ML20210P446
| ML20210P446 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/31/1997 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20210P444 | List: |
| References | |
| NUDOCS 9708270138 | |
| Download: ML20210P446 (415) | |
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{{#Wiki_filter:._ _ _ - 1 1 2 l PDMS SAR UPDATE 2 INSTRUCTIONS Attached is Update 2 to the PDMS Safety Analysis Report (SAR) for Three Mile Island Unit 2.
- Please remove all pages of Update 1 of the PDMS SAR from the PDMS SAR notebook and replace with the enclosed copy of the PDMS SAR, Update 2.
Please contact John Schork, TMI Licensing and Regulatory Affairs at (717) 948-8832 if you have cny questions regarding these instructions. 9 6 9 O 9708270138 970818 ~ PDR ADOCK 05000320 P PDR
v - O TMI-2 POST-DEFUELING MONITORED STORAGE O SAFETY ANALYSIS REPORT AUGU T 99 0 .
TABLE OF CONTENTS CHAPTER TITLE 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 2 SITE CIIARACTERISTICS 3 DESIGN CRITERIA - STRUCTURES, SYSTEMS AND COMPONENTS 4 FUEL 5 PDMS RADIOLOGICAL CONDITIONS 6 DEACTIVATED SYSTEMS AND FACILITIES O 7 OPERATIONAL SYSTEMS AND FACILITIES 8 ROUTINE AND UNANTICIPATED RELEASES 9 DELETED 10 ADMINISTRATWE FUNCTIONS 4 i UPDATE 2 - AUGUST 1997
O . CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLAhT O O - - - - s ,,,,,
. - _ . . ~ _ - . - - .. .....- . . _ - _-- - . .- . - -
I ^ CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
-- TABLE OF CONTENTS SECTION TITI E PAGE
1.1 INTRODUCTION
1.1-1 1.1.1 POST-DEFUELING MONITORED STORAGE 1.1-1 1.1.2 TRANSITION TO POST-DEFUELING MONITORED 1.1 2 STORAGE 1.1.2.1 Prerequisites for PDMS 1.1 2 1.1.2.2 Transition Activities 1.1-3 1.1.3 APPLICABLE REGULATIONS 1.1-5 1.1.4 SAFETY RELATED STRUCTURES, SYSTEMS, AND 1.1 5 COMPONENTS l.1.5 DEVELOPMENT OF ACCEPTABLE OFF SITE DOSE 1.1-6
- 1.1.6 RELATIONS OF Tills PDMS SAR TO EXISTING UNF
- 1.1-6
> UFSAR AND UNIT 2 FSAR 1.2 GENERAL PLANT DESCRIPTION 1.2 1 l.2.1 SITE CHARACTERISTICS 1.2-1 ( 1.2.2 CONTAINMENT SYSTEMS 1.21 1.2.2.1 Containment 1.2-1 1.2.2.2 Containment isolation Valves 1.21 1.2.2.3 Containment Atmospheric Breather 1.2 2 1.2.3 FIRE PROTECTION, SERVICE, AND SUPPRESSION 1.2 2 1.2.4 RADIDACTIVE WASTE MANAGEMENT 1.2 2 1.2.5 RADIATION MONITORING 1.2-2 1.2.6 ELECTRICAL SYSTEMS 1.2-3 1.2.7 PDMS SUPPORT SYSTEMS 1,2-3 1.2.8 FACILITIES AND SYSTEMS RELEASED FOR SITE USE 1.2-3 1.3 MATERIAL REFERENCED 1.3-1 1.4 PRESLHTATION 1,41 1,4-1 GPUN DRAWINGS 1.4-1 1.4-2 ABBREVIATIONS AND ACRONYMS 1.4-1 LIST OF FIGURES i 1.1.2.2.4 Cork Seam Modifications 1.5-1 UPDATE 2 - AUGUST 1997
( CMAPTER1 INTRODUCTION AND GENERAL DESCPJPTION OF PLAST -
1.1 INTRODUCTION
This Post Defueling Monitored Storage (PDMS) Safety Analysis Report (SAR) was submitted in support of the application of GPU Nuclear Corporation (GPU Nuclear) as agent for Metropolitan Edison Company (Met Ed), Jersey Central Power & Light Company (JCP&L), and Pennsyhania j Electric Company (Penclec) for revision of the existing Class 103 license to delineate the non-operating status of the nuclear electric generating station designated as nree Mile Island Nuclear Station Unit 2 i (TMI 2). Here are various sections of this SAR which reference either the Unit 1 UFSAR or Unit 2 FSAR for relevant information These refer to information which has been previously reviewed and approved by the NRC, remains valid, and does not require further review, ne TMI l UFSAR will continue to be updated in accordance with 10 CFR 50.71(e) whereas the TMI 2 FSAR update is no longer required. The nree Mile Island Unit 2 operating license was issued on February 8,1978, and commercial operation was declared on December 30,1978. On March 28,1979, the tuut experienced an accident which resulted in severe damage to the reactor core. TMI-2 has been in a non-operating status since that time, GPU Nuclear conducted a substantial program to defuel the Reactor Vessel and decontammate the facility. As a result, TMI.2 has been defuchd and decontanunated to the extent that the plant is in a safe, inherently stable condition suitable for long-term management and any threat to the public health and safety has been eliminated. This long-term management condition is termed Post Defueling Monitored Storage. 1.1.1 POST-DEFUELING MONrf0 RED STORAGE Post-Defueling Monitored Storage has been established at this time based on three principal considerations:
- 1. The Reactor Vessel and the Reactor Coolant System have been defueled and the core material has been shipped off site.
- 2. Decontanunation has been completed to the extent that further major decontammation programs are not justified on the basis of worker dose.
- 3. A condition of stability and safety has been established such that there is no risk to public health and safety,
, These three criteria are interrelated in that each has some degree of dependence on the other. A - significant amount of decontammation had to be completed prior to the begmnmg of defueling. As defueling progressed and was completed, further decontamination tasks were undertaken which allowed for .he completion of other cleanup aethities. Extenshe decontammation and the completion of defueling were required to establish the inherently stable and safe condition of the facility such that there was no risk to public health and safety. uO 1.1 - 1 UPDATE 2 - AUGUST 1997
Although it would be possible to continue decontammation of the facility, this effort would not appreciably enhance the overall stability and maintainability of the facility nor would it add to the o crall margin of safety for the public However, additional decontamination at this time would result in unnecessary additional occupational exposure to personnel who would be conducting the decontammation tasks. In addition, further decontammatios might require some destructive decontanunation techniques which would be best conducted when the plant is decommissioned. By placing Thil 2 in a long term monitored storage condition at this time (i.e., tmtil the time of decommissioning of Thil-1), it is possible to realize a significant savings in occupational exposure by deferrmg remaining decontammation activities for a number of years. For example, assuming all other factors to be equal, natural decay of the dominant radioactive isotopes (St-90, Cs-137) over a 30-year period would result in the occupational exposure being reduced by a factor of approximately two for any given task. It is possible that additional reductions in occupational exposure also could result from advances in decentammation technology and robotics. Although the exact reductions in occupational exposure are extremely difficult to quantify, it is clear that postponement of the remaming pre-decortunissioning decontamination activities will result in a significant sasings in occupational exposure (Reference Table SA 10, Chapter 5). His estimate was based on information published by the NRC in Supplement I to the Programmatic Emitorntal Impact Statement (PEIS) and adjusted to account for variations in the scope of the work and actual exposure experience during the cleanup. Based on this prelinunary analysis and using revised data, GPU Nuclear, in 1985, more accurately quantified the potential savings in occupational exposure due to the deferral of the remaining decontanunation tasks. The results of this extensive effort are pcesented in Chapter 5, Appendix 5A. Given that occupational exposure remains the singular contributor to emironmental impact, avoidance of unnecessary exposure where the public health and safety are not at risk is appropriate. 1.1.2 TRANSITION TO POST-DEFUELING MONITORED STORAGE The formal transition from the post-accident condition to PDMS required NRC approval. The establishment of PDh1S occurred over a period of time preceding the formal implementation of PDhf' and extending into PDMS. Not all actisities leading to the fmal PDMS configuration were completed pnor to the implementation of PDMS. However, du:ing the transition period, all of the prerequisites to PDMS were satisfied. The following sections outline those conditions which were established prior to the implementation of PDMS and those activities which extended for some period of time subsequent to PDMS implementation. 1.1.2.1 Prerequisites for PDMS He following prerequisites were satisfied prior to the implecantation of PDMS,
- 1. It has been demonstrated that there is ne credible possibihty of nuclear criticality. His condition has been assured by the removtl of abstantial!y all of the fuel from the Reactor Vessel and elimination of all potentially critical configurations The elimmation of any credible possibility of nuclear criticality was demonstrated as a requirement for transition to Mode 2 in accordance with the TMI-2 Recovery Technical Specifications. GPU Nuclear Letters 4410 L-0012 (Defueling Completion Report), C312-92-2M0 (Reacter Vessel Criticality Safety Analyses) and C312-93-2004 (Reactor Vessel Post Defueling Survey Reports) have been submitted to support this conclusion.
O 1.1-2 UPDATE 2 - AUGUST 1997
(/V }j 2. All fuel and core debris which have been removed from the Reactor Vessel and associated systems have been shipped off-site.
- 3. Any potential for a significant release of radioactive material has been eliminated. Radioactive material has been removed and other sources of radioactivity have been isolated so that any potential r dioactive release will be within 10 CFR 50 Appendix ! guidelines for off site dose consequer :es.
- 4. As a preconition to implementing PDh1S, water has been removed to the extent practical from the Reactor Coolant System and the Fuel Transfer Canal, and the fuel transfer tubes have been isolated. To the extent that the Spent Fuel Pools are needed to support Accident Generated Water disposal activities, water may remain in these pools subsequent to the implementation of PDhiS. The treatment and processing of the Accident Generated Water has been completed.
- 5. All radioactive waste from the major cleanup activities has been shipped off-site or has been packaged and staged for shipment off site.
- 6. Radiation within the facility has been reduced, as necessary, consistent with As-low-as-is-reasonably-achievable (ALARA) principles to levels which will allow necessary plant monitoring activities, the performance of required maintenance, and any necessary inspections.
1.1.2.2 Transition Activities p Although the conditions described in Section 1.1.2.1 were established prior to the implementation of (N PDh1S, there were some conditions described in this SAR whict ,u re not essential to the implementation of PDMS and the related activities were conductct' subsequent to the implementation of PDMS. A general description of some of those actisities and conditions follows.
- 1. Decontammation - During the initial stages of PDMS, removal or isolation of small sources of radioactivity or radioactive material will be ongoing.
- 2. Radioactive Waste - Small quantities of radioactive waste will contmue to be generated, accumulated, and packaged during PDMS. Thus, radioactive waste shipments will continue during PDMS.
- 3. PDMS Electrical Modification - The TMI 2 PDMS Electrical Modification results in a safer, more reliable power system than the system existing during Facility Mode 3. Althcugh many electrical loads were eliminated during PDMS preparation, this modification further consolidates loads to produce a simpler, easier to maintain electrical power system. The PDMS -
Electrical Modification wvs completed in July of 1994.
- 4. Cork Seam - The TMI-2 cork seam is a cork-filled construction joint located between the various major structures at TM1-2 (See Figure 1,4-1). During the TMI-2 accident, the cork scam located in the Auxiliary Building Seal Injection Valve Room (SIVR) was contaminated with radioactive wmer. Since the accident, radioactive material has spread along thejoint in -
L one direction into the Annulus, and in the (ther direction into the Auxiliary Building, Sertice i Building and Cc,atrol Building Area. The radioactive contammation is prevented from entering the ground water table by a PVC waterstop and thus represents no threat to the health and l (q) safety of the public. v 1,1-3 UPDATE 2 - AUGUST 1997 l
Modifications have been made to the cork seam to allow periodic monitoring of the water levcis in the joint, to permit periodic water removal, and to prevent water ar.d contammation migration within the cork filled joint. Figure 1.4-1 shows the locations where the cork seam was penetrated and modifications performed. At Ic. cations S-2, S-6, S-7, S-8, and S 9, one inch diameter holes were drilled into the cork seam to a depth of about one inch above the PVC waterstop. A perforated tube was inserted into these monitoring holes and acts much like a comentional well point in providing a means of pumping accumulated water out of the seam in the vicirury. Also, the monitor pipes have been fitted with bubbler devices that allow measuring of the water level in the hole. During the drilling for hole S4, samples of water and cork were removed for analysis. From the data obtained, it was determined that the 40-year Total Integrated Dose to the PVC waterstop in this area would be 6.4E5 rads. His is less than 5% of the dose that would cause degradation of the material. At locations S 1, S-3, S-4, S-5, and S 10, four inch diameter holes were core bored to penetrate the cork and the adjacent concrete slah on both sides of the seam. These holcs extend down to the surface of the PVC waterstop. When these holes were drilled, the bxtcm was carefully probed w th a blunt instrument to verify that the hole had actually reaci.ed me waterstop and to verify that the waterstop had not been breached and the material was still pliable. Where possible the waterstop was visually inspected. He hole was then filled with a moisture activated expandable polyurethane foam material that will bond well with both the concrete and the waterstop material. These locations serve as dams that will block water migration within the cork seam. To prevent any additional spread of contamination frorn the cork seam in the future, work was completed to install a seal over the top of the cork filled joint, he work involved excavating about 3 inches of cork from the top of thejoint along all accessible areas of the full length of the scam. He concrete edges were surface prepped and primed where required, an open cell polyurethane foam backer bar was packed into thejoint, a:'d the top one inch depth of the joint was filled with a pourable polysulfide sealant, ne materials chosen for this modification are expected to provide a very tight bond with the concrete and provide a good pliable contammation barrier over the surface of the cmk seam. The monitoring holes remained intact after installation of the top seal described abose. This will pernut continued monitoring and sampling of the water that may enter the seam, and removal for processing as required. A program is in place for continued monitoring of:he water level in the cork seam by the PDMS staff. If the water levels begm to increase, the water will be pumped out for processing as needed O 1.1-4 UPDATE 2 - AUGUST 1997
I (j' TMI 2 has been maintained in a safe, monitored condition throughout the transition period prior to PDMS, and will be maintamed accordingly following implementation of PDMS even though some transition activities are ongoing. A commitment tracking process has been established to verify the status and completion of all actisities perfonned in preparation for PDMS and during the PDMS transition period to ensure all required activities described in this SAR are completed. 1.1.3 APPLICABLE REGULATIONS GPU Nuclear received an amended facility license for TMI 2 in accordance with the provisiens of Title 10 to the Code of Federal Regulations, Part 50 (10 CFR 50). He provisions of 10 CFR f a t established, were intended to be applicable to an operable nuclear power plant. For this reason, cuy of the requirements originally imposed on TMI 2 no longer apply or can be substantially reduced in scope because of the status of TMI 2 during PDMS. Because nuclear criticabty has been precluded with removal of substantially all of the fuel from TMI-2, and because radiation hazards have been substantially reduced due to the immobilization of essentially all of the radioactivity remaining in the plant, many systems, structures, and components are no longer required and the regulations goveming these systems, structures, and components have a significantly reduced scope of applicability at TMI-2. In order to assure compliance with the appropriate requirements of the regulations in 10 CFR 50, a thorough re iew of these regulations was undertaken. Chapter 3 of this SAR presents the results of that review and serves as the basis for determmmg which regulations have a controlling impact on TMI 2. He determination of applicability does not suggest that some regulations can be ignored. Rather, the A f intent of some regulations can be met with no impact or additional requirements imposed due to the PDMS status of TMI 24 i G 1.1.4 SAFETY-RELATED STRUCTURES, SYSTEMS, AND COMPONENTS There are no structures, systems, or components classified as safety-related at TM1-2 during PDMS. GPU Nuclear procedures defme safety-related structures, systems, and components as those which are necessary to ensure;
- a. He integrity of the reactor coolant pressure boundary,
- b. He capabibty to shutdown the reactor and to maintain it in a safe shutdown condition, or
- c. He capabihty to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparsble to the guidelines exposures of 10 CFR Part 100.
Criterion a requires maintenance of the reactor coolant pressure boundary. Due to the defueled condition of TMI 2, there is no reactor coolant or reactor coolant pressure boundary required. Criterion b requires a capability to shutdown the reactor and maintain it in a safe shutdown condition. In its current defueled state, there are no structures, systems, or components required to maintain a safe shutdown condition. Criterion c requires a capability to prevent or mitigate the consequences of accidents that could result in (q v j potential off site exposures comparable to the 10 CFR Part 100 guidelines. Analysis demonstrates (see Chapters 4 and 8) that there are no postulated events that result in releases greater than 10 CFR 50 Appendix I guidelines. Since 10 CFR 50 Appendix I is more restrictive, there are no postulated events 1.1-5 UPDATE 2 - AUGUST 1997 l l
which could result in exposures comparable to 10 CFR Part 100 guidelines. Due to the non operating and defueled status of TMI-2 during PDhfS, there are no structures, systems, or components which are required to rneet the safety-related criteria. Therefore, there are no structures, systems, or components classified as safety-related at TMI 2 during PDMS. 1.1.5 DEVELOPMENT OF ACCEPTABLE OFF SITE DOSE CRITERIA Various regulations establish permissible limits for off-site radiation exposures resulting from the operation oflicensed nuclear reactors and other nuclear fuel cycle actisities These regulations include 10 CFR 20,10 CFR 50 Appendix I,10 CFR 100,40 CFR 190, and the EPA Protective Action Guidelines. The licensing basis for off site dose criteria for PDMS has been derived from these existirig regulations and applicable precedents. Specifically,10 CFR 50 Appendix I, which is recognized as demonstrably safe with respect to radiological implications, has been established as the PDMS standard. A small fraction (i.e., less than 10%) of the Appendix 1 off-site dose guidelines is expected to be maintained for normal conditions prevailing during PDMS. The potential off-site radiological doses resulting from postulated off-normal conditions will be within the 10 CFR 50 Appendix I guidelines. Due to the non-operating and defueled status of TMI 2, a major radiological release approaching the guidelines of 10 CFR 100 is no longer credible. As noted above, Appendix I guidelines have been selected as the limiting criteria for the evaluation of unanticipated events as an unarguable, demonstrably, conservative basis. This ultra-conservative approach far exceeds the regulatory limits for unanticipated events in operating nuclear power plants. 1.1,6 RELATIONS OF THIS PDMS SAR TO THE EXISTING UNIT I UFSAR AND UNIT 2 FSAR This PDMS SAR makes reference to relevant portions of the Unit 1 UFSAR or the Unit 2 FSAR. The TMl-1 UFSAR will continue to be updated as required, and the updated document will be applicable for those changing site-related conditions that have a bearing on TMI 2. The TMI 2 FSAR will not be updated but will continue to be applied as appropriate to TMI 2 in the PDMS condition. In parti:ular, the bounding conditions in the TMI-2 FSAR as augmented by the PDMS SAR will be used to judge the acceptability of changes, tests, and experiments and the attendant determination of unresiewed safety questions. The TMI 2 FSAR also applies for those areas not addressed by this PDMS SAR. O 1.1-6 UPDATE 2 - AUGUST 1997
(O ) 1.2 GENERAL PLANT DESCRIPTION On March 28,1979, Three hiile Island Unit 2 experienced an accident which severely damaged the reactor core. In the ensuing years, the reactor core has been reme /ed and shipped to the Idaho National Engineering Laboratory for analysis and long term storage. In addition, the facility has been substantially decontaminated and is in a stable and benign condition suitable for long-term management. Three Mile Island Unit 2 was originally designed to comply with the seventy General Design Criteria of 10 CFR 50, Appendix A, dated July 11,1967 and addressed plant design with respect to the Revised General Design Criteria dated July 15, 1971. Due to the defueled and non operating status of TMI 2 during PDMS, many of these criteria no longer apply to the facility. A review of the General Design Criteria, revised as of January 1,1987, is included in Section 3.1 of thn SAR. The general arrangernent of major equipment and structures, including the Reactor, Auxiliary, and Turbine Buildings is shown on GPUN Drawings listed in Table 1.4-1. 1.2.1 SITE CHARACTERISTICS Re site is located on the Susquehanna Rher about ten miles southeast of Harrisburg, Pennsyhania. It is characterized by a 2,000 foot nuntmum exclusion distance; a two mile radius low population zone; sound bedrock as a structural foundation; an ample supply of emergency off-site power and favorable conditions of hydrology, geology, seismology and meteorology. The land within a ten mile radius of the site is t. ed primarily for fanning.
/ \
Rere are two airports within ten miles of the site. Harrisburg Intemational Airport (formerly Olmsted State Airport) is located approximately two and one-half miles northwest of the site, and the Capitol City Airport is located approximately eight miles west-northwest of the site. 1.2.2 CONTAINMENT SYSTEMS The Containment and asseiated systems are used during PDMS as the emironmental barrier for the residual contamination which remains inside the Containment structure. The Containment encloses the areas and systems which contain essentially all of the contammation which could potentially result in off-site exposures. 1.2.2.1 Containment ne primary function of the Contamment during PDMS is as a contanunation bamer. The Containment will prmide shielding of the emironment from the radiation inside the Containment, and will also prmide the means to assure that any effluents from the Containment will be controlled, filtered, and monitored. The Containment is a reinforced concrete structure composed of cylindncal walls with a flat foundation mat and a dome rooflined with carbon steel. The structure provides biological shielding for normal and unanticipated conditions. The steel liner encloses the equipment and systems which remain inside the Containment and ensures that the upper limit of potential leakage of radioactive material will not be exceeded under the worst unanticipated event. A ( l.2.2.2 Contamment Isolation Vahes He Containment isolation vahrs were designed to provide a barrier on the system lines which penetrate 1.2-1 UPDATE 2- AUGUST 1997
the Contamment so that no event can result in loss ofisolation or intolerable leakage, in most cases, the valves are installed both inside and outsidt the Reactor Building on each system line. Only one vah e is required for isolation during PDMS. All valves used for containment isolation during PDMS are normally closed and locked, closed and deactivated or closed and admmistratively maintained closed except for the breather isolation valve ud RB presare indication piping which are normally open 1.2.2.3 Containment Atmospheric Breather ne Containment Atmospheric Breather has been asU to the Containment to provide passive pressure control of the Containment relative to ambient atme.gbrie pressure (via the AFHB) and to establish a "most probable pathway" through which the Containment will " breathe", ne breather is a passive system consisting of a 6 in, diameter duct with a HEPA filter. Providing this filtered pathway will ensure insignificant leakage through any uncontrolled pathway. The Containment Atmospheric Breather is described in more detail in Section 7.2,1.2. 1.2.3 FIRE PROTECTION, SERVICE, AND SUPPRESSION Fire Protection is provided during PDMS to minimize the potential of a release of radioactive material due to a fire in a contammated area, to protect those systems which are mamtained operational during PDMS, and to muumize the liability and property risk from potential fires. Rese objectives have been achieved through a combination of(1) muumizing the potential for a fire by minimizing combustible materials and ignition sources and (2) by providing a system of detection and suppression suitable to deal with any potential fire, 1.2.4 RADIOACTIVE WASTE MANAGEMENT The generation of radioactive waste during PDMS wil' be muumal A small amount of radioactive waste will be generated from the processing of water inleakage to contaminated areas, small decontamination tasks, and surveillance and maintenance aethities. Liquid radwaste will be collected in the various sumps and handled through the liquid radwaste disposal system. Other radwastes will be collected and disposed of as appropriate. l 1.2.5 RADIATION MONITORING During PDMS, radiation monitors will be maintained operational to provide for evaluation of airborne radiological conditions. His requires monitoring the Reactor Building exhaust ventilation and the O l 1.2-2 UPDATE 2 - AUGUST 1997 l l
i l l 1 f N ij s station vent during periods when a ventilation system is operating. The monitors will provide the necessary information to evaluate emironmental releases and air quality conditions in the plant. This monitoring will provide a basis for determmmg the total integrated dose to the pubb . Monitoring and survey data will provide a basis for a trend analysis to ensure that the plant is maintained in a stable condition and enables timely estrective actions, if necessary. 1.16 ELECTRICAL SYSTEMS a During PDMS, portions of the TMI 2 AC and DC electrical systems will be maintained operational to provide reliable power to PDMS support systems, controls, and instrumentation. Electrical eg .ipment that is not required for PDMS support has been deactivated to enhance overall plant safety. T 1.2.7 PDMS SUPPORT SYSTEMS Other systems necessary to support PDMS activities also have Mn provided. The ventiLtion syst ms for the Auxiliary, Fuel Handling, Control and Senice Buildings will be maintained operational te provide ventilation capabilities in those areas. Compressed air, sewers, domestic water, and other systems have been provided for use, as necessary. 1.2.8 FACILITIES AND SYSTEMS RELEASED FOR SITE USE As a result of the accident, unique situations developed which could not be properly managed with the existing facilities or systems which were designed for normal operating power plant use. Several systems were designed and fabricated to process the radioactive wastes resultias from cleanup (Ab) activities, Upon completion of cleanup activitics, several of these facilities were released to general site use (and included under the TMI l license). These systems and facilities include:
- 1. Auxiliary Building Emergency Liquid Cleanup (EPICOR II)
- 2. Waste Handling and Peckaging Facility
, 3. Interim Solid Waste Sterage Facility l 4. Solid Waste Staging Freility l 5. Respirator Cleaning and Laundry Maintenance Facility l 6. Solid Waste Storage B;ilding l 7. Processed Water Storage Facility l l l l l O ? l 1.2 - 3 UPDATE 2 - AUGUST 1997
1.3' _ MATERIAL REFERENCED he following documents are referenced as part o'this application. - Referenced in . !
~ Dxurasw SAR Section l
TMi 2 Final Safety Analysis Report 1.1,1.1.6, 2.4.2, 2.4.5, 2.5, 3.1.1.17, 3.1.2, 3.2.1, 3.2.1,1,-3.2.2.1, - 3.3.1.1, 3.4.2,- 3.5.3.2, 3.6,.3.7, 13.7.2.1.1, 6.0 TMI l Updated Safety Analysis Report : 1.1, 1.1.6, 2.1.3, 2.2, 2.3, [ 2.4.3,3.1.1.17 : NUREG-0683, Supplement No.1, " Programmatic 1.1.1
- Emironmental Impact Statement ...," October 1984
- NUREG-0683, Supplement No. 3, " Programmatic 5A.I.2, 5A.8, 5A.10 1 +
Emironmentallmpact Statement...," August 1989
*Decontammation Task Force Report," by 5 A.2.1, 5 A.8 i P, R. Bengel, et al, December 18,1985 Technical Bulletin 85 1, " Reactor Building SA.4.1
, . General Area Radiation Survey Maps," Rev. 3, February 3,1988 Technical Bulletin 86-10, ""B" Steam Generator 5A.4.1 TLD Characterization," Rev. O, February 18,1986 GPU Nuclear memorandum 9240-88-4372 from SA.4.4 J. E. Tarpinian to D. W. Turner, " Comparison of TMI-2 with other B&W Plants," dated February 2,1988 BBR GMBH Report Number 595-C01 A (82), " Evaluation SA.44 3 of the Dose Rate Data of Various Nuclear Power
. Plants with B&W Nuclear Steam Supply Systems"
~
GPU Nuclear memorandum 6615-90-0188 from S. E. SA.5, 5A.11
' Acker to E. D. Schrull, "PDMS SAR Ch 8 Dose Calculationi," dated November 5,~ 1990 r.
