ML19296B480
| ML19296B480 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/15/1980 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19296B465 | List: |
| References | |
| NUDOCS 8002200637 | |
| Download: ML19296B480 (85) | |
Text
generated when any of the valves open.
Calculations have been made f or saturated, liquid and two phase flow.
A summary of thesa calculations is provided in Appendix 2A.
Test s run by B&W on the electromatic relief valve under reduced flow condiitons have confirmed the validity of this approach.
Because of the straight-forward and well known relationships that exist between flow conditions and dif ferential pressure across the elbow, the signal from one dif ferential pressure transmitter can be confi-dently predicted for any flow conditions. For this reason it ha s been concluded that operating test s, which would be dif ficult since they involve opening the PORV and relief valves, will not be required.
Acoustic monitoring of the electromatic relief valve makes use of well proven equipment and techniques which have been used in the B&W Loose Parts Mor.itoring System.
Test s run on this valve at the B&W Alliance f acility demonstrated that the acoustic monitor-ing system gave satisfactory results.
2.1.1.2.5 Safety Evaluation In st rument taps will be installed on elbows 1, the discharge piping of pressurizer code safety valves RC-RVIA and RC-RVlB and electromatic relief valve RC-RV2.
This piing is classified as N2, Seismic 1.
Analysis has been performed to demonstrate that this modification will not degrade the integrity of the existing pipe. The pipe classification has been maintained up to and including the instrument root valv s.
The mounting of new equipment which will be located it the vicinity of safety related systems has been ana]yzed to ensure that no hazardous missiles will be generated in a seismic event.
It has been concluded that this modification will not degrade any safety related systems.
All of the equipment inside containment for Pressurizer PORV and safety valve detection will be seismically and environmentally qualified. Work is underway to upgrade the portion of the system outside containment. This involves specifying and procuring of additional equipment. This should be installed by January 1981.
2.1.1.2.6 In st rumenta tion The output signalt from the three differential pressure trans-mitters will be displayed on indicators in the control room.
They will be calibrated in " inches of water".
Each signal will also go to an alarm bistable.
A control room alarm will be '
initiated if any of the signals exceed a pre-determined value.
This will alert the operator that one of the valves is open. The dif ferential pressure signal will also be monitored by the plant computer for logging, trending, and alarm f unctions.
The outputs from the accelerometers which will be mounted on RC-RV2 will be processed by monitoring equipment installed in the existing Loose Parts Monitoring Cabinet.
An output signal indicative of flow through the valve will be displayed and recorded locally. A control room alarm will be initiated if flow is detected.
This signal will also be monitored by the plant computer for logging, trending, and alarm purposes.
8002200 Q Q 2.1-4 Am. 12
e.
No Override or Bvpass Capability - This override shall not permit the operator to re-open the valve unless the initiating condition is removed.
If the isolation valves have been re-opened and the initiating condition re-occurs then the valves shall again be isolated.
The containment isolation overrides shall be on an individual signal source basis such that overriding the isolation signal due to one source will still allow the valves to be isolated by a second isolation source if it is activated.
2.1.1.5.2 Design Bases 1.
The diverse containment isolation system shall meet the single f ailure criterion of IEEE No. 279.
2.
Redundancy of sensors, measuring channels, ivgic, and actua-tion devices shall be maintained and not be degraded by the modifica tions.
3.
Electrical independence and physical separation shall be in accordance with IEEE-383, where practicable.
If not po s sible,
existing physical separation criteria will be maintained.
4 Switches, independent of the automatic instrumentation, shall be provided for manual control of all containment isolation valves modified.
5.
Manual testing facilities shall be provided for on-line testing to prove operability and to demonstrate reliability.
Plant operation should not be adversely affected.
6.
All new instrumentation shall meet the environmental and seismic requirement s of IEEE-323.
7.
The status of all containment isolation valves shall be provided in the control room and not be affected by the modifications.
8.
Non-safety related radiation isolation signal will ineet all of the above criteria with the following exceptions:
The system will not be seismically qualified.
a.
b.
Testability requirement s of lEEE-279 will be met to the extent practicable.
2.1.1.5.3 Design Evaluations and Systems Operation In order to cover a broader spectrum of events for which contain-isolation is desirable, the reactor trip signal is used as a ment diverse containment isolation signal.
Since a reactor trip signal occurs on low pressure (1900 psig) it is anticipatory of SFAS and occurs prior to SFAS initiation.
Therefore the NRC 2.1-13 Am. 12
2.1.1.7.5 Design Evaluations Table 3-1 ( supplement I to this RESTART REPORT) indicates that the heaviest loading on one diesel generator during an ESAS actuation would be 2913 Kw and during a loss of of f site power only, the load would be 2817 Kw.
The total load in either case is below the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 3000 Kw.
Since no credit has been taken for the reduction in pumping requirement s following a LOCA and since the diesels 2000 hr rating is not exceeded, the diesel operability will not be affected. A detailed loading study has also verified this fact and testing will be perforced to further verif y this fact.
2.1.1.7.6 Saf ety Evaluation (See also responses to Questions 8 and 10 of Supplement 1, Part 1)
Safety analyses performed on the 177 Fuel Assembly B&W plants have determined that the emergency feedwater systems for a 2772 Mw plant must be capable of delivering 500 gpm (total to both g enera to rs). The basis for this criteria is contained in Volume 1; Section 6 - Supplement 3 of B&W's report entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant *. The analysis submitted by B&W is applicable to TMI-1.
Several studies have also beea performed by B6W for the 177 FA plants on loss of main feedwater transients.
These analysis have demonstrated that 500 gpm or lower auxiliary
(
feedwater flow is adequate following upset transients such as loss of power and the loss of normal feedwater flow.
Therefore, the small break LOCA conditions with a 20 minute delay in auxi-liary f eedwater initiation sets the minimum emergency f eedwater capacity requirements. Considering that TMI-l is only a 2535 Mw, a minimum emergency feedwater capacity requirement of 500 gpm is l
very conservative.
