ML20148B966
| ML20148B966 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/18/1980 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 1-010-22, 1-10-22, NUDOCS 8001240400 | |
| Download: ML20148B966 (81) | |
Text
r._
m ARKANSAS POWER S LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK. ARKANSAS 72203 (501)371-4000 January 18, 1980 1-010-22 2-010-18 Director of Nuclear Reactor Regulation ATTN: Mr. Darrell G. Eisenhut, Acting Director Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Arkansas Nuclear One-Units 1 & 2 Docket Nos. 50-313 and 50-368 License Nos. DPR 51 & NPF 6 Lessons Learned Implementation (File: 1510.3, 2-1510.3)
Gentlemen:
Mr. H. Denton's letter of October 30, 1979, requested Arkansas Power and Light Company to document our method of compliance with the "short-term" Lessons Learned requirements which do not require prior NRC approval. As committed in our November 20, 1979 letter, a description of the methods of implementation / compliance for those items is hereby provided.
Very truly yours,
.D ant.E (. TndA David C. Trimble Manager, Licensing DCT:nak Attachment 6
90030%2 u6 O Y
g1S MGMBEA MCOLE SOUTH UTILITIES SYSTEM
ATTACHMENT 1 i
Item 2.1.1 Emergency Power Supply AH0-1 We have determined that 126 Kw of pressurizer heating is required within two hours of a loss of offsite power to establish and maintain natural circulation of the RCS if forced circulation is lost.
The AN0-1 pres-surizer has available 84 Kw of proportional heaters on each channel of the class 1E power system.
During this January-February 1980 outage, forty-two Kw of backup (on-off) heaters will be added to the swing bus (connected to either one of the diesel generators).
The combination will produce the necessary 126 Kw.
The diesel generator loading was reviewed with the conclusion that the additional load is permissible.
The added 42 Kw will neither effect the diesel loading sequencing nor exceed the seven day rating.
Therefore, these non-class 1E loads will be automatically loaded and will not be shed upon SIAS actuation. The class lE interfaces for main power and control power will be protected by safety grade circuit breakers.
The ANO-1 pressurizer level instrument channels are presently powered from vital instrument buses.
The ANO-1 PORV moti.ve and control power is presently provided from the channel 1 (red) D.C. system. The PORV block valve, following this shutdown, will be powered from the channel 2 (green) A.C. system.
The changeovers of motive and control power from the normal offsite power source to the emergency power source will be automatic.
ANO-2 We have determin'ed that 150 Kw of pressurizer heating is needed within 30 minutes of loss of offsite power to maintain natural circulation.
One Hundred Fifty Kw of proportional heaters is currently powered from each safety bus.
The existing control circuit for the propo'rtional heaters will be modified, during the January-February 1980 shutdown, to allow closing of the power circuit breakers upon loss of offsite power following a Safety Injection Actuation Signal.
The control circuit also contains an undervoltage re-lay which will trip the circuit breakers when the motor control center is transferred to the onsite emergency source.
Manual operation of a handswitch in the Control Room is necessary to close the circuit breaker.
The main and control power interfaces are protected by safety-grade cir-cuit breakers.
The diesel generator loading was reviewed for LOCA, MSLB and Blackout conditions.
It has been concluded that under Blackout conditions (loss of offsite power) the diesel generator rating' will not be exceeded.
Under s
loss of coolant or main steam line breaks, the continuous rating of one diesel could be exceeded, however, neither the seven day nor the two hour ratings were approached or exceeded.
Therefore, the diesel generators are capable of carrying the additional load without compromising their ability to handle the emergency load.
90030M3
1 The AN0-2 design does not incorporate a PORV or block valve.
The AN0-2 pressurizer level indicators are safety-grade, qualified, redundant and powered from safety buses; therefore, they presently meet the necessary requirements.
l Item 2.1.2 - Relief and Safety Valve Testing AND-1 & 2 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 PWR Safety / Relief Valves and Systems," De-cember 13, 1979.
j The EPRI Program Plan provides for a completion of the essential portions of the test program by July, 1981. We will be participating in the EPRI i
program to provide program review and to supply plant specific data as J
required.
Item 2.1.3.a Direct Indication of PORV and Safety Valve Position 1
ANO-1 & 2 An acoustical valve monitoring system is being installed on AN0-1 during 1
January 1980. An acoustical valve monitoring system is being installed on ANO-2 during its January-February 1980 outage.
These systems will provide the operator with an indication of valve position (i.e., open/
closed) on an annunciator panel in the Control Room.
This alarm will sound when either valve is opened.
Redundant sensors which will be located on each valve will transmit re-dundant signals to a signal conditioner located in the Control Room.
Only one of the inputs will be connected to the signal conditioner at a time. However, the operator can manually switch to the other input.
The system is single failure proof from the sensors to the signal conditioners.
In additior., if one signal conditioner fails, the operator can still get a.
position indication by connecting to signal conditioners for one of the other valves.
This acoustical valve position indication equipment will be powered from a ' safety grade power source.
All equipment will also be mounted as seismic class I installations.
In addition, the equipment is environmentally suited for its application.
Sufficient QA documentation, however, does not exist to qualify the equipment seismically or environmentally.
Generic qualification should begin in February, 1980, and will require approximately six (6) months to complete.
Temperature elements downstream of the PORV and safety valves on ANO-1 and downstream of the safety valves on ANO-2 provide backup indication of valve position. These are monitored in the control room and alarm on high temperature.
2 90030164
Item 2.1.3.b - Instrumentation for Dection of Inadequate Core Cooling AN0-1 Our letter of December 13, 1979, to Mr. R. Reid provided the guidelines incorporated into our plant procedures used for detection of inadequate core cooling with currently available instrumentation. A description of the existing instrumentation used is not available at this time, but will be supplied by January 31, 1980.
The design for additional instrumentation for detection of inadequate core cooling is not available at this time, but will be supplied by January 31, 1980.
For indication of primary coolant saturation conditions, we will install two (2) channels of margin to saturation measurement and indication during the January 1980 outage. A functional block diagram of a single channel is shown on the attached figure and is described in the following paragraphs.
For indication of temperature margin to saturation, the calculator selects the highest temperature input.
The calculator utilizes the pressure input as a pointer to locate the corresponding saturation temperature in steam tables resident in the calculator memory. The process temperature is sub-tracted from the saturation temperature and the difference is then available for recording and display.
The pressure margin to saturation calculation is done'in a similar manner.
Each channel calculator receives a single, safety-grade, wide-range (0-2500 psig) primary coolant pressure input from existing buffered outputs present in the Engineered Safeguards System.
One channel is fed its pressure signal from reactor coolant loop A; the other channel drives its pressure input from loop B.
Also, each channel calculator receives two (2) wide-range (1200-9200F)
RTD temperature inputs, one from each reactor coolant hotleg loop.
As referenced in our November 20, 1979, letter to D. G. Eisenhut, the interim installation will use existing non-safety grade temperature signals available in the Non-Nuclear Instrumentation System. This is necessary as qualified RTD bridges will not be available until May 1980.
The temperature inputs will be upgraded to safety grade requirements during the first outage of suf.ficient duration upon receipt of the equipment, but no later than the next refueling outage.
The temperature margin to saturation conditions from each calculator is to be
' continuously recorded on a two (2) channel strip chart recorder located on the main control board. The temperature margin from each calculator may also be individually read from digital indicators mounted in cabinets located in the control room.
By manual selection, pressure margin to saturation may also be read from these digital indicators.
Annunciation of low margin to saturation (temperature only) will be provided on the main control room panels.
Isolation will be provided between the calculator and recorder as well as the calculator an annunciator.
l In addition to the redundant margin to saturation monitoring described above, backup capability already exists on the AN0-1 plant computer.
This computer program functions in a similar manner to the margin calculators.
In addition, primary coolant temperature and pressur.e are directly available to the operators 90030165 3
by means of existing console indicators.
Steam tables are provided in the con-trol room for use by the operators to determine saturation margin manually.
These measures will be used until the upgraded temperature inputs are installed.
The margin to saturation monitoring channels will be installed prior to start-up from the present outage.
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6 90030166 a
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INFORMATION REQUIRED ON THE SUBC00 LING METER Display Information Displayed (T-Tsat, Tsat, Press, etc.)
Tsat - T or P-Psat Display Type (analog, Digital, CRT)
Digital and recorded Continuous or on Demand
- Continuous Single or Redundant Display Redundant Location of Display Control Room t.larms (include setpoints)
To be supplied later Overall uncertainty (oF, PSI)
+ 80F; 45 psi Range of Display Display: 0-199.90F; 0-1999 psig Recorder: 0-1000F Qualifications (seismic, environmental, IEEE323)
Seismic, environmental per IEEE 323 and 344 Calculator Type (process computer, dedicated digital or analog calc.)
Dedicated digital calcula.
If process computer is used, specify availability.
NA l
'(% of time)
Single or redundant calculators Redundant Selection Logic (high'est T., lowest press)
See description j
Qualifications (seismic, environmental, IEEE323)
Seismic, environmental per IEEE 323 and 344 Calculational Technique (Steam Tables, Functional Fit,-
Steam tables ranges)
Input Temperature (RTD's or T/C's)
Rosemount 177GY RTD Temperature (number of sensors and locations)-
Each channel has a hot-leg A and B input for a total of 4. on hotleg piping.
L Range of temperature sensors C)0030167 1200 - 9200 F l'
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Uncertainty
- of temperature sensors (OF at 1)
+ 5.00F Qualifications (seismic, environmental, IEEE323)
Seismic, environmental Pressure (specify instrument used)
Foxboro EllGH-INM2 P.T.
Pressure (number of sensors and locations)
Channel A-one, hotleg A l
Channel B-one, hotleg B
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located in containment Range of Pressure sensors 0-2500 psig Uncertainty
Seismic, environmental Backup Capability Availability of Temp & Press See description Availability of Steam Tables, etc.
See description Training of operators Procedures
- Uncertainties must address conditions of forced flow and natural circulation
- - Includes cumulative uncertainties of all intermediate transmitters, bridges, and signal converters between sensors and calculators.
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Al10-2 Guidelines for detection of inadequate core cooling using currently available instrumentation have been incorporated into operating procedures.
The evaluation for additional instrumentation for detection of inadequate core cooling is complete. The instrument response of a reference C-E plant for events which have the potential for inadequate core cooling is documented in CEN-ll7, " Inadequate Core Cooling -- A Response to NRC IE Bulletin 79-06C, Item 5 For Combustion Engineering Nuclear Steam Supply Sys-tems".