L
~
v, l .3 - 1 = UPDATE 2- AUGUST 1997 4 _- m- r~,- ,
l.3 MRERIAL REFERENCED (Cont'd) Referenced in Document S AR Section Technical Plan TP0/TMI 188, "TMI 2 Cleanup 5 A.6.1.2 Program Post Defueling Monitored Storage," Revision 0, January 1987
- Disposition of the Reactor Building Blockwall," 5A.7.1, Table SA 5, by P. R. Bengel, et al, June 19,1987 (Attachment Table S A-9 to GPU Nuclear Memorandum 4440 87-048)
" Task Force Report - Reactor Building Basement 5A.7.1 Decontamination," November 1987 GPU Nuclear memorandum 9240-88-4521, from D. J. 5A.9.1 Merchant to J. E. Tarpinian, " Review cf 1986 Collective Dose Goal," dated May 9,1988 TP0ffMI-009," Gross Decontammation Experimsnt SA.2.1 Report," Bechtel National, Inc., September 1982 Letter, Travers, W. D. (NRC) to Standerfer, F. R. 31.1.20,31.1.51 (GPUNC), " Approval of Exemption fron.10 CFR 50.61,"
dated December 30,1985 Letter, Snyder, B. J. (NRC) to Kanga, B. K. (GPUNC), 3.1.1.38 "10 CFR 50.49, 'Emironmental Qualification of Electrical Equipment important to Safety for Nuclear Power Plants'," dated July 22,1983 Letter, Stolz, J. F. (NRC) to Standerfer, F. R. 3.1.1.43 (GPUNC), " Issuance of Amendment (TAC No. 65337)," dated May 27,1988 Letter, Snyder, B. J. (NRC) to Hovey, G. K. 3.1.1.43,3.1.1.45 (Met-Ed), Re. Exemption from 10 CFR 50 Appendtx J, dated September 2,1981 Letter, Snyder, B. J. (NRC) to Hovey, G. K. 3.1.1.45 (Met Ed), Re: Relief from the Insenice Inspection Program Requirements of 10 CFR 50.55a, dated April 27,19'l O l 1.3 - 2 UPDATE 2 - AUGUST 1997
=- --
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D 1.3 MATERIAL REFERENCED (Cont'd) > Referenced in Document SAR Section GPU Nuclear letter,4410 90 L-0044, 3.1.1.59 -i' "Decommissionir,g Financial Assurance Certification Report for . . TMI 2," dated July 26,1990' GPU Nuclear letter,4410 90-L-0012, "Defueling 1,1.2.1, 3.1.2.52, 4.0, 4.3.1, Completion Report, Final Submittal," dated 4.3.5 February 22,1990 ALAB-692 dated September 14,1982 3.5.3.2 GPU Nuclear letter, LL2 810191. " Design 3,7.1.2, 7.2.1.1 Pressure for Conteur2nt and Future Mechanical and Electrical Penetration Modifications," dated December 4,1981 Letter, W. D. Travers (NRC) to F. R. Standerfer 3.7.1.2 (GPUNC) " Seismic Design Criteria for Modified p Containment Penetrations," dated April 3,1987 Technical Bulletin 89-08, Revision 0, " Final 4.1.2 Core Material Estimates," October 19,1989 DOE letter WWB 100 85 W. W. Bixby (DOE) to 4.3.2 H. M. Burton (EG&G), " Accountability for the TMI 2 Core," dated October 8,1985 1.etter, B. J. Snyder (NRC) to F. R. Standerfer 4.3.2, 4.3.6 3' (GPUNC), " Approval of Exemption from 10 CFR 30.51, 40.61,70.51(d), and 70.53," dated October 17,1985 NSAC 80-1, " Analysis of ~lhree Mile Island - 4.3.3.1 Unit 2 Accident," Electrical Power Research Institute, March 1980 Rogovin M., et.al., "Three Mile Island, A Report 4.3.3.1 to the Commissioners and the Public," US Nuclear
- Regulatory Commission, Januuy 1980 GPU Nue:=r Procedure 4000-PLN-4420.02, "SNM 4.3J.2 Accountability Plan"
_ 6 TPOfTM1-051, " Location end Characterization of 4.3.3.2 . t ' Fuel Debris in TMI-2," Revision 0, April 1984' 13-3 UPDATE 2- AUGUST 1997
1.3 h1ATERIAL REFERENCED (Cont'd) Referenced in Document SAR Sectipa TFO.W.1-!R "Ex-Vessel Fuel Characterization," 4.3.32 ; Revision 0, July 1984 TP0frMI l 87, " Instrument Selection for Residual 4.3.3.2 Fuel Measurements," Revision 0, January 1987 l GPU Nuclear Procedure 4000 ADM-4420.03, " Review 4.3.3.2 and Quali6 cation of Selected Preliminary Calculations and Characterization Measurements for SNM Documentation" i l Letter, Stolz, J. F. (NRC) to Roche, M. B. (GPUNC), 4.3.5 "Three Mile 151and Unit No. 2 Mode Changes," dated ! April 26,1990 GPU Nuclear merrera: dun, from D. W. Dallcogcc 8.1.3 to T. D. Murphy," Safety /Emironmental Determination and Review of SEEDS," dated August 10,1990 Letter, W. D. Travers (NRC) to F. R. Standerfer (GPUNC), " Seismic Design Criteria for Modi 6ed Containment Penetrations," dated April 3,1987 GPU Nuclear letter, C312-91-2045, "SNM Accountability," 3.1.2.52,4.0,4.3.3.3 transnutting the Auxiliary and Fuel Handling Buildings PDSR, dated June 7,1991 GPU Nuclear letter, C312 912052,"SNM Accountability," 3.1.2.52,4.0,4.3.3.3, tmnsmitting the R.; actor Building Miscellaneous Components PD3R, dated June 18,1991 l l 9 l 1.3 - 4 UPDATE 2 - AUGUST 1997 l
_ . _ _ . _ . _._ _ _ . . . _ . _ .. - . _ _ . . _ . _ _ . ._....m . ____ _ _ _ 4 1.3 MATERIAL REFERENCED (Cont'd) Referenced in , , Document - SAR Section l
- GPU Nuclear letter, C312 912055, "SNM Accountability," 3.1.2.52,4.0,4.3.3.3, j transnstting the Reactor Coolant System PDSR, dated
. July 3,1991 GPU Nuclear letter, C312-91 2064, "SNM Accountability," 3.1.2.52,4.0,4.3.3.3, i
transmitting the 'A' and 'B' Once-Through Steam Generators PDSR, Revision 1, dated July 3,1991
- GPU Nuclear letter C312-93-2004,"SNM Accountability," 1.1.2.1,3.1.2.52,4.0,4,3.3.3 l transmitting the Reactor Vessel PDSR, dated l February 1,1993 GPU Nuclear letter, C312 92 2080, "TMI 2 Reactor Vessel 1.1.2.1,3.1.2.52,4.0,4.3.1, Cnticality Safety Analysis," dated December 18,1992 4.3.5 ,
TMI Radiological Controls Department Procedure, 3.1.1.20, 6610-PLN-4200.01, "Offsite Dose Calculation Manual (ODCM)" , 'TMI 2 Recovery Technical Specifications 1.1.2.1, GPU Nuclear memorandum 6615 92-0160, from S. Acker 5.A.11,8.2.5 to E. Schrull, " Dose Calculation Results per memo C312-92 1945, PDMS S AR Rev.16," dated October 27,1992 GPU Nuclear memorandum 6615 92-0162, from S. Acker 8.2.5
- to E. Schrull, " Additional Dose Calculations per memo C312-92-1045, PDMS SAR Rev.16," dated October 30,1992 GPU Nuclear memorandum 6510 93-0077, from S. Acker 8.2.6 l to E. Schrull, " Dose Calculation Results per memo C312-93-1019, PDMS SAR Rev.17," dated May 21,1993 l.3-5 UPDATE 2 - AUGUST 1997
_ _ _ _ . . . _ . . ~ _ . _ _ _ _ _ _ _ ,
,D i] 1.4 PRESENTATION
-1.4.1 DRAWINGS GPU Nuclear controlled drawings listed in Table 1.4-1 are referenced throughout the text of the PDMS SAR. Copies of the current resision of each drawing are readily available at the Three Mile Island Nuclear Station. GPUN Drawing 2001, P & ID SymbolIdentification, provides explanation for symbols used in non electncal drawings. GPUN drawing 3001, Electrical Symbol L.ist, prosides explanation for symbols used in electrical drawings.
1.4.2 ABBREVIATIONS AND ACRONYMS Abbreviations and acronyms which are used in this document are listed in Table 1.4-2. t O t (N i
\d 1.4-1 UPDATE 2 - AUGUST 1997
TABLE 1.41 GPUN DRAWINGS TITLE GPUN DWG. NO. Site Plan 2E 120-01-001 Reactor Building Basement Flwr 2060 Reactor Building Ground Floor 2061 Reactor Buildmg Operating Floor 2062 Reactor Buildmg Section A-A 2063 Reactor Building Sections 2064 B-B, C-C, D-D Auxiliary and Fuel Handling 2065 Buildmg, Basement and Sub-Basement Floor Auxiliary and Fuel Handling 2066 Buildmg, Ground Floor Auxiliary and Fuci Handling 2067 Buildmg, First Floor Auxiliary and Fuel Handling 2068 Floor Building, Operating Auxiliary and Fuel Handimg 2069 Building, Sections A-A and B-B Auxiliary and Fuel Handling 2070 Buildmg, Section C-C Auxiliary and Fuel Handling 2071 Buildmg, Section D-D Auxiliary and Fuel Handling 2072 Buildmg, Section E-E Control and Senice Building, 2380 Lower Floor Plans Control and Senice Building, 2381 Upper Floor Plans Control and Senice 2382 Buildmg, Sections River Water Pump House 2338 1.4-2 UPDATE 2 - AUGUST 1997
TACLE 1.41 (Cont'd) DRAWING REFERENCES
/ \
l TITIE GPUN DWG. NO. Air intake Tunnel Plan 4039 Turbine Building Basement Plan - East Side 2051 Turbine Building Basement Plan - West Side 2052 Turbine Building Ground Floor Plan - East Side 2053 Turbine Building Ground Floor Plan - West Side 2054 Turbine Building Operating Floor - East Side 2055 Turbine Building Operating Floor - West Side 2056 Turbine Building Section B-B 2057 Turbine Building Sections A A and C-C 2058 Turbine Building Sections D D and E-E 2039 P&lD SymbolIdentification 2001 p Electrical Symbol List 3001 Reactor Building Ventilation and Purge 302-2041 Fire Protection 302-231 Radwaste Disposal Miscellaneous Liquids 302-2045 Radwaste Pumps Seal Water 302-2492 Sump Pump Discharge and Misce!!aneous Sumps 302-2496 Building Air intake, Exhaust, and 302-2219 Radiation Monitoring 13.2 KV One Line Diagram 206201 480 Volt Unit Substation 206202 480 Volt Unit Substation 206203 480 Volt Unit Substation 206204 l ('~x,, 120V Regulated Voltage System 3009 L.) 1.4-3 UPDATE 2 - AUGUST 1997
TABLE 1,4-1 (Cont'd) DRAWING REFERENCES TITI,E GPUN DWG, NO. DC One Line Diagrarn 3019 480V USS 2-38,2-48 One 1.ine Diagram E013 Reactor Building Portable Power Distr. Center 2-E21-011 Reactor Building Portable Power Distr. Center 2-E21-012 Power Distribution Key Diagram 3015 Power Distribution Panel Schedules 3016 Miscellaneous Power Panel Schedules 3017 Sh. I Miscellaneous Power Panel Schedules 3017 Sh. 2 Miscellaneous Power Panel Schedules 3017 Sh 3 Auxiliary Building Heating and Ventilation 302-2042 Fuel Handling Building Ilcating and Ventilation 302-2343 Instrument Air Supply 302-2012 Sht.1 Compressed Air Supply 302-2012 Sht. 2 Service Air 302-2014 Sht. 3 PDMS SAR FIGURES TITLE PDMS SAR FIG. NO. General Area Map 2.1-1 Site Topography 5 Mile Radius 2.1-2 Extended Plot Plan 2.1 3 Flood Water Surface Profiles 2.4-1 Details of Effluent Discharge System 2.4-2 Reactor Building - General Layout 3.7-1 l l Reactor Building Personnel and Equipment 3.7-2 Access Openings Detail 1.4-4 UPDATE 2 - AUGUST 1997
4 TABLE 1.4 2 ABBREVIATIONS AND ACRONYMS ABST Auxiliary Building Sump Tank
- ACES : Automated Cutting Equipment System AEC Atomic Energy Commission
'AFHB Auxiliary and Fuel Handli .g Buildings s AGW Accident Generated Water AISC American Institute of Steel Construction AIT Air intake Tunnel ALARA As Low As Is Reasonably Achic .ble ANS American Nuclear Society ANSI American National Stardards Institute ASME American Society of Mechanic:.1 Engineers ASTM American Society of Testing Materials ATWS Anticipated Transients Without Scram i AWS American Welding Society BBR Brown Boveri Reactor BWST Borated Water Storage Tank B&R Bums and Roe
[ B&W Babcock and Wilcox CACE Containment Air Control Envelope CAS Compressed Air System CCW Closed Cooling Water CCTV Closed Circuit Television CFR Code of Federal Regulations efs/cfm . Cubic Feet Per Second/ Cubic Feet Per Minute Ci Curie CRD Control Rod Drive CRDCS Control Rod Drive Control Sysem
.J 1.4-5 UPDATE 2 - AUGUST 1997 c , , , - ,,-.m. , -,m,. . , . . . . - . - , .y - y. -- --
TABLE 1.4 2 (Cont'd) ADIIREVIATIONS AND ACRONYMS CRDM Control Rod Drive Motor CRDS Control Rod Drive System CS Cork Seam System CSA Core Support Assembly DCR Defueling Completion Report DilCCW Decay Heat Closed Coolms Water DH Decay lleat Removal DOE Department of Energy DOP Dioctyl Phthalate DTA DefuelinETest Assembly DTFR Decontamination Task Force Report DWC Defueling Water Cleanup System EPA Emironmental Protection Agency ESF Engineered Safety Features ETN Exposure Tracking Number FCN Field Change Notice FHB Fuel Handling Building FPPE Fire Protection Program Evaluation FSAR Final Safety Analysis Report FTC Fuel Transfer Canal GDC General Design Criteria GI Gastrointestinal gpm Gallons Per Minute GPU General Public Utilities GPUN GPU Nuclear GPUNC GPU Nuclear Corporation GRC General Review Committee H&V Heating and Ventilation HEPA High Efficiency Particulate Air O 1.4-6 UPDATE 2 - AUGUST 1997
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i N' TABLE 1.4 2 (Cont'd) ABBREVIATIONS AND ACRONYMS IIEU liighly Enriched Uranium lilC liigh Integrity Container liVAC licating, Ventilating, and Air Conditioning IAEA Intemational Atomic Energy Agency INEL Idaho National Engineering Laboratory IOSRO Independent Onsite Safety Review Group 1%TS Industrial Waste Treatment System JCP&L Jersey Central Power and Light LCSA Lower Core Support Assembly LER Licensee Event Report LLD Lowest Level of Detection LOCA Loss of Coolant Accident MDil Mini Decay 11 eat Removal ,Q MET ED Metropolitan Edison Company
%) MIDAS Meteorological Information and Dose Assessment System MPC Maximum Permissible Concentration mph Miles Per liour MSIV Main Steam Isolation Valve MSR Moisture Separator Reheater MU Makeup and Purification System MWiiT Miscellaneous Waste lioldup Tank NFPA National Fire Protection Agency NLB National Liquid Blasting NPDES National Pollutant Discharge Elimination System NRC Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NRJ Nephelometric Turbidiy Index ODCM Off-site Dose Calculation Manual (3 'w/
1.4-7 UPDATE 2 - AUGUST 1997
TABLI:1.4 2 (Cont'd) ABBREVIATIONS AND ACRONYMS OhiB Office of hianagement and Budget ORNL Oak Ridge National Laboratory OTSG Once Through Steam Generator PAF Personnel Access Facility PDhtS Post Defueling hionitored Storage PDSR Post Defueling Survey Report PEIS Programmatic Emironmental Impact Statement PENELEC Pennsylvania Electrie Company PhiF Probable hiaximum Flood PORC Plant Operations Review Committec PORY Pilot Operated Relief Yalve PSAR Preliminary Safety Analysis Report psi Pounds Per Square Inch It"' Piping and Instrument Diagram QA Quality Asst.rance RAF Radiation Area Factor RU Reactor Building RC Reactor Coolant S>mm RCBT Reactor Coolant Bleed Tank RChihi Remote Controlled hiobile hianipulator RCP Reactor Coolant Pump RCS Reactor Coolant System RCTY Remote Controlled Transport Vehicle RPS Reactor Protection System RRY Remote Reconnaissance Vehicle RV Reactor Vessel RWP Radiation Work Permit SAR Safety Analysis Report SCBA Self-Contained Breathing Apparatus O 1.4-8 UPDATE 2 - AUGUST 1997
TABLE 1.4 2 (Cont'd) AllllREVIATIONS AND ACRONYhtS seem Standard Cubic Centimeters per hiinute sefm Standard Cubic Feet Per hiinute SD System Description SDS Submerged Demineralizer System SEEDS Simplified Emironmental Efiluent Dosimetry System SER Safety Evaluation Report SFAS Safety Features Actuation System SFhtL Safe Fuel hiass Linut SFP Spent Fuel Pool SG Steam Generator SISI System In Senice inspection SNht Special Nuclear hiaterial SPC Standby Pressure Control SRP Standard Review Plan
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SSCCW Secondary Side Closed Coolmg Water STP Sewage Treatment Plant TER Technical Evaluation Report TLD Thermoluminescent Dosimeter Thil Three hiile Island Tht!NS Three hiite Island Nuclear Station TPAF Temporary Personnel Access Facihty UFSAR Updated Safety Analysis Report USGS U. S. Geological Survey UThi Universal Transmeridian WDG Waste Disposal Gas WDL Waste Disposal Liquid WDS Waste Disposal Solid WilPF Waste Handling and Packaging Facthty I O 1.4-9 UPDATE 2 - AUGUST 1997
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r r CHAPTER 2 i SITE CHARACTERISTICS r TABLE OF CONTENTS . l' ! SECTION ILILE .PAGE 7 2.1 GEOGRAPHY AND DEMOGRAPHY 2.11 l 2.1.1 SITE LOCATION 2.11. i 2.1.2 SITE DESCRIPTION 2.11
- 2.1.3 POPULATION AND POPULATION DISTRIBUTION 2.1 1 ;
2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND 2.2 1 l-MILITARY FACILITIES 2.3 METEOROLOGY 2.3 1 2.4 liYDROLOGIC ENGINEERINO 2.41 { 2.4.1 l{YDROLOGIC DESCRIPTION 2.4 1 2.4.1.1 Site and Facilities 2.4 1 2.4.2 FLOODS 2.41 2.4.3 FLOODING PROTECTION REQUIREMENTS 2.41 L 2.4.4 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS 2.4 2 2.4.5 GROUNDWATER 2.4 3 , 2.4.6 EMERGENCY OPERATIONS REQUIREMENTS 2.4-4 24.6.1 Flood Protection 2.4 4 . . 2.5 GEOLOGY AND SEISMOLOGY 2.5 1 i l , i UPDATE 2 - AUGUST 1997 l l H
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_ ._ _ . _ _ _ _ . ~ . . _ _ _ . _ _ _ . _ _ . _ . . . . _ _ _ _ . _ _ _ . _ . _ . _ _ _ _ . . _ . . _ _ _ . . _ . . _ . . _ . . _ . _ _ _ . _ _ . _ . 4 i 1 4 e t CHAI'IT.R 2 i TABLE OF CONTENTS (CosW'd) I i LIST OF 11GURES ( h
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j EIGURE NO. TITLE 11GE ; 2.11 GENERAL AREA MAP 2.12 i j- 2.12 SITE TOPOGRAPHY 5 MILE RADI1]S 2.13 2.1 3 EXTENDED PLOT PLAN 2.14 I j 2.41 FLOOD WATER SURFACE PROFILES 2.4 5 2.42 DETAILS OF EFFLUENT DISCHARGE SYSTEM 2.4-6 , t e l i i. l . . r b < i ii UPDATE 2 - AUGUST 1997 i i I l l
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L' fJtAITER 2 SITE CIIARACTERihTICS 2.1 GEOGRAPliY AND DEMOGRAPliY 2.1.1 SITE LOCATION nrec MJe Island is located approximately 21/2 miles south of Middletown, Permsylvania at longitude 76* 43' 30" west and at latitude 40' 8' north. De Unit 2 reactor vessel coordmates are N300, 324.40, E2, 286, 366.04, based on the Pennsylvania State coordinate system (UTM coordinates, Zone 18, 4,446,020 meters north, 353,070 meters east). It is one of the largest of a group of several islands in the Susqueharma IUver and is situated about 900 ft from the east bark it is elongated parallel to the flow of the nver, with its 11,000 ft. in length and 1700 ft. in width. TMI.2 is located in the northem one third of the island. He doutheasterly flowmg Susquchanna IUver makes a sharp change in direction. to nearly due south, in the vicinity of Middletown After this directional change just north of nice Mile Island, the channel widens to approximately 1.5 miles. The Three Mile Island Nuclear Station, Unit 2 is locatW adjacent to Unit 1 in Londonderry Township of Dauphin County, Pennsylvama, about 21/2 miles north of the southem tip of Dauphin County, where p) U Dauphin is coterminal with York and Lancaster counties. Its location with respect to regional topographic and cultural features is shown on Figure 2.1 1 and with respect to local features on Figure 2.12. De station is located on nree Mile Island situated in the Susquehanna Rwer upstream from York llaven Dam. 21.2 SITE DESCRIPTION Figure 2.13 shows the site marked to indicate the area owned or controlled by Met Ed The site consists of 814 acres owned by Met Ed. Met Ed owns a'l of Three Mile Island and all but a small portion on the southern end of Shelley Island. He c clusion area includes portions of Three Mile Island, the twer surface around it, and a Met Ed owned portion of Shelley Island, as shown on Figure 2.1 3. 2.1.3 POPULATION AND POPULATION DISTRIBLrflON ne population and population demographics are given in Section 2.2 of the TM1 Umt i UFSAR. This information is updated as appropriate with the Urut 1 UFSAR updates.
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4 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES The neartry industrial, transportation, arxl military facilities are described in Clapter 2 of the TM1 Unit i UFSAR 1 1 i d i i 1 I l l l 2.2-1 UPDATE 2 - AUGUST 1997
1 0 2.3 h!ETEOROLOGY
%c meteorology for the Bree hiile Island site is given in Section 2.5 of the Thil Unit 1 UFSAR. Since Thil 1 and Thil 2 are on the same site, the rneteorological information for the two units is the same.
He Thil 1 UFSAR is updated periodically in accordance with NRC regulations and contains the currently applicable meteorological information for the Thil site. 4 r (~ 2.3-1 UPDATE 2 - AUGUST 1997 l w-e .--m--, ,w~ - - - -e- g- 7 - + m , - ,p - -.-.
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Lj\ t 2.4 IlYDROLOGIC ENGINEERING 2.4.1 IlYDROLOGIC DESCRIPTION 2.4.1.1 Site and Facihties Thil Unit 2 is located on nree Mile Island in the Susquehanna River about ten miles southeast of Harrisburg. Existing grade elevation at the site is approximately 304.0 ft. based upon U.S.G.S. datum. He site is protected from the design flood condations of 1,100,000 cfs by a system of protective dtkes surrounding Units 1 and 2. These dikes have a maximum crest elevation of 310.0 ft at the north end of the site and a maumum crest elevation of 304.0 ft. at the south enu. He site is engineered to sustain a river flow up to 1,100,000 cfs. Dunng floods exceeding 1,100,000 cfs, all of the necesntry features of the site are protected from flood and wave action associated with these floods up to and including the Probable hiaximum Flood 2.4.2 FLOODS Re description of flooding conditions which were considered in the design of the site and facilitics on the site is
;iven in Section 2.4 of the Thil 2 FSAR. nese floodmg conditions include the Probable hiaximum Flood with p)
(G coincident wind wave actisity, the Probable hiasimum Precipitation, potential dam failures, ice floodmg, and channel diversions. 2.4.3 FLOODING PROTECTION REQUIREh1ENTS nree hiite Island Nuclear Station is situated on a portion of the island that is, under natural conditions, abo'.c the level of an "Agnes" magnitude flood. Natural topography in the main station area is above elevation 300.5 ft , the crest of the Agnes Good at approximately one million cubic feet per second. The design flood for the site was established at 1,100,000 cfs, based upon the provisional Probable hiaximum Flood established by the Anny Corps of Engineers prior to 1969. He hydraulic design of the plant protective dikes is based upon the design flood and includes an ample margin of freeboard. The dikes will be subjected to Good waters rather infrequently, since it would require a repeat of the 1964 flood of 485,000 cfs to reach the elevation of the bottom of the dikes. Even with the Agnes flood, the maximum crest was 3.5 ft. below the lowest south dike elevation of 304 ft, indicating conservatism in design in June of 1969 the Corps of Engineers issued a revised and provisional value of the Probable hiaximum Flood for the Susquehanna River at llarnsburg, which was established as 1,600,000 cfs, as the result of upstream reservoir regulation. This flood at Three hiile Island v,as established at 1,625,000 cfs and the water surface profiles on Figure 2.4 1 were extended by computation to cover this flood magnitude. Unit backmater computations established a PhtF clevation of 308.5 ft on the west side and 308 ft on the east side of the site. Due to wave action, Good orotection is designed to protect against a water level of at least four feet above the PMF water level er 312.5 ft on the west side and 312 ft on the cast side as described in Section 3.4.4. m Various components which have been protected include, but are not limited to the following locations and type ( of Good protection:
'v) 2.4 - 1 UPDATE 2 - AUGUST 1997
O a Fuel llandling Building here are no external operungs in this building requiring flocd protection The Urdt I railroad door, which serves as access to this buildmg, is made watertight. (See the Unit 1 UFSAR, Section 2.6.5.)
- b. Control Building Flood panels are provided for all ground level entrances.
- c. Auxiliary Buildmg A flood panel is provided for the gmund level entrance.
- d. Control Building Area - Access to the tendon gallery is protected by watertight enclosures and flood panels at ground level.
- c. Air intake - Located at an elevation above PMF level and a watertight hatch
- f. General Ground level doors and entrances to the Control Building Area are either water tight or are provided with flood panels. All openings that are potential leaks (e g , ducts, pipes, conduits, cable trays) are scaled.
Unit flood protection will be achieved by instituting operational procedures and actions predicated upon continuous monitonng of upstream river stages and precipitation reports through the River Forecast Center at State College. Operational procedures to establish FLOOD ALERT and EMERGENCY CLOSU1G actions are outlined in Section 2.4.6. Reference Section 7.1.4, Flood Protection Designs and criteria were established for the structures associated with flood protection facilities including ear dikes, channels, and pressure conduits He criteria are conservative and based on sound cisil engineenng practices. The facilities were constructed and will be maintained and inspected consistent with their design as integral parts of a nuclear station. The design of the facilities has been reviewed and approved by the appropriate State and Federal agencies, including the Federal Power Conunission, the Corps of Engineers, and the Water and Power Resources Board of the Commonwealth of Permsylvania, where applicable. 2.4.4 ENVIRONMENTAL ACCEPTANCE OF EFFLUENTS l l Liquid efiluents from TMI l and TMI 2 enter the middle channel of the Susquehanna River through the station discharge pipe located approximately 640 ft downstream frorn the intake structures, as shown in Figure 2.4 2. Processed liquid waste will be discharged on a batch basis. Prior to release, each batch will be sampled and analyzed to determme its radioactivity content. Based upon the activity analysis, the wastes will either be released under controlled conditions or recycled for further processing. The flowTate of waste discharge will be a function of the activity analysis and the flowTate of the discharge from the Unit I mechs.aical draft cooling towers. He ability of the Susquehanna River to disperse and dilute the station discharge stream is dependent upon the magrutude of river water flow. l 2,4-2 UPDATE 2 - AUGUST 1997
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A section of the York llaven Dam blocks the east channel of the Susqueharma Riser at nree Mile Island approximately one mile downtream from TMI 2. He York llaven Darn forms a pool extending approsimately 3 l/2 miles upstream containing a volume of about 10,000 acre feet. As long as the river Dow is 20,000 cfs or less, all the Dow discharges through the York llaven liydro Plant tailrace into the lower section of Conewago Falls. When the river flow is above 20,000 cfs, the excess flow spills over the portion of the main dam upstream of the headrace wall and flows down Grough the Conewago Falls joining the flow from the tailrace at the foot of the dam; the full river flow then continues through the lower section of the falls. The exact extent of the midng of TMINS emuents with the river depends on such factors as station discharge flowrate and the river flowrate. In 1980, an experiment was conducted which tracked the dispersion and c'ilution of a dye from the TMINS discharge. His study showed that plant discharge water and Susquehanna diver water are typically 99% or grea'er mixed before intake by dowmtream users. TMINS emuent seleases are diluted by the effluent of the mechanical draft cooling towers by factors typleally ranging from 167 times to 7600 times, dependmg upon the water intake from the river. The diluted radiological waste is further diluted by the flow of the Susqueharma River. At lvw flow (1700 cfs), the nver would dilute the effluent by a factor of 20. Ilowever, at normal river flow (34,000 cfs), the dilution from the river would be 403 times. For emnent releases a muumum flow rate of 5000 gpm is maintained from the Mechanical DraA Coolmg Towers, however,8,000 gpm MDCT flow is alwayr mintained during IWrS and/or IWFS releases (continuous releases). Dilution credit up to 38,000 gpm is taken for Mechanical Draft Cooling Tower effluent p rates in FSAR calculations.
, t All users of surface water dowmtream of urce Mile Island are also dowstream of York llaven Dam.