As discussed in paragraph 2.1.1.7.3 above, the TMI-1 emergency f eedwa ter system is comprised of two 460 gpm capacity electric pumps and one 920 gpm capacity steam driven (turbine) pump. The addition of the motor driven pumps (automatically) to the diesel block loading sequence and the turbine-driven pump start circuit ensures tha t a single f ailure will not result in less than the minimum required pump capacity being available under all condi-tions including loss-of-of f site power. That is at least two motor driven or one motor driven and the turbine pump will be available under all single f ailure conditions.
2.1-24 Am. 12
2.1.1.8 Leak Reduction Program for Systems Outside Containment A leakage reduction program is being developed consistent wi th the requirement s of NUREG-0578. Babcock & Wilcox (B&W) has been contracted to provide assistance in developing and accomplishing the overall program.
Metropolitan Edison Company, assited by B&W will implement the leakage reduction program in three distinct phases.
In phase I, the scope, plan and development will be a c compli shed.
This will include:
1.
Determination of which systems need to be included in the program and which systems may be excluded.
2.
Determination of where the leakage should be measured on each system.
3.
Determination of the best method to measure leakage.
4 Determination of the system and plant conditions during the leakage measurement.
5.
Development of a testing cocedure for each sy st em,
6.
Development of a method to collect and present the data such that meaningful recommendations can be made.
7 Develop a schedule and frequency for data collection to improve the consistency of sample results.
In phase 2, the actual leakage measurement test s will be performed for those systems identified. And in phase 3, the data collected during the tests will be evaluated and the necessary corrective actions performed.
The results of these phase 2 tests will be reported to the NRC within 60 days of completion of phase 3.
Phase 1 is expected to commence the week of December 3,1979 and is estimated to last approximately 3 weeks. Phase 2
- will commence immediately af ter completion of phase 1 dependent upon system availability.
Total duration is expected to be approximately three weeks. Phase 3 will commence immediately af ter completion of phase 2 with an expected duration of four to six weeks.
Af ter the program's initial implementation, Metropolitan Edison Company will initiate a Preventive Maintenance Program which will perform periodic leak tests of the systems defined in the initial program.
2.1-29a Am. 12
Those systems whic i could contain highly radioactive fluids during a serious transient or accident have be.en listed in Table 2.1-4 as being within the scope of the N' REG-0578 J
Leakage Reduction Program.
Table 2.1-4 also includes a summary description of the test method for each system.
In some cases, wholly new procedures are required, while in other cases minor revisions to existing surveillance procedures will suffice. With the exception of the Reactor Building Integrated Leak Rate Test, all Leakage Reduction Program Surveillance tests will be perf ormed on a refueling interval frequency.
Table 2.1-5 lists those systems presently excluded from this program and the reasons for their exclusion.
Phases #2 and 3 of this program will be completed prior to IMI-l Restart. The results of the Phase #2 tests will be reported to the NRC within 60 days of completion of Phase 3.
In addition to the above, review and inspection of release paths, as identified in IE Circular 79-21 and exemplified by the North Anna Unit 1 incident, were conducted.
No modifica-tions to existing systems and/or equipment were deemed to be necessary as a result of this review.
There were, however, some minor maintenance items identified, such as the need for installation of additional pipe caps or blanks on the down-stream side of some system vent, drain or test isolation valves.
These corrective measures will be completed prior to restart of TMI Unit 1.
2.1.1.9 Automatic Closure of the Pressuri:er PORY Block Valve This Section has been deleted.
2.1-29b Am. 12
2.1.2 Long Term Modification 2.1.2.1 Post Accident Monitoring 2.1.2.1.1
System Description
Certain post accident monitoring capability will be provided in compliance with Reg. Guide 1.97, Rev.2 as discussed below.
Pending tne availability of appropriately qualified instrumentation and equipment, the following modifications will therefore be completed by January 1, 1981.
The conceptual design will be provided for NRC review by January 1, 1980.
Containment Pressure - Continuous containment pressure indication will be provided in the control room esing a range from -5 psig to three times the design pressure of the containment. The pressure indication will be safety grade and will meet the design and qualification requirements of Reg. Guide 1.97.
Redundant indication of pressure will be provided.
Containment Water Level - Continuous containment water level indication shall be provided in the control room. A safety grade wide range indicator from the bottom of containment to a level of 10 feet will be installed in accordance with the requirements of Reg. Guide 1.97.
In addition, a narrow rsnge indicator from the bottom to the top of the sump with continuous indication in the control room shall be installed which meets the requirements of Reg. Guide 1.89 and is capable of being periodically tested.
Containment Hydrogen Indication -Safety grade continuous indica-tion of containment hydrogen will be provided in the control room. The range of indication will be 0-10% concentration assuming commercial availability over this range.
High Range Containment Radiation Monitor - Two safety grade containment radiation monitors that are physically separated shall be provided with recording display and continuous indicator presentation in the control room. The range of this monitor shall be 107 R/hr and shall detect photon radiation down to 60 Kev.
The design of the radiation monitors shall be provided in accordance with Reg. Guide 1.97 Rev. 2 (Dec. 1979). To our knowledge, manufacture of appropriately qualified equipment to satisfy these requirements will commence by July 1980.
They will be installed and operational by January 1,1981.
High Range Ef fluent Monitor - One high range effluent monitors intended as the Long-Term modification shall be oeprational for each normal gas release point by January 1, 1981.
The range of these monitors shall be as follows:
o Undiluted Containment Exhaust 10 uci/cc 5
2 o Main Steem Lines 10 uci/cc o Auxiliary & Fuel Handling Building Exhaust 3
10 uci/cc o Condenser OFF GAS Exhaust 2
10 uC1/cc 2.1-30 Am. 12
Regulatory Guide 1.97 Rev. 2 (Dec. 1979) shall be utilized in the design guidance of high range effluent monitor. Vital bus power shall be employed for each system's modular assembly with the normal power supplying the monitor pumps with diesel gener-i ations as back ups.
Further descriptions of increased range capabilities are provided in Section 2.1.2.
j i
High Range Ef fluent Radio Iodine & Particulate Sampling Analysis -
The existing sampling system will be expanded and will
{
include the addition of silver zeolite cartridges. The system assign and operation will both decrease the activity on the cartridges so they can be handled and will decrease the xenon l
to iodine ratio.
Counting of the cartridges will be by use of Nal crystal connected to a single or dual channel analyzer with l
appropriate window and discrimination settings for th 364 Kev gamma of I-131, or by use of a GELI/MCA system.