The con-clusion reached in CEN-ll7 is that there currently is sufficient instrumentation in the plant to be used to detect inadequate core cooling. ANO-2 has updated its emergency procedures and associated operator training based on recommended guidelines from Combustion Engineering.
We have received several variations of a conceptual design for Reactor Vessel Water Level Indication from C-E as part of the C-E Owner's Group effort.
This design (s) is currently being evaluated as to feasibility and value in assessing the extent of core uncovering during an accident.
We will advise you by Feb-ruary 1,1980, of the results of our evaluation.
However, as stated in the above discussion, it is not expected that C-E's contention of no need for this instrumentation will be invalidated.
For indication of primary coolant saturation conditions, we will install two channels of saturation margin measurement and indication.
A functional block diagram of a single channel is shown on the attached figure and is described in the following paragraphs.
For indication of temperature margin to saturation, the calculator selects the highest temperature input.
The calculator utilizes the pressure input as a pointer to locate the corresponding saturation temperature in steam tables resident in the calculator memory.
The process temperature is sub-tracted from the saturation temperature and the difference is then available for recording and display. The pressure margin to saturation calculation is done in a similar manner.
The pressure input for each subcooled margin calculator is derived from redundant, safety-grade, wide-range (0-3000 psig) pressurizer pressure transmitters. These transmitters also provide the pressure signals to the Plant Protection System. The two temperature inputs for each calculator are from redundant, safety-grade, wide-range (150-7500F) TH0T RTD's in each loop.
Redundancy requirements are satisfied by separate and redundant subcooled margin monitoring channels.
Tne temperature margin to saturation from each calculator is to be con-tinuously recorded on a strip chart recorder located on the main control board. The temperature margin from each calculator may also be displayed on a digital indicator in the back of the control room.
Optionally, the pressure margin to sattiration may be displayed on the indicator.
Separate annunciators are provided on the main control room panels.
Isolation is provided between the calculator and recorder as well as the calculator and annunciator.
Although not required upon completion of installation, a backup to the redundant channels of subcooled margin monitoring is provided by means 90030170
of an existing program on the ANO-2 plant computer.
This program functions in a similar manner to the margin calculators previously described.
In addition, primary coolant temperature and pressure are directly available to the operators by means of existing indicators.
Steam tables are pro-vided in the control room for use by the operators to determine saturation margins.
The subcooled margin monitors will be installed during the outage beginning on.or before January 31, 1980.
e 90030171 9
9
INFORMATION REQUIRED O!1 THE SUBC00 LING METER Display Information Displayed (T-Tsat, Tsat, Press, etc.)
Tsat - T or P-Psat Display Type (Analog, Digital, CRT)
Digital and recorded Coritinuous or on Demand
- Continuous Single or Redundant Display Redundant Location of Display Control Room Alarms (include setpoints)
To be supplied later Overall uncertainty (OF, PSI) 1 50F, i 46 psi Range of Display 0-199.90F; 0-1999 psi Qualifications (seismic, environmental, IEEE323)
Seismic, environmental per IEEE 323 and 344 Calculator Type (process computer, dedicated digital or analog calc.) Dedicated digital calcula.
If process computer is used, specify availability.
NA
(% of time)
Single or redundant calculators Redundant Selection Logic (highest T., lowest press)
See description 1
Qualifications (seismic, environmental, IEEE323)
Seismic, environmental per IEEE 323 and 344 Calculational Technique (Steam Tables, Functional Fit, Steam tables ranges)
Inout Temperature (RTD's or T/C's)
Rosemount RTD j
Temperature (number of sensors and locations) 2 Hotleg/ loop (1 from each loop / calc.)
Range of temperature sensors 150-7500F lo 90030172
Uncertainty
- of temperature sensors (OF at 1) 0.750F Qualifications (seismic, environmental, IEEE323)
Seismic, environmental Pressure (Specify instrument used)
Rosemount 1153A Pressure (number of sensors and locations) 1 pressz/ calc.
0-3000 psig Range of Pressure sensors Uncertainty
Post LOCA environ-mental, seismic Backup Capability Availability of Temp & Press ite Description Availability of Steam Tables etc.
See Description Training of Operators Procedures
- Uncertainties must address conditions of forced flow and natural circulation.
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Item 2.1.4 Diverse Containment Isolation ANO-1 We have reviewed the Reactor Building Isolation System design and procedures and find the following automatically actuated valves which provide penetration isolation (valve numbers in parentheses):
Reactor Coolant Pump) Controlled Bleed-off (CV-1270, CV-1271, CV-1272, 1)
CV-1273, and CV-1274.
5 2)
Reactor Coolant Letdown'(CV-1214, CV-1216, and CV-1221).
3)
Intermediate Cooling Water to Control Rod Drives (CV-2235).
4)
Chilled Water to Reactor Building Coolers (CV-6202).
5)
Intermediate Cooling Water to Reactor Coolant Pump Motors and Lube Oil Coolers (CV-2234).
6)
Intermediate Cooling Water to Letdown Cooling and Seal Water Coolers (CV-2233).
7)
Chilled Water from Reactor Building Coolers (CV-6203 and CV-6205).
8)
Intermediate Cooling Water from Reactor Coolant Pump Motors, Lube Oil Coolers and Control Rod Drives (CV-2220 and CV-2221).
9)
Fire Water (CV-5611 and CV-5612).
10)
Nitrogen Supply to Quench Tank (CV-1667).
11)
Reactor Building Sump Drain to Reactor Auxiliary Sump (CV-4446 and CV-4400).
12)
Quench Tank Drain to Auxiliary Building Equipment Drain Tank (CV-1053 and CV-1052).
13)
Reactor Building Vent Header to Auxi,liary Building Gas Collection Header (CV-4803 and CV-4804).
14)
Reactor Building Air Particulate Monitor to Auxiliary Building Gas Collection Header (CV-7453 and CV-7454).
15)
Reactor Building Purge (CV-7402, CV-7404, CV-7401, and CV-7403),
16)
Quench Tank Gas Sample (CV-1054 and CV-1845).
Items 1 and 2 above are isolated upon receipt of a Low Reactor Coolant System Pressure Signal ($ 500 psig) or High Reactor Building Pressure (E psig) Signal.
1 4
Item 3 through 16 isolate upon a High Reactor Building Pressure Signal.
The valves noted in Items 3 through 8 are normally open during power operation, as these valves are in systems which provide support to systems within the Reactor Building.
The valves noted in Items 9 through 16 are normally closed 90030175 13
during operation and require specific manual operation for opening, i.e., there is no automatic actuation of any of these valves to permit a connection from the Reactor Building Atmosphere to the Auxiliary Building Atnosphere or environ-ment.
In addition, to provide full opening of the penetration, specific manual operation of at least two valves is required, each of which requires a separate and deliberate action.
In the case of Items 3 through 8, it is felt that in order to ensure that a " normal", orderly cooldown ensues following receipt of an ES Signal (Low ReactorCoolantSystemPressure);totakemaximumadvantageofsystemsavail-able without unnecessarily proceeding to degraded modes; and'to ensure no un-warranted equipment damage, no changes to the Reactor Building Isolation Sys.:em are needed.
In reference to Items 9 through 16, based on the fact that these valves are normally closed and that specific manual action is required to breech Reactor Building isolation, no changes to the Reactor Building Isolation System are needed.
However, to provide for an increased margin of safety, we have prepared a design change to provide the valves in Items 9 through 16 above with an ES Signal to isolate on Low Reactor Coolant System Pressure (g 1500 psig).
This design change will be implemented during the January-February 1980 out-age.
Specific manual operation will be required to reset individual valves even though the ES channel has been reset.
90030176 14
At40-2 We have reviewed the Containment Isolation Actuation System (CIAS) design and procedures and have listed below the automatically actuated valves which pro-vide penetration isolation (valve numbers in parenthesis):
Category I 1)
Chemical and Volume Control System Letdown (2CV-4821-1 and 2CV-4823-2).
Category II 2)
Chilled Water Supply to containment Coolers (2CV-3852-1).
3)
Chilled Water Supply from Containment Coolers (2CV-3850-2 and 2CV-3851-1).
4)
Component Cooling Water to Reactor Coolant Pump Coolers (2CV-5236-1).
5)
Component Cooling Water from Reactor Coolant Pump Coolers (2CV-5254-2 and 2CV-5255-1).
Category III 6)
Containment Vent Header (2CV-2400-2 and 2CV-2401-1).
7)
Reactor Coolant System and Pressurizer Sample (2SV-5833-1 and 2SV-5843-2).
8) flitrogen Supply to Safety Injection Tanks (2CV-6207-2).
9)
Quench Tank Liquid Sample (2SV-5878-1 and 2SV-5871-2).
10)
Safety Injection Tank Sample (2SV-5876-2).
11)
Quench Tank Makeup Water Supply (2CV-4690-2).
12)
Containment Sump Drain (2CV-2060-1 and 2CV-2061-2).
13)
Containment Purge Inlet (2CV-8289-1, 2CV-8284-2 and 2CV-8283-1).
14)
Containment Purge Outlet (2CV-8291-1, 2CV-8286-2 and 2CV-8285-1).
15)
Low Pressure liitrogen Supply (2CV-6213-2).
16)
Reactor Drain Tank Drain (2CV-2202-1 and 2CV-2201-2).
90030177 1
15
Category IV 17)
Reactor Coolant Pump Controlled Bleedoff (2CV-4847-2 and 2CV-4846-1).
18)
Steam Generator Sample (2CV-5852-2 and 2CV-5859-2).
19)
Air Particulate Monitor in Hydrogen Purge System (2SV-8231-2, 2SV-8273-1 and 2SV-8271-2).
20)
Air Particulate Monitor in Containment Atmosphere Sample (2SV-8261-2, 2SV-8265-1 and 2SV-8263-2).
21 )
Fire Water Supply (2CV-3200-2).
Item 1 (Categ'ory 1) above is isolated upon receipt of a Safety Injection Actuation Signal (SIAS) or_ a Containment Isolation Actuation Signal (CIAS).
SIAS is generated when Reactor Coolant System pressure is less than or equal to 1740 psia or when Containment Building pressure is greater than or equal to 18.4 psia. CIAS is generated when Containment Building pressure is greater than or equal to 18.4 psia.
Items 2 through 21 isolate upon receipt of a CIAS.