Therefore, it is assured that mixing of station emuent and river water flow will occur prior to use. 2.4.5 GROUNDWATER nree Mile Island Nuclear Station is located in the Triassic lowland of Pennsylvania, a region often referred to as the Gettysburg Basin. He island was formed as a result of fluvial deposition by the Susquehanna River. It is composed of sub-rounded to rounded sand and gravel, containing varying amounts of silt and clay. Soil depths vary from approximately 6 feet at the south of the island to about 30 feet at the center of the island. The site is underlain by Gettysburg shale which is at approximately 277 feet elevation. There are two difTerent water bearing zones at TMINS. One is comprised of the unconsolidat.J materials overlying the Gettysburg shale (bedrock), and the other is comprised of the bedrock. Permeabilities in geologic materials on nifNS vary, however, gmundwater discharges into the Susquehanna River and does not communicate with off site groundwater supplies. O 2,4-3 UPDATE 2 - AUGUST 1997
Ilydrostatic pressure of the water table on the east and west shores of the river should prevent the island O groundwater and the station discharge f om contmunications with onshore groundwater. Therefore, groundwater effluents from ThflNS cannot impact the quality of groundwater off site. Additionally, the tntium concentrations in the ThilNS groundwater are well below 10 CPR 20 regulatory lirnits and will not adversely affect the Susquehanna River A more thorough desenption of the groundwater characterirties and related features are given in Section 2.413 of the Umt 2 f SAR. 24.6 Eh1ERGENCY OPERATIONS REQUIREhtENTS 2.4.6.1 Flood Protection Although the flood protection design features of the station are based on the Ph1F, the ernergency operational procedures are based on forecasts received from the Federal State River Forecast Center, Nstional Weather Senice, State College, Pennsylvania. Communications are normally by phone or through civil defense radio as a backup. The emergency procedure for the station will go into effect when the Federal-State River Forecast Center forecasts tn 36 hours, a river flow of 350,000 cfs A flood ALERT will be initiated when a 36 hour forecast of 640,000 cfs is received and a Eh!ERGENCY CLOSURE when a 36 hour forecat of 940,000 cfs or greater is received Dunng the EhtERGENCY CLOSURE, flood panels will be moved mto place. 2.4-4 UPDATE 2 - AUGUST 199
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2.5 GEOLOGY AND SEISMOLOGY The geology and seismology fu the Three Mile Island site has been reviewed and acceptrA by the NRC based on the infonnation presented in Section 2.5 of the Unit 2 FS AR. 2.5 1 UPDATE 2 - AUGUST 1997 Iuii ull II
J O l i CIIAPTER 3 DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS
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CllAPTER 3
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DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMpONE?US ( TABLE OF CONTENTS i SECTION HILE PAGE l 3,1 REGULATORY CONFORMANCE 3.1 1 3.1.1 CONFORMANCE WITil 10 ClK PART 50 3.11 GENERAL PROVISIONS 3.1.1.1 10 CTR $0.1. Basis. Purpose, & Procedures Applicable 3.11 3.1.1.2 10 CFR $0.2 Definitions 3.11 3.1.1.3 10 CFR $0.3. Interpretations 3.11 3.1.1.4 10 CFR 50.4. Communicauons 3.11 3.1.1.4a 10 CIR $0.S . Deliberate Misconduct 3.12 3.1.1.5 10 CII 50.7. Employee Protection 3.12 3.1.1.6 10 CFR 50.8. Infonnation Collection 3.12 Requirements: OMB Approval 3.1.1.6a 10 CFR 50.9. Completeness & Accuracy ofInformation 3.12 U REQUIREMEtU OF LICENSE. EXCEPTIONS 3.1.1.7 10 CFR 50.10 License Required 3.12 5.1.1.8 10 CFR $0.11. Exceptions & Exeruptions from 3.12 Licensing Requirements 3.1.1.9 10 CFR 50.12 Specific Exemptions 3.12 3.1.1.10 10 CTR 50.13 Attacks & Destructive Acts by Enemies 3.12 of the United States; and Defense Activities CLASS!TICATION AND DESCRIPTION OF LICENSES 3.1.1.11 10 CFR 50.20 Two Classes of Licenses 3.1 2 3.1.1.12 10 CFR $0.21 - Class 104 Licenses; for Medical 3.1 3 Therapy and Research and Development Facilities
/~'s i UPDATE 2 - AUGUST 1997
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CHAPTER 3 . t TABLE OF CONIT.NTS (Cont'd) SliCDOB IHLE PAGE i 3.1.1.13 . 10 CFR 50.22. Class 103 Licenses;for 3.13 l Comunercial & Industnal Facilides 3.13 ! 3.1.1.14 10 CFR 50.23 Constmetion Permits APPLICATIONS FOR LICENSES, FORM. CONTENTS. INELIO1BILITY OF CERTAIN APPLICANTS ! 3.1.1.15- 10 CPR 50.30. Filing of Applications for 3.13 l
- Licenses; Oath or Amtmation 3.1.1.16 10 CFR 50.31 Con 61aing Applicadons 3.13 }
10 CFR 50.32 . Elindnation of Repetition 3.13 i 3.1.1.17 e i i 3.1.1.18 - 10 CFR 50.33 . Centents of Applicadons, 3.13 > General Information i 3.1.1.19_ 10 CFR 50.33a . Information Requested by the 3.13 . t Attorney General for Antitrust Resiew 3.1.1.20 10 CFR 50.34 Contents of Applications; 3.14 ' TechnicalInformation , 3.1.1.21 10 CFR 50.34a . Design Objectives for Equipment to Control Releases 3.1 9 ; of Radioactive Material in Emuents. Nuclear Power Reactors ; 3.1.1.22 19 CFR 50.35 . lasuance of Construction Permits 3.1 9 ! 3.1.1.23 10 CFR 50.36 - To-hnical Specifications 3.1 9 i- 3.1.1.24 10 CFR 50.36a . Technical Specifications on 3.19 i Emuents from Nuclear Power Reactors 3.1.1.25 10 CFR 50.36b . Emironmental Conditions 3.1 10 - 3.1,1.26 10 CFR 50.37 - Agreement Limiting Access to 3.1 10 Fxstricted Data ; L i s 1 r
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_ ,_. ..__,_,m., _mm.- me, , . - , , _ _ , _y,r ,, ,,,,._,,..,y.w, . , , - , ., . , . _ ,,,_w., ,,,._,y,,_mi ,
i i CilAP1TiR 3 [ TA13LE OF CONTENTS (Cont'd) EliO108 1111 1 1%DL 3.1.1.27 10 Cm 50.38 Inclipbility of Certain Applications 3.110 31.128 10 CFR 50.39 Publi: Inspection of Applications 3.1 11 STANDARDS FOR LICENSES AND CONSTRUCTION PERMITS 3.1.1.29 10 CR 50.40 Common Standards 3.1 11 3.1.1.30 10 CFR 50.41 Additional Standards for Class 104 Licenses 3.110 3.1.1.31 10 CFR 50.42 Additior al Standards for Class 103 Licenses 3.1 11 3.1.1.32 10 CFR 50 43 Additional Standards and Prmisions Affecting 3.1 11 Class 103 Licenses for Commercial Power 3,1.1.33 10 CFR 50.44 Standards for Combustible Oas Co trel Sptem 3.1 11 in Light Water Cooled Power Reactors 3.1.1.34 10 CFR 50 45 Standards for Construction Pernuts 3.1 11 3.1.1.35 10 CFR 50.46 Acceptance Cnteria for Emergency 3.1 12 Core Cooling Systems for Light Water Nuclear Power Reactors (/ 3 1.1.36 10 CFR 50 47 Emergency Plans 3.1 12 3.1.1.37 10 CFR 50.48 Fire Protection 3.1 12 3.1.1.38 10 CFR 50 49 Emironmental Qualification of 3.1 12 Electric Equipmeat important to Safety for Nuclear Power Plants ISSUANCE, LIMITATIONS, & CONDITIONS OF LICENSES & CONSTRUCTION PEI8 TITS 3.1.1.39 10 CFR 50.50 Issuance of Licenses and Construction Permits 3.1 13 3.1.1.40 to CFR 50.51 Duration of License, Renewal 3.1 13 N {O iii UPDATE 2 - AUGUST 1997
CHAPTER 3 , l
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, TABLE OF CONTENTS (Cont'd) SECTION IHLE P. AGE 3.1.1.41 10 Cm 50.52. Combining Licenses 3.1 13 3.1.1.42 10 CFR 50.53 Jurisdictional Limitations 3.113 3.1.1.43 10 CFR 50.54 Conditions of Licenses 3.113 3.1.1.44 10 CFR 50.55. Conditions of Construction Permits 3.1 17 3.1.1.45 10 CFR 50.55a . Codes and Standards 3.117 3.1.1 46 10 CFR 50.56. Conversion of Construction Permit to License. 3.1 18 or Amendu.cnt of License 3.1.1.47 10 CFR 50.57. Issuance of Operating License 3.118 3.1.1.48 10 CFR 50.58. llearings and Report of the Audsory 3.118 Committee on Reactor Safeguards 3.1.1.49 10 CFR 50.59. Changes, Tests and Experiments 3.119 3.1.1.50 10 CFR 50.60 Acceptance Criteria for Fracture 3.1 19 Prevention Measures for Light Water Nuclear Power Reactors for Nonnal Operation ( 3.1.1.51 10 CFR 50.61 Fracture Toughness Requirements 3.120 for Protection Against Pressurized Thermal Sho:k Events 3.1.1.52 10 Cm 50.62. Requirements for Reduction of Risk 3.119 from Anticipated Transients without Scram (ATWS) Events for Light Water Cooled Nuclear Power Plants 3.1.1.53 10 CFR 50.63 Loss of Alternating Current Power 3.120 3.1.1.54 10 CFR 50.64 . Limit tions on the Use of Highly Eruiched 3.1 20 Uranium QIEU) in Domestic Non. Power Reactors 3.1.1.54a 10 CFR 50.65. Requirernents for Monitoring the Effectivenets 3.120 of Maintenance at Nuclear Power Plants" INSPJICTION, RECORDS, REPORTS, NOTIFICATIONS 3.1.1.55 10 CFR 50.70. Inspections 3.120 ,f%.; iv UPDATE 2 - AUGUST 1997
I l l \ CHAPTER 3 O TABLE OF CONTENTS (Cont'd) t SECTION TITLE PAGE 13.1.1.56 10 CFR 50.71 - Maintenance of Records, Making of Reports 3.121 3.1.1.57 10 CFR 50.72 Imnediate Notification Requirements for Operating 3.1 21 Nuclear Power Reactors 3.1.1.58 10 CFR 50.73 - Licensee Event Report System 3.121 3.1.1.58a 10 CFR 50.74 - Notifications of Change in Operator or 3.1 21 Senior Operator Status" 3.1.1.59 10 CFR 50.75 - Reporting and Recordkeeping for 3.1 22 Deconunissioning Planning US/lAEA SAFEGUARDS AGREEMENT 3.1.1.60 10 CFR 56.78 -Installation Information and Wrification 3.1 22 TRANSFERS OF LICENSES-CREDITORS RIGIITS SURRENDER OF LICENSES 3.1.1.61 10 CFP 50.80 - Transfer of Licenses 3.1 22 3.1.1.62 10 CFR 50.81 - Creditor Regulations 3.1 22 (/ 3.1.1,63 10 CFR 50.82 - Applications for Termination of Licenses 3.1 22 AMENDhENT OF LICENSE OP CONSTRUCTION PERMIT AT REQUEST OF HOLDER 3.1.1.64 10 CFR 50.90 - Application for Amendment of 3.1 22 License or Construction Permit 3.1.1.65 10 CFR 50.91 - Notice for Public Cornment; 3.1-22 State Consultation 3.1.1.66 10 CFR 50.92 Issuance of Amendment 3.1-22 C C v UPDATE 2 - AUGUST 1997
CHAlrfER 3 ( V o) TABLE OF CONTENTS (Cont'd) REVOCATION, SUSPENSION, MODIFICATION, AhENDMEC OF 13 CENSES AND CONSTRUCTION PERMITS, EMERGENCY OPERATIONS BY THE COMMISSION SECTION TITLE PAGE 3.1.1.67 10 CFR 50.100 Revocation, Suspension. Modification of 3.1 23 Licenses and Constru: tion Permits for Cause 3.1.1.68 10 CFR 50.101 Retaking Possession of Special Nuclear Matenal 3.1-23
$.1.1.69 10 CFR 50.102 Commission Order for Operation After 3.1-23 Revocation 3.1.1.70 10 CFR 50.103 Suspension and Operation in War 3.1 23 or National Emergency BACKFITTING 3.1.1.71 'O CFR 50.109 - Backfitting 3.1 23 ENFORCEhEh7 3.1.1 72 10 CFR 50.110 - Violations 3.1 23 3.1.1.73 10 CFR 50.111 Criminal Penalties" 3.1 23 v 3.1.1,74 10 CFR 50.120 - Training and Qualifications of Nuclear Power 3.1 23 Plant Personnel 3.1.2 GENERAL DESIGN CRITERIA 3.1 24 3.1.2.1 Criterion 1 - Quality Standards and Records 3.1 24 3.1.2.2 Criterion 2 - Design Bases for Protection Against Natur:u Phenomena 3.1 24 3.1.2.3 Criterion 3 - Fire Protection 3.1 25 3.1.2.4 Criterion 4 - Emironmental and Missile Design Bases 3.1-25 3.1.2.5 Criterion 5 - Sharing of Strustures, Systems, and Components 3.1-26 t
V si UPDATE 2 - AUGUST 1997
CHAP'IER 3 - j- TABLE OF CONTENTS (Cont'd) s
= SECTION IIILE EAD.E _-
d-3.1.2.6 - Criterion 10 Reactor Design _ 3.126 3.1.2.7 Criterion 11 - Reactor inherent Protection - 3.126 r 3.1.2.8 - Criterion 12 - Suppression of Reactor Power Oscillations _ 3.1 26 ; 3.1.2.9 - Criterion 13 -Instrumentation and Control 3.127
- 3.1.2.10 - Criterion 14 - Reactor Coolant Pressure Boundary 3.1 27 3.1.2.11 Criterion 15 - Reactor Coolant System Design 3.127 1 3.1.2.12 Criterion 16 Containment Design 3.1-27 3.1.2.13 - Criterion 17 - Electric Power Systems 3.128 3.1.2.14 Criterion 18 -Inslution and Testing of Electric Power Systems 3.1 29
? 3.1.2.15 Criterion 19 - Control Room 3.1 29 3.1.2.16 Criterion 20 - Protection System Functions 3.1 29 3.1.2.17 Criterion 21 - Protection System Reliability and Testability 3.1 30
- 3.1.2.18 Criterion 22 Protection System Independence 3.1 30 3.1.2.19 Criterion 23 - Protection S3 stem Failure Modes 3.130 3.1.2.20- Criterion 24 - Separation of Protection and Control Systems 3.130 3.1.2.21 Criterion 25 - Protection System Requirements 3.1 31 for Reactisity Control Malfunctions 3.1.2.22 Criterion 26 - Reactivity Control System - Redundancy and Capability 3.131 3.1,2.23 Criterion 27 - Combined Reactivity Control Systems Capability 2.1 31 3.1.2.24 Criterion 28 ReactivityLimits 3.1 31 t
i vii UPDATE 2 - AUGUST 1997
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' CHAl'IER 3
- TABLE OF CONTENTS (Cont'd)
SECTION TTTLE PAGE 3.1.2.25 Criterion 29 Protection Against Anticipated 3.132 Operational Occurrences 3.1.2.26 Criterion 30 - Quality of Reactor Coolant 3.1 32 Pressure Boundary c 3.1.2.27 Criterion 31 Fracture Prevention of Reactor 3.1 32 Coolant Pressure Boundary
-3.1.2.28 Criterion 32 Inspection of Reactor Coolant 3.1 33 Pressure Boundary 3.1.2.29 Criterion 33 Reactor Coolant Makeup 3.133 3.1.2.30 Criterion 34 Residual Heat Removal 3.1 33 1- 3.1 33
-3.1.2.31 Criterion 35 Emergency Core Cooling 3 1.2.32 Criterion 36 Inspection of Emergency Core Cooling System 3.1 34 3.1.2.33 Criterion 37 - Testing of Emergency Core Cooling System 3.1 34 3.1.2.34' Criterion 38 Containment Hea, Removal 3.1-34 3.1.2.35 Criterion 39 Inspection of Containment Heat R. moval System 3.1 35 Criterion 40 Testing of Containment Heat Removal System 3.1-35 f 3.1.2.36 3.1.2.37 Criterion 41 - Containment Atmosphere Cleanup 3.1 35 3.1.2.38 Criterion 42 -Inspection of Containment Atmosphere 3.1-36 Cleanup Systems 3.1.2.39 Criterion 43 - Testing of Containment Atmosphere Cleanup Systems 3.1-36 3.1.2.40 Criterion 44 Cooling Water 3.1-36
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- TABLE OF CObrTENTS '(Cont'd)
SECTION IIIII; PAGE 3.1.2 41 ' Criterion 45 + Inspection of Cooling Water System - 3.1 37 3.1.2.42 Criterion 46 Testing of Cooling Water System - 3.137 3.1.2.43 - Criterion 50 - Containment Design Basis 3.1 37
! 3.1.2.44 . Criterion $1 Fracture Prevention of 3.138 Containment Pressure Boundary 3.1.2.45 Criterion 52 Capab:lity for Containment 3.148 Lealage Rate Testing 3.1.2.46 Criterion 53 Provisions for Ceatainment Testing and In.,pec ion -3.1 38 _;
3.1.2.47 Criterion 54 Piping Systems Penetrating Containment ' 3.1-39 3.1.2.48' Criterion 55 - Reactor Coolant Pressure Boundary 3.139 Penetrating Containment . 3.1.2.49 Criterion $6 Pnmary Containmer.t Isolation 3.140 3.1.2.50 Criterion 57 Closed System isolation Valves 3.140 ( 3.1.2.51 Criterion 60 Control ofReleases of Radioacthe Materials to the Emironment 3.1-41 3.1.2.52 Criterion 61 - Fuel Storage and Handling and Radioactivity Control 3.1-41 3.1.2.53 Criterion 62 - Prevention of Criticality in Fuel Storage and Handling 3.1 41 3.1.2.54 Cnterion 63 Monitoring Fuel and Waste Storage 3.142 3.1.2.55 Criterion 64 - Monitoring Radioactidty Releases 3.1-42 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AhT) 3.2-1 COMPONENTS 3.2.1- SEISMIC CLASSIFICATION 3.2 1 < f 1 i t C\ l l
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. 4 CHAPTER 3 c TABLE OF CONTENTS (Cont'd) ,
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SECTION TITLE PAGE 4 3.2.1.1 . _ - Seismic CategoryI 3.2 1 3.2.1.2 PDMS Category !! '3.2 2
~3.2.1.2.l~ Design Basis 3.2 2 .
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- 3.2.1.2.2 PDMS Category 11 Structures 3.2-2 3.2.1.2.3 . PDMS Category Il Systems. 3.2 2 ,
3.2.2 SYSTEM QUALITY GROUP Cl.ASSIFICATION 3.2 3 3.2.2.1 Identification of Safety Related Systems and Components 3.2-3 3.3 - WIND AND 'IDRNADO LOADINGS -3.3 1 1 -. 5 i 3.3.1 WIND LOADINGS 3.3 1 3.3.1.1 Design Wind Velocity 3.3 1 i 3.3.2 ' TORNADO LOADINGS 3.31 3.3.2.1 ' Applicable Design Parameters 3.3 1 3.3.2.2 Tornado Missiles 3.3-1 3.4 - WATER LEVEL (FLOOD) DESIGN 3.4 1 3.4.1 FLOOD ELEVATION 3.4 1 3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD - 3,41 CALCULATIONS 3.4.3 FLOOD FORCE APPLICATION 3.4 1 e 3.4.4 - FLOOD PROTECTION 3.4-1 3.5 MISSRE PROTECTION CRITERIA 3.5 1 3.5.1 MISSILE LOADINGS AND BARRIERS 3.5 1 ~
'3.5.2 - hDSSILE SELECTION - 3.5-1 O
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CHAP 7ER 3 I TABLE OF CO! GENTS (Cont'd) dj SECTION TITLE PAGE I- :3.5.3 SELECIED MISSILES 3.51 l 3.5.3.1 Tornado Generated Missiles 3.51 3.5.3.2 . Aircraft impact - -3.5l'
-3.5.4 BARRIERDESIGN PROCEDURES 3.5 2 3.5.4.1 Overal: Structural Effect ~3.5 2 i -
3.5 2 3.5.4.2 Missile Penetration (Localized Effect) 3.5.5 MISSILE BARRIER FT.ATURES 3.5-2 3.6 SEISMIC DESIGN 3.6 1 3.7 DESIGN OF PRINCIPAL BUILDING STRUCTURES 3,7 1 f 3.7.1 CONTAINMENT BUILDING 3.7 1 3.7.1.1 Structure Description -3.7-1 3.7.1.2 - Liner Plate and Penetrations 3.7 1 3.7,2 OTHER PRINCIPAL STRUCTURES . 3.7 3
- 3.7.2.1 Description of Structures 3.7 3
'3.7.2,1.1 Auxiliary Building 3.7-3 3.7.2.1.2 FuelHandling Building 3.7-3
-3,7.2.1.3 Control and Services Buildings 3.7 3 3.7.2.1.4 Control Building Area 3.7-4 r 3.7.2.1.5 Air intake Tunnel 3.7-4 i
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- 1 CHAPTER 3 - i e
-TABLE OF CONIENTS (Cont'd)
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. LIST OF TABLES i- ,
TABLE NO. TITLE PAGE i 3.21- PDMS CATEGORY 11 SYSTEMS 3.2-4 -l i- !. 3.3 1 TORNADO GENERADID MISSILES 3.3 2
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3.7-1 MODIFIED CONTAINMENT PENETRATIONS 3.74 } F Y r t t- - 4 4 ? i i i t 1-2; 4 4 4 i i i E 1_ 4 E i a:
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1 3 .- _. _ ( i ( -- xii UPDATE 2 - AUGUST 1997
-*vwmT4s-y
- 5, e e W? *s"Nymw- W' - e"' TT*MM 6 spa's yv-** -+ +tg rgra -
CHAPTER 3 TABLE OF CONTEhTS (Cont'd) f LIST OF FIGURES FIGURE NO. 'JIILE 1%QE 3.71 REACTOR BUILDING GENERAL LAYOUT 3.74 1, 3.7 2 REACTOR BUILDING PERSONNEL AND 3.7 5 EQUIPhENT ACCESS OPENINGS DETAIL 3.7-3 CONTAINhENT WALL PENETRATION DETAILS 3.7-6 3.74 SECTION TIIROUGH11E PLANT 3.7-7 STRUCTURES "A A" 3.7.5 SECTION TIIROUGli TIE PLANT 3.7 8 STRUCRTRES 'B-D" 3.7 6 SECTION TIIROUGl! TIE PLANT 3.7-9 STRUCTURES "C C" ?: 'D-D"
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[ xiii UPDATE 2 - AUGUST 1997
. CHAPTER 3
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DESIGN CRITERIA - STRUCTURES, SYSTEMS, AND COMPONENTS
. J 31 REGULATORY CONFORMANCE 3.1.1 CONFORMANCE WITH 10 CFR Pan 50 Three Mile Island Nuclear Station ' Unit 2 was originally designed to conform to the Regulations of 10 CFR Pan 50, including the General Design Criteria of Appendix A. On March 28,1979, the unit i
experienced an accident which severely damaged the reactor core. Subsequently, the core was removed and shipped off-site. The removal of the reactor core and the revision of the license to a non-operating license have changed the function of the facility from an operating nuclear power plant to one of . management and maintenance. Dese characteristics substantially alter the applicabili'y and the requirement for nree Mile Island Unit 2 conformance with the regulations of 10 CFR Pan 50. The degree and manner of addressing those regulations which are applicable to Three Mile Island Unit 2 during Post Defueling Monitored Storage are described in the following sections. In addition, based on this evaluation, a request for relief from the insenice inspection requirements of 10 CFR 50.55a and an exemption from the requirements of 10 CFR 50.60 are also provided. The regulations which have been
- reviewed are those wisch were revised as of knuary 1,1995.
3.1.1.1 10 CFR 50.1 Basis, purpose, and procedures applicable, Article 50.1 describes the basis, purpose, and procedures of 10 CFR Pan 50 and gives notice that persons may be individually subject to NRC Enforcement Actions for violations of Article 10 CFR 50.5 s " Deliberate Misconduct." No exceptions are taken to the provisions of this article. 1 3.1.1.2 10 CFR 50.2 - Defmitions Anicle 50.2 proWes defmitions of terms used throughout 10 CFR 50. No exceptions are taken to the provisions of this anicle. 3.1.1.3 10 CFR 50.3 - Interpretations. Anicle 50.3 delegates the responsibility for interpretations of 10 CFR 50 to the General Counsel. No exceptions are taken to the provisions of this anicle. 3.1.1.4 10 CFR 50.4 - Communications. Article 50.4 describes communications requirements for nuclear power plants. No exceptions are taken to the provisions of this article. r l'
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I-3.1-1 UPDATE 2 - AUGUST 1997 i l . . .
3.1.1.4a .10 CFR 50.5 - Deliberate Misconduct Article 50.5 prohibits deliberate misconduct as dermed in this article by Licensees, contracters, sut - contractors, and their respective employees and provides for appropriate enforcement action. No exceptions are taken to the provisions of this article. 3.1.1.5 10 CFR 50.7 Employee Protection Anicle 50.7 describes provisions relating to tir protection of employees at nuclear facilities. No exceptions are taken to the provisions of this anicle. 3.1.1.6 10 CFR 50.8 - Information Collection Requirements: OMB Approval Anicle 50.8 states that the OMB has approved NRC information requirements. No exceptions are taken to the provisions of this article. 3.1.1.6a *10CFR50.9 - Completeness and Accuracy ofInformation" Anicle 50.9 provides requirements for licensees to maintain and to proside the NRC complete and accurate information. No exceptions are taken to the provisions of this anicle. 3.1.1.7 10 CFR 50.10 - License Required Article 50.10 describes the requirements and restrictions of a license obtained pursuant to 10 CFR Part
- 50. No exceptions are taken to the provisions of this article.