The expanded portion of the sampling system would be placed in service follow-ing an accident and will be located in an applicable area exhibit-l ing low background.
The system will be on site and operable by l
1 January, 1981.
I l
Prior to incorporation of the expanded sampling system, procedures k
will be developed for the use of silver zeolite cartridges and l
normal particulated filters for sampling with a NaI detector and a single or dual channel analyzer for iodine and gross particulate release rate determination.
Specific details to insure exposures are maintained as low as reasonably achievable will be incorporated into the procedures.
These procedures will be available for NRC review prior to restart or 1 October 1980 whichever occurs first.
2.1.2.2 RCS Venting 2.1.2.2.1
System Description
Vents will be provided for the reactor coolant system in order to ensure that natural circulation and adequate core cooling can be maintained following an accident. The vents will be located at the top of the pressurizer and at the top of both candy canes using existing penetrations. The discharge from the vent will be directed to the reactor coolant drain tank.
The reactor coolant system venting modification will be a safety grade design and will be single failure proof for both isolation and venting of the reactor coolant system. Control and position indication for the power operated vent valves will be provided in the control Pending the availability of the required safety grade room.
equipment to accomplish this modification, implementation can be completed by January 1, 1981.
2.1.2.2.2 Design Evaluation Babcock & Wilcox is currently completing the analysis to justify the adequacy of the proposed vent locations and size.
It is 2.1-31 Am. 12
anticipated that the unique design of the Babcock & Wilcox nuclear steam supply system which allows venting the high point s in the loops themselves will preclude the necessity for venting the reactor vessel head.
The analysis will be provided for NRC review by January 1, 1980.
Once the analysis of the adequacy of the conceptual design for RCS venting is complete, procedural guidelines will be prepared in suf ficient time to train operating personnel on the proper use of the new venting system.
The conceptual design plan will be confirmed by January 1,1980, following the completion of the final generic B&W recommendations on the reactor coolant system venting modification.
2.1.2.3 Plant Shielding Review 2.1.2.3.1
System Description
A design review of the plant shielding for radiation f rom systems out side containment will be completed by January 1, 1980.
This review will be aimed at assuring access to vital equipment and as suring that vital equipment is qualified to function in the general area radiation levels.
This review will consider those systems which may contain liquid or gaseous input from the primary system during an accident situation including the follow-ing: low pressure injection recirculation, containment spray recirculation, high pressure injection recirculation, process sampling, makeup and letdown, waste gas.
Sources for each of these systems will be developed utilizing decay f actors consis-tent with the time the systemw ill be operated and dilution factors consistent with accident scenarios during which the system will be operating. Field run piping will be considered.
2.1.2.3.2 Design Basis The source term to be used for shielding calculations shall be as follows:
Liquid Systems:
Noble Gas - 100% of core inventory Halogens - 50% of core inventory Others - 1% of core inventory Containment Air:
Noble Gas - 100% of core inventory Halogens - 25% of core inventory The criteria for limiting general area radiation levels in order to assure personnel access to vital equipment will be as follows:
Areas requiring continuous occupancy - <15 mr/hr 2.1-32 Am. 12
Control Room Operation Support Center (TM1-1 Health Physics)
Technical Support Center (Mod / comp room and cooldown from outside control room panel)
Areas requiring possible f requent access - <100 mrem /hr (Once or more per each 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shif t)
Radiochemistry Laboratory H2 Recombiner Control Panel Liquid Waste Disposal Panel For all other areas, shielding will be provided as required to keep personnel exposures less than 10CFR20 and to maintain the integrated dose to vital equipment below that for which the equipment has been qualified.
The integrated dose to vital component s and equipment will be determined using the calculated radiation levels and the required length of service of each component and piece of equipment post accident.
2.1.2.4 Post Accident Samoling Capability 2.1.2.4.1
System Description
Post accident analysis of reactor coolant samples and the con-tainment atmosphere is recognized as a means to better define core damage and anticipate the need for remedial actions. The TMI-1 capabilities for post accident sampling will be modified as necessary to provide key sample results on an on-line basis and to provide backup confirmatory sampling capability within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of directing that a sample be taken.
The key parameters to be monitored with on-line instrumentation include containment hydrogen concer. ration, reacotr coolant boron concentration and letdown failed 'uel monitors. The on-line hydrogen monitoring capability Fas been previously described in this restart report. An on-lir.e boronmeter will be installed.
The conceptual design and schedule for installation will be forwarded to the NRC by March 1,1980. The existing reactor coolant system letdown monite,rs will remain on scale with up to 107. f ailed fuel based on the FSAR definition of failed fuel.
This existing monitor is deemed adequate as an indicator that significant core damage has occurred.
A design and operational review of the reactor coolant and containment atmosphere sampling systems shall be performed.
Modifications shall be completed as necessary to ensure that personnel can obtain samples under accident conditions without incurring a radiation exposure to any individual in excess of 3 rems to the whole body and 18 3/4 rems to the extremities. The source terms to be considered shall be those previously listed under the Design Basis of Plant Shielding.
In addition, a design and operational review of the radiological spectrum analysis facilities and the chemical analysis facilities will be conducted 2.1-33 Am. 12
in order to identify any additional design features or shielding require >
ensure that confirmatory samples can be obtained and analyzec
. thin the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period previously mentioned.
The chemical analyses to be considered shall include both boron and chloride analysis.
The results of the design and operational review and conceptual design for required modifications will be forwarded to the NRC by March 1,1980.
RCS sampling can be accomplished within one hour and analzsed in an additional hour under normal circumstances. Met-Eds experience sam-ling under accident conditions at TMI-2 has taught us that slow deliberate steps are necessary to prevent personnel overexprosures.
Since an earily indication of significant fuel failure is obtcined f rom the f ailed fuel monitor and an on-line baronometer is being provided it is considered unnecessary and imprudent to attempt to draw a confirmatory sample in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. All necessary and appropriate operator and emergency actions can be taken based on the earily indica tions.
In addition chloride analysis provides inform-ation which is only useful in the long term and therefore is not immediately necessary.
2.1.2.5 Reactor Coolant Pump Trip on HPI 2.1.2.5.1
System Description
The purpose of this proposed modification is to provide automatic trip of the Reactor Coolant Pumps when degraded primary system conditions associated with a LOCA have been detected.