The valves noted in Items 2 through 5 (Category II) are normally open during power operation since they are in systems which provide support to needed systems within the Containment Build-ing. The valves noted in Items 6 through 16 (Category III) are normally closed during power operation and are only opened periodically by specific manual operation, i.e., there is no automatic opening of any of these valves.
The valves noted'in Items 17 through 21 (Category IV) are normally open during power operation, but are not necessary to be open following receipt of a SIAS.
Since Items in Category II are providing support to systems within the Contain-ment Building, the valves should stay open upon receipt of a SIAS to prevent un-necessary equipment damage.
The systems represented in Category II contribute to a " normal", orderly cooldown following receipt of a SIAS.
Items in Category III are normally closed during power operation and specific manual operation is required to open them.
Furthermore, to'cause full opening of the penetration, specific manual operation of at least two valves is required, each of which requires a specific and deliberate action.
Based on this f'act, no changes to the Containment Building Isolation System are needed.
Items in Catagory IV are normally open during power operation and specific manual operation is required to close these valves following receipt of a SIAS.
Each of the Category IV systems.were reviewed and it has been verified that a direct connection between the Containment Building atmosphere and the Auxiliary Build-ing atmosphere or the environment does not exist while these penetrations are open.
Based on this fact, no changes to the Containment Building Isolation System are needed.
However, to further increase the margin of safety, a design change will be implemented during the January-February 1980 outage for items in Category III and IV to add a SIAS to those valves.
This design change will provide an 90030178 16
additional degree of assurance that no release path to the environs exists upon receipt of a SIAS without a concurrent CIAS.
The valves selected to receive SIAS will continue to receive CIAS. The ESFAS-response of the valves will not be affected by the additional actuation sig-nal.
Resetting of the actuation signal will not automatically reset the valves.
Item 2.1.5.a Dedicated H2 Control Penetrations ANO-1 The original ANO-1 license required redundant, safety-grade and dedicated hydrogen purge systems which have been in place since the unit was licensed.
Therefore, no changes are necessary to meet the requirenents of NUREG-0578.
ANO-2 The ANO-2 license is based on redundant, safety-grade, qualified in-contain-ment hydrogen recombiners and a safety grade, dedicated hydrogen purge system.
Therefore, no changes are necessary to meet the requirements of NUREG-0578.
Item 2.1.5.c - Recombiner Procedures ANO-1 The hydrogen purge system at ANO-1 for hydrogen control inside containment was used as the design basis for licensing. The procedures for use of this system have been thoroughly evaluated in light of the NUREG-0578 requirements.
There-fore, our present procedures satisfy the requirements of this item.
ANO-2 The hydrogen recombiners located inside containment for hydrogen control were used as the design basis for licensing. The procedures for use of this system have been thoroughly evaluated in light of the NUREG-0578 requirements.
There-fore, our present procedures satisfy the. requirements of this, item.
Item 2.1.6.a - Systems Integrity for High Radioactivity A40-1 & 2 Arkansas Power and Light has undertaken a program to reduce leakage of all the applicable systems given on page A-26 of NUREG-057E.
The affected systems are being' tested during the January-February 1980 out-ages for leakage using either a volumetric makeup or direct leakage collec-tion / measurement method. An initial leak test is being performed to determine the "As Found" condition. An inspection is being made of the mechanical joints and mechanical interfaces (Flanges, Unions, Valve Bonnets, Packing, etc.).
Identified leakage of 0.01 cc/ min will be repadred/ reduced as practical and a second leak test will be performed.
The initial leak testing and repairs capable of being made during operations have been completed for ANO-2.
All ANO-1 systems will be inspected, tested, and repaired prior to achieving criticality following the current outage.
90030179
--,_,.n 7
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AP&L will perform a review of the leak test results and any identified source of leakage. This review will be aimed at identifying needed improvements in our present Preventative Maintenance program.
In addition, the leak tests will be re-performed at least once every 18 months.
The systems to be tested are:
Dirty Liquid Radwaste Clean Liquid Radwaste Gas Radwaste Makeup & Purification Decay Heat System RB Spray System North Anna Incident Evaluation ANO-1 & 2 As per procedures for Reactor-Turbine Trip (and HPI), the Volume Control Tanks of both Units 1 and 2 remain aligned to a charging pump, and equilibrium make-up is provided to maintain VCT level.
It is possible to isolate the VCT in Unit 2 from the charging pumps and take suction from the Refueling Water Tank (though it is not in the Reactor-Turbine Trip Procedure to do so).
Should the liquid safeties lift at the VCT, the relief path is to one of four Boron Management Holdup Tanks, with a capacity of 51,270 gallons each.
These tanks relieve to one another, and gases are expelled through a pressure control valve to the Gas Collection Header.
Therefore, no liquid or gaseous releases to the reactor auxiliary building atmosphere would be expected.
All tanks and pumps associated with contaminated materials are hard piped to either the Gas Collection Header if they are sources of aerated gases or the Waste Gas System if they are sources of hydrogen gas.
The Waste Gas Systems in both units are currently being investigated for possible design improvements to further ensure that inadvertsnt releases can be avoided. The schedule for completion of the investigation is an-ticipated by March 1,1980.
The results of this investigation and the extent of modification proposed will dictate our completion of hardware changes.
2.1.6.b - Design Review of Plant Shielding of Spaces for Post-Accident Ooerations ANO-1 & 2 Our plant shielding review has been completed. Applicable portions of this report entitled " Design Review of Plant Shielding and Sampling Capabilities in Response to NUREG-0578" are attached.
For accident conditions which assume a Regulatory Guide 1.4 release of fission products, radiation levels through-out the reactor auxiliary building are in excess of General Design Criteria A by a considerable amount.
However, it should be noted that the results of this review are misleading due to the non-existence of a mechanistic sequence of events leading to 100% fuel chadding failure and, therefore, the resulting radiation source term which form the basis of this report.
90030180 18
We intend to administratively control access to certain areas of the reactor auxiliary building based on assessment of the situation through the plant hardware installed.or being installed (i.e., margin to saturation meters, incore thermocouples, high range containment radiation monitors, etc.) as a result of the TMI-2 event and NUREG-0578.
The severity of the situation will be detailed by this instrumentation and will dictate those systems which i
will remain operational following the incident.
For the most severe incident, we envision that only those systems designed for post-accident conditions (i.e., ECCS systems such as High Pressure Safety Injection, Low Pressure Safety Injection and Reactor Building Spray) will be operat-ional, thereby, eliminating a large part of the reactor auxiliary building Which will be affected by any high radiation dose levels which may exist.
Based on the above, a.nd because of changes initiated in plant design and operator training as a result of experiences gained from TMI-2, we do not reasonably expect the original shielding design of the plant to be inadequate.
Therefore, no plant shielding modifications are proposed at this time; how-ever, if our review indicates shielding modifications may be feasible, it shall be considered.
Item 2.1.7.a - Auto Initiation of the Auxiliary Feedwater System ANO-1 & 2 This item is not required to be addressed by the January 2 confirmatory orders.
However, based on your January 18, 1980 letter of our proposed design, we will install the Emergency Feedwater Control System as described in our letters of '
October 31, 1979, and December 18, 1979, prior to startup from the January 1980 outage.
Item 2.1.7.b - Emergency Feedwater Flow Indication
-ANO-1 At the present time, the ANO-1 emergency feedwater system has an orifice plate with one differential pressure transmitter and one control room panel mounted flow indicator for each emergency feedwater line to the two steam generators.
This system will be upgraded, during the January 1980 outage, by adding tv!o (2) differential pressure transmitters powered from separate emergency power supplies.
Four indicators will be located in the control room supplying two indicators for each steam generator.
Sketc h 2.1.7-1 shows a simplified lay-out of the basic equipment.
These changes will meet your single failure criteria, testability, accuracy, and safety-grade power supply requirement.
These changes also meet the long-term safety-grade requirements except for the four (4) panel mounted indicators which will be installed prior to Jan-uary 1,1981.
ANO-2 The ANO-2 emergency feedwater system is designed to meet BTP 10-1 Rev.1, and as such currently meets both short and long-term requirements of NUREG-0578. Therefore, no modifications are necessary.
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2.1.8.a - Improved Post-Accident Sampling ANO-1 & 2 A design and operational review of the reactor coolant and containment atmosphere sampling systems has been performed.
Applicable portions, of a report entitled
" Design Review of Plant Shielding and Sampling Capabilities in Response to NUREG-0578" are attached. As can be seen from this report, when Regulatory Guide 1.4 assumptions are utilized for release of fission products and gases to the coolant, personnel could not obtain samples for r.adiological and chemical analyses without exceeding General Design Criteria 19 guidelines.
However, the results of this review are misleading in that there is a non cristence of a mechanistic sequence of events leading to 100% fuel cladding failure and the resulting radiation source terms which form the basis of this report and prohibit work in the reactor auxiliary building.
Our existing sampling system is undergoing review (to be completed by April 1, 1980) and will be modified to the extent possible to reduce the postulated radiation dose to personnel and still maintain the capability to obtain meaning-ful samples which will totally reflect the continuities of the sampled environ-ment. Varying levels of sample dilution and remote operation are envisioned as possible modes for achieving this end.
This design, however, may not totally reflect dose reduction to GDC 19 levels, but will reflect a reasonable attempt to comply with the NUREG-0578 requirements.
Based on changes made to plant design and operator training which have resulted from the TMI-2 event, we reasonably expect that cladding failure will not exceed the 1% failed fuel criteria to which the plant was initially designed.
- However, changes to the post-accident sampling system will be initiated and completed by January 1,1981, to alleviate potential for postulated employee over exposure during the course of these activities.
i 2.1.8.b - Increased Range of Radiation Monitors ANO-1 & 2 The following responses cover the sections listed under the NUREG-0578 item for radiation monitoring by January 1,1980.
The numbering system below coincides with the requested response format.
1.
Noble Gas Effluents 90030183 a.
System / Method Description i)
The instrumentation to be used will consist of a detector with a remote readout in one of the Control Rooms. The detector (s) will consists of GM and/or Ionization chambers.
The detectors energy response for gaseous radiation will be at least +20 from 60 kev to 7 2 MeV. One or two detectors will~be used as necessary toprovidefor[cceasuring Xe-133 gas concentration from 100 to 103 ;Cl as a minimum.
The readout device' will be capable of accepting the signal generated by the detector (s).