3.1.1.8 10 CFR 50.11 - Exceptions and Exemptions from Licensing Requirements Article 50.11 describes cenain agencies of the Federal Government not required to obtain a license pursuant to 10 CFR 50. No exceptions are taken to the provisions of this article. 3.1.1.9 10 CFR 50.12 - Specific Exemptions Paragraph 50.12 allows the Commission to grant exemptions from the requirements of regulations. No exceptions are taken to the provisions of this anicle. 3.1.1.10 10 CFR 50.13 - Attacks and Destructive Acts by Enemies of the United States; and Defense Activities Anicle 50.13 exempts Licensees from having to provide design features to protect a facility from sabotage and other destructive acts. No exceptions are taken to the pro isions of this anicle. 3.1.1.11 10 CFR 50.20 -Two Classes of Licenses Article 50.20 provides for two classes oflicenses; class 103 and 104. TMI-2 currently possesses a class 103 license. O 3.1-2 UPDATE 2 - AUGUST 1997
3.1.1.12 10 CFR 50.21 - Class 104 Licenses; for hiedical Therapy and Research and Development (3 Facilities
- i. }
V The class oflicense described in Article 50.21 does not apply to Thil 2. 3.i.2.13 10 CFR 50.22 - Class 103 Licenses; for Commercial and Industrial F.cilities The class oflicense described in Ar;icle 50.22 is applicable to Thil 2 and is the class oflicense under which the facility is currently licensed. Thil 2 will continue to be licensed as a Class 103 facility but the license has been modified to recognize the unique PDhtS condition of the plant. 3.1.1.14 10 CFR 50.23 Construction Permits Article 50.23 describes to whom and when a construction permit will bc issued. No exceptions are taken to the provisions of this article. 3.L 1.15 10 CF 50.30 - Filing of Applications for Licenses; Oath or Affirmation Article 50.30 desen ,es requirements for filing for amendments to a facility license. No exceptions are taken to the provisions of this article. 3.1.1.16 10 CFR 50.31 - Combining Applications Article 50.31 states that applications for licenses may be combined No exceptions are taken to the provisions of this article. O Q 3.1.1.17 10 CFR 50.32 - Elimination of Repetition Article 50.32 states that applications may reference information in other applications. No exceptions are taken to the provisions of this article. In fact, the Th112 FSAR and the Thil ! USAR are referenced in the application. 3.1.1.18 10 CFR 50.33 - Contents of Applications; General Information Article 50.33 specifies general requirements for applications. No exceptions are taken to the provisions of this article. 3.1.1.19 10 CFR 50.33a -Information Requested by the Attomey General for Antitrust Resiew Article 50.33a specifies infonnation required to be filed by applicants who are filing construction permit applications for nuclear power reactors, uranium enriclunent or fuel reprocessing plants. No exceptions are taken to the provisions of this article. [} 3.1 - 3 UPDATE 2- AUGUST 1997
3.1.1.20 .10 CFR 50.34 - Contents of Applications; Technical Information Paragraph 50.34(a) describes requirements for a Preliminary Safety Analysis Report (PSAR) to be filed with an application for a construction permit. Since TM12 has completed the construction permit process and receited a license, the requirements of this paragraph are no longer applicable to TMI 2. Paragraph 50.34(b) requires that each application for a license to operate a facility shall include a Fmal Safety Analysis Report (FSAR) and describes the required contents of that FSAR. This paragraph requires the FSAR to include information that describes the facility, presents the design basis and presents a safety analysis of the structures, systems and components of the facility as a whole. The original TMI-2 liceme application included the required FSAR which addressed the requirements of paragraph 50.34(b). Although TMI 2 has addressed the requirements of paragraph 50.34(b) in the FSAR, it is useful to review each of the FSAR requirements in the context of Post Defueling Monitored Storage. A paragraph by paragraph review of the content requirements of the FSAR, considering the circumstances which exist during Post Defueling Morutored Storage, follows: 50.34(b)(1) Paragraph 50.34(b)(1) requires the FSAR to include all current information, such as the results of emironmental and meteorological monitoring programs, which has been developed since issuance of the construction permit, relating to site evaluation factors identified in Part 100 of this chapter. His information was included in the FSAR and is bounding for Post Defueling Monitored Storage. 10 CFR Part 100 evaluation factors generally include such assumptions as a substantial meltdown of the reactor core with subsequent release of appreciable quantities of fission products for site evaluation. While these kinds of assumptions were used in the initial siting evaluation of TM1-2, *. hey do not apply during Post Defueling Monitored Storage. Due to the non operating and defueled status ofTMI 2, a major nuclear event can no longer occur at TM12. In order to assure that TM12 does not present any significant risk to tie public during PDMS,10 CFR 50 Appendix 1 release guidelines (see S AR Table 8.1-4) have been selected as the limiting criteria for the evaluation of routine and unanticipated releases. %erefore, although the evaluation criteria of paragraph 50.34(b)(1) are no longer directly applicable during PDMS, comparable evaluatiom have been reviewed using the linuting criteria specified above. In addition, TM1 1 has in place an extensive Radiological and Emironmental Monitoring Program for the TMl site. Since TMI 2 is on the same site and will utilize relevant site information (i.e., the ODCM), TMI-2 will utilize the latest applicable site-related emironnantal and meteorological information. 50 34(b)(2) Paragraph 50.34(b)(2) requires descriptions and analysis of structures, systems and components. Although the specific requirements of this paragraph cannot be complied with as written due to the unique condition of TMI-2 during PDMS, the intent of this paragraph has been addressed by providing descriptions of the structures, systems and e O 3.1 - 4 UPDATE 2 - AUGC i 1997
components that provide necessary protective functions during PDMS. *lhe primary protective (q ,J
)
function of structures, systems, and components during PDMS is the isolation from the environment of the contamination which remains at TM12. 50.34(b)(2)(i) Paragraph 50.34(b)(2)(i) requires the discussion of several items (e.g., reactor core, reactor coolant system, emergency systems) for nuclear reactors. Due to the non-operating and defueled status of TMI-2 during PDMS, the facility is no longer a " nuclear reactor." Although many of the items referenced in this paragraph will not exist or have no function during PDMS, the intent of this paragraph has been addressed by presiding descripdons of those items which are required during PDMS, See the discussion of paragraph 50.34(b)(2)(ii). iQ14(b)(2)(ii) Paragraph 50.34(b)(2)(ii) requires the discussion of a number ofitems (e g., ventilation systems, control systems, waste handling) for facilities other than nuclear reactors. Although TM12 was originally licensed as a nuclear reactor, during PDMS the facility will not function as a nuclear reactor. Due to the unique condition of TMI-2 during PDMS, the provisions described in this paragraph more accurately portray requirements for TMI 2 than does paragraph 50.34(b)(2)(i). The intent of the provisions of this paragraph, as they relate to TMI-2 has been addressed in this document with the additional consideration that TMI 2 was originally licensed as a nuclear power reactor. t 50 34(b)(3) Paragraph 50.34(b)(3) requires that the kinds of radioactive materials produced and the means for controlling and limiting effluents and exposures be described. Although there will be no radioactive materials produced during PDMS as would normally occur at an operating power reactor, there will be some radioactive waste generated as well as the residual contanunation that remains within the faci lity. Although the specific requirements of this paragraph cannot be complied with as wntten due to the unique condition of TMI 2 during PDMS, the intent of the provisions of this paragraph has been addressed by prosiding descriptions of the kinds of radioactive materials which remain at the facility and the means for controlling and limiting effluents and exposures to these materials to within the limits of 10 CFR 20. 50.34(b)(4) Paragraph 50.34(b)(4) requires a fmal analysis and evaluation of the structures, systems and components which relate to the protection of the public from the consequences of normal operation, transients, and accidents. Although the specifics of these requirements as given in paragraph 50.34(b)(4) do not apply for PDMS, the intent of these requirements has been addressed for the limited number of postulated events and the insignificant risk to public health and safety has been demonstrated. v 3.1 - 5 UPDATE 2- AUGUST 1997
l iaragraph 50.34(b)(4) also requires an analysis and evaluation of ECCS cooling performance following postulated loss of coolant accidents in accordance with the requirements of 10 CFR 50.46. As 10 CFR 50.46 does not apply to Thil-2 in its current defueled condition, an analysis of ECCS cooling performance is not provided. 50.34(bMS) Paragraph 50.34(b)(5) requires the description and evaluation of programs to resolve safety questions identified at the construction permit stage. Rese requirements were addressed in the FSAR and do not apply to Thil 2 during PDMS. }0.34(bM6Vi) Paragraph 50.34(b)(6)(i) requires information conceming the applicant's organizational structure, allocations of responsibilities and authorities, and personnel qualifications requirements. Although the actual orgamzation and responsibilities will be substantially different than that of a normally operating power plant, the requirements of this paragraph as they apply to Thil-2 during PDhtS have been addressed in Section 10.5. 50.34(by5)(ii) Paragraph 50.34(b)(6)(ii) requires a description of the managerial and admimstrative controls used to satisfy applicable requirements of 10 CFR Part 50 Appendix B. Appendix B establishes quality assurance requirements for actisities affecting the safety-related functions of those structures, systems and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. During PDMS, Thil 2 has no structures, systems or components classified as safety-related and, therefore, the requirements of paragraph 50.34(b)(6)(ii) and Appendix B do not apply to Thil-2 during PDhtS. Due to the unique condition of Thil-2 during PDhiS, the specific requirements of this paragraph are not directly applicable; however, Thil-2 has addressed the intent of this paragraph by establishing a quality assurance program similar to that described in Appendix B for activities such as radioactive waste shipping ond conformance with 10 CFR 20 requirements as well as for all activities which arejudged to be within the intent of this paragraph. 50 34(bM6Viii) Paragraph 50.34(b)(6)(iii) requires information concerning plans for preoperational testing and initial operations. Rese requiremc.nts do not apply to ThfI-2 during PDMS. 50 34(bM6Kiv) Paragraph 50.34(b)(6)(iv) requires information conceming plans for conduct of normal operations, including maintenance, surveillance and periodic testing of structures, systems and components. During PDMS, operations are limited to those activities O 3.1 - 6 UPDATE 2 - AUGUST 1997
related to monstonng and maintaming the facility in a stable condition. Although the specific requirements of this paragraph do not apply due to the unique condition of TMI 2 during PDMS, the intent of these provisions has been addressed by prosiding information conceming ' activities appropriate during PDMS. 50.346V6Vv) +
-Paragraph 50.34(b)(6)(v) requires information conceming plans for coping with emergencies, {
which shall include the items specified in Appendix E. Emergency plarming requirements are
- based on the assumption of the potential necessity to notify the public of the existence of, or
- - potential for significant off site releasesf Appendix E recognizes that emergency platming i needs are different for facilities that present less risk to the public. Due to the non operating and defueled status of TMI 2 during PDMS, there is no potential for any significant off-site 4 radioactive release. Further, due to the existence of TMI-l'on the same site, emergency planning requirements for the site are dommated by TMI 1. Herefore, the limited emergency planning necessary to accommodate the existence ofTMI 2 on the same site as TMI-l has been incorporated in an integrated corporate Wergency plan. - _
50.346M6Xvi) i Paragraph 50.34(b)(6)(vi) requires information concerning proposed technical specifications
- prepared in accordance with the requirements of Article 50.36. Due to the unique condition i of TMI 2 during PDMS, the specific requirements of Article 50.36 are not applicable; however, the intent of this article has been addressed. Rev. O of the PDMS S AR prosided
' draft Technical Specifications (Tech, Specs.). The NRC subsequently, issued TMI 2 Technical Specifications as Appendix A to the Possession-Only License for PDMS. He 3 draft Tech Specs in Chapter 9 have been deleted from the PDMS SAR to prevent confusion between the draft and the actual Tech. Specs. 50.346V6)Mi) l Paragraph 50.34(b)(6)(vii) requires information conceming th: construction of multiunit power plant sites. Rese requirements are not applicable to TMI 2 during PDMS. i 50.346X7) g Paragraph 50.34(b)(7) requires the SAR to include the technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this F chapter. De tecimical qualifications of GPU Nuclear, which are applicable to activities related to the unique PDMS conditions, are prosided in Section 10.5. 4 4 - 3.I' : UPDATE 2 - AUGUST 1997
50 34(b)(8) Paragraph 50.34(b)(8) requires the SAR to include a description and plans for irnplementation of an operator requalification program. The operator requalification program shall, as a nummurn, meet the requirements for those programs contained in Appendix A of Part 55 of this chapter. Due to the nonoperating and defueled status of TM12 during PDMS, the requirements for licensed reactor operators do not apply and consequently the requirements for operator requalification also do not apply. 50.34(b)(9) Paragrap'. 50.34(b)(9) requires a description of protection provided agamst pressurized thermal shock events. Due to the non-operating and defueled status of TMI 2 during PDMS, the requiremmts of paragraph 50.34(b)(9) do not apply. In addition, TMI-2 was granted an exemption to 10 CFR 50.61 (Reference 3.1-1) which acknowledged that TMI 2 need take no measures to protect against pressurized thermal shock. 50 34(e) Paragraph 50.34(c) requires each application for a license to operate a production or utilization facility to include a physical security plan. Due to the unique condition of TMI 2 during PDMS, the specific requirements of this paragraph are not applicable; however, the intent of the requirements has been addressed in this S AR He secunty provisions necessary for TMI 2 have been provided by locating the unit inside the same protected area as TMl Unit I and the provisions incorporated in the TM1 site security plan referenced in Section 10.2. 50 34(d) Paragraph 50.34(d) requires that each application for a license to operate a production or utilization. facility that is subject to Article 73.50, Article 73.55, or Article 73.60 shall include a licensee safeguards contingency plan in accordance with the criteria set forth in Appendix C to 10 CFR Part 73. The safeguards contingency provisions necessary for TMI-2 are provided by being located inside the same protected area as TMI l and are incorporated in the safeguards contingency plan for the TMl site. See Section 10.2. 50 34(e) Paragraph 50,34(c) requires that each applicant for a license to operate a production or utilizatiori facility who prepared a physical security plan, a safeguards contingency plan, or a guard qualification and training plan shall protect the plans and other related Safeguards Information again:t unauthorized disclosure in accoidance with the requirements of 10 CFR 73.21 as appropriate. Due to the non-operating and defueled O 3.1 - 8 UPDATE 2 - AUGUST 1997
1 status of TMI 2 during PDMS and the location of TMI l on the same site, overall security p will be controlled by the site security plan. All security activities establishid in accordance l i,v)c with the regulations in 10 CFR Part 50 will be protected against unautho.ized disclosure in accordance with 10 CFR 73.21. 1Q14I0 Paragraph 50.34(f) establishe; TMI-related requirements for a specific group of plants. TMI 2 is not included in this group of plants; therefore, tlus paragraph does not apply to TMI2. 50.34M Paragraph 50.34(g) requires applicants for operating licenses docketed after May 17,1982, to include SRP evaluations with their license applications. As this application is not requesting an operating license for TMI 2, this paragraph does not apply to TMI-2. 3.1.1.21 10 CFR 50.34a - Design objectives for equipment to control releases of radioactive material in efiluents-nuclear power reactors. Article 50.34a establishes requirements for radioactive effluent control descriptions in construction permit and operating license applications. Due to the unique condition of TMI 2 during PDMS, the specific requirements of this article are not applicable; howrver, there will be linuted radioactive effinents to the emironment during PDMS. Descriptions of the equipment to monitor and control those releases are provided consistent with the intent of this article. ID 3.1.1.22 10 CFR 50.35 - Issuance of construction permits. Article 50.35 establishes ruluirements for the Commission with respect to the issuance of construction permits and defines the limitations of the construction permit. No exceptions are taken to the provisions of this article. 3.1.1.23 10 CFR 50.36 Technical specifications Article 50.36 establishes requirements for Technical Specifications. No exceptions are taken to the pro isions of this article. 3.1.1.24 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.
/~~'s (V )
3.1 - 9 UPDATE 2- AUGUST 1997
50.36da) Paragraph 50.36a(a) establishes requirements for effluents for operating reactors. Although TM12 is not an operating re. actor and the requirements of this paragraph cannot be complied with as written, the effluenti Juring PDMS will be controlled and limited to very low values. De intent of the provisions of this paragraph is addressed by providing effluent limits and the description of how these limits will be met in Chapters 7 and 8. 50 36a(aXD Paragraph 50.36a(a)(1) requires that procedures be developed for the control of emuents and that equipment installed in radioactive waste systems pursuant to 50.34(a) be maintained and used. Procedures in use will be in place for the control of effluents durmg PDMS. The TMI-2 equipment that will be used to process radioactive wastes during PDMS will be maintained and is described in Section 7.2.3. 50.36afa)(2) Paragraph 50.36a(a)(2) requires that each licensee subnut annual reports on effluents and prepare estimated public dose from those efIluents. These requirements are applicable to TMl 2 during PDMS. 50 364(h) Paragraph 50.36a(b) establishes guidelines for limiting radioactive effluents and references 10 CFR 20.106 and 10 CFR 50 Appendix I as applicable in limiting effluents. Rese requirements are applicable to TM12 during PDMS. 3.1,1.25 10 CFR 50.36b - Emironmente Conditions Article 50.36b establishes that the NRC may specify conditions as part of the license to protect the emironment, No exceptions are taken to the provisions of this article. 3.1.1.26 10 CFR 50.37 - Agreement Limiting Access to Restricted Data Article 50.37 establish requirements for access to Restricted Data. No exceptions are taken to the provisions of this article. 3,1.1.27 10 CFR 50.38 -Incligibility of Certain Applicants Article 50.38 establishes that certain persons are not eligible to apply for or obtain a license. No exceptions are taken to the prosisions of this article. O 3.1 - 10 UPDATE 2 - AUGUST 1997
3.1.1.28 10 CFR 50.39 Public Inspection of Applications Article 50.39 states that applications and documents submitted to the Commission may be made available for public inspection. No exceptions are taken to the provisions of this article. 3.1.1.29 . 10 CFR 50.40 - Common Standards Article 50.40 establishes guidelines for the Commission in determi. ting if a License will be issned to an applicant. No exceptions are taken to the provisions cf tidr article. + 3.1.1.30 - 10 CFR 50.41 Ad&tional Standards for Class 104 Licenses Article 50.41 establishes ad6tional standards for class 104 IWases for the Commission to use in determuung if a license will be issued to an applicant The cass oflicense described in this article does not apply to TMI 2.
- 3.1.1.31 10 CFR 50.42 - Aditicaal Standards for Class 103 Licenses Article 50.42 establishes additional standards for class 103 liwnses for the Commission to use in determining if a license will be issued to an applicant. No exceptious are taken to the provisions of this article.
3.1.1.32 10 CFR 50.43 Additional Standards and Provisions Affecting Class 103 Liccares for Conuncreial Power f Article 50.43 establishes additional standards and provisior.s for class 103 licenses. No exceptions are taken to the provisions of this article. 3.1.1.33 10 CFR 50.44 - Standards for Combustible Gas Control System in Light Water-Cooled Power Reactors Article 50.44 trpecifically exempts plants that have permanently ceased operations fi om the requirement to establish a combustible gas control system to be used in the event of a LOCA. This exemption applies to TMI 2 during PDMS. Thus, no exceptions to the provisions of this article are necessary. _ 3.1.1.34 l_0 CFR 50.45 - Standards for Construction Permits 4 Article 50.45 establishes standards for the issuance of a construction permit. No exceptions are taken to the provisions of this article. ( F 3.1 - 11 UPDATE 2- AUGUST 1997
3.1.1.35 _10 CFR 50.46 - Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors Article 50.46 specifically exempts plants that have permanently ceased operations from the regt.irement for emergency core coating systems for light water nuclear power reactors This exemption applSs to TMI 2 during PDMS. Thus, no esceptions to the provisions of this article are necessary. 3.1.1.36 10 CFR 50.47 - Emergency Plans Article 50.47 establish <.s requirements for the content and criteria for accep:ance of emergency plans. Emergency planning requirements are based on the assumption of the potential necessity to notify the public of the costence of, or potential for significant off-site releases. Appendix E recognizes that emergency planning needs are different for facilities that present less risk to the publi: Due to the non-operating and defueled status of TMI-2 during PDMS, there is no potential for any significant off site radioactive release. Due to the existence of TMI l on the same site, emergency planning requirements for the site are dommated by TMI 1. Therefore, the limited emergency planning necessary to accommodate the existence of TMI 2 on the same site as TMI l has been incorporated in an integrated corporate emergency plan. See the discussion of pamgraph 50.34(b)(6)(v). 3.1.1.37 10 CFR 50.48 Fire Protection Article 50.48 establishes fire protection requirements for plants that have permanently ceased operation. These requirements are applicable to TM12 during PDMS. 3.1.1.38 10 CFR 50.49 - Environmental Qualification of Electric Equipment important to Safety for Nuc! car Power Plants Article 50.49 specifically exempts plants that have permanently ceased operations from the requirements to establish a program for the qualification of electrical equipment important to safety. This exemption applies to TMI 2 during PDMS. Thus, no exceptions are taken to the provisions of this article. O 3.1 - 12 UPDATE 2 - AUGUST 1997
! 3.1.1.39 ,10 CFR 50.50 - Issuance of Licenses and Construction Permits ('s Article 50.50 states that the Commission will issue a license or construction permit with such conditions and limitations as it deems appropriate. No exceptione are taken to the provisions of tlus article. 3.1.1.40 10 CFR 50.51 - Duration of License, Renewal Article 50.51 establishes the duration oflicenses issued by the Commission. No exceptions are taken to the provisions of this article. 3.1.1.41 10 CFR 50.52 Combining Licenses Anicle 50.52 establishes that the Commission may combine licensed activities in a single license. No exceptions are taken to the provisions of this anicle. 3.1.1,42 10 CFR 50.53 Jurisdictional Limitations Anicle 50.53 establishes jurisdictional limitations on licenses. No exceptions are taken to the prosisions of this anicle. 3.1.1.43 10 CFR 50.54 - Conditions of Licenses Article 50.54 establishes a series of conditions applicable to holders of a license. Due to the non-operating and defueled status ofThil-2 during PDhtS, many of these requirements do not apply. The applicability of each paragraph of Anicle 50.54 has been addressed in the following resiew.
,O ! 50.54M L
Paragraph 50.54(a) requires that each nuclear power plant or fuel reprocessing plant licensee subject to the critena of 10 CFR Part 50 Appendix B implement a quality assurance program pursuant to paragraph 50.34(b)(6)(ii). Appendix B establishes quality assurance requirements for the safety-related functions of those structures, systems and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. During PDh1S, Th112 will not have any structures, systems or components classified as safety related and, therefore, p, i V 3.1 - 13 UPDATE 2 - AUGUST 1997
the requirements of paragraphs 50.54(a)(1),50.34(b)(6)(ii) and Appendix B do not apply to TMI-2. Howewr, the intent of this article has been addressed by establishing and maintaining a quality assurance program similar to that described in Appendix B for TMI-2 actisities. 50 54(b) throuch 50.54(h) Paragraphs 50.54(b) through 50.54(h) establish general limitations on licenses. No exceptions are taken to the prosisions of these paragraphs. 50.54(i) throuch 50 54(m) Paragraphs 50.54(i) through 50.54(m) establish requirements related to reactor operators and senior reactor operators. As discussed in License Amerximent No. 30 (Reference 3.13), these requirements are specified for fueled reactors. As the TMI 2 reactor has been defueled, the requirements of these paragraphs do not apply to TMI 2 durin'g PDMS. Also see Section 3.1.1.20 regarding paragraph 50.34(b)(8). 50.54(n) Paragraph 50.54(n) states that "The licensee shall not, except as authorized pursuant to a construction permit, make any alteration in the facility constituting a change from the technical specifications previously incorporated in a license or construction permit pursuant to Article 50.36 of this part." No exceptions are taken to the provisions of the article. 50.54(o) Paragraph 50.54(o) specifically exempts the primary reactor containments of plants that have permanently ceased operation from the requirements of 10 CFR 50 Appendix J. This exemption applies to TMI-2 during PDMS. Thus , no exceptions to the provisions of this article are necessary. 50.54(n) Paragraph 50.54(p) requires that a licensee prepare and maintain safeguard contingency plan procedures and provides for revisions to those procedures. He safeguards contingency provisions necessary for TMI 2 are provided by being located inside the same protected area as TMI-l and are incorporated in the safeguards contingency plan for the TMi site. See Section 10.2. O 3.1 - 14 UPDATE 2 - AUGUST 1997
70.54(a) ON Paragraph 50.54(q) requires that a licensee shall follow and maintain emergency plans which meet (V) + requirernents of paragraph 50.4',(b). This paragraph also defines requirements for revising
- emergency plans. Due to the existence of Thu l on the same site as TMI 2, emergency planning requirements for the site are dominated by ThD 1. Therefore, the limited emergency planning necessary to accommodate the existence of ThC 2 on the same site as ThD 1 has been incorporated in an integrated corporate emergency plan.
M. ilio Paragraph 50.54(r) establishes requirements for test reactors. These requirements do not apply to ThD 2. 5054(s) Paragraph 30.54(s) requires each licensee w ho is authorized to possess and'or operate a nuclear power reactor to submit radiological emergency plans of state and local governmental entities to the NRC. All radiological emergency planning provisions necessary for Thil 2 have been incorporated in the Th81 site emergency planning process, includmg the provisions of paragraph 50.54(s). 50.54(t) Paragraph 50 54(t) establishes requirements for the development, revision, implementation and maintenance of the emergency preparedness program for nuclear power reactors. Emergency preparedness requirements applicable to TM12 are incorporated in the emergency preparedness program established for the TMI site. See Section 10.3. p) i 5054(u) V Paragraph 50.54(u) requires each licensee to submit emergency plans in accordance with 10 CFR 50.47(b) and Appendix E. Article 50.47 establishes requirements for the content and criteria for acceptance of emergency plans. Emergency planning requirements are based on the assumption of the potential necessity to notify the public of the existence of, or potential for significant off-site releases. Apper. dix E recognizes that emergency planning needs are different for facilities that present less risk to the public. Due to the non operating and defueled status of TMI 2 dunng PDMS there is no potential for any significant off-site radioactive release and due to the existence of TM1 1 on the same site, emergency planning requirements for the site will be dominated by TMI 1. Therefore, the limited emergency planning necessary to accommodate the existence of ThD 2 on the same site as ThD 1 has been incorporated in an integrated corporate emergency plan. See Section 3.1.1.20 regarding paragraph 50.34(b)(6)(v). 50,54M Paragraph 50.54(v) reqtures that each licensee shall ensure that physical secunty, safeguards ! contingency and guard qualification and training plans and other related safeguards information are protected against unautnorized disclosure in accordance with the requirements of 10 CFR 73.21 as appropriate. To the extent that TM12 possesses the above information dunng PDMS, it will be protected from unauthorized disclosure in accordance with 10 CFR 73.21. See paragraphs 50.34(c), 50.34(d) and 50.34(e).
- p. \
I
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N.-) l 3.1 - 15 UPDATE 2 - AUGUST 1997
50 54(w) ParaFraph 50.54(w) requires that each electric utility licensed under this part for a production or utilization facility of the type described in paragraph 50.21(b) or paragraph 50.22 shall by June 29,1982 take reasonable steps to obtain on-site property damage insurance available at reasonable costs and at reasonable terms from private sources. The appropriate insurance has been acquired and will be maintained for TMI on a site-basis. jo.54(x) and 50.54M Paragraph 50.54(x) allows a licensee to take action which departs from a license condition or technical snecification in an emergency when this act ion is immediately needed to protect the health and safety of the public. Paragraph 50,54(y) requires that for plants that have permanently ceased operation, any action taken pursuant to paragraph 50.54(x) be approved, as a minimum, by a licensed senior operator or a certified fuel handler prior to taking the action. The provisions of this article have limited applicability to TMI 2 during PDMS. Due to the non-operating and defueled status of TMI 2 during PDMS, there are no postulated events which could affect public health and safety in such a manner. In addition, the technical specifications will be oflimited scope and it is not anticipated that a condition will exist such that it could become necessary to take action that departs from either a license condition or a technical specification to protect public health and safety. Since TMI 2 will not have licensed senior reactor operators or certified fuel handlers during PDMS, if an extremely unlikely event were to occur necessitating deviation from the technical specifications the action would have to be approved by senior management. 50 54(z) Paragraph 50.54(z) requires cach licensee to notify the NRC Operations Center of the occurrence of any event specified in 10 CFR 50.72. Due to the non-operating and defueled status of TMI 2 during PDMS, there are very few potential events which would require reporting under 10 CFR 50.72. Ilowever, to the extent that reporting is required under 10 CFR 50.72, the requiren. cats of this paragraph are applicable. See Section 3.1.1.57 regarding paragraph 10 CFR 50.72. SIL54fm0 Paragraph 50.54(aa) establishes that the licensee must meet Sections 401(a)(2) and 401(d) of the Federal Water Pollution Control Act. No exceptions are taken to the provisions of this article. 50 54(bh) Paragraph 50.54(bb) requires licensees of operating nuclear power reactors to acquire NRC approval of the program to fund, manage, and transfer irradiated fuel upon expiration of the reactor operating license. It further requires that Licensees that ceased operation prior to April 4,1994 to submit their spent fuel management fundmg phm by April 4,1996. As the irradiated fuel which comprised the TMI-2 reactor core has been transferred to the possession of the Department of Energy no funding plan is required for TMI-2. A letter documenting this position was submitted to the NRC on May 31,1994 (C311-94-2077). 3.1 - 16 UPDATE 2- AUGUST 1997
l 50.54(cc) 73 Paragraph 50.54(cc) requires licensee untien notifications of the appropriate NRC Regional (v) Administrator of certain bankruptcy filings. No exceptions are taken to the prmisions of this article. 50.54fdd) Paragraph 50.54(dd) allows licensees to take reasonable actions that depan from a license condition or a Technical Specification under certain conditions during a National Security Emergency. No exceptions are taken to the prmisions of this anicle. Sli4Rd Paragraph 50.54(ec) allows licensecs, authorized to possess by product and special nuclear material to receive back low level waste (LLW) separated at the plant and shipped off-site for processing No exceptions are taken to the provisions of this anicle. 50.54(f0 Paragrap'o 50.54(f0 establishes requirements for plant shutdown and subsequent restart if tibratory ground motion exceeds that for an operating basis earthquake for future nuclear power stations As TMI 2 is currently licensed (i.e permanently shutdown) the requirrments of this paragraph do not apply to TMI 2. 3.1.1.44 10 CFR 50.55 - Conditions of Construction Permits g~g Article 50.55 establishes terms and conditions of construction permits No exceptions are taken to the prmisions of this article. (V) 3.1.1.45 10 CFR 50.55a . Codes and Standards Article 50.55a requires each operating license for a nuclear power facihty be subject to *e insenice testing and inspection requirements of paragraphs (0 and (g) to Anicle 50.55a and that each construction permit be subject to the remaming paragraphs of the article. As this application is not for a con <truction permit, paragraphs (f) and (g) of Article 50.55a are the only portions of the article potentially applicable to TMI-2 in PDMS. Paragraphs (f) (1) and (g) (1) are the para;raphs which apply to Th0-2. These paragraphs require that safety related pumps, valves and components (' including supports) meet the requirements of paragraphs (f) (4) and (5) or (g) (4) and (5). These paragraphs define the insenice testing and inspection requirements and proside a mechanism for relief from impractical requirements based on a satisfactory demonstration of this action to the Nuclear Regulatory Commission. Relief from these requirements was sought and for the most pan grsoted early in the Th0-2 Cleanup Program. The NRC granted the following relief for ThD-2 (Reference 3.1-5):
- 1. He prmisions ofIWA-2400 of Section XI of the ASME Boiler and Pressure Vessel Code (code) 1974 edition, Summer 1975 Addenda for extending the inspection intenst for a period of tirr.c equivalent to the shutdown period of ThD-2 wu applicable.
- 2. Testing of pumps in accordance with Section IWP-3400 of the code was only required for those pumps specified in the Recovery Technical Specifications.
(O) v 3.1 - 17 UPDATE 2 - AUGUST 1997 l
- 3. Category A ulves wo e dermed to be contamment isolation valves. The NRC agreed that these valves should not be exercised. Ilowever,10 CFR Part 50 Appennx J Type "C" testing was repired for any contamment isolation valve that was opened ard subsequently closed in order to verify its contamment isolation function.
- 4. Category B and C valves in systems outef senice need not be tested, however, Category B and C valves in safety related systems in senice should be exercised at least once per 92 days.