This will be accomplished by requiring that RCP trip be initiated when the Engineered Safeguards System has actuated Safety Injection and an increaseing RC void fraction has been detected as indicated by low RC pump motor current. The proposed logic will preclude RC pump trip during those events such as severe overcooling or very small breaks where maintenance of forced cooling is very d esi ra ble.
The conceptual design described in this section is being submitted for NRC review and comment and will be implemented subject to concurrence of the NRC Staf f.
2.1.2.5.2
.De sign Bases Analysis has shown that a certain range of small primary breaks may result in unacceptable clad temperatures if t'te R.C. Pump are tripped at a time when the R.C. System void fraction has achieved a high level. To prevnt these detrimental consequences, the proposed control scheme will promptly trip the R.C. Pumps when R.C. system conditions indicate that a small break in this range may be in progress.
(Until this modification is in place procedures will specify operator action to manually trip the RCP's upon actuation of Safety Injection). The system shall actuate when High Pressure Injection has been initiated and the R.C. System void fraction has reached a nominal value which indicates that a high void fraction may develop.
It is also very desirable, although not necessary, to avoid initiation during transient s such as overcooling and very small breaks where
", pump trip is not required so that forced R.C. System ci rc ula t can be maintained. The proposed system will meet both of
' ese criteria.
2.1-34 Am. 12
2.1.2.5.3 Systec Design The R.C. Pumps will be tripped on a coincident detection of High Pressure Injection by the ESFAS and low R.C. Pumps Motor Current in at least two of the four R.C. Pumps. Thi s mean s t ha t for the special case of two numps operation, the pumps will be tripped on HPI alone. Redundant sensors will be used for pu=p current on each R.C. Pump Motor. Redundant trip signals will be derived for each motor.
Electrical separation will be provided for redundant sig nal s.
No surge f ailure in the proposed system shall prevent a trip when required.
No single failure in the system shall result in the trip of more than one R.C. Pump.
The actuation system will be designed to be operable af ter a seismic event. However, the R.C. Pump Motor Switchgear is not sei smi-cally qualified. Provisions for on-line surveillance testing will be included.
The operator will be able to restart a pump after trip by manual means.
2.1.2.5.4 Design Evaluation The proposed uesign concept assumes that pump motor current can be used in combinaion with HPI actuation to detect the need for an R.C. pump t rip.
Supporting evidence for this concept has been generated by two EPRI sponsored test programs.
In 1973, Babcock 6 Wilcox in conjunction with the Bingham-Willamette Company, conduc ted a te st program to investigate the single and two phase performance of a one-third scale reactor coolant pump using air-wa ter mixtures. The result of these tests were reported under a contract with EPRI in 1977 (Ref.1).
Testing performed by CREARE under an EPRI contract using a 1/20 scale pump also shows a substantial decre'ase in torque at void fractions above 20%. Preliminary results of this testing were reported in the 6th Water Reactor Safety Research Information Meeting in November, 1978 (Ref. 2).
Since torque is directly related to pump motor current, the use of pump motor current as an indicator of the fluid void fraction, based on the ref erenced experimental data, is appropriate.
The proposed design meets the dual requirements of reliability initiating a trip when required, and not degrading plant availa-bility through inadvertent t ri ps.
2.1.2.5.5 Safety Evaluation The system will be f unctional during a seismic event (except for the R.C. Pump Motor Switchgear), will be testable and will meet single failure criteria for actuation.
No single failure will result in trip of more than one pump. Redundant circuits will be separated.
Where the new system interfaces within existing safety systems, care will be taken in the design to assure tha t there will be no degradation of existing safety functions.
2.1-35 Am. 12
2.1.2.
5.6 REFERENCES
1.
1/3 Scale Air-Water Pump Progra=, Pump Performance Data, EPRI NP-160, Vcl. 2, Oct. 1977.
2.
EPRI/CREARE 1/20-SCAI.E TWO PRASE PUMP PERFOR%\\NCE PISULTS, P.
W.
Runstadker, Jr. and W.
L. Swift, CREARE Incorporated, Present at the 6th Water Reactor Safety Research Information Meeting, Nat. Bureau of Standards, Garbersburg, MD, Nov. 6-9, 1978.
2.1-36 Am. 12
2.1.2.6 Auxiliary Feedwater System Auto start of the emergency-feedwater (EFW) System is being implemented in two phases:
1.
Control Grade Auto Start - This is a non-safety related initiation as described in paragraph 2.1.1.7 and it is a short-term approach, 2.
Safety Grade Auto Start - This will be a long-term modification where the initia-tion will meet the requirements for Class IE system and the system is functionally described below.
1.
The safety grade EFW auto start when implemented will automatically initiate the syste= on presence of the following conditions with or without the availability of the off-site power:
Loss of both normal feedwater pumps, or Loss of all four reactor coolant pumps, or Low differential pressure between the normal feedwater and main steam lines at each steam generator, The system initiation on low steam generator level will even-tually be added.
This will be done after the necessary analysis and engineering has been completed to insure that this signal will give a satisfactory actuation and will not interact with other plant functions. Loss of normal feedwater pumps is detected by differential pressure switches across each pump (two switches per pump, i.e.,
one switch per train).
The model of differential pressure switches used for this application has been seismically tested. These switches have temperature limits of -60 tr. 200*F.
Since they will be loca-ted in Turbine Building which is a non-seismic building, the switches will be tied into their respective EFW initiating circuits (Train A&B) through buffer devices and thus the switches will be treated as safety grade items to the extent possible.
2.
All cables associated with the initiating logic will be quali-fied for Class IE application and the initiations will be designed to meet single failure criteria. All circuits will meet the regulatory crit; cia for separation of Class lE cir-cuits.
3.
The initiating logic will include hardware for the following purposes:
Latching mechanism to seal-in the actuation Manual Reset Capability Testability of the initiating circuit 4.
Indication will be provided in the control room to identify the source of the initiation.
5.
Annunciation will be provided in the contcol room to alarm:
Auto start of the EFW system. This will be a common alarm for both the trains.