The readout will read in mR or R per unit time. The range of the readout will, as a minimum, correspond to 102 to 103 LCi/ec for Xe-133.
A visual conversion aid to convert dose rate to pCi/cc, such as 21
a graph, chart, etc., will be located at the readout device.
Calibration of the device prior to its operation will be accomplished using a known Cs-137 standard and standard techniques.
The system will be checked against a remotely installed source monthly, and calibrated every three months.
ii) During an emergency, the Reactor Building Purge Systems i
1 and Spent Fuel Ventilation Systems will not be operated.
Also, other systems that may create a release path in the Unit 2 Auxiliary Building Extension (e.g., AN0 Penetration Room Ventilation System) will be isolated.
Therefore, the building release paths will be restricted to the Auxiliary Building Exhausts on Units 1 and 2.
Because a more acces-sible location is needed for particulate and iodine col-lection media, a new isokinetic sampling system will be installed.
Isokinetic sample nozzles are being installed in the Unit 1 Auxiliary Building Exhaust duct and in the Unit 2 Auxiliary Building Exhaust duct.
These lines are routed down into the Spent Fuel area where the iodine and particulate sample media will be located.
The sample lines will be tied together and valved such that a dual set of filters and one pump can be used for either Unit.
A fixed sample volume will be defined by use of shielding.
The detector will also be shielded from interference from back-ground radiation since this will De a low background area.
These design modifications will be completed during the January 1980 outage.
m 90030184 22 E
iii) The readout will be located in either Unit 1 or Unit 2 Control Rooms, iv) The system will provide continuous readout in the Control Room.
v) The system will be powered by vital AC power in the Control Room.
b.
Procedures for conducting all aspects of the measurement /
- analysis, i) Exposure is minimized since the readout is in the Control Room.
ii) Calculations such as those found in the " Reactor Shielding Design Manual" by Rockwell were used to determine the radiation intensities at the detector.
Previous Reactor Building air samples during operation plus the gas activities given in the Unit 1 FSAR show that 96% to 98% of the initial noble gas activity will be Xe-133. Calculations are also being made for periods following an accident when the isotopic mix will change.
in addition, provisions will be in-corporated to allow grab sampling for noble gas so that calculations can be verified or modified if necessary after the isotopic mix' starts to change.
iii) Readout will be monitored at least every fifteen minutes by an individual located in the Control Room.
iv) The system will be calibrated to dose rate from Cs-137 using the calibration sources available at ANO.
Initial calibration may be performed by the vendor.
2.1.d.c - Improved Iodine Instrumentation ANO-1 & 2
. As noted in the November 11, 1979 response to Lessons Learned Task Force Recommenaations, ANO has nine portable air samplers and procedures for obtaining ana performing spectral analysis of the samples, and therefore, presently satisfies the requirement.
Item e - Reactor Coolant System Venting The following are descriptions of the venting systems for ANO-Units 1 & 2 which are proposed for implementation as Category B items.
ANO-1 The hot leg venting system will consist of piping, solenoid operated valves and instrumentation designed to permit the main control room operator to remotely vent the high points of the hotleg piping and the top o(j ft33 0185 r
23
' pressurizer to containment during post-accident conditions. The system will be designed to vent superheated steam, steam water mixtures, water, 0p, fission gases, helium, nitrogen and nydrogen as high as 2500 psia and 670 The system will be safety grade with the same qualifications as were accepted for the RCS at the time of licensing.
Redundant vent paths, each path to consist of two valves in series powered from the same class lE power supply, will be provided at each hot leg vent. The redundant path at each location will be powered from a different class lE power supply. A single safety grade vent path will be added to the pressurizer; the PORV and block valve will provide a redundant vent pathe Each vent will be seismically qualified.
Each vent path will be ' capable of venting a gas volume of at least 1/2 the RCS volume in one hour.
The RCS mass loss from an open vent path will be less than the definition of a LOCA in 10 CFR 50, Appendix A.
To minimize inadvertant vent opening, power will be removed from the valves during normal operation.
Opened / closed indication will be provided in the con-trol room for all power operated valves.
Analyses demonstrating that venting will not result in violation of combustible gas concentration limits and procedural guidelines for the operators' use of the vents will be provided to you by June 1,1980.
These analyses are intended to support your review and approval of our design.
ANO-2 The RCS high point venting system will consist of piping, solenoid operated valves and instrumentation designed to permit the main control roca coerator j
to remotely vent the reactor vessel head and the top of the pressui uer to containment during post-accident conditions.
The system will be designed
.to vent superheated steam, steam water mixtures, water, fission gases, helium, nitrogen and hydrogen as high as 2500 psia and 7000F.
l The system will be safety grade with the same qualifications as were accepted for the RCS at time of licensing.
Redundant vent paths, each path to con-sist of two valves in series powered from the same class lE power supply, will be provided at the reactor vessel head and the top of the pressurizer:
The redundant path at each location will be powered from a different class lE power supply.
Each vent will be seismically qualified.
Each vent path will be capable of venting a gas volume of at least one-half the RCS volume in one h,our.
The RCS mass loss from an open vent path will be less than the definition of a LOCA in 10 CFR 50, Appendix A.
To minimize inadvertant opening, power will be removed from the valves during normal operation.
Opened / closed indication will be provided in the control room for all power operated valves.
Analyses demonstrating that venting will not r,esult in violation of com-bustible gas concentration limits and procedural guidelines for the operators' use of the vents will be provided to you by June 1, 1980.
These analyses are intended to support your review and approval of our design.
l c)D0301%
24
Item 2.2.1.a - Shift Supervisor's Responsibilities AU0-1 & 2 Once per year, the'Vice-President, Generation and Construction, will issue a management directive to the personnel primarily responsible for plant operations and safety, which will emphasize that the primary management responsibility of the shift supervisor is for the safe operation of the plant. This directive will also clearly establish the shift supervisor's command duties under all plant conditions.
The first of such directives has oeen issued for 1980.
Plant procedures are being reviewed and modified, as appropriate, to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined to effect the establish-ment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the Control Room relative to other plant management personnel.
Particular emphasis is placed on the following:
a.
The responsibility and authority of the shift supervisor is to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when on duty in the Control Room.
The idea is reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the Control Room.
b.
The shift supervisor, until properly relieved, will remain in the Control Room at all times during accident situations to direct the activities of Control Room operators.
Persons authorized to relieve the shift supervisor are specified.
c.
If the shift supervisor is temporarily absent from the Control Room during routine operations, the Plant Operator who is the lead control room operator will be designated to assume the Control Room command function.
These temporary dut'ies, re-sponsibilities, and authority are clearly specified.
Training programs for shift supervisors will emphasize and reinforce the responsibility for safe operation and the management function the shift supervisor is to provide for assuring safety.
Ine administrative duties of the shift supervisor have been reviewed by the Director of Generation Operations. Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant will be delegated to other operations personnel not on duty in the Control Room.
Procedures to implement the above have been completed and training provided.
Item 2.2.1.b - Shif t Technical Advisor AND-1 & 2 I
90030187 as O
~
The Shift Technical Advisor (STA) function has been added (as required by NUREG-0578) as a method of improving the plant staff's capabilities to respond to abnormal operating conditions and for improving the evaluation of operating experience.
The Shift Technical Advisor personnel have been selected from our site engineering staff personnel. As of January 1,1980, thirteen graduate engineers have been selected from the normal plant staff.
The present complement of STA's was selected from the plant Engineering and Tech-nical Support Group.
They will serve a 24-hour duty day c7n a rotating basis and will be onsite at all times during their duty. ' Sleeping quarters have been added on the plant site for their use and they will be available to the control room within 10 minutes of being called by the Shift Supervisor.
There are two major functions to be provided by the STA: Accident Assessment and Operating Experience Assessment.
An essential ingredient for the satis-factory performance of these functions is that the efforts of the STAS must be dedicated to concern for the safety of the plant.
In the accident assessment fuction, the STA's duties will be primarily for diagnosis of off-normal events and to advise the Shift Supervisor.
The Shift Supervisor has primary responsibility for the safe operation of his unit and must judge for himself as to the validity of advice given by the STA.
Primary responsibility for the Operating Experience Assessment function at Arkansas Nuclear One will lie with the Plant Performance group. This group consists of three graduate engineers with Nuclear / Mechanical / Electrical backgrounds and having power plant experience.
In addition, reviews of operations and recommendations to improve safety are the responsibility of all Shift Technical Advisors.
In the operating experience assessment function, the STAS will evaluate plant operations, plant design, 6cd operating events from a safety point of view.
Examples of assignm: 9ts which may be used to accomplish this objective include:
(1)
Review of operating and maintenance records to detect unsafe practices, potential equipment failures or reliability problems.
(2)
Review of plant transients to detect needs for procedure or equipment changes.
(3) Review of reportable occurrences from Arkansas Nuclear One and from other plants with similar designs to cetect developing problems.
(4) Review of operating, maintenance, quality control, and sur-veillance testing procedures and practices to detect possible safety problems / improvements.
90030188 26
~
(5)
Discussions with operating and maintenance personnel and evaluation of their comments regarding plant problems po-tentially affecting safety.
(6) Prepare written safety evaluations and recommendations re-sulting from the reviews outlined in (1) through (5).
In order to be able to provide meaningful and accurate evaluations, the STAS should have knowledge and training in at least the fol, lowing general areas: mathematics, reactor physics, chemistry, materials, thermodynamics, fluid mechanics, heat transfer, electrical engineering, instrumentation and controls, as well as experience and training in plant design, control room layout and reactor operations.
Due to the incompleteness of the diverse background desirable to pro-vide the STA functions by most or all of our present STAS, the first year (1980) must be dedicated largely to training.
Emphasis in most cases is needed in specifics of plant design features, plant layout, and reactor operations. h erefore, much of the time and duties of the STAS will be spent in activities closely resembling that for Reactor Operator trainees.
The program, as presently formatted, will encompass three categories of training.
Category A consists of a lecture series and on-the-job train-ing (0JT) in plant structures, systems, component design and layout.
Category B consists of a lecture series and 0JT on the functions and capabilities of Instrument and Control Systems.
It will include train-ing-provided by CE and B&W and simulator training.
Category C consists of a lecture series and 0JT on Plant Response and analyses of transients and accidents.
It also will include training by B&W and CE and simulator training.
Item 2.2.lc - Shift and Relief Turnover Procedures ANO-1 & 2 AP&L has reviewed and revised, as appropriate, plant procedures for shift and relief turnover.