- 5. The Mini Decay Heat Removal System was to be handled as a separate action.
During PDMS, the relief granted by the NRC still applies. In addition, based on the following justification, no further insenice inspection is required.
- 1. The PDMS , Technical Specifications require no pumps to be operable. Therefore, based on the existing relief, testing of pumps in accordance with IWP 3400 would not be required during PDMS.
- 2. Performanexf Type "C" testing of containment isolation valves (Category A valves) in accordance widi 10 CFR Part 50 Appendix J is not required at TMI-2. The NRC granted TMI-2 an exemption from Type "C" tening (Reference 3.1-4).
- 3. DurMg PDMS, there will be no safety related systems at TMI 2. Therefore, performance of testing of Category B and C valves is not required based on the relief granted by the NRC.
- 4. No testing is required for the MDHR System as it will be deactivated for PDMS (see Section 6.25). Article 50.55a further requires in paragraph 50.55a (g) (6) (EE) (A) an augmented exammation of the reactor vessel to look for degradation of reactor vessel materials in accordance with Section XI Division 1 of the ASME Code. As the TMl-2 Reactor Vessel is no longer a pressure retaining component this examination is not applicable to the TMI-2 as described in paragraph IWB-1200 " Components Subject to Exanunations" of Section XI Division 1 of the ASME Code, Therefore, as discussed above, complete relief from the insenice testing and inspection requirements of 10 CFR 50.55a during PDMS is appropriate.
3.1.1.46 10 CFR 50.56 - Conversion of Construction Permit to License; or Amendment of License Article 50.:i6 establishes that the Commission will, in the absence of good cause shown to the contrary, issue a license ci anebent of a hcense as the case may be. No exceptions are taken to the provisions of this article. 3.1.1.47 10 CFR 50.5 ? Issuance of Operating License Article 50.57 establishes the standards the Commission shall use in determuung the issuance of an operatmg license. No exceptions are taken to the prosisions of this article. 3.1.1.48 10 CFR 50.58 - Hearings and Report crlthe Advisry Committee on Reactor Safeguards 3.1 - 18 UPDATE 2 - AUGUST 1997
= -
50.58(a) Paragraph 50.58(a) establishes tnat each application for a construction permit, an operating ( license, or an amendment to the construction permit or oper. sting license may be referred to D the Advisory Committee on Reactor Safeguards. The report from the Advisory Committee on Reactor Safeguards will be made part of the public record No exceptions are taken to the provisions of this article. 50.58(b) Paragraph 50.58(b) establishes that the Commission may hold hearings on each application for a construction permit or an operating license for a production or utilization facility of the - type described in 10 CFR 50.21(b) or 10 CFR 50.22. No exceptions are taken to the provisions of this paragraph. 3.1.1.49 10 CFR 50.59 - Changes, Tests and Experiments Arti>.;1e 50.59 establishes the requirements for changes, tests or experiments that affect the facility. No 2 exceptions are taken to the provisions of this article. 3.1.1.50 10 CFR 50.60 - Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation 10 CFR 50,60 specifically esempts plants that have permanently ceased operation from the requirement that alllight water nuclear power reactors meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary as set forth in Appendices G and H to 10 CFR 50. This esemption applies to TMI.2 during PDMS. Thus, no s exceptions are taken to the provisions of this article. i
- g 3.1 - 19 UPDATE 2 - AUGUST 1997
1 l 3 1.l.51 10 CIR 50 61. Fracture tougimess requirements for protection against pressunted thermal shock events Anicle 50 61 specifically etranpts plants that haic permanently acased operations from the requirements for protection against pressurized thennal shock in pressurized uater nuclear power reactors This esemption applies to Th11-2 during pDhtS. Thus, no esceptions to the prvvisions of this article are necessar). 3.1.1.52 10 CPR 50 62 Requirements for reduction of nsk from anticipated transients without scram (ATWS) events for light water cooled ruclear power reactors Artic!c $0 62 specifically esemptions plants that haie permanently ecased operations from the requirements to have equipment to addrers ATTS rients This esemption applies to Th112 during 8'D31S. Thus, no esceptions to the prmisions of this article are necessar). 3.1.1.53 10 CIT,50 63 Loss of alternating current poucr. Anicle 50.b3 requires that each light water cooled nuclear power plant licensed to operate be able to withstand and recovet from a station blackout event. Since this applicauon chminates the legal audiont) to operate the Thil 2 facility from the beense, a subsequent license application would be necessar) to tesume operation Thereforc dunng PDh15 the requirements of this anicle are not applicable to Thil 2. 3.1.1.54 10 CIR 50 64 Limitations on the use of highly enriched uranium GIEU)in domestic non-power teactof s Anicle 50 64 establishes requirements for the issuance oflicenses to use highly enriched uranium fuel in non-po./cr reactors. No exceptions are taken to the prmisions of this anicle. 31.1.54a 10CTR50 65 Requirements for hionitoring the Effectn eness of hiainternnce at Nuclear power plants' Article 50.65 specifically esempts plants that haic permanently cer , trations from the seguirement that each holder of a license to operate and monitor the perfonna. . or conditions of stnictures, sptems and components against licensed established goals ulth the etreption of structures, sy stems and components associated with storage, control, and maintenance of spent fuelin a safe condition. Th1I 2's fuel has been shipped off site; thus the requirements of this article t.o not epply to T3112. 3.1.1.54b 10 CFR 50.66 - Requirements for Therinal Annealing of the Reutor Pressure Vessel Article 50.66 preides a consistent set of requirements for tL use 64 thennat annealing to mitigate the effects of neutrun irradiation. Due to the defueled non-operating status of Thll 2 in PDhtS, the Th112 Reactor Vessel will not be thennall 3annealed. Thus the requiretuents of Ok8s article do not apply. 3.1.1.55 10 CFR 50.70. Inspections. Anicle 50.70 establishes requirements to pennit NRC inspectors to maintain activities at cach nuc..ar power plant site. During PDhtS, Thil 2 will be required to support NRC inspection actinues to the extes.: determined necessary by the NRC. No exceptions are taken to the pimisions of this anicle. O 3.1 20 UPDATE 2 - AUGUST 199'
l 3.1.1.56 10 CFR 50.71. hiamtenance of records, making of reports. Os Article 50.71 establishes requirements for facihty records and updating the Safety Analysis Reports. De requirements of these paragraphs apply to Th112 during PDhtS. 3.1.1.57 10 CFR 50 72 Immediate notification requirements for operating nuclear power reactors. With the exception of paragraphs 50.72(bf(2)(iv)(A),50.72(b)(2)(is)(B),50.72(b)(2)(v), and 50.72(b)(2)(vi) the requirements for notification address events or situations which are related to the operation of the power plant and conditions which do, or may compromise the safe operation of the plant or cask storage of spent fuel on site. Since Thil 2 will be specifically precluded from nie operation of the plant during PDhtS, the requirements of those paragrephs which relate to power plant operation will not apply. Sirrularly, since Thil 2's fuel has been shipped off site, and there is no cask storage on site the requirements of the paragraph related to cask storage does not apply. Paragraphs 50 72(b)(2)(iv)(A) and 50.72(b)(2)(iv)(B) establish requirements for reponing liquid and gaseous efIluents which exceed levels established by these paragraphs. These requirements are applicable to Thil 2 during PDhtS. Paragraphs 50.72(b)(2)(v) and 50 72(b)(2)(vi) require the reporting of any event requiring the transport of a radioactively contaminated person to an off site medical facility for treatment and any event or situation, related to the health and safety of the public or onsite personnel, or protection of the envirotunent, for which a news release is planned or notification to other govcmment agencies has been or will be made. Rese requirements are also appbcable to Th112 during PDhtS. With the exception h of subparagraph 50.72(a)(4) which is not applicable to Th112, required notifications, will be made in accordance with paragraph 50.72(a). 3.1.1.58 10 CFR 50.73 - Licensee Event Report System Article 50 73 requires that the holder of an operating license for nuclear power plant (licensee) shall submit a Licensee Event Report (LER) for any event of the type _ desenbed in this paragraph within 30 days after the discovery of the event. he requirements of this article are appheable to Th112 during PDhtS. 3.1.1.58a " Article 50.74 - Notification of Change in Operator or Senior Operator Status" Article 50.74 requires each Licensee to notify the conunission of a change in status of sny licensed operator or senior operator. As the Thil 2 reactor has been defueled and the requirement to maintain licensed operators and senior operators at Thil 2 has been climinated (Reference 3 1 3) this requirement is not applicable to Thil 2 in PDhtS. t Nv h 3 1 - 21 UPDATE 2- AUGUST 1997
3 1.1.59 10 CFR 50.75 Reporting and accordkeepinE or f decomrnissioning planning. Anicle 50.75 establishes requirements for prosiding reasonable assurance to the NRC that funds will be available for decommissioning. No exceptions are taken to the provisions of this article. Additionally, Reference 3.1-6 provided the decommissioning fundmg plan for TMI 2. 3.1.1.60 10 CFR 50.78 Installation Infonnation and Venfication Article 50.78 requires that, "Each holder of a construction pemut shall, if requested by the Commission, submit installation infonnation on Form N 71, pennit verification thereof by the International Atomic Energy Agency, and take such other action as may be necessary to implement the US/lAEA Safeguards Agreement, in the manner set forth in Articles 75.6 and 75.11 through 75.14 of this chapter." No exceptions are taken to the provisions of this article. 3.1.1.61 10 CFR 50 80 Transfer of Licenses Article 50,80 specifies requirements for transferring a license from one entity to another. No exceptions are taken to the provisions of this article. 3.1.1.62 10 CFR 50 81 Creditor Regulations Article 50.81 defines the rights and restnctions applying to any creditor relative to any license issued by the Comntission. No exceptions are taken to the provisions of this article. 3.1.1,63 10 CFR 50.82 Termination of Liccases Article 50.82 defmes the requirements for tenmnatmg a heense. No exceptions are taken to the prosisions of this article. 31.164 10 CFR 50.90 - Application for Amendment of License or Construction Pennit Article 50.90 esthblishes that a holder of a license must file an application for an amendment, desenbing the changes desired, if the license holder wishes to amend the license. No exceptions are taken to the provisions of this article. 3.1.1.65 10 CFR 50.91 Notice for Pubhc Comment, State Consultation Article 50.91 establishes requirements applying to the Conunission and TMI 2 regarding the application for an amendment to a 10 CFR Part 50 license following permanent removal of the fuel. The rcquirements of this article apply to TM12. 3.1.1,66 10 CFR $0,92 Issuance of Amendment Article 50.92 establishes the standards by which the Commission detennines if no significant hazards exist for a license amet.dment. The licensee must file a no significant hazards analysis with each amendment aopFeation usire the standards set forth in Article 50.92 as required by Article 50.91. De requirements of thir article apply to TMI 2. O l l 3.1 - 22 UPDATE 2 - AUGUST 1997 l
3.1.1.67 10 CFR 50.100 Revocation. Suspension, hiodtfication, of Licenses and Construction Pernuts for Cause Article 50.100 provides that the Ccanmission may revoke, suspend, or modify a license or coratrvetion pennit for any material fala statement or for other reasons specificd in Article 50.100. No exceptions are taken to the prosisions of this article. I 3.1.1.68 10 CFR 50.101 Retaking Possessinn of Special Nuclear hiatenal Article 50.101 establishes that the Commission may cause the retaking or possession of special nuclear matenal upon revocation of a heense. No exceptions are taken to the provisions of this article. 3.1.1.69 10 CFR 50.102 Commission Order for Operation after Revocation Article 50.102 establishes that the Commission may, by following the requirements of Article 50.102, order operation of a facihty whose license has been revoked No exceptions are taken to the ptovisions of this article. 3.1.1.70 10 CFR 50.103 Suspenrion and Operation in War or National Emergency Article 50.103 establishes that the CornnJssion has, upon declaration of war by the Congress, certain rights regardmg the suspension and/or operation of nuclear power plants licensed by the Commission. No exceptions are taken to the prmisions of this article. 3.1.1,71 10 CFR 50.109 - Backfitting Article 50.109 defines backfittmg and defines requirements the Comnussion must meet regarding backfitting. No exceptions are taken to the praisions of this anicle. 3.1.1.72 10 CFR 50.110 Violations Article 50.110 establishes acr. ions the NRC may take regarding violations of any provision of the Atomic Energy Act of 1954, as amended, or Title 11 of the Energy Reorganization Act of 1974, or any , regulation or order issued therender. No exceptions are taken to the provisions of this article. 31.1.73 " Article 50.111 Cnminal Penalties" Article 50.111 defmes which articles of 10CFR Part 50 are subject to criminal sanction as dermed in the Atomic Energy Act of 1954. No exceptions are taken to the provisions of this article. 3.1.1.74 " Article 50.120 Training and Qualifications of Nuclear Power Plant Personnel" Article 50.120 defmes training program requirements for cach holder of an operating license. As Thll-2's License has been modified to Possession Only for PDhtS the requirements of this article are not applicable. Appropriate training and qualification requirements for personnel supportmg actisities at Thti 2 are dermed in Section 6.0 of the Thti 2 Technical Specification. O V 3.1 - 23 UPDATE 2 - AUGUST 1997
3.1.2 GENERAL DESIGN CRITERIA The Three hiile Island Nuclear Station Urit 2 was designed and emstructed in accordance with the 70 gereral desiga entena as lined in Appendix A of 10 CFR 50 dated July 11,1967. A discussion of each entenon, demonstrating how the pnncipal design features or design loses meet these cntena, is presented in Section 3.1.1 of the Thil 2 FSAR. The general design enteria in Appendix A were inised by the AEC on July 15,1971. The design and purchase of many Three hitle Island Unit 2 components were completed prior to the issuance of these inised general design criteria These inised criteria, as they applied to the original design of the plant, are addressed in Section 3.1.2 of the Thti 2 FSAR During the PDhiS period, fulfillment of snany of the general design enteria in Appcadix A of 10 CFR 50 are not necessary or appropriate, departure from the criteria are idenufied and justified herein Other of the criteria are appbcable only to a scry limited degree. Cnteria widch address such requirements as containment, quality standards. and natural phenomena are exarnples of those enteria w hich apply only to a lirnited degree dunng PDhtS. Since the plant was originally designed and constructed in accordance with these enteria and since neither the accident not acthides during the recovery period significandy degraded the plant with teepect to the capabilities required during PDhtS, the facility, as it exists, is designed and constructed to standards wluch far escced the requirernents for PDhtS. Each of the general design enteria in Appendix A of 10 CFR 50, at snised on January 1,1987, and the necessary and appropriate degree of applicability during PDhis is discussed in the following sectiont 31.2.1 Criterion i . Quahty Standards and Records Structures, systems, and components important to safety shall te designed, fabncated, erected, and tested to quahty standards commensurate with the imponance of the safety funcuon to be performed Where pencrally recogrured codes and standards are used, they shall be identtfied and evaluated to detennine their applicab6t), adequacy, and sufficiency and shall be supplemented or modtfied as necessary to assure a quahty product in keepmg with the required safety function. A quality assurance program shall te established and implemented in order to provide adequate assurance tiuit these structures, systems, and components will sausfactonly perform their safety functions Appropnate records of the design, fabrication, erection, and tesung of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear pourt unit licensee throughout the hic of the unit. Dixussion Due to the unique condition of Thil 2 during PDhiS, the specific reqmrements of Cntenon I are not applicable, however, the intent of Criterion I has been addressed recogniting that the degree of quality assurance nctessary to assure that the required capabihties are maintained dunng PDhtS is far less extensive than that which was originally required for Thil 2. A quality assurance program has teen estabbshed and will be maintained commensurate with the functional requirements of PDhtS. The Quality Assurance Plan for PDhtS is teferenced in Section 10.1. 3.1.2.2 Cnter.on 2 . Design Dases for Protection against Natural Phenomena Structures, systems, and components imponant to safety shall be designed to w1thstand the effect of naniral phenomena, such as earthquakes, tornadoes, harncanes, floods, tsunami, and seiches without loss of capability to perform their salety functions. The design bases for these structures, sy stems, and components shall reflect: (1) Appropnate consideration of the most snere of the natural phenomena that have been historically reponed for the site and surrounding area, with 9 3.1 - 24 UPDATE 2 - AUGUST 1997
sufficient truirgin for the linuted accuracy, quantity, and period of time in which the historical data havs < r been accumulated, (2) appropriate combinatiorts of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed Djintition Due to the unique corxhtion of TMI.2 during PDhtS, the specific requirements of Cntenon 2 are not applicable; however, the intent of Criterion 2 has twn addressed by recognizing that the level of protection from natural phenomena required dunng PDMS is that which is required to maintain the isolation of the contamination which remains at the facihty. Rete are no active functions required to be performed by any system to provide the protection from natural phenomena dunng PDMS. Por example, all that is required during a seismic event is that the structure or system remain intact. Rose structures, systems, and components necessary for the level of protection reciuired for PDMS were originally designed and constructed to enteria which excced the requirements for PDMS. This level of prottetion is more than adequate to meet the functional requirements for protection from natural phenomena during PDMS. 3.1.2.3 Cnterion 3 - Fire Protection Structares, systems, and cornponents irnportant to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
- Noncombustible and heat resistant matenals shall be used whenever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropnate capacity and capability shall be provided and designed to rnittirnize the adverse effects of fires on structures, systems, and componmts important to safety. Pire fighting systems shall be designed to assure that their rupture or inadvenent operation does not significantly impair the safety Os capability of these structures, systems, and components.
DEmioD Due to the unique condition of T. Nil 2 during PDMS, the specific requirements of Criterion 3 are not applicable, however, the intent of Cnterion 3 has been addressed by recognizing that the requirements for fire protection during PDMS are based on industrial safety and insurance requirements. A fire protection prograin has been established and will be mamtained commensurate with the industrial safety and insurance requirements and to protect those systems important to PDMS. He Fire Protection System is desenbed in Section 7.2.2. 3.1.2.4 Cnterion 4 Emironmental and Missile Design Bases Structures, systems, and components important to safety shall be designed to accomrnodate the effects of and be compatible with the emitorunental conditions associated with normal operation, maintenance, testmg, and postulated accidents, including loss of coolant accidents. Dese structures, systems, and components shall be appropriately protected against dynamic effect, includmg the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures ati from events and conditions outside the nuclear power unit, llowever, the d.mamic effects associated with postulated pipe ruptures of primary coolant loop piping in pressurized water reactors may be excluded from the design basis when analyses demonstrate the probability of rupturing such piping is extremely low under design basis conditions. C (\ 3.1 - 25 UPDATE 2 - AUGUST 1997
i l l
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Dhwula Due to the non-operatmg and defueled condition of Bil 2, the requirements of Cnterion 4 associated with the dynamic effect, including the effects of inissiles, pipe widpping, and discharging fluids, that rnay result from equipment failures and from events and conditions outside the nuclear power umt do not apply. Structures, ry stems and cornponents rehed upon to prmide protection from the effects of and required to be compatible with emironmental conditions associated with PDhiS operations, maintenance, testing. and postulated unanticipated esents are appropnately designed to accommniate effects associated with these actmties. 31.2.5 Criterion 5 Sharing of Structures. Systena, and Cornponents Structures, systenu, and components important to safety shall not be shared among nuclear power uruts unless it can te shown that such shanng will not significantly impair their abihty to perform their safety function, including. in the event of an accident in one unit, an orderly shutdown and cool down of the remaining units. DiKunb** Due to the non+perating and defueled condition of Thil 2, there are no important to safety functions associa'.ed with any TM14 stiucture, system, or component. The required TM) 1 safety functions associated with the few structures and systems shared tw Thil 1 and 3112 are independent of any Thil 2 function for the respective structure or sy stem 31.2.6 Critermn 10 Reactor Design he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that rpecified acceptable fuel design linuts are not exceeded during any condition of normal operation, including the effects of anticipated operational occunences. DinWslin Due to the non operatmg and defueled condition of TMI 2, the requirements of Criterion 10 are not applicable. 3.1.2.7 Cntenon 11 Reactor inherent Protection The reactor core and associated coolant systems shall be designed so that in the power operaung range the net effect to the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reacthity. D16@Hi@ Due to the non operatmg and defueled condition of Bil 2, the requisements of Criterion 11 are not applicable. 3.1.2.8 Criterion 12 - Suppression of Reactor Power Oscillations ne reactor core and associated coolant, control, and protection rystems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be rehably and readily detected and suppressed. O 3.1- 26 UPDATE 2 - AUGUST 1997
DiKmaion /^ Due to the non-operating and defueled condition of TMI 2, the requirements of Criterion 12 are tot ( applicable. 3.1.2.9 Criterion 13 Instrumentation and Control lostrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, includmg those variables and systems that can affect the fission process, the insegnty of the reactor core, the reactor coolant pressure boundaiy, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within presenbed operation ranges. Discussion Due to the non operating and defueled condition of TM12, the requirements of Criterion 13 are not applicable. 3.1.2.10 Criterion 14 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. DiscussioD Due to the non-operating and defueled condition of TMI 2, the requirements of Criterion 14 are not applicable. 3.1.2.11 Cnterion 15 - Reactor Coolant System Design
%c reactor coolant system and associated audliary, control, and protection systems shall be designed with suflicient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of nonnal operation, including anticipated operational occurrences.
Digussion Due to the non<>perating and defueled condition of TMI 2, the requirements of Criterion 15 are not applicable. 3.1.2.12 Criterion 16 - Containment Design Reactor containment and associated systems shall be prmided to establish an essentially leak tight barner against the uncontrolled relcase of radioactivity to the emironment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require. m 3.1- 27 UPDATE 2 - AUGUST 1997
DiKu11120 The Containment and t.ssociated systems are maintained during PDMS to prevent the uncontrolled release of the contamination which remains inside the Containment. In addition, the Containment series as the pntnary emironmental barrier for the radioactive materials inside the Contamment. Ahhough the Containment will not be maintained during PDMS to the same degree ofleaktightness as during power operation, there will be essentially no uncontrolled or unmorutored leakage. Normally, all efiluents to the emironment will be through the Containment Atmospheric Breather System via the Auxiliary Building or the Containment Purge System, toth of which have llEPA filtered systems Leakage to the emironment from other pathways has been dcaonstrated, by analysis (see Section 7.2.1.2) and by periodic testing, to be a very small portion of the overall leakage from the Containment. He Containment Atmospheric Dreather Systern controls the Containment effluents during passive storage periods consistent with the "rnost probable pathway" concept referred to in Regulatory Guide 1.86 "Tenninstion of Operating License for Nuclear Reactors." 3.1.2.13 Criterion 17 - Electric Power Systems An onsite electric power system and an off site electne power system shall be prosided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assunung the other system is not functioning) shall be to provide sufficient capacity and capabihty to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integnty and other vital functions are maintained in the event of postulated accideres ne onsite electric power supplies, includmg the battenes, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assunung a single failure. Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to muumize to the extent practical the likelihood of their sunuttaneous failure under operating and postulated accident and em tronmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in suflicient time following a loss of all onsite alternating current power supplies and the other off-site electne power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, contamment integrity, and other sital safety functions are maintained. Prmisions shall be included to minimize the probability oflosing electne power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies. Discussion Due to the non+perating and defueled condition of TMI 2, the requirements of Criterion 17 are not applicable. Ilowever, capabilities for electric power are maintained during PDMS commensurate with the electric power requirements necessary for PDMS actisities. 3.1 - 28 UPDATE 2 - AUGUST 1997
3.1.2.14 Criterion 18 - Inspection and Testing of Electric Power Syrtems O (' Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing ofirnportant areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite pourt sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power urut, the off site power system, and the onsite power system Discussion Due to the non operating and defueled condition of TMI 2, the requirements of Criterion 18 are not applicable. 3.1.2.15 Criterion 19 Control Room A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, includmg loss of coolant accidents Adequate radiation protection shall be provided to permit access and occupancy of the control room under aMdcnt conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalcut to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures. Discussion Due to the non operating and defueled condition of TMI 2, the requirements of Criterion 19 are not applicable. 3.1.2.16 Cnterion 20 Protection System Functions The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactisity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operation occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety. DiKunico Due to the non-operating and defueled conchtion of TMI 2, the requirements of Criterion 20 are not applicable. v 3.1 - 29 UPDATE 2- AUGUST 1997 A
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3.1.2.17 . Criterion 21 Protection System Rehability and Testability The protection system shall be designed for high functional reliability and insenice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from senice of any component or channel does not result in loss of the required rmmmurn redundancy unless the acceptable reliabihty of operation of the protection system can be othensise demonstrated De protection system shall be designed to permit periodic testing ofits functioning when the reactor is in operation, includmg a capabilit., to test channels independently to detennine failures and losses of redundancy that may have occurred Dlituitico Due to the non-operating and defueled condition of TMI 2, the requirements of Cnterion 21 are not applicable. 3 1.2.18 Criterion 22 - Protection System Independence The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other dermed basis. Design techniques, such as functional diversity or dacrsity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function. DiscussicD Due to the non-operating and defueled condition of TMI 2, the requirements of Critenon 22 are not applicable. 3.1.2.19 Criterion 23 - Protection System Failure Modes ne protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other dermed basis if conditions such as disconnection of the system, loss of energy (e g., electne power, instrument air), or postulated adverse emironments (e g , extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. Dncinien Due to the non operating and defueled condition of TMl 2, the requiremcnts of Cnterion 23 are not applicable 3.1.2.20 Criterion 24 - Separation of Protection and Control Systems The protection system shall be separated from control systems to the extent that failun. of any single control system component or channel, or failure or removal from senice of any single protection system component or channel u hich is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be Imtited so as to assure that safety is not significantly impaired. 3.1-30 UPDATE 2 - AUGUST 1997
DiKuisjon
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Due to the non-operating and defueled condition of Thil 2, the requirements of Cnterion 24 are not (' applicable. 3.1.2.21 Cnterion 25 Protection System Requirements for Reactivity Control hialfunctions j The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malftmetion of the reactisity control systerns, such as accidental withdrawal (not ejection or dropout) of control rods. D1Kussion Due to the non-operating and defueled condition of Thil 2, the requirements of Criterion 25 are not applicable. 31.2.22 Criterion 26 - Reactisity Control Systern Redundancy and Capability Two independent reacthity control systems of different design principles shall be provided One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design lunits are not exceeded The second reactisity control system shall be capable of reliably controlling the rate of reacthity changes resulting from planned, normal power changes (including xenon bumout) to assure acceptable fuel design limits are not f exceeded one of the systems shall be capable of holding the reactor core subentical under cold i conditions. JhKumDD Due to the non-operating ar.d defueled condition of Thil 2, the requirements of Criterion 26 are not applicable. 3.1.2.23 Cnterion 27 - Combined Reactisity Control Systems Capabihty The reactivity control systerns shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropnate margin for stuck rods the capability to cool the core is maintained. Discussion Due to the non-operating and defueled condition of Thil 2, the requirements of Criterion 27 are not applicable. 3.1.2.24 Criterion 28 - Reactivity Limits The reactisity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactisity accidents can
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[O 3.1 -31 UPDATE 2 - AUGUST 1997
ueither (1) result m damage to the reactor coolant pressure boundar> greater than hrrutext local yibiding r,or (2) sufficiently disturb the core, its rupport structures of other reactor pressure vessel internals to impair significantly the capabihty to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), in:t dropout, steam line rupture, changes in reactos coolant ternperature and pressure, and cold water addition. DittuitiOD Due to the rnn operating and defueled condition of Th!I 2, the requirements of Cnterion 28 are not applicable. 31.2.25 Cnterion 29 Protection Against Anticipated Operat;onal Occunences A protection and reactisity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occunences. D.lituitiOB Due to the non operating aad defueled condition of Tht! 2, the requirements af Cnterion 29 are not applicable. 31.2.26 Criterion 30 - Quahty of Reactor Coolant Pressure Bourxiary Components vhich are part of the reactor coolant pressure boundary shall be designed, fabrica ed, erected, and tested to the highest quality standards practical Means shall be provided for detecting and, to the extent practical, identifyint; the location of the source of reactor coolant leakage. HiKnitiOD Due to the nor operating and defueled condition of Tht! 2, the requirements of Cnterion 30 are not applicable. 3.1.2.27 Cnterion 31 Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed with sutlicient margin to assure that when stressed under operatmg, maintenance, testing, and postulated accident conditions (1) the boundary behaves m a nonbrittle manner and (.?) the probabihty of rapidly propagating fracture is minimized The design shall reflect consideration of smice temperatures and other conditions of the loundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties m determining (1) matenal properties, (2) the effects ofirradiation of matenal properties, (3) esidual, steady state and transient stresses, and (4) size of flawt DiKustico Due to the non+perating and defueled condition of Thil 2, the requirements of Cnterion 31 are not apphcable. O 3.1 -32 UPDATE 2 - AUGUST 1997
3.1.2.28 C iterion 32 Inspection of Reactor Coolant Pressure Boundary O Components which are part of the reactor ccalant pressure boundary shall be designed to permit (1) periodic inspection and testing ofimportant areas and features to assess their structural and leaktight integnty, and (2) an appropriate material surveillance prograrn for the reactor pressure vessel DEunieri Due to the non-operating and defueled condition of TMI 2, the requirements of Criterion 32 are not applicable. 3.1.2.29 Cnterion 33 Reactor Coolant Makeup A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided The system safety function shall be to assure that specified acceptable fuel design lunits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure 1 oundary and rupture of small piping or other small components which are part of the boundary. 'Ihe system shall be designed to assure that for onsite electric power system operation (assuming off site power is not available) and for off site electric power system operation (assuming onsite power is not asailable) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant imentory during normal reactor operation. DiKunico Due to the non operating and defueled condition of TMI.2, the requirements of Criterion 33 are not applicable. 3.1.2.30 Cnterion 34 Residuallleat Removal A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor ccre at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be prmided to assure that for onsite electric power system operation (assuming off site power is not available) and for off site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assunung a single failure. DiKunipD Due to the non+perating and defueled condition of TMI 2, the requirements of Criterion 34 are not applicable. 31.2.31 Criterion 35 Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided The system safety function shall be to transfer heat from 'he reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal water reaction is limited to negligible amounts, j 3.1 - 33 UPDATE 2 - AUGUST 1997 l
Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, _ and containment capabilities shall be provided to assure that for onsite electric power system operation (assurning off site power is not available) and for off-site electric power system operation (assurning onsite power is not available) the systern safety function can be accomplished, assuming a single failure. DJitu11100 Due to the non operating and defueled condition of TA112, the requirements of Cntenon 35 are not applicable 3.1.2.32 Criterion 36 -Inspection of Emergency Core Coolmg System The emergency core cooling system shall be designed to pennit appropnate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water mjectioa nozzles, and piping, to assure the integrity and capabihty of the system. DiKui$ IGD Due to the non operating ano aefueled condition of Th112, the requirements of Cntenon 36 are not applicable. 3.1.2.33 Criterion 37 Testing of Emergency Core Cooling System The emergency c,re cooling system shall be designed to pennit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integnty ofits componerus, (2) the operability and performance of the active components of the system, and (3) the operabihty of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that bnngs the system into operation, includmg operation of applicable pouions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. Discissieu Due to the non-operating and defueled condition of Thil 2, the requirements of Cnterion 37 are not apphcable. 3 1.2.34 Criterion 38 - Containment fleat Renwval A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pre .sure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabihties shall be provided to assure that for onsite electnc power system operation (assuming off site power is not available) and for off site c!ectric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure. Durn 31100 Due to the non-operating and defueled condition of Thil 2, the requirements of Criterion 38 are not applicable. 3.1 - 34 UPDATE 2 - AUGUST 1997
3.1.2.35 Criterion 39. Inspection of Containment }{ cat Removal System The containment heat removal system shall be designed to permit appropriate periodic instation of important components, such as the torus, sumps, spray nor21:s, and piping to assure the integrity and capability of the system. : Discussion Due to the non+perating and defueled condition of Thti 2, the tequirements of Cntenon 39 are not , apphcable. 3.1.2.36 Cnterion 40 - Testing of Containment licat Removal System The containment heat removal system shall be designed to pernut appropriate periodic pressure end functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to the design as practical the performance of the full operational sequence that brings the system into operation, includmg operation of applicable portions of the protection system, the transfer between normal and emergency power sources and the operation of the associated cooling water systern. Dncussion Due to the non+perating and defueled condition of Thil 2, the requirements of Criterion 40 are not applicable. ( 3.1.2.37 Critenon 41 Contamment Atmosphere Cleanup Systems to control fusion product, hydrogen, oxygen, and other substances which may be released in the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associa'.ed systems, the concentration and quality of fission products released to the emironment following postulated accidents, and to control the concentration of hydrogen er oxygen and other substances in the containment atmosphere followmg postulated accidents to assure that containment integnty is maintained Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming off site power is not available) and for off site electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure, Discussing Due to the non operating and defueled condition of Thil 2, there are no postulated accidents dunng PDhf S which could result in the generation of fission products, hydrogen, oxygen or other substances which would require Con *.ainment atmosphere cleanup systems as described in Criterion 41. Therefore, I design of the Containment atmosphere cleanup system for Thti 2 dunng PDhtS in accordance with l Criterion 41 is not applicable. See the analysis in Section 8.2. l [ () V l 3.1 - 35 UPDATE 2 - AUGUST 1997
3.1.2.38 Criterion 42 Inspection of Containtnent Atmosphere Clearmp Systems he containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection ofirnportant components, such as f her frames, ducts, and piping to assure the integrity and capabihty of the systems. Ditanico Due to the non-operating and defueled condition of Thfi 2, there are no postulated accidents during PDhtS which could result in the generation of fission products, hydrogen, oxygen or other substances which would require containment atmosphere cleanup systems as descnbed in Cnterion 41. Herefore, design of the Contaminent atmosphere cleanup system in accordance with Criterion 42 is not applicable to Thil 2 dunng PDhtS. 3.12.39 Criterion 43 - Testing of Containtnent Atmosphere Cleanup Systems The containtnent atmosphere cleanup systerns shall be designed to pentut appropnate penodic pressure and functional testmg to assure (1) the structural and leaktight integnty of its components, (2) the operability and perfenutnce of the active components of the systems st:h as fans, filters, dampers, pumps, and valves sud (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the s) stems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems. E1G11 ion Due to the non operattng and defueled condition of Thil 2, there are no postulated accidents dunng PDhtS which could result in the generation of fission products, hydrogen, oxygen or other substances which would require containment atmo.phere cleanup systems as desenbed in Critenon 41. Herefore, design of the Containnunt atmosphere cleanup system in accordance with Cnterion 43 is not applicable to Thil 2 during PDhtS. 1 1.2.40 Cnterion 44 Coolmg Water A system to transfer heat from structures, systerns, and cornponents important to safety, to an ultimate heat sink shall be prosided De system safety function shall be to transfer the combined heat load of these structures, synems, and components under normal operation and accident conditions. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabihties shall be provided to assure that for onsite electric power system operation (asrumirg off site power is not available) and for off site electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assunung a single failure. LhtGMien l Duc to the noneperating and defueled condition of Th112, the requirements of Cnterion 44 are not applicable. l O 3.1 -36 UPDATE 2 - AUGUST 1997
3.1.2.4 i Criterion 45. Inspect on of Cooling Water System The cmling water system shall be designed to permit appropnate periodic inspection of important comporents, such as heat exclumgers and piping, to assure ti e integnty and capability of the system i Danttii2D ; Due to the non operating and defueled condition of TM12, the requirements of Criterion 45 are not applicable. 3.1.2 42 Criterion 46. Testing of Coolmg Water System The coolmg water system shall be designed to permit appropriate periodic pressure and functional resting to assure (1) the structural and leaktight integrity of its components, (2) the operabihty and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the perfonnance of the full operational sequence that brings the syste n into operation for reactor shutdown and for loss +f<oolant accidents, including operation of applicable portions of the protection system and the transfer betwem normal and emergency power sources. Disniden Due to the non-operatmg and defueled condition of TMI 2, the requirements of Cnterion 46 are not applicable. 3.1.2.43 Cnterion 50. Containment Design 11 asis
" Die reactor containment structure, includmg access openings, penetrations, and the containment heat remo.al system shall be designed so that the containment structure and its intemal cornpartments can accommodate, without exceedmg the design leakage rate and with suflicient margin, the calculated pressure and ternperature conditions resulting from anyloss of coolant accident This margin shall reflect consideration of(1) the efTects of potential energy sources which have not been meluded in the determination of the peak conditions, such as energy in steam gercrators and as required by article 50.44 energy from metal water and other chemical reactions that may result from degradation but not total failure of cmergency core cooling functioning, (2) the hmited experience and experirnental data available for derming accident phenomena and containment responses, and (3) the conservatism of the calculational model and input parameters.