2.1-37 Am. 12
2.1.2.7
, Increased Range of Radiation Monitors (2.1.8.b) 2.1.2.7.1 The existing Radiation Monitoring System provides in-line monitoring capability for ef fluents f rom:
a) Auxiliary and Fuel Handling Building (RM-A8) b) Reactor Building Purge (RM-A9) c) Condenser Of f-Gas (RM-AS)
Discharge f rom Waste Gas Decay Tanks are monitored by RM-A7 prior to combination with other exhaust and af ter dilution by RM-A8.
The Reactor Building Hydrogen Purge System discharge is monitored by the normal purge system monitor RM-A9.
The mo nitors, RM-A8 and RM-A9 are manufactured by Victoreen, I nc. and consis t of:
a) A fixed filter particulate monitor; Beta scintillation 10 com/ain detector; sensitivity approximately 1.5 x 10 C1/cc 6
based on SR-90; full range 1 x 10 cpm.
b) A Fixed Charcoal Filter Iodine Monitor; NaI detector with 9 cpm / min fixed window; Sensitivity approximately 1.3 x 10 C1/ce; fill range 1 x 10 cpm.
c) A gross gaseous monitor; Bata Scintillation detector;
'E Sensitivity approximately 4 x 10 I"11 ##"8*
C /ce; 1x 10 cpm.
d) Air sampling pu=p with normal sample flow of approximately 1 cubic foot pe r minute.
Radiation monitor RM-A5 has only a gross gaseous monitor (c above) situated on the discharge of the condenser vacuum pumps, exhausting to the suction of the vacuum pucps.
Flow through the monitor is regulated to maintain approximately 500 cc/ min.
All monitors have Control Room readout and recording.
2.1~38 Am. 12
2.1.2.7.2 Long Term Modifications Increased range capabilities will be furnished for each of the effluent monitors described above (RM-AS, RM-A9, RM-A5) and the Main Steam lines.
For the Long Term Modification additional monitoring ranges will be provided utilizing ionization chambers for the Reactor Building Purge Exhaust, the Condenser 0FF-GAS Exh aus t, the Main Steam Lines.
The Auxiliary and Fuel Handlirg Building Exhaust will have extended monitoring ranges incorp-orating a G.M. device. The sensitivity of the individual units will be determined by standard volume source calculations.
The sensitivity will assure that release rate of:
5.600,000 C1/see f rom Auxiliary & Fuel Handling Bulding.
2,300,000 C1/sec f rom Reactor Building Purge.
1400 C1/see f rom Condenser Off-Gas based on maximum flow rates from each release path.
2500 Ci/sec f rom a single steam generator can be detected.
The ins tallation of each monitor will include evaluation of the position of the monitor relative to other potential radiation sources and shielding necessary to minimize the effect of sources other than sample lines on the response of the monitor and recording.
For each of the monitors described, the following applies:
Each will be powered f rom vital power, thereby providing redundancy in power supply.
Establishing sensitivities will be correlated to solid source calibrations.
Procedures defining calibration method and f requency will be written to assure proper response of the ins truments.
Emergency procedures will be written to the use of the radiation instrumentation in conjunction with flow infor-nation to determine release rate.
Emergency Plan implementing procedures describe the dissemination of information obtained f rom monitors.
Procedures and evaluations will be available for NRC review prior to 1 January 1981.
2.1-39 Am. 12
2.1.2.7.3 Short Term Modifications Increased range capabilities will be furnished for the Reactor Building Purge Exhaus t, the Conde nser Off-Gas Exhaust, and the Main Steam Lines as a short term modification.
This Short Term Modification will consist of G.M. Tubes or ionization daambers affixed to each of the ef fluent release paths described i
in 2.1.2.1.1 (only one detection system will be provided for j
each OTSG).
Remote readout will be provided to areas, which are I
habitable during an accident. The Long Tern Modification for the Auxiliary & Fuel Handling Building is projected to be com-pleted by 1 June 1980.
If a Long Term Modification is not available by start-up, a Short Term Modification utilizing a G.M. tube or ionization chamber will be incorporated.
All devices will have necessary shielding if background effects are considered excessive.
The installation of each monitor will include evaluation of the position of the monitor relativve to other potential radiation sources and shielding necessary to minimize the effect of sources other than sample lines on the response of the mo nitor.
The sensitivity will assure that release rates of:
5,600,000 C1/sec f rom Auxiliary & Fuel Handling Bldg.
2,300,000 C1/sec f rom Reactor Building Purge.
1400 Ci/sec f rom Condenser Off-Gas based on maximum flow rates froc each release path.
2500 Ci/see f rom a single steam generator can be detected.
The range of these monitors is identical to the range capa-bility of the long term modification.
For each of the monitors described, the following applies:
Each will be powered f rom normal power with battery backup.
Established sensitivities will be correlated to solid source calib ra tion.
Procedures defining calibration methods and f requency will be written to assure proper response of the i ns t rume nts.
Emergency procedures will be written to the use of radia-tion ins trumentation in conjunction with flow information to determine release ra te.
Emergency Plan implementing procedures describe the dis-semination of information obtained f rom the monitors.
Procedures and evaluations will be available for NRC review prior to restart of Unit I or 1 October 1980, whichever occu rs fi rs t.
2.1-40 Am. 12
'lllREE MIL SLAND UNIT NO. 1 TABLE 2.1-4 LEAKAGE REDUCTION PROGRAM TEST SUfDtARY SYSTEMS TEST METiiOD ACCEPTANCE CRITERIA 1.
Makeup and Purification Reactor Coolant System Leak Rate Calculation la t e r (Surveillance Procedure #1303-1.1) monitors the leakage rate of this system daily during reactor operatlun. On a not-to-exceed-re-fueling interval basis, Makeup Pump IC will be run to subj ect the normally static-Illgh Pressure Injection Piping up to MU-V16C and MU-V16D to operating pressure for lealage rate calculations.
2.
Liquid Waste Disposal A new surveillance procedure has been draf ted Later which will collect and measure leakage on the Reactor Coolant letdown piping from MU-V8 to the Reactor Coolant Bleed Tank. This test will be performed each refueling interval.
3.
Decay Heat Removal Leakage from this system is collected and Leakage shall not exceed measured using Surveillance Procedure six gallons per hour.
