provides the best available mixing with containment air (see C.2 above). The solenoid operated 1667 060
Page 20 of 23 relief valve vents directly to the vapor container via the rupture disc on the discharge piping. Air mixing is provided by the cooling ducts which are off the discharge of the post accident fans.
Cooling is provided by the metal vapor container which is the ultimate heat sink.
C.11 The inadvertent opening of the reactor vessel vent is an unlikely occurrence because of system design and administrative controls placed on system operation. The inadvertent opening of this vent does not constitute a LOCA. Direct position indication is sufficient to determine inadvertent operation.
The inadvertent opening of the pressurizer solenoid relief valve exceeds the makeup capability of the charging system. An acoustic accelerometer provides indication of inadvertent operation. The submittal of the analysis for this type event was addressed in Westinghouse Owners Group letter 0.G.-26, to the NRC dated December 20, 1979 (C. Reed to D. Ross).
1667 061
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Page 22 of 23 ATTACHMENT Section 2.2.2.b Onsite Technical Support Center (TSC) c Yankee has established an Onsite Technical Support Center (TSC) at the Yankee Rowe plant in accordance with NRC requirements.
The TSC has been established in the Operations Department Technical Assistant's office.
This' location is separate from, but in close proximity, to the Control Room.
The TSC has the same environment as the Control Room and, therefore, will remain
, habitable to the same degree as the Control Room for postulated accident conditions.
Yankee Rowe has established plans and procedures for engineering /
management support and staffing of the TSC. Means of communication between the TSC, the Control Room, the Emergency Coordination Center and the NRC have been established.
Since the TSC and the Control Room share the same environment, monitoring for both direct radiation and airborne radioactive contaminents, already established for the Control Room, will apply to the TSC as well. Yankee Rowe has designated action levels to define when protective measures should be taken.
Personnel manning the TSC will have access to technical data including, but not limited to, emergency procedures, as built drawings of structures, systems and components as well as direct display via closed circuit television of plant parameters necessary for the assessment function. A closed circuit television monitor has been installed in the TSC with a dedicated camera in the Control Room.
This camera is remotely operated from the TSC and will display and transmit plant parameters on the Main Control Board to the TSC.
Long Range Plans For Upgrading the TSC Yankee is currently evaluating long range plans for the Technical Support Center. These plans tentatively include establishing communications between the TSC, The Emergency Coordination Center and Yankee Westboro engineering offices via computer terminals.
This data link will give the Westboro engineering office vital information pertaining to plant parameters and enable the Westboro engineering office to give the plant close engineering and management support in the event of an emergency at the plant. This data link will enable implementation of emergency support procedures now established a,t corporate headquarters.
The following will be implemented as soon as practical, but no later than January 1, 1981.
The Technical Support Center will be upgraded to house 25 people, necessary engineering data and information displays.
Staffing levels and disciplines reporting to the TSC for different types of emergencies will be established by procedure.
This physical size upgrade may require the TSC to be relocated.
The center will remain on site within the plant security boundary.
Shielding and other environmental habitability requirements for postulated accident conditions will be established and implemented.
The center will be activated at the discretion of plant management or the Shift Supervisor for any off normal situation that requires activation of an emergency plan action level. Vital plant parameters will be displayed via the closed circuit television from the main Control Room.
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As described previously, the closed circuit television monitoring system is capable of accessing and displaying plant parameters independent of activities in the Control Room.
The power supply for the TSC television monitor is reliable and of a quality compatible with TSC instrumentation requirements.
The power supply is continuous once the TSC is activated.
Technical data required by TSC personnel for the accident assessment function will be established and included at the TSC.
The closed circuit television will display those items listed under 2.2.2.b.7 with the exception of offsite radiological considerations. Under this category, meteorological data is automatically transmitted to the Westboro Environmental Laboratory.
The TSC will have access to this information via the computer terminal data link between the Westboro engineering office and the TSC.
The offsite radiation levels as determined by survey teams will be available to the Technical Support Center through the offsite Emergency Coodination Center.
The Emergency Coordination Center will, at the request of the TSC, transmit this data via the computer terminal data link, or by telephone.
The Technical Support Center will be housed in a structure that has been built in accordance with sound engineering practice with consideration to natural phenomena that may occur at the site.
Protection from radiation hazards, includi6g direct radiation and airborne contaminants as per General Design Criteria 19 and SRP 6.4, will be provided to ensure TSC habitability for postulated accident conditions. The TSC will be modified such that personnel inside the TSC will not receive doses in excess of those specified in CDC 19 and SRP 6.4.
Permanent monitoring systems with local alarms to warn personnel of adverse conditions will be provided in the TSC.
These monitoring systems will monitor direct radiation dose rates as well as airborne radioactivity concentrations inside the TSC.
Procedures will be established to specify appropriate protective actions to be taken in the event of high dose rates or airborne radioactive concentrations exist inside the TSC.
A permanent ventilation system which will include particulate and charcoal filters will be evaluated for the Technical Support Center.
When implemented spare parts will be readily available and procedures in place for replacement of failed components during an accident.
An evaluation of emergency power supply requirements will be conducted in conjunction with other TMI and SEP related modifications.
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