Diguiston Due to the non+perating and defueled condition of TMI 2, the requirements of Cnterion 50 are not applicable. llovvver, the intent of Critenon 50 has been addressed by considering that there are functional requirements required to be provided by the Containment during the PDMS period which are different, and substantially less than those described in Cnterion 50. The Containment, during PDMS, is required to function as the primary barrier between the radioactive contamination inside the Containment and the emironment. It provides this function m three ways 1) it mmimizes the uncontrolled migration of contamination from inside the Containment to the emirunnent,2) it functions as an emtlope to control the release of Containment atmospheie effluents to the emironment, and 3)it functions as primary shieldtng for the radioactive materials inside the Containment.
- 3.1 37 UPDATE 2 - AUGUST 1997
Tie Contaimnent was origitally designed and constructed to ti e criteria descrited in Cnterion 50. Smce the design basis requirements for the Containmeat dunng PDMS are substantially less than that stquired by Criterion 50, the Containment isolation capabilities were not degraded by either the accident or the recovery actinues, the Containment is capable of meeting the requirernents of PDMS. (See Section 7.1.2) 3.1.2 44 Cntenou $1. Tracture Prevention of Containment Pressure 11oundary Tie reactor containment toundary shall be designed with sufficient margm to assure that under operation, maintenance, tesung, and postulated accident conditions (1)its ferntic materials behase m a nonbnttle mannet and (2) the probability of rapidly propagating fracture is mimrnized The design shall reflect coraideration of sersice ternperatures and other cond;tions of the containtnent toundary matenal dunng operadon, mainterance, tesung, and pastulated accident condiuons, and the uncertainties in detennining (1) matenal properues. (2) residual, steady state, and tiarnient stresses, and (3) siit of flaw s DiiGMiUD Due to the non opcraung and defueled condition of TMI 2, the requirements of Cnterion $1 are not appbcable. 3.1.2 45 Cnterion 52. Ctpability for Containment Leakage P. ate Tesung The reactor containment and other equipment which may be subjected to nntaintnent test condiuons shall be designed so that penodic integrated leaka re rate testing can te conducied at containment design pressure. DilWish!!)
'the requircinents for leakage rate testmg canability as established in Cntenon 52 assume the possibihty of pressurimuon events and the cornoquent potential for leahage of fission products. Due to the non operaung and defueled condition of TMI 2, there are no eg ents uidch can result in sigmficant pressuritation of the Containment and threaten its isolation capabilities (See Cnterion 50) During passive pressure control, the Containment will be conunually vented to the APill3 through a passive breather system. During operation of the Containtnent Purre System, a filtered, monitored exhaust path is prm1ded via the stauon vent to the atmosphere. Therefore, no sigmficant pressuritation of the Contamment could occur dunng this mode of operation liased on the above conditions, no leak rate testing at containraent design pressure is required during PDMS, 3 1.2.46 Cntenon $3 Prmasions for Containment Testing and Inspection
'The reactor containment shall be designed to penmt (1) appropriate periodic inspection of all important areas.
suc h as penetradora, (2) an appropnate surveillance program, and (3) penodic testing at contamment design pressure of the leaktightness of penetrations which have resilient seals and expansion trilows. DiKituiOD The Contaitunent has teen designed and constructed in accordance with Criterion 53. Due to the unique condition of TM12 during PDMS, the specific requirements of Cnterion $3 are not applicable, however, the intent of Cnterion $3 has been addressed by providing the appropriate surveillance acuvities ba J on the Containment isolation requirements for PDMS as descrited in Section 7.2.1, O 3.1-38 UPDATE 2 - AUGUST 1997
J p 3.1.2.47 Cnterion 54 Piping Systems Penetrating Containment Piping systems penetrating primary reactos containment shall be provided with leak detection, isolation,
~
omd containment capabihties having redundancy, niiabihty, and performance capabilit es which reflect the importance to safety ofisolating these piping systerns. Such pipinc systems shall be designed with a capabihty to test penodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. , DiKMilco Due to the non operating and defueled condition of TMI 2 during PDMS, the specific requirements of Cnterion $4 regarding leak detection, isolation, and containment capabilities are not applicable. Ilowever, piping systems penetrating Containment have been isolated and will be maintained isolated during PDMS as desenbod in Section 7.2.1. 3.1.2.48 Cntenon 55. Reactor Coolant Pressure Boundary Peaetratmg Containment Each line that is part of the reactor coolant pressure boundary and that penetratt primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation pro isions for a specific class oflines, such as instnuaent lines, aic acceptable on some other defined basis:
- 1. One locked clcsed isolation valve inside and one locked closed isolation valve outside contaitunent; or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside containment; or
- 3. One locked closed isolation valve inside and one automatic isolation valve outside containment. A siinple check valve may not be used as the automatic isolation valve outside containment; or
- 4. One automatic isolation valve inside and one automatie isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.
Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety. Other appropriate requirements to minimize the probability of consequences of an accidental rupture of these lines or oflines connected to them shall be provided as necessary to assure adequate safety. Detennination of the appropriateness of these requiremeats, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation vahrs and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site emiron<
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3.1 -39 UPDATE 2- AUGUST 1997
Discussion Due to the non operating and defueled condition of Thil 2 there is no reactor coolant pressure boundary, thuefore, the specific reqairernents of Criterion $5 are not applicable. Ilowrver, the intent of Cnterion 55 les been addressed for PDhtS. All piping which penetrates the Containment has been isolated as desmbed in Section 7.2.1. 1 1.2.49 Criterion $6 Primary Contairunent isolation Each line that connects directly the containment atmosphere and penetratt. primary reactor containtnent shall be prmided with containment isolation valves as follows, unless it can be demonstrated that the contairunent isolation provisions for a specific class oflmes, such as instrument lines, are acceptable on Some other dermed basis:
- 1. One locked closed isolation valve inside and one locked closed isolation valve outside containment; or
- 2. One automatic isolation valve inside and one locked closed isolation valve outside contairunent; or
- 3. One locked closed isolation valvt inside and one automatic isolation vr tre outside containment. A simple check valve may not be used as the automatic isolation vahe outside contairunent; or
- 4. One autorrntic isolatiori valve inside and one Du!Omatie isolation valve outside containitient . A simple check valve may s.ot be used as the automatic isolation valve outside c,.it tintnent.
Isolation vah es outside containment shall be located as close to contamment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety. DJKtlitiOH Due to the unique condaion of Th112 during PDhtS, the specific requiiements of Criterion 56 are not arplicable; however, the intent of Criterion 56 has been addressed for PDhtS. Piping systems penetrating Containment have been isolated outside Contain...ent and will be maintained isolated Due to the non-operatmg and defuele nndition of Th112, one closed isolation valve outside Centainment on each piping penetration provw suitable Containment isolation during PDh15. See Section 7.2.1. 3.1.2.50 Criterion 57 Closed System Isolation Valves Each line that penetrates primary reactor containment and is e ter part of the reactor coolant pressure boundary or connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of rernote manual operation. This valve shall be outside containment and located as close to the containment as prvctical. A simple clwk vaht may not be used as the automatic isolation vah c. O 3.1 40 UPDATE 2 - AUGUST 1997
1 l l DiKunioD
,m (V) Due to the unique conition of Th!I 2 during PDhtS, the specifie requirements of Criterion 57 are not applicable; however, the intent of Criterion 57 has been addressed for PDhtS. All piping systems penetrating Containment have been isolated as desenbed in Secuon 7.2.1.
3.1.2.51 Cntenon 60 Control of Releases of Radioactive hiaterials to the Emironment The nuclear power unit design shall include rneans to control suitably the release of radios.ctive matenals in gaseous and liquid emuents and to handle raicactive solid wastes produced during normal reactor operation, including anticipated operational occunences. Sumcient holdup capacity shall be prosided for retention of gaseous and liquid emuents contatrung raicacuve materials, panicularly where unfavorable site environmental constions can be expected to impose unusual operational Imutations upon the release of such emuents to the emironment. Dixunjen Due to the unique constion of Th112 dunng PDhtS, the specif e requirements for Criterion 60 are not applicable, however, the intent of Criterion 60 has been addressed by provisng means to suitably control releases of radioactive materials to the emitortment dunng PDhtS. 3 1.2.52 Criterion 61 Fuel Storage and llandling and Radioactivity Contrcl ne fuel storage and handhng, ra6oactive waste, and other systems wiuch may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to pernut appropriate perio&c inspection and testing of
")
(O components important to safety, (2) with suitable shielding for radtation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having rehability and testability that reflects the irnportance to safety of decay heat and other r-i641 heat removal, and (5) to prevent significant reduction in fuel storage coolant incr.;wy under accident conditions. D1Kuulen Due to the non.cperating and defueled condition of Th112, the requirements of Cnterion 61 with regaids to fuel handling and storage are not applicable. Ilowever, small quantities of residual fuel remein in vanous locations within the Reactor Coolant System and in other areas of the Reactor Building; the Post Defuehng Survey Repons (Referenecs 3.1 8 through 12) identified the quantity of residual fuel in each defined location and addressed the potential for fuel relocation. As discussed in Section 4.3 4, the enticahty analyses provided in the Defuelmg Completion Report (DCR) (Reference 3.17) and in GPU Nuclear letter, C312 92-2380, dated December 18,1992 (Reference 3.1 13) demonstrated that criticahty has been precluded at Th112, Finally, personnel accessibility, potential exposure, and other protective features for the residual fuel and other radioactis e material are provided consistent mth the requirements of Criterion 61. 3 1.2.53 Cnterion 62 - Prevention of Criticality in Fuel Storage and llandhng ( Cnticality in the fuel storage and handling system shall be prevented by physical systems or processes, l l preferably by use of geometrically safe configuratiocs. l l l 3.1 - 41 UPDATE 2 - AUGUST 1997
1 DiK @ilen P e to the non-operating and defueled condition of Thil 2, the requirements of Criterion 62 are not applicable. See Section 4.3. 3.1.2.54 Cnterion 63 - hiomtorms Fuel and Waste Storage Appropriate s) stems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capabihty and excessive radiation levels and (2) to initiate appropriate safety actions. DIKuitieu Due to the nn-operating and defueled condition of Th112 there will not be any materials which can generate sufficient decay heat to require residual heat removal capabilities. Therefore, the requirements of Criterion 63 are not applicable. 31.2.55 L,terion 64 - hionitoring Radioacthiry Releases hicans shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation ofloss of-coolant accident fluids, effluent discharge paths, and the plant ernirons for radioactivity that may be released from normal operation, includmg anticipated operational occurrences, and from postulated accidents. DEttulen Due to the unique condition of Th112 danng PDhtS, the specific requirements of Critenon 64 are not applicable; however, the intent of Cnterion 64 has been addressed by prosiding means to monitor radioactivity releases, as described in Sections 7.2.1.2 and 7.2.4, commensurate with the plant condition during PDhtS. O l 3.1-42 UPDATE 2 - AUGUST 1997
PIIIEMCES 3.11 Letter, Travers, W. D. (NRC) to Standerfer, F. R. (GPUNC), ' Approval of Exempdon from 10 CFR s $0.61,* dated Decemter 30,1985. 3.1 2 letter, Snyder. D. J. (NRC) to Kanga D. K. (GPUNC), *10 CFR $0.49, 'Emironrnental Qualification of Electrical Equipment important to Safety for Nuclear Power Plants',' dated July 22. 1983. 313 Letter Stolz, J. F. (NRC) to Standerfer. F. R. (GPUNC), *lssuance of Arnendment (T AC No. 65337).* dated hiay 27,1988. 3.1-4 Letter, Snyder, D. J. $RC) to llovey, O K. (hiet Ed), Re: Exemption frorn 10 CFR 30 Appendix J, dated Septernber 2,1981. 3.1 5 Letter, Snyder, D. J. (NRC) to llovey, G. K, (Met Ed1, Re: Rehef from the Insenice inspection Program Requirements of 10 CFR $0.55a, dated April 27,1981. 3.16 GPU Nuclear letter,4410 90 L 0044,'Deconunissionirig Financial Assurance Certification Report for ... Thil 2
- dated July 26,1990.
3.1 7 GPU Nuclear letter,4410 90 L 0012, 'Defueling Completion Report, Final Submittal," dated February 22,1990. 3.1 8 GPU Nuclear letter, C312 91204$ 'SNh1 Accountability," transmitting the Auxiliary and Fuel llandhng Duildings PDSR, dated June 7,1991. 3.1 9 GPU Nuclear letter, C312 912052, 'SNht Accountabilit)," transmitting the Reactor Buildmg hitscellaneous Cornponents PDSR. dated June 18,1991. 3110 GPU Nuclear letter, C312 91205$, 'SNh! Accountability ' transmitting the Reactor Coolant System PDSR, dated July 3,1991. 3.1 11 GPU Nuclear letter, C312 91 2064, *SNh! Accountabilit),' transmitting the 'A' and 'D' Once. Through Steam Generators PDSR, Revision 1 dated July 3,1991. 3.1 12 GPU Nuclear letter, C312 9.%2004, *SNh1 Accountability, trans.* titting the Reactor Vessel PDSR, dated February 1,1993. 3.1 13 GPU Nuclear letter, C312 92 2080, 'Thil 2 Reactor Vessel Cnticahty Safety Analyses,' dated December 18,1992 r ( 1 1 3.1 43 UPDATE 2 - AUGUST 1997
3.2 CLASSIFICATION OF STRUCTURES SYSTEh15, AND COh!PONENTS 3.2.1 SEIShilC CLASSIFICATION The unit structures, systems, and components have been classified accordmg to their function and the degree ofintegnty required to protect the public. Structures, components, and systems are classified for seismic design purposes as either Category I or Category 11. ne onginal seismic design criteria for structures, cornponents, and systems utilized during PDhtS is given in Section 3.2 of the Thll 2 FSAR. 3.2.1.1 Seismic Category I Seismie Category I structures, systems, and components, including instruments and controls, are those which are necessary to ensure:
- a. He integrity of the teactor coolant pressure boundary,
- b. He capabihty to shutdown the reactor and to rnaintain it m a safe shutdown condition, or
- c. De capability to prevent or mitigate the consequences of accidents which could result in potential off site exposures comparable to the guideline exposures of 10 CFR Part 100.
%e first enterion requires the reactor coolant pressure boundary to be ensured. Due to the non-operating and defueled condition of Tht! 2, there is no reactor coolant pressure boundary.
The second criterion requires the capabihty to shutdown the reactor and maintain it in a safe shutdown f N condition Due to the non operating and defueled condition of TMI 2, there is no reactor to shutdown and maintain in a r.afe shutdown condition. In addition, the enticahty analysis discussed in Section 4.3.4 demonstrates that enticality has been precluded at %112. The third criterion requires the capability to prevent or mitigate the consequences of accidents that could result m potential off site cAposures comparable to the 10 CFR Part 100 guidelines. Analysis demonstrates (see Chapters 7 and 8) that there are na postulated events that could result in releases greater than 10 CFR $0 Appendix I hmits, therefore there are no postulated events which could result in exposures comparable to 10 CFR Part 100 g.iidelines. Due to the non operating and defueled status of Tht! 2 dunng PDh1S, there are no structures, systems, or components v hich are required to meet the enteria of seismic Category i nerefore, there are no structures, systems, or cornponents classified as seismic Category I at Th112 during PDh1S. nose structures, systems, and cornponents which were originally designed as seisnue Category I and the applicable design criteria are described in Section 3.2.1 of the Thil 2 FSAR. fN O 3.21 UPDATE 2 - AUGUST 1997
3.2.1.2 PDMS Category 11 Those atmetures, components and systems which are relied upon for the isolation of residual contamination from the environment and for the prevention of an uncontrolled release of radioactivity during a seismic event have been designated as PDMS Category 11. 3.2.1.2.1 Design Basis All structures, components, and systems which wrre designed to seismic Category I or seismic Category 11 are in conformance with the seismic loadmg requirements of the Uniform Buildmg Code for Zone 1. Those structures, systems, and components that were originally designed and fabncated to seismic Category I or seismic Category 11 requirements have not had their origmal structural design capability sigruficantly degraded and will meet the seismic loadmg requisements of the Uniform Building Code for Zone 1. 3.2.1.2.2 PDMS Category 11 Structures Structures designated as PDMS Category 11 are the Reactor Buildmg, the Control and Senice 11uildmgs, the Auxiliary Buildmg, and the Fuel llandling Building. These five structures were originally designed to scistrde Category I requirements. 3.2.1.2.3 PDMS Category !! Systems As a result of the accident on March 28,1979, various onginal plant systems were contammated with radioactive materials from the reactor core. These systems have been decontaminated to the extent practical. Ilowever, a number of these systems still contain some degree of residual contamination and provide an initial barner for control of the contamination which remains within the respective syster'- The Containment provides a second barrier for those systems or partial systems located within tir stnseture. The structural capabihties of the syuems as originally designed to seisnue Categord and seismic Category 11 crit ria have not been degraded significantly and these capabilities proside the necessary degree of contamination control. These PDMS Category 11 systems and their original seismic design entena are listed in Table 3.2 1. Other piping =ystems designated as PDMS Category 11 which were not designed to seismic Category I or scismic Category 11 cnteria also provide an irutial barner for control ofinternal contamination durmg the PDMS period. Since these systems were designed to function under equal or higher pressure and temperature conditions, they will maintain structural integnty under the conditions which exist durmg PDMS. In addition, deactivated systems have been drained to the extent practical, thus, mininuzing contaminated leakage from these systems. A breach in system piping which has been deactivated would expose only a very small portion of the piping systems intemal surface contamination and as deactivated systems are not pressurized, there is no motive force to expel contammation Therefore, no further design, analysis, testing, or surveillance is required to assure the structural integnty of those systems relied upon during PDMS. These PDMS Category 11 systems are listed on Table 3.2 1 and are identified as not being designed to seismic enteria (N/S). O 3.2 - 2 UPDATE 2 - AUuUST 1997
3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION j f 3.2.2.1 Identification of Safety Related Systems and Components l Due to the non<perating and defueled status of TMI 2 during PDMS there are no safety related rystems or components. 'the original quality group classifications and other design criteria for the ! original plant systerns listed in Table 3.2 1 are given in Section 3.2 of the TMI.2 FSAR. TABLE 3.21
)
PDMS CATEGORY 11 SYSTEMS l 2 FSAR IN OPERATIONAL / SYSTEM SE!SMIC Cl.,ASj CONTAINMENT DEACTIVATED Core Flood I") PART?) DEACT. RH Spray i PART. DEACT. Decay Heat Removal i PART. DEACT. Steam Generator Recirculation N/S NO DEACT. ! Spent Fuel i PART. DEACT. Main Steam I") PART. DEACT. Decon Process Water N/S PART. DEACT. Dewatering Station N/S NO DEACT. Fuel Transfer Canal N/S PART. DEACT. Fill and Drain Makeup & Purification i PART. DEACT. Nuclear Plant Nitrogen 11 PART. DEACT. Nuclear Sampling I"' PART. DEACT. Temporary Nuclear Sampling N/S") NO DEACT. Steam Generator Secondary Vent I & 11 YES DEACT. and Drain inside Containment ! Waste Disposal Gas 1* PART. DEACT. Defueling Water Cleamg N/S PART. DEACT. Reactor Coolant i YES DEACT. Waste Disposal Solid i PART, DEACT. Feelllandling i PART. DEACT. Submerged Demineralizer N/S NO DEACT. ( Reactor Building Sump . N/S PART, OPER; (l.crel Measurement) 3.2 - 3 UPDATE 2 - AUGUST 1997
,ewww--mwe y w y-g y py- *-vg-e- y- ---
yv-wy y-i-g-ie-- au, e
- - - - - eieu.-- .es -w.e-a,e- ge ws w ,- ge,ys.gi e - we. a.%- - w yy y y p- ---eg-,rg g.,e9g.,p.g .,pg*rgim..-s-se gyf-gm + ee m,s y-9g--,-wwy,9w,
TAllLE 3.21 (Cont'd) PDMS CAT EGORY 11 SYS1 EMS ISAR IN OPERAT!ONAIJ SYSTEM SEISMIC Cl, ASS CONTAINMENT DEACTIVATED Sump Water Sucker N/S PART. DEACT. Processed Water Storage N/S NO DEACT. and Distribution Canister lading Decon N/S NO DEACT. Defu?ing* N/S YES DEACT, Deca Filter Skid N/S NO DEACT. Sludge Transfer N/S PART. DEACT. Reactor !!uilding Purge 11 NO OPER. Exhaust Upstream ofIlEPA Filter Waste Disposal Liquid I&il PART, OPER Km N/S Non-scismic PART. Partially I LI . 0; e a i al MQ1IS: (1) Execpt for N2 supply lines from NM U 26 and 27 to CF V 114 and i15. (2) All of system is in containment except for 1 inch fill and drain lines (3) Seismic I up to MSIVs and EF P 2. All else is Seismic 11. Portions of Seismic 11 piping are contarninated (4) Majority of system is Seismic 1. Seismic 11 portion is very small and contains a small pump and associated 1/2 inch tubing. (5) System consists of mainly 1/2 inch arv; ~/8 inch diameters. (6) System was designed and constructed Seismic I except for 2 flexible connections to the secondary vent and drain system. A few non-seismic modifications were made dunng the course of the cleanup effort, however, the majority of the system remams seismically qualified (7) System consists of the defuelmg equipment that remains in the RY. O 3.2 - 4 UPDATE 2 - AUG'JST 1997
3J WIND AND TORNADO LOADINGS 3.3.1 WIND LOADINGS 3.3.1.1 Design Wind Velocity Recurence intervals, data sources and the history of occurrence of high winds, hurricanes and torraloes m the vicinity if the site r.re discussed in Sections 2.3.1 and 2.3.2 of the TMl 2 FSAR. Lased on studies by the Weather Bureau, wind speeds 30 feet above the site grade are expected to exceed 78 mph once in 10^ years. llence, all structures were designed for forces associated with a wind of 80 mph auf, with limited exceptions as described in SAR Section 3.7.2.1.1, exist as originally 9' designed. Additional design information is given in Section 3.3 of the TMI 2 FSA" r $s 3.3.2 TORNADO LOADINGS 3.3.2.1 Applicable Design Parameters The forces due to tornado loadmg have been assumed as the forces associated with a wmd iming a velocity of 360 mph. ne 360 mph velocity is considered as the resultant of a 300 mph tangential and 60 mph translational velocity of the storm. A differential pressure of 3 psi between inside and outside has been considered in the design of the Reacto Building. All other structures designed for tornado loadings are provided either with adequate areas of openings to relieve the differential pressure of 3 psi in 3 seconds or are designed to withstand an extemal pressure of 3 psi. \\ 3.3.2.2 Tomado Missiles i ( The missiles assumed to have been generated by the tomado event are listed in Table 3.3-1. ney include items such as siding panel, pipe, and steel plate which could be detached from structures under tomado-associated loadmgs. A1: these missiles have been investigated for their ability to penetrate exterior concrete walls and slabs. He maximum penetration of a concrete barrier was found to be 30 in. De mirdmum thickness of structures designed to withstand tomado generated missiles is 36 in. Their efTects on the buildings designed for tomado loadings were considered together with the effects of loading,s described in Section 3.3.2.1. Thus, with limited exceptions as described in SAR Section 3.7.2.1.1, the capabihty of buildings designed for tornado loadings will not bejeopardized as a result of flymg objects from other structures under a tomado event. K 6: 1 3.3-1 UPDATE 2 - AUGUST 1997
l 1 O TABLE 3.3-1 TORNADO GENERATED MISSILES Impact Impact Ele.2 tion Impact Weight Area Abirre Grat V .locity
,{lbl, (Lq.,[t) ,_(h; , inn.h)
- 1. Utihty pole 1200 1.0 23 200
- 2. Passenger Auto 2000 25 25 100
- 3. Passenger Auto 4000 30 3 100
- 4. Concrete Fragment 4500 30 5 60 10' dia. x 3' x 4"
- 5. Steel plate 1000 1.0 10 200
- 6. Crated motor 1000 15 10 200
- 7. Pipe,4" dia. x 250 0.15 25 200 12'-0"
- 8. Wood plank i10 0.33 any 360 4" x 12" x 12'-0" height
- 9. Street light fixture 25 0.5 any 360 height D. Crushed rock,1-1/2" .25 0.01 any 360 height 11, Sidmg Panels 400 0.04 any 360 1.5' x 50' height
- 12. Tools 125 0.1 any 360 height
- 13. Ductwork 150 0.3 any 360 height
- 14. 8' Handrail 50 0.05 any 360 Section height O
3.3 - 2 UPDATE 2 - AUGUST 1997
~_ _ _ __.