These revised procedures are consistent with the clarification of this recommendation provided at our Regional Meeting.
1.
Procedure (s) have been provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to assure a complete and effective turnover. The following items have'been included in the procedure (s):
a.
Assurance that critical plant parameters are within allow-able limits; b.
Assurance of the availability and proper alighment of all systems essential to the prevention and mitigation of operation permitted by the Technical Specifications.
For such systems and components, the length of time in the degraded mode will be compared with the Technical Specifications action statement; 90030189 27
~
c.
Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications.
2.
Procedure (s) have been provided to assure a complete and effective turnover by the offgoing to the oncoming auxiliary operators and technicians.
These procedure (s) address any equipment under main-tenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient; and 3.
A system has been established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verificationofsystemalignments).
The reviews, peocedural modifications, and training have been completed.
Item 2.2.2.a Control Room Access ANO-1 & 2 AP&L has reviewed the plant procedures for Control Room access.
We have implemented procedures which will limit Control Room access during an emergency. These procedures include the following:
1.
Administrative procedures that establish the duthority and responsibility.
of the person in charge of the Control Room to limit access.
2.
Procedures that establish a clear line of authority and responsibility in the Control Room in the event of an emergency. The line of succes-sion for the person in charge of the Control Room has been established and limited to persons possessing a current senior reactor operator's license.
The plan clearly defines the lines of communication and authority for plant management personnel not in direct command of operaticns, including those who report to stations outside of the Control Room.
These procedures have been implemente,d and training provided.
Item 2.2.2.b - On-Site Technical Support Center (TSC)
ANO-1 & 2 AND has completed the following items per NUREG-0578:
A.
A TSC has been established on the 4th floor of the Plant Administrative Building which is located south of the Turbine Building.
The TSC presently serves as the Planning and Scheduling facility and contains approximately 1000 sq. ft, work space, including a 150 sq. ft office for the Duty Emergency Coordinator.
B.
Procedures have been revised to reflect the engineering / management support and staffing for the TSC. The TSC staff is shown on Figure 2.2.6-1.
90030190 28
C.
Direct communications between the TSC and the Control Room, Near Site Emergency Operations Center and the 14RC has been established.
D.
Portable monitoring equipment has been provided for near the Technical Support Center direct radiation and airborne radioactive contaminants with tarning capability if radiation levels in the TSC are approaching potentially dangerous levels.
The TSC will be evacuated based on rec-ommendations of the Health Physics personnel prior to reaching dangerous radiation levels.
E.
Technical Data (such as layout drawings, P & ID's, electrical schematics, Isometric drawing and Tech Manuals) is available from the plant's Records Management System.
This data is located in the Administrtion Building within one floor o'f the TSC.
Plant parameters will be monitored by CRT display in the TSC from the plant computer.
F.
Accident accessment from the Control Room, should the TSC become un-inhabitable, will be conducted using present procedures.
In the event the TSC becomes uninhabitable, the Duty Emergency Coordinator, Operations Superintendent, Operations & Maintenance Manager, Engineering and Tech-nical Support Manager, Technical Analysis Superintendent, and Plant Analysis Superintendent will evacuate to the Control Room.
The remain-ing members of the On-Site Technical Support Center Engineering / Management staff will evacuate to the flear Site Emergency Operations Center, which is located approximately 0.65 miles north of the plant, with continued direction being provided from the TSC personnel located in the Control Room.
If the TSC becomes uninhabitable, technical data such as plant
' layout drawings, P & ID's and electrical schematics will be relocated to the flear Site Emergency Operations Center.
Item 2.2.2.c - On-Site Operations Support Center Ai40-1 & 2 The On-Site Operations Support Center (OSC) is located in the vicinity of the On-Site Technical Support Center.
Operations Support Personnel will be located in the OSC for response to the control room and/or'TSC needs.
Telephone communications presently exist with the control room and the TSC.
l 4
90030191 29
ATTACHMENT 2 3.0 PLANT SHIELDING FOR POST-ACCIDENT OPERATIONS This section provides an assessment of. potential post-accident shielding problems associated with those systems outside of the containment structure which might contain large radioactive inventories.
The purpose of recommendation 2.1.6.b is to facili-tate post-accident operations using systems that 5ay contain high levels of radioactivity and to ensure that safety equipment in proximity to the resulting radiation fields is not unduly degraded.
Corrective action can consist of design change, additional fixed or portable shielding, post-accident procedure optimization, or equipment upgrading.
For these systems Section 3.1 contains an estimate of potential radiation source terms, Section 3.2 identi-fies the access requirements, and Section 3.3 indicates the post-accident shielding requirements necessary to accommodate these access requirements.
Section 3.4 provides a discussion of poten-tial equipment problems which may result from the high radiation levels.
3.1 SOURCE TERMS In response to the NRR Lessons Learned Task Force Section 2.1.6.b, estimates for source terms and doses have been made for the fol-lowing systems:
Makeup and purification / Chemical and volume control e
Decay heat removal / Residual heat removal e
Reactor building spray e
Safety injection e
e Liquid and gaseous radwaste Sampling o
90030192 3-1 e --
- v-
i
' The initial primary coolant and containment sump fission product inventories are listed on Table 3-1.
These figures reflect the assumptions set forth in subsections 3.1.1 and 3.1.2.
3.1.1 PRIMARY COOLANT ACTIVITY In accordance with Regulatory Guide 1.4, it is assumed that 100%
of all noble gases, 50% of all iodine, and 1% of all remaining radioactive particul.ates are dissolved into the primary coolant.
The components that may be affected by these source terms include:
e Letdown heat exchanger, piping and valves, o
Reactor makeup pumps, piping and valves, o
Decay heat pumps, heat exchangers, piping and valves if operating in a shutdown decay heat remeval mode (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter accident).
3.1.2 CONTAINMENT SUMP ACTIVITY It is assumed that primary coolant entering the containment will release 99.9% of the noble gases to the containment atmosphere; the primary coolant that collects in the containment sump will then contain 0.1% of the noble gases,, and all of the radioactive iodines and particulates.
The components that may be affected by these source terms include:
e Decay heat pumps, heat exchangers, piping and valves (oper-ating in the containment recirculating mode),
Safety injection pumps, piping and valves (operating in o
the containment sump recirculating mode).
e Containment spray pumps, piping and valves.
90030193 3-2
Table 3-1.
Fission Product Inventory Primary Coolant and Containment Sump, Units 1 and 2 Fission Product Inventorv, Curies / Gram Isotone Primary Coolant Containment Sumo Unit 1 Unit 2 Unit 1 7 Unit 2 1311 1.3 x 10-1 1.7 x 10-1 1.3 x 10-1 1.7 x 10-1 2.0 x 10-1 2.6 x 10-1 2.0 x 10-1 2.6 x 10-1 1321 3.0 x 10-1 3.9 x 10-1 3.0 x 10-1 3.9 x 10-1 1331 3.5 x 10-1 4.5 x 10-1 3.5 x 10-1 4.5 x 10-1 1341 2.7 x 10-1 3.5 x 10-1 2.7 x 10-1 3.5 x 10-1 1351 83mKr 2.7 x 10-2 3.5 x 10-2 2.7 x 10-5 3.5 x 10-5 85mKr 8.9 x 10-2 1,1 x 10-1 8.9 x 10-5 1,1 x 10-4 85 r 2.2 x 10-3 2.8 x 10-3 2.2 x 10-6 2.8 x 10-6 K
87 r 1.6 x 10-1 2.1 x 10-1 1.6 x 10-4 2.1 x 10-4 K
88 r 2.5 x 10-1 3.2 x 10-1 2.5 x 10-4 3.2 x 10-4 K
131mXe 2.3 x 10-3 3.0 x 10-3 2.3 x 10-6 3.0 x 10-6 133mXe 1.3 x 10-2 1,7 x 10-2 1.3 x 10-5 1.7 x 10-5 133 e 5.4 x 10-1 6.9 x 10-1 5.4 x 10-4 6.9 x 10-4 X
135mXe 1.4 x 10-1 1.8 x 10-1 1.4 x 10-4 1.8 x 10-4 135 e 1.1 x 10-1 1.4 x 10-1 1.1 x 10-4 1.4 x 10-4 X
5.4 x 10-1 6.9 x 10-1 5.4 x 10-4 6.9 x 10-4 138Xe Solids 1.5 x 10-1 1.9 x 10-1 1.5 x 10-1 1.9 x 10-1 Based on a Unit 1 reactor coolant mass of 5.35 x 105 lb grams) a Unit 2 reactor coolant mass of 4.619 x 105 (2.43 x 108 lb (2.1 x 108 grams), and a Unit 2/ Unit 1 power ratio of 1.11.
j of the total reactor core' fission product inventory, the primary coolant is assumed to contain 50% of the iodines, 100% of the noble gases, and 1% of the particulates.
The containment sump liquid is assumed to contain 50% of the, iodines, 0.1% of the noble gases, and 1% of the particulates.
90030194 3-3
0 t
3.1.3 DOSE RATE CALCULATIONAL BASIS The dose rate as a function of time, shown on Figure 3-1, was cal-culated using the CYLDOSE code.
CYLDOSE calculates the linear attenuation, scatter buildup, and resulting tissue dose rate from a cylindrical gamma radiation source.
Multiple source materials and' shield materials may be specified.
Dose points may be se-lected anywhere along the side of the source or at its end; a line source approximation is used for dose points along the side, whereas a truncated cone source approximation is used for dose points at the end.
For convenience of calculation the gamma energy emitted by the source (s) is divided into groups and each group is designated by a number and a group average energy.
Cal-culations may be done considering one or a combination of these groups.
Source strengths associated with these energy groups may be read into the code as data or calculated by the code.
3.2 AREAS AND COMPONENTS REQUIRING ACCESS FOR POST-ACCIDENT OPERATIONS In developing the system and component access requirements, par-ticular attention was directed at the systems for which the shielding evaluation was performed.
Additional components within the Auxiliary Building that are likely to require some access are tabulated in Subsection 3.2.7.
3.2.1 UNIT 1 REACTOR COOLANT SYSTEM SAMPLING EQUIPMENT 3.2.1.1 Reactor Coolant Sameline The valves used to draw a sample of the reactor coolant are all located near the sample sink; a motor operated valve, CV-1814, is controlled from control room panel C-lS.