- 1303-11.16 and meets the require nents of Technical Spec ification 4.5.4 Decay Heat Removal System Leakage. This test will be performed each refueling interval.
4.
Reactor Building Spray Surveillance Procedure #1300-3A A/B is being I.ater revised to include a 1-hour leak rate determination with the Reactor Building g
Spray Ptrmps recirculating to the Borated Water Storage Tank. This test will be
[
performed each refueling interval.
5.
Waste Gas Disposal A new surveillance procedure has been draf ted later to determine the leakage rate of the entire Waste Gas System. Nitrogen will be used in lieu of helium when performing this test.
The test will be performed each refueling
TIIREE ?!ILE.
AND UNIT NO. 1 TABLE 2.1-4 (Cont 'd. )
LEAKAGE REDUCTION PROGRAff TEST SUtttARY SYSTEMS TEST IIETIIOD ACCEPTANCE CRITERIA 6.
Reactor Coolant Sampling A new surveillance procedure has been later draf ted to determine the leakage rate from the Reactor Coolant Sampling System.
Th is test will be performed in conjunction with the taking of a coolant sample on a re-fueling interval frequency.
7.
Reactor Building Containment The Containment Build ing leakage rate is See Technical Specif1-determined by the performance of Surveillance cation 4.4.1 Containment Procedure #1301-11.18, Reactor Building Irakage Tests Local Leak Rate Testing; #1303-6.1, Reactor Building Integrated Leak Rate Testing; and
- 1303-11.24, Reactor Building local Leakage-Penetration Pr essurization. Test frequencies are as specified in Technical Specification 4.4.1 Containment Leakage Tests.
8.
Containment Monitoring Leakage Measurements are provided by the See Technical Specifi-performance of Surveillance Procedure cation 4.4.1 Containment
- 1303-11.18, Reactor Building Local Leak Irakage Tests Rate Testing on a refueling interval frequency.
9.
Fluid Block Leakage measurements are provided by the See Technical Specifi-performance of Surveillance Procedure cation 4.4.1 Containment g
- 1303-11.23, Reactor Building Local Leakage-Leakage Tests Fluid Block on a refueling interval f requency.
10.
Miscellaneous Waste Leakage measurements will be incorporated la t er Storage Tank into the nitrogen leak test of the Waste Gas Disposal System described in Item 5 above.
THREE PILE ISLAND UNIT NO. 1 TABLE 2.1-5 SYSTEMS NOT !NCLUDED IN LEAKAGE REDUCTION PRO EAM SYSTEMS PEASCN F0F E.XCLUSION 1)
Liquid '..' acte Disposal (except Except for those portions subject to tr.e Reactor Ccolant letdo.n pioing nitrogen test cf the Waste Gas Disposal f r0r.'G '!5 to Reactor Coolant Eleed Tanks)
System, this system will not be tested.
The untested portions of this sytem are not required to adjust I,eactor Coolant System inventory during a serious accicent er trar.sier.: and 'could only e usc: -
deliberate fashion during the long-term recovery phases.
2)
Hydrogen Recembiner This piping is not presently instclied in TMI-1.
Upon installation, tne applicable portions of this system will be tested for leakage and on a refueling interval therea f ter.
3)
' aste Gas Disposal '.'ent to This piping is not presently installed in Contain: cent TMI-1.
Upon installation the applicable portions of this system will be tested for leahage and on a refueling interval thereafter.
4)
..ecctor Building Sumo Sampling This piping is not presently installed in TMI-1.
Upon installation the appiitcble portions of this system will be tested for leakage and on a refueling interval the rea f te r.
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In addition to the benefits discussed above, the venturis provide two added benefits.
First, they balance flow of the injection legs under all other small break conditions such that TMI-1 flow split will be within the bounds of the generic analysis (i.e., 70%/3-% flow split).
Secondly, the cavitating can be relaxed.
8.3.16 Large Break Lcss of Coolant Accidents (Reference FSAR Section 14.2.2.3) 1.
Description Break sizes in the reactor coolant system (RCS) greater than 0.5 ft.2 are classified as large break loss of coolant accidents (LOCA's).
These breaks involve rapid depressuri-
=ation of the RCS and are accompanied by rapid increases in containment pressure. Offsite doses are calculated from the design basis radioactivity release to containment, and the design basis containment leak rate.
2.
Acceptance Criteria 1.
Peak fuel clad temperature does not exceed 2200*F.
ii.
The core is maintained in a coolable geometry.
iii.
Local fuel cladding oxidation (metal water reaction) shall not exceed 0.17 times the total cladding thick-ness of.05 times the total cladding mass.
iv.
Offsite doses are within the limits specified by 10 CFR 100.
3.
Mitigation 1.
Core flood tank actuation at 600 psig to establish water inventory.
ii.
Low pressure injection system flow below 200 psig to establish core cooling for the remainder of the accident.
iii.
Building spray addition to put iodine in solution with the containment water volume thus decreasing release to the environment.
iv.
Containment leak tightness to limit radioactivity releases.
v.
Switchover of the decay heat removal system suction source to the containment building sump on low-low BWST level.
4.
Conclusion The calculated off site dose resulting from the design basis LOCA may increase as a result of the deletion of sod!um thiosulfate from the building spray system.
Doses will still 8-17 Am. 12
TABLE 8A-1 PROPOSED PITRAN/GPU-01 ANALYSES OF TMI-l I.
Description Results Case 1:
Lase (1840 gpm EFW, Show plant response to LOOP &
normal 6 augmented makeup transition to natural circulation.
Case 2:
1840 gpm EFW and L*; decay Examine overcooling potential Figure SA-3 heat (max cooldown case) with 200% EFW and minimum decay heat.
Case 3:
Stuck open OTSG safety Examine plant performance for first valves (17% of design 10 minutes of secondary side flow) depressurization using one OTSG model.
3a: Same as Case 3 but with Figure 8A-2a no EFW Case 4:
100% EFW and flow limita-Examine long-term plant response with Figure SA-S tion EFW flow limitation and LOOP from 100% power.
Case 5:
1% decay heat with 1145 Examine effect of flow limiters in Figure SA-4 gpm EFW maximum EFW system on max cooldown case.
Case 6.