3.4 WATER LEVEL (FLOOD) DESIGN p) i
'v' Flood history, flood design consideration, design of hydraulic facilities and emergency operation requirements for the unit facilities are discussed in Section 2.4.
3.4.1 FLOOD ELEVATION The foundation mats, exposed walls and flood pancis of those stmetures designed to withstand floods are designed to withstand the hydrostatic pressures associated with a water level of 31 l'-0". 3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATIONS Due to wave action, the maximum water level considered is 4' 0" higher than the PMF elevation expec*ed around the station facilities. See Section 2.4.3 of the TMI 2 FSAR. No other phenomena have been considered in the design load calculations. 3.4.3 FLOOD FORCE APPLICATION Static forces on the flood panels and vertical walls are calculated considering hydrostatic pressures. Dynamic forces on the flood panels and vertical walls are calculated considering wave action. The buoyancy forces on the mats are calculated considering the uplift due to the height of water above the bottom of the mats. 3.4.4 FLOOD PROTECTION O Unit design is based on a water elevation of 308'-6" on the west side and 308'-0" on the cast side under h flood conditions. Structures which originally contained Engineered Safety Feature equipment are scaled against entry of flood water to an elevation of 312'-6" on the west side and 312'-0" on the east side. Complete protection has been provided at the exterior faces of these structures. The waterstops between adjacent building walls and mats are capable of withstanding a maximum water head of 45 feet which is in excess of maximum associated head for the flood level. He exterior sliding doors and flood panels are provided with watenight seals. l0 l O 34-1 UPDATE 2 - AUGUST 1997
3.5 MISSILE PROITCTION CRITERIA (~\) s The missile protection criterion (i.e., GDC 4; see Section 3.1.2.4) is based on precluding damage to structures, systems, and components important to safety. During PDMS, there is no equipment that is impodant to safety; therefore, the consequences of a missile impact are limited to physical damage. With limited exceptions as described in Section 3.7.2.1.1, all contammation isolation structures are protected from a loss of function due to damage frorn potential extemal missiles as described below. Due to the non-operating and defueled status of TMI-2 during PDMS, there are no postulated internally generated missiles which would have any effect on the capability of structures to provide their required contamination isolation function. 3.5.1 MISSILE LOADINGS AND BARRIERS ne exterior walls of the Reactor Buildmg. Auxiliary and Fuel Handling Buildings and the Control & Senice Buildmgs were designed to withstand the effects of extemal missiles. 3.5.2 MISSILE SELECTION ne potential sources of extemal missiles have been investigated for those buildmgs designed for missile protection. Missiles generated by emironmental conditions and a postulated aircraft i.npact have been considered. The critical possible missiles have been selected and analyzed. They are described in the following paragraphs. 3.5.3 SELECTED MISSILES 3.5.3.1 Tomado Generated Missiles ( The residual contarnination which remams at TMI 2 is contained within buildmgs that have been designed to withstand tomado generated missiles as discussed in Section 3.3.2.2.3 3.5.3.2 Aircraft Impact he probability of aircraft impact on the station has been studied. The results of evaluations show that the probability of an aircraft impact on the unit is very low (see Section 2.2.3.1.2 of the TMI 2 FSAR and Reference 3.5 1). In addition, all contamination isolation structures' were designed to withstand a hypothetical aircraft incident. The impact loadings on these structures are assumed to come from:
- 1. An aircraft of 200,000 lbs. traveling at 200 knots impacting on an effective area of 19 ft.
diameter
- 2. An aircraft of 300,000 lbs. traveling at 200 knots impacting on an area of 16 ft. dian.eter
- 3. An object of 6000 lbs. traveling at 200 knots impacting or. an effective area of 5 ft. diameter
- 4. An object of 4000 lbs. trawling at 200 knots impacting on an effective area of 3 ft. diameter, ne results of these analyses provide assurance that with limited exceptions as described in S AR Section 3.7.2.1.1 the hypothetical objects wou'd not perforate or cause collapse of any of
,, these structures.
/ L )\
" 1 he "Other Buildings" from Table 5.3-6 that contain less than lpCi residual contaminatian are excluded from this discussion.
3.5-1 UPDATE 2- AUGUST 1997
3.5.4 DARRIER DESIGN PROCEDURES With limited exceptions as described in SAR Section 3.7.2.1.1, structures are protected against the effects of all possible missiles, described above, by concrete barriers. The effects depend on the missile features such as weight, shape and velocity, as well as on the barrier features such as material properties and geometry of the barrier. These effects can be classified as follows: 3.5.4.1 Overall Structural Effect The overall structural effect of a missile impacting on a structure is evaluated by the equivalent static load produced by the missiles. This is done by determuung the deflection resulting from dissipation of the kinetic energy and fmding the equivalent static load or directly from the kinetic energy of the missile and the magnitude of the missile penetration. 3.5.4.2 Missile Penetration (Localized Effect) The following empirical formulas luwe been used to determine the missile penetration into the concrete
- barner,
- a. Modified Petry Formula
- b. Army Corps of Engineers
- c. Ballistic Research Laboratory.
Computations have been nude for all nussiles described previously using these formulas. The results show the Pctry formula as least conservative, and the Ballistic Research Laboratories formulas as most conservative. Based on these considerations, it has been ascertained that with lunited exceptions as described in S AR Section 3.7.2.1.1, a potential missile would not jeopardize the function of the structures analyzed. 3.5.5 MISSILE BARRIER FEATURES AJ shon in Figures 3.7 1,3.7-2 and 3,7-3, the exterior concrete walls and slabs of structures were designed to act as barriers against external missiles. O 3.5 - 2 UPDATE 2 - AUGUST 1997
3.5 MISSILE PRDTECTION CRITERIA The missile protection criteriore (i c., GDC 4; see Section 3.1.2.4)is based on precluding dam?ge to O structures, systems, and components important to safety. During PDMS, there is no equipment that is important to safety; therefore, the consequences of a missile impact are limited to physical damage. With limited exceptions as described in Section 3;/.2.1.1, all contamination isolation structures are protected from a loss of function due to damage from potential external missiles as described below. Due to the non-operating and defueled status of TMI-2 during PDMS, there are no postulated intemally generated missiles which would have any effect on the capability of structures to provide their required , contamination isolation function. 3.5.1 MISSILE LOADINGS AND BARRIERS The exterior walls of the Reactor Building, Auxiliary and Fuel 11andling Buildings and the Control & Senice Buildings were designed to withstand the effects of external missiles. 3.5.2 MISSILE SEl.ECTION The potential sources of extemal missiles have been investigated for those buildings designed for missile protection. Missiles generated by emironmental conditions and a postulated aircraft impa:t have been considered. The critical possible missiles have been selected and analyr.ed. They are described in the following paragraphs. 3.5.3 SELECTED MISSILES O G 3,5.3.1 Tornado Generated Missiles The residual contamination which remains at TMI-2 is contained within buildings that have been designed to with:tand tornado generated missiles as discussed in Section 3.3.2.2.' 3.5.3.2 Aircran impact The probability of aircraft impact on the station has been studied The results of evaluations show that the probability of an aircraft impact on the unit is very low (see Section 2.2.3.1.2 of the TMI 2 FSAR and Reference 3.5-1). In addition, all contamination isolation structures' were designed to withstand a hypothetical aircran incident. The impact loadings on these structures are assumed to come from: 1, An aircran of 200,000 lbs. traveling at 200 knots impacting on an efTective area of 19 R. diameter
- 2. An aircran of 300,000 lbs. traveliag at 200 knots impacting on an anca of 16 ft. diameter
- 3. An object of 6000 lbs. traveling at 200 knots impacting on an effective area of 5 R. diameter
- 4. An object of 4000 lbs. traveling at 200 knots impacting on an etTective area of 3 ft. diameter.
The results of these analyses provide assurance that with limited exceptions as described in SAR Section 3.7.2.1.1 the hypothetical objects would not perforate or cause collapse of any of these structures. t ) 2
*Ihe "Other Buildings" from Table 5.3-6 that contain less than 1 Ci residual contamination are excluded from this discussion.
3.5-! UPDATE 2 - AUGUST 1997
6 m--- s A M m e m.a omJ esa32-m,4snt.- Ja a am ka oe==-A-me,nu w e ma ,+a-,,a.4m.mAwms--s em M L-ms.- ,,Asem--a m u An, m.,A wn-ahm2me-AwM*Oann,Le.m-au.a.rM.mn'm.e--"m&="--=waw-u-=' 'um= mm '
=-.mmmma='--~~4 PEFERENCES l 3.5 1 ALAB-692 dated September 14,1982.
I l d 1 t i i \ O 1 l i t O i 4 3.5 - 3 UPDATE 2 - AUGUST 1997
3.6 SEISMIC DESIGN A Re Reactor Building, the Auxiliary and Fuel 11andling Buildings, Control and Senice Buildings, and other buildings were originally designed to Category I criteria for seismic design. Dese criteria may be found in Section 3.7 of the TMl-2 FSAIL Due to the non-operating and defueled status of TMI 2 during PDMS, it is not required that these buildings meet the seismic Category I criteria. The seismic criteria in the Uniform Building Code for Zone 1 is suitable for the PDMS period. It is assumed that the structural capability of the buildings as originally designed has not been significantly degraded and would fully comply with the criteria in the Uniform Building Code. Here are no systems which are required to meet the seismic Category I criteria. It is assumed that the systems which were designed to the seismic Category I entena and contain residual contammation will retain their structural capability through the PDMS period and continue to isolate internal contamination from the extemal emironment. C' ( k b 3.6 - 1 UPDATE 2- AUGUST 1997
3.7 DESIGN OF PRINCIPAL BUILDING STRUCTURES r% His section addresses the design of the principal buildings which sen e protective functions during (V) PDMS. These structures were originally designated as Category I structures and were designed to provide the protective functions associated with an operating nuclear power plant. During PDMS, the protective functions provided by these structures are very limited compared to those for which the buildmg was originally designed. His section presents a summary of the criteria necessary to pro ide the protective functions required during PDMS. The detailed design criteria for the Category I structures is given in Section 3.8 of the TMI 2 FSAR. 3.7.1 CONTAINMENT DUILDING 3.7.1.1 Structure Description The Containment Building is a reinforced concrete structure composed of a cylindrical wall with a flat foundation mat and a dome roof. The foundation slab is reinforced with conventional mild steel reinforcing. The cylindrical wall is prestressed with a post tensioning system in the vertical and hoop directions. The dome roofis prestressed utilizing a three-way post tensioning system. The inside surface of the Containment Building is lined with a carbon steel liner to ensure a high degree ofleak tightness. He nominal liner plate thickness is 3/8 in. for the cylinder ,1/2 in. for the dome and 1/4 in. for the base. The foundation mat is bearing on rock and is 1I ft 6 m. thick with a 2 ft. thick concrete slab above the base liner plate. He cylinder has an inside diameter of 130 ft , a wall thickness of 4 ft and a height of 157 ft. from the top of the foundation slab to the spring line. He roofis a shallow dome havmg a large O radius of i10 ft., a transition radius of 20 ft. 6 in., and a thickness of 3 ft. 6 in. Typical Containment Ij s Building details are shown in Figure 3.7-1. 3.7.1.2 Liner Plate and Penetrations The liner plate has been designed to function as a leak-tight membrane. Nominal steel plate thicknesses for the plate are 1/2 in. for the dome and the ring girder,3/8 in. for the cylindrical wall and 1/4 in. for the base, The thicknesses are locally increased at penetrations, polar crane supporting brackets and at equipment support locations. The hner is anchored to the concrete shell by means of structural tees (ST 31) running vertically and meridionally along the cylinder and dome, respectively. The anchorage system is designed to prevent clastic instability of the liner. A continuous system of steel channel is welded c,ver all liner weld seams on the face of the base line, and over all inaccessible liner weld seams (on the exposed face of the liner) of the wall. The liner plate and the concrete shell are penetrated by the openings as described below:
- a. Equipment Access Hatch and Personnel Air Locks An equipment hatch with an inside diameter of 23 ft has been provided to allow passage oflarge equipment and components into the Containment Building. The equipment hatch is an intecral part of the Containment and is bolted to a flanged n
(v) 3.7 - 1 UPDATE 2 - AUGUST 1997
i i 1
. steef sleeve anchored in the concrete Contamment wall and welded to the Containment liner so that the equipment hatch can be removed and reinstalled from the outside of the Ccotainment Building. One personnel air lock is a removable unit penetrating through the equipment hatch. The equipment hatch, consisting of the hatch cover, the attached air lock, and the flanged sleeve, is a complete gas-tight unit when field assembled.
A second personnel air lock is welded to the Containment liner, detailed layouts of these oper.ings are shown in Figure 3.7-2.
- b. Piping and Electrical Penetrations The piping penetrations are normally anchored at the Containment Building shell.
Typical piping ducts and penetrations are shown in Figure 3.7 3. All penetrations are of the double barrier type. Where temperature considerations require, the second barrier of the piping penetration includes expansion bellows. The electrical penetrations are welded closures.
- c. Modified Penetrations Several penetrations were modified during the cleanup period. These penetrations were modified in accordance with criteria submitted to the NRC (Reference 3.7-1).
Piping and electncal penetrations were designed and modified to withstand at least 5 psig and tested to hold 1.2 to 1.5 times this pressure for not less than 10 minutes in accordance with ANSI B31.1. Fluid hard piped piping was designed, modified, and tested in accordance with ANSI B31.1. The leakrate from modified penetrations, includmg flange and isolation valve leakage, was limited to 100 seem per 1 inch of pipe diameter. A complete listing of all containment penetrations is provided in PDMS SAR Table 7.2-2 which includes the status of each for PDMS. The containment penetrations that are in a modified configura* ion for PDMS are listed in Table 3.7-1 which also provides details of the modnations. For this review, a modification to a continment penetration was included if the modification was made to the area from the inboard isolation valve through the penetration to and includmg the outboard isolation valve. Modifications to containment penetrations during the recovery penod have been accomplished such that the combined modified cross sectional area does not exceed the NRC safety evaluation requirement of 40 square feet (Reference 3.7-2). The Containment Atmospheric Breather has been added to the Contamment to proside passive pressure control of the Contamment relative to ambient atmospheric pressure and to establish a "most probable pathway" through which the Containment will
" breathe." This addition ensures that the Containment structure will not experience significant pressure ditTerential to threaten the structural capability of the contammation boundaries provided by the Contamment. Further discussion of the Breather is provided in PDMS SAR Section 7.2.1.2.
O 3.7-2 UPDATE 2 - AUGUST 1997
/^N f 3.7.2 OTIER PRINCIPAL STRUCTURES (d 3.7.2.1 Description of Structures The physical layout of other principal structures for the unit is shown in Figures 3.7-4 and 3.7 5. Following is the physical description of these structures. All these structures are constmeted of reinforced concrete. 3.7.2.1.1 Auxiliary Building ne Auxiliary Buildmg was designed to house the components of auxiliary systems required for reactor coolant purification, conditioning, reprocessing and cooling, radw2ste processing and engineered safety features, he Fuel Handling Building and this building are physically attached by means of a common structural wall. A vertical air intake shaft is also attached to the east wall of this buildmg. The Auxiliary Building is rectangular in plan and has three main floors. At the east exterior wall, a large door opening is located at the grade level. This door opening is not protected from an aircraft impact loading or external missiles. In addition, concrete drilling operations in a small exterior wall section of the Auxiliary Building damaged six (6) reinforcing steel bars, two (2) of which impact the ability of the Auxiliary Buildmg to meet TMI 2 FSAR design criteria for aircraft impact. Nonetheless, the wall presently meets normal non-operating load requirements. These two (2) areas comprise the limited exceptions to wind and tornado loadings and missile protection alluded to in SAR Sections 3.3 and 3.5. To the south of this building is the Senice Building and a portion of the Control Building Area. There is one foot separation between these structures. All floors of this building are slab-beam and flat slab [J] construction. 3.7.2.1.2 Fuel Handling Building The Fuel Handling Building was designed to accommodate the storage of new and spent fuel. In addition, it provides an area to accommodate the storage of packaged waste prior to otT-site disposal. This building is physically connected to the Auxiliary Building on the east side by means of a common wall. He Fuel Handling Building of Unit 1 is located on the north side. The clear distance between these two buildings is 3 in. One bridge crane, common to both buildings, is used for fuel handlmg. Here is no wall separating these structures at the top floor (above elevation 347'-6"). The Containment structure is located on the south side of this building and is separated by a clear gap of 2 in. Two stainless steel Imed, reinforced concrete fuel storage pools are located in the Fuel Handling Buildmg. 3.7.2.1.3 Contrgl and Senice Buildines he Control Buildmg and the Senice Building are separated by a common wall. The Control Buildmg houses a control room, a battery room, a cable room and a mechanical equipment room. The Senice - Building provides access to the Reactor Building, the Auxiliary Building and the control room. He Senice Buildmg serves as the Rad. Con. Control Point for Unit Two. He Control and Senice Buildings are rectangular buildings with a common
- }
v 3.7-3 UPDATE 2 - AUGUST 1997
foundaticut mat. Floors of the control building are supported by interior walls only and a peripheral gap of 4" between edges of the floors and the inside face of the exterior walls has been provided to create a structural separation between the exterior structure and the interior structure. He purpose of this separation is to protect vital and sensitive control room equipment from dynamic aircraft impact loadmg to which the exterior walls and the roof could be subjected. Door openings and other penetrations in exterior walls of the control building that are susceptible to aircraft loading have been shielded by reinforced concrete shield walls. The Control and Senice Buildings are surrounded but physically separated from the Auxiliary Buildmg, the Control Building Area and the Turbine Buildmg. 3.7.2.1.4 Control Bulkline Area The Control Buildmg Area is located between the Reactor Buildmg and the Turbine Building and accommodates the main steam relief and isolation valves and cmergency steam generator feedwater pumps. This is a one story underground structure from the Turbine Building on he south end At the north end, this structure follows the curvature of the outside face of the Containment Structure On the west side, a metal frame structure called the CACE is situated at ground level Protecting the RB equipment hatch. 3.7.2.1.5 Air intake Tunad ne air intake tunnel provides a passage for air to be distributed into the following structures: Reactor Building, Auxiliary Building, Fuel Handhng Building, Control Buildmg, Senice Building and Control Building Area. He air intake tower at an end of the tunnel extends above the grade level. The bottom of the air intake extends 12' -0" below the bottom of the tunnel providmg a large well for the accumulation of rain water inleakage. Access to the tunnel is provided from the Senice Building and from the air intake tower. One manhole from the grade level to the tunnel floor has been pro ided for emergency egress. Near the Auxiliary Buildmg, the tunnel bends 90 degrees upward and changes into a rectangular concrete plenum. The plenum extends above the grade level and is closed off at the Auxiliary Buildmg rooflevel. He supply air enters the Senice Building and the Auxilian Building through the openings in walls of the plenum. O 3.7-4 UPDATE 2 - AUGUST 1997
- . - . . . . . - . - - - . . - . . . . . . - . ~ . _ . _ . . - _ . _ -. - . - . . - - ~ .
i i i i -. e , . REFERENCES i- , s 3.71 GPU Nuclear lener LL2 810191, " Design Pressure for Contamment and Future Mechanical l and Electrical Penetration Modibcations," dated December 4,1981. ! l i-3.72 Letter, W. D. Travers (NRC) to F. R. Standerfer (GPUNC), " Seismic Design Criteria for i Modified Containment Penetration," dated April 3,1987. U U 4 4 1-e f i-i j i 4 i a 1 e .L a 1 l 9, I 4 l 4 1 4 3.7-5 UPDATE 2 - AUGUST 1997
,+4 -__.,,; , _. , . - .-- . ,,,-,,_.._4,,... , . , , , , - - . _ , _ , . . . , . ,
TABl.E 3.7-1 MODIFIED COSTAINMENT PENETRATIONS PENETRATION NUMBER MODIFICATION R-401 This penetration was modified in late summer 1979 to allow access to obtain samples of the Reactor Building (RB) sump water. Followmg successful completion of the sampling program, further changes were made to use the penetration for RB water level measurement by addition of a manometer system to the sampling tube. A further modification was made to proside a more permanent closure in consideration of future potential increase of RB water level to the extent of floodmg the penetration. This fmal modification removed the 12 inch gate valve and special cover assembly outboard of the valve and wdded a closure assembly to the penetration. R 561 The penetration was modified to provide a flow path into the RB for a high pressure decontammation water supply and additional senice air. This penetration modification consisted of replacing an existing 10 inch penetration with a 10 inch flange with three piping connections on the outboardside containing isolation valves and hose connections on the inboardside. R 545 The spare penetration was modified to provide a flow path into and out of the RB for the DWCS. This penetration modification consisted of adding a double valve pressure boundary on the outboard side and providmg piping for a hose connection on the inboard side. R 536 This spare penetration was modified to provide a flow path into the RB for Plasma Arc Nitrogen as a quenching gas for PCI cutting operations. This penetration modification consisted of addmg a two inch pipe through the penetration with isolation valves on the outboardside, including penetration test connections and hose adapter on the inboardside. R 508 This electrical penetration was modified to provide an access for ROSA cables awl CCTV coaxial cables. This penetration modification consisted of boring and tapping holes in the existing flange for cable routing, resealing and testing for leaks. R 554 The penetration was modified to provide a source of clean compressed air for use with pneumatic controls and operators. This penetration modification consisted of replacmg two (2) outboard containment isolation valves with three (3) vahts and adding a flow limiter and quick disconnect on the inboard side. l O 3.7 - 6 UPDATE 2 - AUGUST 1997 l
D TABLE 3.71 (Cont'd) ( \ MODIFIED CONTAINMENT PENETRATIONS PENETRATION NUMBER MODIFICATION R-555(c) De penetration was modified to provide axess for a Piasma Arc Cutting machine ground cable. this penetration has had a blind flange installed. R 562 'Ihe penetration was modified to provide a flow path for sludge transfer from the RB to the spent resin storage tank in the Auxiliary Building. This penetration modification consisted of adding a double valve pressure boundary on the outboard side and piping for a hose connection on the inboard side. R 565 The penetration was modified to provide a means of transferring shield water to the contai unent sump. This penetration modification consisted of adding a piping spool assembly to the outboard side of the penetration. R-626 The penetration was initially modified to insert an antenna and camera arrangement into Containment. These were subsequently removed and the penetration was modified to allow pumping out of the RB basement, This penetration modification consisted of installing a new spool piece and piping. For PDMS, a flange was bolted over the penetration with a single pipe centered in the flange. The pipe has an isolation valve with a cap installed on the end of the pipe.
" Tying in with the pipe" can be effected in one of two ways:
1 Cutting the existing pipe completely and addmg a "T" fitting; or
- 2. Cutting a circular hole in the existing pipe and weldmg a new pipe to it.
In either case, the end resut cf two parallel flow paths where one previously existed. , 1
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3.7-7 UPDATE 2 - AUGUST 1997
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CHAPTER 4 FUEL l
1
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CilAPTER 4 FUEL TABLE OF CONTENTS SECTION IIILE j'iGI
4.0 INTRODUCTION
.01
4.1 BACKGROUND
INFORMATION 4.11 1 4.
1.1 DESCRIPTION
OF THE MARCH 1979 ACCIDENT 4.11 4.1.2 POST + ACCIDENT CHARACTERIZATION 4.12 4.1.4 ' Core inventon 4.12 i 4.1.2.2 Upper Core Region 4.12 4.1.2.3 MidCore Region 4.12 4.1.2.4 Lower Core Region 4.12 4.1.2.5 lower CSA Region 4.13 4.1.2.6 lower llead Region 4.13
- 4. ' ' 'I Core Former Region 4.13 ,
.4.1.2.8 EvVessel 4.13 4.2 FUEL RELATED ACTIVITIES 4.21 4.
2.1 DESCRIPTION
OF FUEL REMOVAL ACTIVITIES 4.21 4.2.1.1 Scope of Defueling 4.21 4 4.2.1.2 Defueling Equipment 4.2 1 4.3 SNM ACCOUNTABILITY AND CRITICALITY 4.31 - SAFETY ANALYSIS i 4.
3.1 INTRODUCTION
. 4.31 ;
4.
3.2 BACKGROUND
4.31 O ; i UPDATE 2 - AUGUST 1997 _,.1,_, - u.,.m., - ~,..--,y, , - . , ,,,.m. r-&-* - -- - , , - - - - - - - - - -----v~w w r----------++-+-
t O CilAPTER 4 TABLE OF CONTENTS (Cont'd) SECTION TITLE PAGE . i 4.3.3 SNM ACCOUNTAltiLITY PROCESS 4.3 2 4.3.3.1 Classification of Plant Areas - 4.32 4.3.3.2 SNM AccountabilityMethods 4.32 4.3.3.3 Documentation 4.3 3 , 4.3,4 FINAL SNM ACCOUNTAlilLITY 4.3-4 I 4.3.5 CRITICALITY ANALYSIS 4.3-4 i 4.3.6 CONTROL OF SNM DURING PDMS 4.3 5 APPENDIX 4A DEFUELING EQUIPMENT 4A 1 Shielded Rotatable Work Platfonn 4A 1 Defueling Canisters 4A 1 Defueling Vacuum System Airlift Systems 4A 3 E Core Bore Equipment 4A 3 "The Automated Cutting Equipment System (ACES) 4A 3 Long-11andled Tools 4A 4 6 I li UPDATE 2- AUGUST 1997 l 1 i l
i i i i
. CHAPTER 4 4
TABLE OF CONTENTS (Cont'd) i LIST OF TABLES T4BLE NO. TITLE EMiE 4.11 CORE MATERIALINVENTORY 4.15 43 6 l 4.31 - FINAL SNM INVENTORY BY LOCATION Reactor Buildmg J 43 2 FINAL SNM INVENTORY BY LOCATION Auxiliary and Fuel 4.3 7 !