All other valves are j
manually operated.
i 90030195 3-4
+
FIGURE 3-1(a)
RELATIVE DOSE RATE AS A FUNCTION OF TIME O - 30 DAYS UNIT 1/ UNIT 2 100
~
10 1,L w
2=
8 s
5 302 SUMP WATER a
PRIMARY COOLANT 10 3 1 DAY 1 WEEK 1 MONTH 104 O
100 200 300 400 500 600 700 TIME AFTER SHUTDOWN (hours) l 90030196 3-5
FIGURE 3-1(b)
RELATIVE DOSE RATE AS A FUNCTION OF TIME O-24 HOURS
' UNIT 1/ UNIT 2 100 i
i i
1 SUMP WATER s
8 8
10'l -
g P
5 E
PRIMARY COOLANT 1
I I
'I I
10 2 O
5 10 15 20 25 l
90030197 TIME AFTER SHUTDOWN (hours) 3-6
+
3.2.1.2 RCS Decay Heat Removal System Samplina Valves SS-41A SS-41B 3.2.2 DECAY HEAT REMOVAL OPERATIONS 3.2.2.1 Decav Heat Durina Cooldown Valves Other Ecuiement BW-8A Pump P34A DH-1A Pump P34B DH-15 DH-19 BW-8B DH-1B 3.2.2.2 Decay Heat Coolant Purification i
Valves MU-7 MU-5 MU-6 ABD-15 ABV-3 DH-4B DH-5B DH-6 (Throttling)
DH-4A DH-5A 1
3-7 90030198 1
l e
3.2.3 LOSS OF COOLANT /RC PRESSURE OPERATIONS 3.2.3.1 Ruoture Greater Than HPI Capacity Valves Other Equipment MU-13 Purge Dampers (actuate from ventilation control panel)
DH-7A Isolate DH rooms by closing watert'ight doors DH-7B MU-14 MU-15 MU-16 MU-17 ABS-13 ABS-14 3.2.3.2 Rupture Within HPI Capacity Valves Other Equirment MU-13 Same as 3.2.3.1 DH-7A DH-7B MU-14 MU-15 MU-16 MU-17 MU-23 MU-24 MU-25 MU-26 ABS-13 ABS-14 MU-21A, B&C 9
3-8
3.2.4 UNIT 2 REACTOR COOLANT' SAMPLING 3.2.4.1 Reactor Coolant HPI Lec for Licuid Valves Other Eculement 2PS55 Sample sink 2PS56 2PCV5922 2PS59 2PS57 3.2.4.2 Isolation of 2T120 for Total Gas Analysis Valves Other Ecuionent Valves listed in 3.2.4.1 Sample sink 2PS93B 2PS93A 2PS91 3.2.4.3 Samoline of the Shutdown Cooline System Valves 2BS18A 2BS18B 2PS62 2PCV5923 2PS63
'90030200 3_,
- '3.2.5 SHUTDOWN COOLING SYSTEM OPERATIONS 3.2.5.1 Initiation of Shutdown'Coolina Valves 2SI-1A 2SI-1B 2SI-2A 2SI-2B 2SI-4A 2SI-4B 2SI-5A 2SI-5B 2SI-20 (Throttle) 3.2.5.2 SDC Purification Valves 2CVC-147 2CVC-146 2SI-35 2SI-34 3.2.6 UNIT 2 LOSS OF REACTOR COOLANT OPERATIONS n the event of a rupture greater than the capacity of the high pressure injection pump, post-LOCA containment radiation menitor-ing will be required in accordance with Appendix A of OP 2202.06.
In addition, ECCS pump rooms must be isolated by closing the water-tight doors and SDC pump room floor drains 2 ABS-5 and 2 ABS-6.
Case II (rupture within HPI pump capacity) also requires post-LOCA containment radiation monitoring.
9003020I 3-10
l l
' Valve operations listed in OP 2102.10, Plant Shutdown and Cooldown
~
will also have to be performed during the loss of reactor coolant event, in accordance with procedure OP 2202.06.
3.2.7 Based on experience at TMI, the following areas and/or ccm-i ponents may also require post-accident access:
o Control Room Liquid Radwaste Control Panel / Equipment e
Gaseous Radwaste Control Panel / Equipment e
Solid Radwaste System Control Panel / Equipment e
Auxiliary Feedwater Pumps / Valves e
Auxiliary Shutdown Panel (s)'
e Component Cooling Water Pumps e
Auxiliary Building Ventilation System (fans, filters, e
adsorbers) e Control Room HVAC Systems Cable Spreading Room e
Electrical Equipment /Switchgear Rooms o
e Diesel Generators 3.3 ACCIDENT SCENARIOS - UNITS 1 AND 2 The dose contributions from the subject systems were calculated for several different accident scenarios.
Depending on the sce-nario and operating mode, components in these systems will contain varying amounts of activity.
By examining each scenario, the associated dose rates and attendant operational problems, the need fcr modification ~of equipment, shielding and/or operating proce-dures can be identified.
3.3.1 FUEL FAILURE EVENT WITHOUT COINCIDENT LOCA I
In this condition, it is assumed that the primary coolant will attain the activity level specified on Table 3-1.
Since a LOCA condition does not exist, no containment isolation signal is gc30202 1
3-11
generated (assuming the absence of any other isolation signal).
Consequently, primary coolant activity would be introduced into the Makeup and Purification System (Unit 1) or Chemical and Volume Control System (Unit 2), including the following components:
o Reactor makeup pumps / charging pumps e
Purification demineralizers e
Reactor coolant filters Seal water cooling system e
e Letdown coolers Associated valves and piping e
(Note:
It is recognized that Unit 2 has provisions for radiation monitoring equipment to sample and monitor the activity in the reactor coolant upstream of the reactor coolant filters.
A high-activity signal from this monitoring equipment will alarm in the control room.
This same equipment could, on a high activity alarm, provide a signal to isolate the letdown line.
It is not clear whether this automatic isolation feature presently exists.
Unit 1, at least, does not appear to have this feature.)
In this acde, the makeup and purification system /CVCS can continue
'ro operate for purposes of makeup, letdcwn, and their associated functions.
By examining Tables 3-2 and 2-3, it is noted that the makeup and letdown loops are responsible for major dose contribu-tions within the auxiliary building.
The attendant dose problems impact most of the analyzed access points (refer to dose point numbers on Tables 3-2 and 3-3; these numbers correspond to the locations marked on Figures 3-3 through 3-12).
The initial dose rates will decrease as a function of time in accordance with the curves shown on Figures 3-1(a) and (b).
gog30203 3-12
3.3.2 FUEL FAILURE COINCIDENT WITH LOSS OF COOLANT In the event of a loss of ccolant, engineered safety features will initiate safety injection and containment isclation.
As a result of automatic containment isolation, the flow of primary coolant to the makeup and purification system will be terminated.
The introduction of containment sump water into these systems is ex-pected to cause the following problems:
Unit 1 Operation in the containment sump recirculation mode e
will make decay heat valves DH-1A and DH-1B inaccessible for manual realignment if required.
As noted on Table 4
3-2, dose rates range up to 2 x 10 R/hr, depending on I
location relative to the recirculation equipment.
gg030204 3-14
o With containment sump recirculation in operation, the liquid and gaseous radwaste system control panels will be inaccessible.
High dose rates will be encountered both at the panels and along the access routes to the panels.
Unit 2 Primary coolant and containment sump wat'er activity levels e
will cause high doses in sump recirculation / decay heat valve operating areas; manually-operated valves that will be inaccessible include:
2SI4A, 2SI43, 2SIlA, 2SIl3, 2SI5A, and 2SISB.
These valves would normally require alignment when initiating decay heat removal system opera-tion.
3.3.3 FUEL FAILURE WITH COINCIDENT LOCA; LETDOWN IN OPERATION This scenario is identical to that discussed.in subsection 3.3.2, with the additional condition that primary coolant.is being brought out of containment through the letdown loop.
The contain-ment isolation / loss of letdown discussed in 3.3.2 above would present some operational difficulties by impeding the use of some functions (e.g.,
degasificaton, boron control); thus, it may be desirable to utilize the letdown feature.
The radiation hazards and problems that will be encountered are essentially a combina-tion of the disadvantages of the first two scenarios discussed.
Table 3-2 indicates that the worst case dose figures could apply ih this operational mode.
It should be further noted that these figures only reflect the " shine" doses from ecuipment, and do not consider possible airborne (immersion) doses.
If Figure 3-2 (effect of lead shielding) is examined in light of these calculated doses, it is readily seen that in order to reduce doses to acceptable levels (100mr/hr), substantial thicknesses of c;gg30205 3-15
FIGURE 3-2 EFFECT OF LEAD SHIELDING 100 10 1 LEAD SHIELD
/
.SdURCE /
' DOSE POINT 10 2 (At Surface of Lead)
E
- %18" Concrete Wall u.
z O
C SHIELDING CONFIGURATION W=
C 10 3 Oc 10'4 10 5 I
I I
I 0
2 4
6 8
10 h
LEAD THICKNESS (inches)
)
\\
3-16
' lead shielding will be required.
In addition to the obvious prob-lems of physical installation, the possible use of such quantities of lead shielding raises additional questions regarding the struc-tural adequacy of the buildings (to support the additional loads).
3.4 POTENTIAL EQUIPMENT PROBLEMS The integrated dose received by safety-related instrumentation should not be sufficient to result in any degradation in perform-ance, because the equipment that is expected to perform under such extreme accident conditions is generally qualified for service in high radiation fields.
Because of their sensitivity to ionization energy, electronic devices are not' normally located in areas which may be subjected to high radiation fields; however, this must be verified.
The following items are prone to deterioration when subjected to significant ionization energy:
e Rubber, Teflon, and plastic components e
Glass (transmittance degradation) e Electrical components (insulation and solid state devices)
Using Regulatory Guide 1.4 assumptions, the integrated 30 day dose increase facter to any component over the actual normal operation integrated dose received by that component, is estimated to be 1.4 x 104 For ccmponents in extremely high radiation areas, it is assumed that under normal operating conditions that doses received by the ccmponents would be no greater tha 0.1, rad /hr.
Therefore, the conservatively estimated 30 day dose recieved by the same equip-ment, based on Regulatory Guide 1.4 assumptions, is less that 1.0 x 106 rads.
gQ30 -
3-17
Electronic equipment employing solid-state devices is subject to i
fallure under these dose conditions; in addition, the high radia-tion field will cause device performance characteristics to change regardless of the total integrated exposure.