2 OTSG model with stuck Evaluate effect of non-symmetric open relief valves cooldown on secondary side.
Case 7:
2 OTSG model 6 no EFW Evaluate non-symmetric loss of to 1 OTSG heat sink.
Case 8:
EFW in superheat region Evaluate effect on transition to natural circulation when model puts EFW in superheat region rather than downcomer.
Shows more realistic plant response.
Case 4:
Low OTSG level due to Evaluate effect of loss of natural failure in level con-circulation due to inadequate EFW troller and skewed power level in OTSG.
Power distribution level skewed to top to minimize natural circulation driving head.
Case 10: No EFW Evaluate plant response to loss of Figure 8A-11 heat sink under natural circulation conditions.
Am. 12
TABLE SA-1 PROPOSED RETPAN/G?U-01 ANALYSES OF TMI-l (Continued)
Reference I.
Loss of Offsite Power (LOOP) (Cont.)
Description Results Case 11: No EFW with PORY stuck Establish time until RCS becomes open saturated when forced flow is unavailable.
Case 12: 2 loop model with no EFW Study plant response to a complete or steam relief on 1 side loss of secondary heat sink re-sulting from complete instantaneous unavailability of an OTSG Case 13: Controlled cooldown Cooldown of plant initiated from stable plant condition following a LOOP.
Case 14: Controlled cooldown using Same as 13, but with only one OTSG one OTSG.
available for cooldown.
Ic. 12
TABLE SA-1 PROPOSED RETR /d/GPU-01 Ls'ALYSES OF D I-l (Continued)
Reference II.
Station Elackout Description Results Case 1:
Ease Look at long-term plant response Figure SA-7 to event, including voiding in the RCS LOOP.
Case 2:
1 gpm pressurizer leakage Effect on plant response due to Figure SA-12 and letdown isolation at cooldown of pressurizer steam 20 minutes space.
Maximizes expected cooldown.
2a:
1 gpc pressurizer leaks-a, Figure SA-13 letdown isolation at 26 minutes and minimum decay heat b
Am. 12
TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF TMI-l (Continued)
III.
Loss of Feedvater Description Results Case 1: 460 gpm EFW flow l 1.2 ANS Examine plant response to operation Figure SA-9 decay heat of only one motor-driven EFW pump.
la: 460 gpm EFW flow &.85 ANS Figure SA-10 decay heat lb: 460 gpm EFW flow 6 1.2 ANS decay heat & trip of 2 RC pumps.
Case 2:
EFW flow limitation Look at plant transition to stable Figure SA-6 shutdown with EFW flow limited.
Case 3:
Base case 1840 gpm EFW Show plant response to LOFW with Figure 8A-5 6 no equipment failures no equipment failures and no operator actions.
Case 4:
Failure of EFW level Evaluate ef fect of continuous EFW Figure SA-1 control & 1% ANS decay heat flou causing overcooling of OISG.
Case 5:
LOFW during plant startup Look at plant response just below turbine trip threshold with no trip on LOFW.
Case 6:
Case 2 using 2 LOOP model Compare 1 & 2 LOOP results.
Case 7:
Partial LOFW (pressurizer Compare response to complete LOFW spray causes limitation on
& evaluate ICS response.
(Picked to pressure increase - trip represent worst case partial LOFW.)
on variable low pressure)
Case 8: No EFW Look at time available before HPI must be initiated.
Case 9:
Turbine stop valve fails Limiting overcooling case following open a LOFW.
Case 10: Manual actuation of HPI Evaluate effect on pressurizer level of operator initiating a second HPI pump immediately after reactor trip.
Evaluate time before level goes off-scale (high).
Am. 12
TABLE 8A-1 PROPOSED RETRAN/ CPU-01 ANALYSES OF TMI-1 (Continued)
III. Loss of Feedwater (Cont.)
Description Results Case 11: Loss of feedwater with Quantify benefit of anticipatory trip on high pressure reactor trip.
Case 12: Overcooling subsequent Evaluar.e ef fect of tripping RC to LOFW which causes SFAS pumps following an LOFW event and initiation subsequent cooldown.
Case 13: Spurious feedwater rupture Evaluate event which causes loss detection signal to one of normal and emergency feedwater OTSC.
to one OTSG.
Am.
12
TABLE SA-1 PROPOSED RETRAN/GPU-01 ANALYSES OF TMI-l (Continued)
IV.
Feed Line Break Accident Description Results Case 1:
EFW initiation on Demonstrate plant can tolerate feed / steam LP the design basis feedline break accident.
Case 2: No EFW Evaluate latest time at which EFW can be initiated for design basis feed line break.
Det ermine what plant parameters can effec-tively be used to initiate EFW besides feed steam LP.
Case 3:
Partial break which causes Evaluate spectrum of LOFW events gradual loss of OTSG level and ability of plant to respond and no feed / steam LP to these breaks.
signal Am. 12
TABLE 8A-1 PROPOSED RETR /J;/ CPU-01 A'iALYSES OF "MI-l (Continued)
V.
Steam Line Break Accident Description Results Case 1:
Lenchmark of TMI-2 docket Establish benchmarked code for analysis use on TMI-1.
Case 2:
TMI-l design basis analysis Determine if means are available (no subsequent failures) to establish long-term safe shutdown.
Case 3:
Break from startup condi-Determine if startup from low tion with flooded nozzles OTSG level is necessitated by saf ety concerns.
Case 4:
Break spectrum Evaluate if plant can respond to a spectrum of steam line breaks and satisfy all safety criteria.
Am. 12
TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF TMI-l (Continued)
VI.
Reactor Trip Description Results Case 1:
Base case plant response Establish nominal plant response to reactor trip from to a reactor trip due to a nominal conditions and turbine trip.
normal feedwater available Case 2: Deletion of feec pump Determine value of pump kicker kicker circuit circuit in aiding a stable plant shutdown following trip.
Case 3:
Reduction of turbine bypass Determine if turbine bypass capacity from 22.5% to 15%
capacity should be returned to 15% capacity.
Case 4:
Sensitivity to changes in turbine bypass control setpoint.
Case 5: Failure of turbine trip to Evaluate plant response to delayed cause reactor trip reactor trip on high pressure.
Case 6a: Failure of normal level Investigate overcooling potential controller after a reactor from normal feedwater addition trip assuming loss of (loss of OTSG pressure causes feedwater heating feedwater isolation).
6b: Same as 6a, but without Investigate overfill potential from loss of feedwater heating normal feedwater.
(Isolation on low pressure does not occur if feedwater heating is not lost.)
Am. 12
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RESPONSE TO OUESTION 20. SUPPLEMENT 1, PART 1 TABLE 20-1 Information on the Subcooline Meter Plant Name: Three Mile Island, Unit No. 1 Vendor:
Foxboro (Calculator)
Keference for Information-Display Information Display Tsat Margin Display Type Dicital Continuous or en Demand Continuous Single or Redundant Display Sincle (Selectable from redundant inputs)
Location of Display Control Room Console Alarms Low Tsat marcin Overall uncertainty 23 F 0
Range of Display 1000F superheat to 4000F Subcooline Qualifications Seismic Calculator Type Analoc If process computer is used specify availability Not applicable Single or redundagt calculators Redundant Selection Logic Highest T.
Qualifications (Seismic, Environ, IEEE 279 Calculation Technique Steam Table Approximation For computing saturation temperature or pressure, the steam table saturation curve will be synthesized with 0.5% accuracy by means of a " Signal Characterizer" module.
This is a solid state func-tion generator which has the capability of simulating the charac-teristic curve of a process by means of a number of straight line segments.
Up to eight segments may be used. The slope and inter-section of each segment are individually adjustable.
Am. 12
RESPONSE TO OUESTION 20. SUPPLEMENT 1, PART 1 TABLE 20-1 (Cont ' d. )
Innut Temperature RTD's Temperature 4 sensors, T hot Range of Temperature sensors 1200 - 9200F Uncertainty of temperature sensors (OF at 1) 60F Qualifications (seismic, environmental, IEEE 323) control Grade (Short Term)
Pressure (Specify instrument used)
Foxboro EllGH Pressure (Number of sensors and locations)
Two, PZR Press.
Range of Pressure sensors 0-2500 PSIC Uncertainty
Seismic, Environmental Backup Capability Availability of Temp & Press Individual press. & temo. indic available in Control Room Availability of Steam Tables etc. Steam tables available on control panel and in computer Training of opergpors In accordance with operator retraining program Procedures Procedures will be cenerated for use of instruments Other Tsat will be independentiv coicouted by plant computer Am. 12
SUPPLEMENT 1. PART 1 QUESTION:
27b.
Valve M'J-V3 does nat receive a diverse safety grade automatic isolation signal.
This is unacceptable.
Modify your design accordingly.
RESPONSE
The letdown isolation system is a closed, Seismic Category I system outside containment.
In accordance with SRP 6.2.4, items II.3.c and e, remote manual isolation is required for this system. Automatic letdown isolation has been provided on TMI-1 in order to relieve the operator of one immediate manual action after trip.
Consideration was given to automated isolation of h0-V3; however, the desire for automatic isolation was considered less important than the availability of the letdown path after isolation.
Since failure of MU-V3 to reopen would result in loss of the letdown path, it was maintained as a remote manually operated valve. This philosophy of isolation is similar to the treatment of ECCS lines, where the need for injection and need to maintain containment isolation are balanced by providing remote manual isolation.
Plant operating procedures will be developed so that the operator makes an evalation of the consequences of reopening this line after a reactor trip.
In addition, high radiation will isolate letdown by closing MU-V2A, B.
RCS inventory control is maintained with the makeup and letdown systems following reactor trip.
Once pressurizer level is established, makeup and letdown are used to maintain reactor coolant pump seal flow cooling.
This process introduces about 30 gpm into the RCS and necessitates about 45 gpm letdown in order to maintain proper inventory.
If letdown flow was lost, water could be supplied to the makeup pumps either via the boric acid transfer pumps to the makeup tank or directly from the berated water storage tank.
In both instances, however, there would be a net addition of water to the RCS with an increasing pressurizer level of 1-2 inches a minute.
Eventually, seal injection would have to be terminated.
The RC pump seals can be protected from damage without seal injection if ICCW flow is available to the thermal barrier coolers. This mode is not the preferred manner of operation, however.
Furthermore, the ICCW lines are subject to jet impingement and pipe whip.
The ICCW lines could be disabled, in which case RCP seal protection would be dependent seal injection.
In summary, failure of the MU-V3 valve resulting in loss of letdown would have a detrimental effect on plant safety.
Redundant means of seal inj ection would be lost and RCS inventory control would also be lost.
It was concluded that the letdown function is important to orderly plant shutdown; and, given the Icvel of leakage protection from the letdown line, that a redundant isolation signal was counter productive to plant safety.
Am_ 12
SUPPLEMENT 1, PART 1 QUESTION:
27c.
Valves IC-V2, IC-V3, IC-Y4 and IC-V6 do not receive a diverse safety grade automatic isolation signal unless a safety grade line break isolation signal is included in the design.
Sufficient details have not been provided to ascertain if this design feature is incorporated in your design.
If the line break detection feature is incorporated in your design, provide the details supplemented with sufficient electrical drawings to allow an independent evaluation of whether or not the design meets IEEE Std. 279-1971.
If the line break detection feature is not incorporated in your design, the design is unacceptable and must be modified accordingly.
RESPONSE
The rationale for the choice of containment isolation signals is illustrated in Figure 2.1-S.
One option available is to make the ICCK lines inside containment Seismic Category I and protected from the effects of pipe whip and jet impingement.
Remote manual isolation would then be required in accordance with SRP 6.2.4.
Pipe whip / jet impingement surveys are now complete.
The results of the survey showed that these lines cannot be adequately protected against pipe whip and jet impingement, therefore, a line break detection isolation system is to be provided.
A conceptual description of this new feature will be frovided by February 29, 1980.
e Am. 12 9
SUPPLEMENT 1, PART 1 QUESTION:
45.
(Order Item 1(d))
Your response to this item indicates that procedures have been or are still being revised.
Provide the procedures developed to define operator action during small break LOCA's.
RESPONSE
EP 1202-6 has been provided for your review.
This procedure defines operator guidance during small break LOCA.
In addition, the B&W Guidance (B5W Document 69-1106001) used in developing the small break LOCA procedures is attached.
e Am. 12 7