It is recommended, therefore, that a survey be made to establish the locations of electronic equipment with respect to (potential) high radiation areas.
1 I
ego 3-18
Table 3-2.
Unit 1 Post-Accident Dose Rates at Selected Locations
~
i-Initial Dose Dose Point Description Immediately Following Accident 1
Adjacent to primary coolant. filter 6500 R/hr; attributable to letdown and containment air sampling line source term station El. 335-354 2a Adjacent to primary coolant filter 400 R/hr; attributable to letdown and makeup pump P360C (pump running) line source term El. 335-354 2b Corridor adjacent to makeup pump 100 R/hr P36C area (pump running)
El. 335-354 i
3 At elevator El. 335-354 0.2 R/hr from letdown line/ makeup y
pump, plus 20 R/'ar from recircu-lation system at El. 317.
1 R/hr r
in decay heat mode, assuming initial 24-hour decay.
4 Demineralizer valve operating area 23 R/hr 5 '
Makeup pumps valve operating area 640 R/hr, excluding gaseous and liquid waste lines 6
a.
Corridor adjacent to makeup 1 to 600 R/hr, depending on loca-pumps tion relative to opening to valve operating area b.
Inside makeup pump cubicle 500,000 R/hr ao cp 7.
a.
Corridor adjacent to reactor 7,200 R/hr, assuming 50% liquid, C],
coolant makeup tank 50% gas; approximately 6 times L
normal gaseous activity due to
[3 noble gas buildup.
The activity levels will be produced only if c;)
O letdown is used (i.e.,
activity can be minimized if tank is isolated).
7 Table 3-2.
Unit 1 Post-Accident Dose Rater at Sclected Locations (Continued)
Initial Dose Dose Point Description (Immediately Following Accident) 8a Outside door of "A"
train decay 1 x 104.R/hr, from containment sump heat /LPSI recirculation recirculation operation system, near unshielded recirculation line No. GBC-3-4"-
8b Same location, except operating 40 R/hr.
in decay heat mode only 8c Same location, containment spray Less than 1 R/hr.
in location 8d Same location, with unshielded rc 70 R/hr.
circulation line No. GCB-3-4 y
excluded to 9a Outside door of "B"
train decay 1500 R/hr.
Of this figure, 1300 heat LPSI recirculation R/hr is due to the exposed "B"
system; containment sump recir-train recirculation line No.
culation and containment spray GCB-3-4" is due to the "A"
systems operating train recirculation 9b-Same location, except operating 1 R/hr in-decay heat mode.
.10a Vicinity of decay heat, LPSI, 900 R/hr, with containment sump O
recirculation equipment recirculation and containment CD spray operating; 0.3 R/hr with jj$
decay heat operating only CD pa 10b Near sta irway No.1, El. 317-335, 30 1700 R/hr in containment sump ft. away from unshielded recircu-recirculation mode; less than CD -
lation line GCB-3-4" 0.3 R/hr in decay heat mode only.
10c Inside containment recirculation In excess of 20,000 R/hr., after equipment cubicle, at decay heat 24-hr. decay.
Recirculation and gq[&
pump manual valves.
containment spray-in operation.
1 i J
l.
l l
l
~.
l Table 3-2.
Unit 1 Post-Accident Dose Rates at Selected Locations (Continued)
Initial Dose' Dose Point Description (Immediately Following Accident) 11
' Control room, above makeup
'15.4 R/hr from makeup tank, 7.3 R/hr tank T-4 from vacuum degasifier, and an ad-ditional 1.4 R/hr from line IICC,
l 3"
(assuming that letdown is in operation).
12 Control room, in area of main 5.0 x 10-4 R/hr from makeup tank, control boards 1.8 x 10-4 R/hr from vacuum de-gasifier (assuming that letdown i
is in operation).
13 Diesel generator area 17-20 R/hr, from makeup tank.
An additional 81 R/h,r from vacuum
~
y degasifier, if this equipment u
is in operation.
14 Proposed. remote sample station 7 mR/hr, from makeup tank location (refer to Section 4) 15 Vacuum degasification tank valve 160 R/hr from degasification room equipment, if operating.
16 Vicinity of waste gas surge tank 16 R/hr from tank T-17; 14,000 T-17 R/hr from line HRC-2-2";
8,500 R/hr from line HRC-2-2.5";
520 R/hr from lins'HSC-1-3" co 17 Vicinity of tank T-12D 290 R/hr from tank T-12D; 1,700 cp R/hr from line HSC-1-3";
cd 1,700 R/hr from valves in the t/2 adjacent v'alve operator corridor.
CD r0
,~.
18 Vicinity of filter F-16.
29 R/hr, from filter.
=
Table 3-2.
Unit 1 Post-Accident Dose Rates at Selected Locations (Continued)
Initial Dose (Immediately Following Accident)
Dose Point Description 19 Vicinity of gas decay tank 53 R/hr from Tank T-18C; 2,800 R/hr from line T-18C HRC-2-2.5"; 6,400 R/hr from line HSC-1-3".
20 Outside shield wall, vicinity 20,000 R/hr from line HSC-1-3";
of tank T-18C 7,300 R/hr from line H'RC-2-2.5";
1 R/hr from-tank T-18C.
From 1200 R/hr. at stairway to 3000 Liquid Waste Control Panel R/hr. walking to panel; 100 R/hr.
at panel during containment sump recirculation mode.
Gaseous waste control panel 1 R/hr. at panel when operating in decay heat mode only.
1700 R/hr. at panel when operating containment spray and contain ment sump recirculation.
1 R/hr. at panel when operating in decay heat mode only.
9
- I e
C2 C)
C2 N
_ _ =.
~
Table 3-3.
Unit 2 Post-Accident Dose Rates at Selected Locations Initial Dose Dose Point Description Immediately Following Accident 30 Control room, above volume control 300 R/hr from volume control tank tank (VCT).
31 Control room, near main control 1.5 x 10-4 R/hr, primarily from boards the VCT.
32 Decontamination room adjacent to 10 R/hr, from the vacuum degasi-control room fier tank, if degasification equipment is operating.
33 Electrical equipment room adjacent 1820 R/hr, from VCT.
to cable spreading room 34 Access corridor, elevation 372',
750-800 R/hr, from VCT.
w between diesel generators and E
cable spreading room.
w
^
35 Stairway.eas.t of diesel generators 1.8 mR/hr, from VCT.
(elevation 372').
36 Diesel generator room.
2.4 R/hr, from VCT.
37 Stairway entrance near hot machine 0.3 R/hr from VCT; 50 R/hr from shop (el eva t ion 354').
the vacuum degasification CD
~
equipment, if operating.
CJja 38 Sample room, elevation 354'.
0.2 R/hr from VCT;~2 R/hr from the c7 vacuum degasification equipment, rs) if operating.
td 39 Corridor adjacent to VCT, 1,820 R/hr, from VCT; 500 R/hr from c]evation 354'.
degasification equipment, as noted above.
40 Stairway leading to south 900 R/hr from VCT; 1 R/hr from degasi-3 piping penetration area.
fication equipment, and 160 R/hr from the seal water heat exchanger, if operating.
b
Table 3-3.
Unit 2 Post-Accident Dose Rates at Selected Locations (Continued)
Initial Dose Dose Point Description Immediately Following Accident l
41 East of letdown heat exchanger, 10 R/hr from heat exchanger; 1000 R/hr l
outside shield wall.
from the "B"
train shutdown cooling.
equipment on Elevation 317 below.
42 Wall adjacent to charging pump 65 R/hr from charging pump "B",
if 2P36B on elevation 335.
running.
The "B"
train shutdown cooling equipment contributes the same dose as at point 41.
43 a.
Corridor at entrance to 110 R/hr (55 R/hr per pump), plus an charging pumps 2P36B and additional localized dose of 1000-2P36C; both pumps running.
2,000 R/hr from the unshielded " win-dow" to lines 2HCD-2-4" and 2HCD-63-3".
w I
b.
Passageway to pump 2P36A 1,000 R/hr from line 2HCD-63-3", plus and tank room 2054 on ele-54 R/hr from pump 2P36A.
vation 335.
c.
At stairway entrance, ele-4-5 R/hr.
vation 335.
44 Elevation 317, in valve corridor 160,000 R/hr, from containment con ta in ing two shutdown cool-spray system; 80,000 R/hr after ing manual valves.
24-hour decay.
(Shutdown cooling will not be initiated until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has elapsed.)
ao 45 Inside area containing train "B"
106 R/hr.
c7 cp shutdown cooling, containment tr spray, HPSI/LPSI equipment.
a G2
{jf a.
Outside steel door to area de-7,000 R/hr.
scribed for point 45.
2 yr
)
.x b.
Outside shield wall enclosing 300 R/hr.
s above area.
J (2
' s@
t
i.]
Table 3-3.
Unit 2 Post-Accident Dose Rates at Selected Locations (Continued)
Initial Dose Dose Point Description Immediately Following Accident i
47 Same as 46(a) and (b), except train See doses under 46(a) and (b).
I "A" equipment considered.
48 Door outside IIPSI pump 2P89C area.
150 R/hr.
49 Entrance to stairway at elevation 20 R/hr.
317.
l 50 Unshielded letd,own line 2HCD-63-3" 9,000 R/hr.
at elevation 335.
6 i
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t................
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ATTACHMENT 3 gu j 4.0 IMPROVED POST-ACCIDENT SAMPLING' CAPABILITY l 4.1 EXISTING PRIMARY COOLANT SAMPLING SYSTEMS 4.1.1 UNIT 1 The present sampling station for Unit 1 is located on elevation 354 adjacent to the reactor coolant makeup tank and waste gas compressors. This sa$pling station permits the taking of reactor coolant samples from the letdown line, pres 5,orizer water space, pressurizer steam space, and decay heat cooler outlet. 4.1.2 UNIT 2 The present sampling station for Unit 2 is also located on eleva-tion 354. Unlike Unit 1, this sampling station is farther away from the reactor coolant makeup tank but is also adjacent to the waste gas compressors. This station permits the taking of reactor coolant samples from a reactor coolant hot leg, pressurizer surge line, pressurizer steam space, decay heat removal / safety injection system, and letdown / purification system. 4.1.3 COMMON FEATURES OF UNITS 1 AND 2 Because the sampling station of Unit'2 is adjacent to the sampling station of Unit 1 and connected by a. passageway, the count room and chemical lab in Unit 1 also serves Unit 2. 4.2 REQUIREMENTS FOR SAMPLING: PRE-AND POST-ACCIDENT 4.2.1 NORMAL SAMPLING Primary coolant sampling is performed on a regular basis; sampling pro.vides a feedback control feature essential for long-term reac-tor operation. Based on analysis of the sampled primary coolant, action can be taken to adjust boron levels and water chemistry. 90030234 ~ 4-1 e
In addition, the integrity of the reactor core can be assessed by measuring therpresence of fission products in the primary coolant. Those items that must be determined by sampling and the justifica-tion for each are listed below: Boron concentration .- part of the reactivi,ty control and safe shutdown margin' Hydrogen concentration - hydrogen is required to keep the oxygen concentration to a minimum ~ Oxygen concentration - high oxygen concentration can lead to corrosion and crud buildup on the core and core internals Chlorides - the presence of chlorides should be kept to a minimum to prevent em-brittlement and cracking of metals Radionuclices (noble - their presence in the reactor cool-gases, iodine, and ant indicates a degree of fuel-rod cesium) failure 4.2.2 POST-ACCIDENT PRIMARY COOLANT SAMPLII;G Following'an accident, primary coolant sampling will be required to provide the following data: Boron concentration - to ascertain the safe shutdown reactivity margin; sample should be taken within 1 hour after accident 90030235 4-2 4
Radionuclide analyses - to determine degree of core damage; (Noble gases,' Iodine, sample should be taken within Cesium and other non-2 hours of accident volatile fission pro-ducts Chlorides - NRC requirement; sam le should be taken within or.e shif t 4.3 SAMPLING PROBLEMS ASSOCIATED WITH HIGH-ACTIVITY REACTOR COOLANT-4.3.1 HAZARDOUS DOSES TO PERSONNEL There are essentially two major dose-related problems to solve in order to sample.the reactor coolant with high activity levels resulting from a fuel-failure event. The first problem is that (in Unit 1) dose levels (from the reactor coolant makeup tanks) at the sampling stations and in the access corridors of the aux-iliary building can preclude access to the Unit 1 sampling station. In Unit 2, the dose levels are considerably lower; access to and occupancy of the Unit 2 sampling stations is possible for short periods of time. Any reactor coolant leakage into the auxiliary building (e.g., from the reactor coolant makeup pump seals) will result in airborne activity that will 'cause large whole-body (im-mer.sion) doses. The second major problem is that the present sampling equipment and reactor coolant sample itself will cause hazardous doses on the order of 2,000 to 4,000 rem /hr at 1 meter away from the equip-ment. The airborne activity that could result from attempting to take a sample would also cause an extremely hazardous condition. 90030236 4-3
'4.3.2 POST-ACCIDENT DOSES AT PRIMARY COOLANT SAMPLE STATION In the event of a loss-of-coolant accident, automatic containment isolation will cause isolation of the letdown line. This isola-tion is desirable in that it will prevent high-activity reactor coolant from entering the makeup and purification system (i.e., the regenerative heat exchanger, reactor coolant fil!ters, deminer-alizers, makeup tank, and makeup pumps). Unless there is leakage past the isolation valves, the total dose to the existing sampling stations from the adjacent reactor makeup tank and waste gas comp-ressors would be essentially the same as before the accident. Un-er these conditions, it would be possible to get to the sampling station, but the doses from the sample and sampling equipment would still be hazardous. If, however, post-accident reactor coolant is processed by the letdown system (which may be required for degassing operations or for boron control), then the doses from the reactor makeup tank in the corridors and sampling room (Unit 1) would be extremely hazardous, as shown on Table 3-2. The doses at the Unit 2 sampling stations are significantly less.haz-ardous, as shown on Table 3-3. 4.3.3 PRIMARY COOLANT SAMPLING LOCATIONS 4.3.3.1 Unit 1 Unit 1 has three reactor coolant sampling locations that draw reactor coolant into the sampling station. Samples are normally taken from the letdown system just upstream of the domineralizer. Because the letdown system is automatically isolated on a safety injection signal, it will not be possible to take a sample from this point.unless the letdown flow is resumed. However, with the letdown system isolated, samples can still be taken from the pressurizer liquid, steam space, and decay heat cooler sample lines if modifications are made. Because molecular diffusion will tend to keep concentrations of dissolved chemicals 4-4 90030237 i i
uniform throughout the reactor coolant' system, the steady-state concentrations of boron, nuclide's, and other~ chemicals in the pressurizer liquid leg should be essentially the same as the steady-state concentration in the main reactor' coolant loops. Therefore, by allowing time to pass for concentration gradients to equalize, samples extracted from the pressurizer liquid phase should be representative of the reactor coolant that would nor- ~ mally be sampled from the letdown system. The pressurizer liquid sample is brought via 3/8-inch tubing into the sampling station where it is first cooled by the~ shielded pres-surizer sample cooler. It is estimated that the personnel dose from this shielded cooler is approximately 600 to 1,000 rem /hr (in addition to any other contributing doses). Therefore, it is con-cluded that even with the letdown leg isolated (which would allow personnel to enter the sampling station without being subjected to hazardous doses from the reactor makeup tank), the direct shine doses from the pressurizer sample cooler would be very hazardous, in addition to the doses that would be received from the sample specimen (bomb). It is recognizec that the pressurizer liquid sample is not the most desirable sample location for radionuclides because the sam-ple does not come from the main reactor coolant loop.
- However, short of making any modifications to the Unit 1 reactor coolant piping (e.g., providing for a sample point in the hot leg), the pressurizer liquid sample is the most attractive alternative.
4.3.3.2 Unit 2 Unit 2 has a greater number of locations where reactor coolant can be drawn for samples. These locations are the reactor coolant letdown system, reactor coolant hot leg, pressurizer surge line, pressurizer steam space, and safety injection / shutdown cooling system. All of these samples are taken in the Unit 2 sampling room. As discussed above, it is possible to occupy the Unit 2 l 90030238 4-s l l
. sampling ro,om for short periods of time. However, the' same haz-ards encountered with the sample cooler and sample specimen in Unit 1 will also be present in Unit 2. In addition to the pres-surizer liquid sample, Unit 2 is provided with the capability for taking samples from one of the reactor coolant system hot legs and from the safety injection / shutdown cooling system. ? 4-6 90030239
letdown system will be automatically isolated if an acci-dent occurs. However,*even though letdown flow is termi-nated,'the boronmeter should still provide a reading that is indicative of the reactor coolant boron concentration before any borated water is injected into the reactor coolant system. This information can provide important data for prediction of the shutdown margini 1 j 4.5 CONTAINMENT AIR SAMPLING 4.5.l CONTAINMENT AIR SAMPLING EQUIPMENT 4.5.1.1 Radionuclide Analysis The present containment air sampling equipment for Unit 1 and 2 draws containment air through a particulate paper and charcoal cartridge for a specific time period. The cartridge then removed and is taken to a count room, where the accumulated activity is counted, both on the filter paper (for particulate) and on the charcoal (for iodine). 4.5.1.2 Hydrogen Monitoring In the event of a loss-of-coolant accident, the reactor core could reach temperatures leading to a Zircaloy-water reaction, and hy-drogen gas could be released to the containment (in addition to the normal hydrogen that would be liberated from the primary cool- . ant). It is desirable to monitor the containment for hydrogen so that the hydrogen recombiners can be adequately employed to main-tain the hydrogen concentration at safe levels, and to provide indication of a possible Zircaloy-water reaction. 4 b 4-15 90030240
) '4.5.2 UNIT 1. CONTAINMENT AIR SRMPLING STATION The Unit 1' containment air sampling station is on elevation 335 (in a radiation Class IV area) near the containment wall, opposite the shield walls for the reactor coolant filters. Approximately 30 4 feet from the sampling station in the same room is the shielded i seal water injection filter and other decay heat an$ high pressure injection piping. The estimated maximum dose received in attempt-ing to reach the sampling station is 6500 R/hr in the walkway corridor (from the reactor coolant filter) plus an additional dose of approximately 1000 R/hr from the seal water injection filter and other nearby piping. If the containment air sampling equip-ment is moved approximately ten feet to the other side of the adjacent 2.5-ft.-thick shield wall (separating this area from the ~ adjacent radiation Class II zone), then the dose at this new sam-pling location would be reduced to about 6.5 R/hr, which is pri-marily due to the reactor coolant filter. The present nuclide sampling apparatus is as described in subsec-tion 4.5.1.1. However, based on information supplied to date, hydrogen monitorf.ng is performed by physically obtaining a con-tainment air sample in an evacuated bottle (approximately 100cc in volume), using sample equipment at the same location; the bot-t1e is then taken to a laboratory for analysis. It is estimated that the dose from a 100cc glass bottle (contain-ing a containment air sample taken immediately af ter the accident) at 0.1 meter is approximately 116 R/hr; 1.10 R/hr at 1.0 meter. Two hours after the accident, this same sample would give a dose at 0.1 meter of approximately 20 R/hr. 4.5.3 UNIT 2 CONTAINMENT AIR SAMPLING STATIONS Unit 2 has two separate containment air sampling stations. One station is a continous air monitor located on elevation 360 in piping penetration room 2084, adjacent to the shielded seal water U 4-16 90030241
.o heat exchanger and also decay heat and high/ low pressure injection piping. The dose from this equipment is estimated to be on the order of 5,000 to 10,000 R/hr. The other containment air sampler is a bottle sampler located near i the spiral staircase in containment penetration area 2081, on ele-vation 356. This area is shielded from equipment and piping con- ) taining reactor coolant so that shine doses are not significant and personnel can enter this area to take a sample. However, the dose received from the sample bottle itself will be of concern; the dose from a 100cc bottle is approximately the same as that identified in subsection 4.5.2. Unit 2 also has containment air hydrogen analyzers located in the reactor coolant sampling room. Fortunately, the Unit 2 sampling room is sufficiently shielded from the volume control tank to allow access to the sampling station from several different en-trances. The dose received at the stairway entrance leading to the sampling room is on the order of 3 R/hr (from the VCT) but approximately 50 R/hr from the vacuum degasification tank, if the degasification system is processing reaci:or coolant. Within the sample room, the doses from the volume control tank and vacuum degasification tank (if operating) are 0.2 R/hr and 2 R/hr, respec-tively. These dose levels may be tolerate.d for a short period of time while obtaining a hydrogen analysis of the containment air. Doses received from the containment hydrogen analyzer itself have not been estimated. 90030242 o 4-17 , - -,}}