- Handling Buildings 4
l 8 1 4 ] i - i I i J l l iii UPDATE 2 - AUGUST 1997 l l
CHAPTER 4 TABLE OF CONTENTS (Cont'd) LIST OF FIGURES FIGURE NO. TITLE fAGI 431 SNM ACCOUNTABILITY LOCATIONS REACTOR BUILDING 4.3 8 282' 6" EL 4.3 2 SNM ACCOUNTABILITY LOCATIONS REACTOR BUILDING 4.39 30$'-0" EL 4.3 3 SNM ACCOUhTABILITY LOCATIONS REACTOR BUILDING 4.3 10 347'-6" EL, 4.34 SNM ACCOUNTABILITY LOCATIONS AUXILIARY / FUEL 4.311 IIANDLING BLDO, 4.34 SNM ACCOUNTABILITY LOCATIONS AUXILIARY / FUEL 4.3 12 IIANDLING BLDO. 4.3 4 SNM ACCOUNTABILITY LOCATIONS AUXILIARY / FUEL 4.3 13 11ANDLING BLDO. O -4.35 SNM ACCOUNTABILITY LOCATIONS AUXILIARY / FUEL 4.3 14 1IANDl.ING BLDO. 280' 6" EL 4.3 6 SNM ACCOUhTABILITY LOCATIONS AUXILIARY / FUEL 4.3 15 11ANDLING BLDO. 305'-0" EL 4.3 7 SNM ACCOUNTABILITY l OCATIONS AUXILIARY / FUEL 4.3 16
- 11ANDLING BLDO. 328'-0" EL 4.38 SNM ACCOUNTABILITY LOCATIONS AUXILIARY / FUEL 4.3 17 iLANDLING BLDO. 34T 6" EL APPENDIX 4A 4A 1 SillELDED WORK PLATTORM ASSEMBLY 4A 5 4A 2 . FUEL CANISTER - 4A 6 4A 3 KNOCKOUT CANISTER 4A 7 4A 4 FILTER CANISTER 4A 8 iv UPDATE 2 - AUGUST 1997
i i l CHAPTER 4 j
. FUEL !
i 4.0 D(TRODUCTION j 4 1his chapter summarizes the conditions and methities associated with the TMI 2 core subsequent to the accident on March 28,1979. ! Section 4.1 provides a brief description of the accident and the post accident core conditions as they ! became known through de defuelms process. Section 4.2 provides surnmary descriptions of the major j defueling methities, the background ar.d understandmg of the core conditions which prevailed aner the l ! accident, and tlw basis for the rniew and approval of the fuel related conditions of TMI 2 during i PDMS This discussion is relevant because the core debris which remains during PDMS is directly : dependent on the core conditions subsequent to the accident and the level of success achined by the methities undertaken to remove the core debris. The residual fuel conditions are provided in the Post-4 Defueling Survey Reports (PDSRs), this information is summarized on Tables 4.3 1 and 4.3 2. 'Ihe . analysis demonstrating assured suberiticality dunng PDMS is provided in tlw Defueling Completion ' Report (DCR) and the criticality safety analyses presented in GPU Nuclear letter C312 92 2080, dated December 18,1992 and summarized in Section 4.3. t t i l 9 1 L ) 4 t L e d 4.0 - 1 - UPDATE 2 - AUGUST 1997
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p 4.1 DACKGROUND INFORMATION 4.
1.1 DESCRIPTION
OF TIIE MARCil 1979 ACCIDENT 1 he March 1979 accident was initiated by cessation of secondary feedwater Dow. The steam generator ! boiled dry, and the resultant reduction of pnmary-to secondary heat exchange caused the primary coolant to heat up, surge into the pressurizer, and inercase the prinury system pressure. ne Pilot Operated Relief Yah c (PORV) opened to relieve pressure but failed to close when the pressure decreased ne first 100 minutes of the accident can be characterized as a small break loss of coolant accident with resultant loss of pntnary coolant and decreasing pressure. It differed from the scenario ; expected during a LOCA in that the pressurizer liquid level indication remained high his was ! interpreted by the reactor operator as indicating that the Reactor Coolant System was full of water ; when, in fact, the RCS was continually voiding. Up to about 100 minutes into the accident, the core was still covered with sufficient water to be cooled. ne Reactor Coolant Pumps were turned off at 100 minutes and core heat up began as the water level decreased to elevations below the top of the core. By 150 minutes, a Zircaloy-steam exothertrue reaction was occurring, which increased the core heat-up rate. Consequently, Zircaloy meltmg temperatures were exceeded, resulting in relocation of the molten Zircaloy and some liquefied fuel to the lower core regions, solidtfying near 4 s Vant interface. His condition prevailed until 174 minutes, at which time a large region of consowe p/sded core matenal existed m the lower, central regions of the core, Coolant flow through this wMA:J material was probably negligible. De intact fuel rod stubs in the lower core region indicate tha' omy the lower portion of the core remained cool. O A Reactor Coolant Pump was tumed on brie 0y at 174 minutes and coolant was pumped into the Reactor Vessel. He resultant thernal mechanical forces generated from the rapid steam formation are believed to have shattered the odlized fuel rod remnants in the upper regions of the core, forming a rubble bed on top of the consolidated core materials. The consolidated core naterials continued to heatup during the next $0 minutes (174 to 224 min.), even though coolant delivery to the Reactor Yessel from the purnp transient and emergency core coolmg injection is estimated to have covered the core by approximately 210 minutes. By 224 minutes, much of the non cooled consolidated region had reached temperatures sufficient to melt the U Zr-O ternary miouie. On-line TMI 2 data recorded during the accident suggests that the crust surrounding the consolidated core failed between 224 and 226 minutes into the accident. Dased on the end-state core and core support assembly configuration and supporting analysis of the degraded core heat up, it is believed that as the crust failed, molten core material migrated to the lower intemals. The majority of the molten matenal flowed down through the region of the southeastern assemblies and into the core bypass region. A portion of the molten core material flowed around the bypass region and migrated down into the lower intemals and lower head region Limited damage to the CSA occuned as the core material flowed to the lower plenum. It is estimated that about 17-20 tons of material relocated to the lower intemals and lower head region Several incore instrument guide tubes were melted but overall vessel integnty was maintained throughout the accident, b s 4.1 - 1 UPDATE 2 - AUGUST 1997
O 4.1.2 POST ACCIDENT CilARACTEldZATION De post accident cond tions inside the TMI 2 Reactor Vessel, as cunently understood, are summanzed in the following sections. A detailed desenption of the con &tions in the Reactor Vessel is given in Reference 4.1 1. 4.1.2.1 Core Imentory he original core inventory included approximately 207,300 lb of fuel (i c., UO2) and 75,400 lb of structural and absorber matenal for a total of 282,700 lb. Including the rnatenal added as a result of darnage to the Reactor Yessel intemal components due to the dynanues of the accident and the material generated by defuelmg nethities, the total post accident core debris material was estimated to be 296,100 lb. His estimate does not include any portions of the CSA and gnd structures beneath the core that may have partially melted and mixed with the core material. See Table 41 1. Due to the accidcat progression, some of this core matenal was relocated within and outside of the Reactor Yessel Each of the relevant areas is discussed separately. 4.12.2 Upper Core Region This region covers approximately the top 8 A. of the onginal core length (13.8 A.). The bottom boundary of this region is the layer of resoli&Oed material identined during debris bed probing and obsentd dunng the core bore aethities ne material in tids region consisted of partial, standmg penpheral assemblics, partial assemblies hanging from the underside of the plenum, loose debiis, fuct assembly and control element structural com . A such as end Ottings and spiders, fuel rods of sarious sizes, and loose debns consisting of rea um aaterial and fractured fuel rods and pellets. 4.1, Mid Core Region nis region includes the layer of resoli&fied matenal referenced above (Section 4.1.2.2). This region was sunounded by standmg peripheral assemblies and supported by partial fuel assemblics, nis region consisted of a large resolidified mass that was relatively thin at the periphery and very tidck near the center. %e thickness varied from about 13 in. at the periphery to about 60 in. at the center. Composition of this damaged material included a monolithic resolidified mass with pores, rubble that was fused together by once-mohen matenal, buried end fittings, and other structural material. 4 1.2 4 lower Core Region ne geometrically intact partial assemblies at the bottom of the core varied in length from 9 in. near the center to full length at the periphery outboard cft % resolidirH nt.uerial. Only two assernblies were fulllength and with better than 90% of the fuel rod cross-section. Most of the lower portions of these fuel assemblies were ductile while upper portions were bnttle, indeating a higher degree of oxidation. On the castem :ide of the core near the major relocation path, a number of the partial assemblies were brittle near the bottom with ductile fuel rods on the upper portions of the stub assemblies. Several assemblies on the cast side were resoli&fied masses with no identiAable fuel rods. 4.1 - 2 UPDATE 2 - AUGUST 1997
f t
?
4.1.2.5 lower CSA Region ! his region consists of the area between the bottom of the lower end fittings and the 2 inch tidek flow distributor, ne mWority of the regi m just beneath the core region contained only fme loose debris with _ no structural damage. ne region outboard of the fuel region under the flow bypass region contained a large number of resolidified columns which were created by the downward flow of mehed core material ,
- from the bypass region. ne eastern part of the lower internals contamed a large amount of resolidified material under several fuel assemblies and under the bypass region. His area ns the major relocation path for the mehed core material which flowed to the lower head region. There was some structural damage in this area with some melting ofincore guide tubes and support posts. ,
4.1.2.6 lower Head Region ne lower head region encompassed the space between the flow distnbutor and the spherical lower head. A large quantity of material was relocated to this region during the accident. He material appeared to be distributed nonuniformly. Particle sizes in the loose portion ranged from fine dust to approumately 8 in. diameter nodules. Dere was much less loose matenal in the north quadrant of this region than in the other areas. . t j 4.1.2.7 Core Former Region His region includes the area between the baffle plates and the thennal sideld in the upper CSA. Some damage (melting) to the baffle plates occurred on the eastem side. Inspections of this region revealed a large mass of material between the baffic plates and the core barrel. Material was obsen ed ; on the various core former plates around the core circumference with most of the material accumulated in the east and the north No physical damage of the thermal shield and core barrel was obsened. Resolidified material was seen at a number oflocations just below and penetrating through flow holes , below the core former plates. 4.1.2.8 Ex Vessel ; his region consisted of any area outside the boundaries of the Reactor Vessel where core material had been transported. His included the Reactor Coolant System and associated components, and the Reactor Building and Auxiliary and Fuel Handling Buildmg sumps. During the accident, small > quantities of core debris were relocated throughout the RCS and support systems. -i f 4.1 3 UPDATE 2 - AUGUST 1997
+
4 h i
%.- v,-rsr ov-,.e -mmE-,-....,%wm-,,ww.,e,..wc +.e.~m .%e .rz,,w..,-ev,-v. -y,,m.<, w a , ..,,,,.,,..m -.,-,,-*-,-9.m,m.-w.,um.--r-t MTv----*r e-- & w
l l l REfrardres i , i 4.11 Technical Bulletin 89-08, Revision 0, " Final Core Matenal Estimates,* October 19,1989 l 1 i i l l i l O' 4 1
. _ . _ _ . . _ . .-_ _ . _ _ _ _ - _ ._. _ _ _ ~. . _.._. _ _ _ . _ _ -
i TABLE 4.1 1 ! CORE MATERIAL INVENTORY 4 Original Leadian .Weimht (Ib) i Total U02 (177 assemblies) 207,300 j r 51,100 : TotalZircaloy 4 ) Absorber and Other 24,300 j Structural Material i Total Original Core Material 282,700 i Post Accident Weinht (Ib) . Original Mass 282,700 i Material Melted from Plenum Upper Core 500 Tie Plate and Grid Pads Material Melted from Bame Plates, 600
\ Core Former, and incore Assemblies i
02 Due to Zr Oxidation 7,700 i Defueling generated Material 4,600 r 2 %,100 ! Total Estimated Post accident Core Debris MaterialInventory I d h 7 r V 4.1 - 5 - UPDATE 2 - AUGUST 1997
-*=eg -g---eemp, w .+r-+ .g-wpwww-rupi-m.-m.1geep-j q ,r- &-egt -is,g,--gr-,,sw!gye,--g g w- riq -
4.2 TUEL RELATED ACTIVITIES 4.
2.1 DESCRIPTION
OF FUEL REMOVAL ACTTVITIES ' nis section prmides a summary description of the acthities associated with the removal of the damaged reactor core from the Reactor Vessel and the shipping of the core debris to the storage site in Idaho. 4.2.1.1 Scope of Defueling The aethities associated with the defueling of the TMI.2 Reactor Vessel were primarily the removal of r core material from the Reactor Vessel, encapsulation of these materials within specially designed canisters, and placement of the canisters into the storage racks located in Spent Fuel Pool "A". De . canisters were sabsequently shipped to Idaho National Engineering Laboratory for analysis and storage. ne defueling process was dnided into four major activities. These four aethities are described below.
- 1. Initial defueb - nis task involved the remmal of loose debris which was readily accessible and small enough for easy handling and placement in canisA ' without sizing. His debris was located on the top of the core region and readily accessible once the Reactor Vessel head and plenum assembly were removed from the Reactor Vessel.
The loose debris was composed primarily of fuel element end fittings, small pieces of fuel rods, intact and broken fuel pellets, fuel fmes, and other nuscellaneous debris. 4 ,
- 2. Core recion defueling - his task invoked the removal of the debris remaining in the core region after completion of the initial defueling. This phase differed from the initial defueling in that significant sizing operations had to be performed on much of the remaining debris prior to placing the pieces in the canisters. His involved the bonng and breaking up of the solid fused portion of the core and sizing any pieces too large to be placed in canisters. Once the large or fused portions of the core were sized and !
I placed in canisters, the stubs of damaged fuel assemblies were also removed and placed in canisters. Finally, general cleaning of debris from the lower gnd assembly completed the core region defueling.
- 3. Other vessel recions defuelig his task involved the removal of core debris from the upper and lower core support assembly, the core formers, and the lower Reactor Vessel head. The core debris in these locations consisted of debris which relocated from the original core boundaries during defueling operations or flowed, as molten material, out of the core region durmg the accident. He defueling of these locations involved the disassembly and removal of reactor intemals to prmide access to the fuel. Tte core debris was sized, as required, for placement in canisters.
- 4. Evvessel defueline This task involved the removal of core debris from some locations with!u the Reactor Coolant System, outside the Reactor Vessel. Small quantities of fuel were located in the steam generators, pressurizer, other Reactor Coolant System components, and Reactor Coolant System piping.
4.2.1.2 Defuehng Equipment ne unique defucting operations required the design and fabrication of special equipment and tools.
.,- Although not directly applicable to this SAR, the major equipment and tools are described in Appendix
/ 4A for historical reference.
p 4.2 - 1 UPDATE 2 - AUGUST 1997
l l 4.3 SNM ACCOUNTABILITY AND CP.!TICALfiY SAFE'IY ANALYSIS 4.
3.1 INTRODUCTION
ne purpose of this section is to desenbe ths TMI 2 SNM Accountabihty Program and summarize the enticahty safety analyses presented in the DCR (Reference 4.310) and in GPU Nuclear letter, C312-92 2080, dated December 18,1992 (Reference 4.311) His section identifies the methods and sequence of events for residual SNM accountability; the Quahty Assurance program applied to the SNM measurements; the areas, systems and components that were assessed for residual quantities of SNM; and the areas, systems and components that did not require SNM assessment. ne quantity of fuel (i c., 00:) remaimng at TMI 2 is a small fraction of the initial fuel load. As a result of TMI 2 defueling and decontamination activities, apprournately 99% of the fuel wu removed ) and transferred to DOE and'or licensed burial facilities. ! The fmal resuhs of the SNM Accountability Program are based on a comprehensive post-defueling j survey of the TMI 2 facility. De post defueling survey consisted of a review of the TMI 2 plant to identify areas that could contam SNM and areas unlikely to contain SNM. He quantity of SNM was deterrnined in each area that was identified to have SNM present. His section desenbes the process by , which the post defueling survey was conducted and summarizes the results of the survey. Fmally, this section pusents a summary of the enticality safety analyses presented in References 4.3 10 and 4.311 which demonstrated that a criticality event could not occur in TMI 2.
43.2 BACKGROUND
s' ne March 1979 accident resulted in significant damage to the reactor core with a subsequent re! case of fuel and fission products into the Reactor Coolant System and other connected systems. The core was reduced to fractured ruel pellets, resolidified fuel masses, structural metal components, loose rubble and partial fuel assemblies ne ges.eric term used to refer to the post accident core material is core debris. The core debris removed from the TMI 2 facility was loaded into special carusters for shipment to the DOE INEL facility in Idaho. Each shipment was accompanied by a Nuclear Material Transaction Report (DOE /NRC Form 741) which recorded the net weight of the contents of each canister. Fuel accountability by the normal method, i c., accounting for individual fuel assemblics, was not possible. Since the canisters were filled with a mixture of SNM, other materials, and water, there was no practical or feasible method to detennine the exact SNM content in each canister. A statement to that effect was included on cach DOE /NRC Form 741. In October 1985, GPU Nuclear, the U.S. Department of Energy and the U.S. Nuclear Regulatory Commission entered into an agreement (References 4.31 and 4.3 2) that fmal SNM accountabihty for TMI 2 would be performed after defuelmg was completed. The accountability would be based upon a thorough post defuehng survey of TMI 2 which would quantify the amount of residual SNM in plam systems and components. Implied in this agreement was an understanding that the post defueling survey would invoht all areas, structures t,..tems and components where SNM could reasonably be suspected to have been deposited as a result of the 1979 accident and subsequent cleanup actisities. A b 4.31 UPDATE 2- AUGUST 1997
I J 4.3.3 SNht ACCOUNTA111LITY PROCESS 4.3 3.1 Classificaten of Plant Areas The entire Thil 2 plant was ininted to detennine where SNh1 could have been deposited as a result of the 1979 accident and subsequent cleanup activities. Each area was classified into one of three categories: CATEGORY l Locations where SNat was highly probable CATEGORY 2 1.ocations where it was possible that SNhl could be deposited CATEGORY 3 Locations where it was unlikely that SNh1 was deposited Category I locations required that measurements or, in selected cases, analysa, be perfonned for SNht. Category 2 areas were considered to have a lower probabihty for fuel deposits, but were asser, sed in the same marmer as the Category 1 areas. Category 3 areas were determined not to require SNh1 assessment based on analyses of the Th112 accident (References 4 3 3 and 4.3-4) and review of cleanup activities. 4.3.3.2 SNh1 Accountabihty hiethcds SNh1 accountabihty for Th112 was completed in accordance with the SNh1 Accountabihty Plan (Reference 4.3 5) Several plant areas and components weic characterned for SNh1 deposits pnor to initiation of the formal SNh1 Accountabihty Program in some cases, ALARA considerations, the quality of the previous measurements, and lack of actions potentially affectmg SNh1 deposits warranted their use. '!hese measurements were independently iniewed in accordance with Reference 4.3 9 to ensure sufficient data existed to meet SNht accountabihty QA standards. In all cases, the quantity of residual SNh1 was determined through rneasurements, sampling. inspection, or ens:ineering analysi:. MEAEUEMENTS In most cases, measurements were performed in indnidual locations after planned cleanup actisities were completed within the area. In some areas, as stated above, it was determined that the cleanep activit i es did not materially affect the original SNht measurements which were then used for SNh1 accountability. The post-defueling suney required the application of several measurement teclutiques. Technique selection for an individual measurement depended upon the geometry of the component / system or area to be assayed, physical access limitations, radiological conditions, personnel exposure considerations and the probable quantity of SNh1 in the area Where required or desirable, the measurements also imulved use of more than one measurement technique. Since the final SNh1 accountability actisities were classified as "important to Safety," measurements conducted for SNht accountability were performed using QA approved procedures. Ganuna scintillation spectrometry using sodium iodide detectors accounted for the majonty of the early work. Later measurements involmd the use of high purity germanium detectors, which allowed greater resolution for the tracer isotopes ofinterest. Other measurement techniques included alpha O 4.3 - 2 UPDATE 2 - AUGUST 1997
scans using proportional detectors and gross ganuna measurement techniques using collimated Geiger. hiueller detectors. He endfitting and dry vessel measurements were completed using neutron intenogation techniques. Detailed descriptions of the measurement techniques and selection criteria can be found in References 4.3-6 through 4.3 8. SAh!P1ING To obtain additional isotopic and volutnetric information for use with the other anal 3sis techniques, samphng of suspected fuel locations was performed. Solid and liquid samples were obtained from various areas and components to obtain isotopic, composition, and density data for use with measurements and visual inspections. Scrape samples were taken of rnetal surfaces (i c., rnanways, piping, filter housings) to detenrune film depositions. nese samples were analyzed using either on site or off site facilities, applying QA approved procedures.
%$.lJAUNSPECT10E in areas where measurement was not practical, video camera probes were used to estimate the volume of material remaining in the subject area Using the volumetric data generated through sampling, a fuel quantity was assigned.
ENGINEERING ANALYS$ In the latter part of the project, several areas that had not been measured were estimated using a flow-path analysis. The flow-path analysis was perfonned by examination of possible SNh1 introduction pathways into an area through plant systems dunng the accident or subsequent cleanup actisities. 4.3.3.3 Documentation \ ne quantity of residual SNh1 in cach location was documented in a GPU Nuclear engineering calculation. He overall results are provided in References 4.313 through 4.317 and summarized in Tables 4.31 and 4.3 2. Figures 4.31 through 4.3 8 provide the locations of residual SNh1 in the Reactor, Auxiliary, and Fuel llandling Buildings. He engineering calculations were based on geometric configuration, analysis of the measurement data, instrument calibrations, capabihties and
;cfonnance. Also included in the calculations were any specific assumptions made based on review of earlier measurements and the relevant history of that location during the accident and cleanup. All SNh1 engincenng calculations were produced and approved in accordance with approved procedures.
He engineenns calculations, in tum, provide the quantity of SNh1 for a specific area, system or component that is outlined in the PDSR. Each PDSR contains:
- a detailed description of the area, system or component
- its role in the accident ancor cleanup aethities
- the rationale supporting a conclusion as to whether contained residual SNh1 exists and, if so, a summary of the appropriate SNh1 engineering calculations L) 4.3-3 UPDATE 2 - AUGUST 1997
. applicable photographs and'or drawings of the area
- an assesstnent of residual fuel The PDSRs were forwarded to the NRC (e g, References 4.313 through 4.317). ne completed PDSRs fonned the basis for the fmal TMI 2 SNM inventory detailul on 'lables 4.3 1 and 4.3 2.
4.3.4 FINAL SNM ACCOUNTABILITY Tmal accountability was performed by summing the residual fuel quantities identified in the PDSRs and reporting the results as the remaining plant inventory of Special Nuclear Material. He amount of fuel shipped to the DOE INEL was determined by subtracting the sum of the fmal plant inventory and the amount of SNM shipped as radioactive wane from the pre accident plant inventory of SNM, as corrected for decay in the most recent SNM Material Balance Report. PRE ACCIDENT REPORTED INVENTORY (corrected for decay)
- FL1In Plant inventory SNM shipped as Samples /Radwaste
= SNM S!!!PPED TO INEL (m canisters) ne resulting SNM inventory was reported on the PDMS SNM Material Balance Report (DOE /NRC Form 742) 4.3.5 CRITICAlflY ANALYSIS O
The inherent enticahty safety of the residual fuel during the PDMS period has been demonstrated by ' Reference 4 310 which was submitted to support the transition from Mode I to Mode 2 in accordance with the TMI 2 Recovery Technical Specifications and by Reference 4.311 which evaluated RV suberiticahty based on an mereased RV fuel estirnate. He enticality analyses presentcd in References 4.3 10 and 4.3-11 addressed the quandty of residual fuel in each defmed location and the potential for fuel relocation. The analyses estimated the quantity of fuel remaining, its location, its dispersion within the locadon, its physical form (i e., film, finely fragmented, mtact fuel pellets, resolidified), its mobility, the piesence of any mechanism that would contribute to the mobility of the material, the presence of any moderatmg or reflecting material, and its potential for a entical event. Each issue was addressed to the extent appropriate for a given quantity of fuel. He NRC staff concurred with the enticality analyses presented in the DCR via their April 26,1990 letter (Reference 4.312) stating "no objections" to the TMI 2 transition from Facility Mode I to Facility Mode 2. A reanalysis of the RV steady state and accident enticahty safety evaluadons was necessitated by an increase in the estimated quantity of fuel remaining in the RV above that assumed for the DCR. A conservative criticality rnodel was used to bound the most credible fuel configuration. These analyses have demonstrated that criticality has been precluded as a result of the extensive TMI 2 defueling effort. His conclusion was based on three evaluations: the Safe Fuel Mass Limit determination, the bounding Reac:or Yessel steady state criticality calculations, and the potential for l O 4.3 - 4 UPDATE 2 - AUGUST 1997
i criticality under accident conditions, in fact, it was demonstrated that no physically achievable quantity I of residual core debris could result in a critical fuel con 6guration. nerefore, criticahty is precluded for ('} all credible conditions. Althwgh not needed to assure reactivity control over the long-tenn, as an
\
additioral conservative measure, a stable and insoluble neutron poison, consisting of 1400 lbs of Boron Silicate glass shards, was added to the bottom head of the RV. 4.3.6 CONTROL OF SNht DUldNO PDhtS Control of SNht at Thil 2 dunng PDhtS relies upon isolation boundanes and control of access to cornponents which contain SNht. Isolation boundanes will be naintained, as necessary, to prevent relocation of significant SNht quantities The Reactor Coolant Systern, which contains the largest quantity of SNht, will be drained to the cuent practical and isolated within the Containment Buildmg.
%cre will be no physical inventory of SNht quantities at Thil 2 dunng PDhtS because the remairung materials are oflow enrichment, highly radioactise and relatively inaccessible. The NRC has granted Thil 2 an exemption from the 10 CFR 70 $1(d) physical inventory requirements (Reference 4.3 2),
llowever, all shipments of accountable quantities of SNh1 from Tht! 2 during PDhtS will be reponed as required on DOE /NRC 74 i Nuclear hiaterial Transaction Reports. O) (m I \ U 4.3-5 UPDATE 2 - AUGUST 1997
MEh'GS 4.3 1 DOE letter WWD 100 85, thsby. W. W. (DOE) to Durton,11 M (EG&G),
- Accountability for the TMI 2 Core,* dated October 8.1985 4.3 2 Letter, Snyder, D. J. (NRC) to Standerfer, F. R (GPUNC), " Approval of Exemption frorn 10 CFR 30.51,40.61,70 51(d) and 70.53,* dated October 17,1985 4.3 3 NSAC 801, " Analysis of "Direc Mile Island . Urut 2 Accident,* Flectncal Power Research Institute, March 1980 4.34 Rogodn M , et al., "Three Mile Island, A Report to the Comrmssioners and the Public," US Nuclear Regulatory Commission, January 1"'9 4.3 5 GPU Nuclear Procedure 4000 PLN-4420.02, "SNM Accountabihty Plan
- 4.3 6 TP0rFMI 051,
- Location and Characterization of Fuel Debris in TMI 2," Revision 0, April is'44 4.3 7 TP0/FMI 124, "Ex Vessel Fuel Characterization," Revision 0, July 1984 4.3 8 TP0ffMI 187, " Instrument Selection for Residual Fuel Measurements," Revision 0, January 1987 4.3 9 thU Nuclear Procedure 4000-ADM 4420.03, " Review and Quahfication of Selected Preliminary Calculations and Charactenration Measurements for SNM Documentation" 4.3 10 GPU Nuclear letter, 4410 90.L-0012, "Defueling Completion Report. Final Submittal," dated February 22,1990 4.3 11 GPU Nuclear letter, C312 92 2080, "TMI 2 Reactor Vessel Cnticahty Safety Analyses," dated December 18,1992 4.3 12 Letter, Stolz, J. F. (NRC) to Roche, M. D. (GPUNC), "Three Mile Island Umt No. 2 Mode Changes," dated Apnl 26,1990 4.3 13 GPU Nuclear letter, C312 912045, "SNM Accountabihty," transnutting the Auxiliary and Fuelllandling Buildings PDSR, dated June 7,1991 4.3 - $a UPDATE 2 - AUGUST 1997
- IEf1RENCf3 (Cont'd1 4314 GPU Nuctur letter, C312 9120$2, *SNht Accountability,* transmitting the Reactor fluilding Miscellaneous Components PDSR, dated June 18,1991 4.3 l$ GPU Nuclear letter, C312 9120$$, "SNhl Accountability,* transtrutting the Reactor Coolant Sy stem PDSR, dated July 3,1991 4316 GPU Nuclear letter, C312 912064, "SNhl Accountability," transtmtting the 'A' and 'tr Once Through Steam Generators PDSR, Revision 1, dated July 3,1991 .
4.3 l*/ Gl'U Nuclear letter, C312 93 2004, "SNh1 Accountability," transnutting the Reactor Vessel PDSR, dated February 1,1993 4.3 - $b UPDATE 2- AUGUST 1997 b
TAllLE 4.31 FINAL SNhl INVENTORY llY LOCATION Reactor Iluilding I OCATION LESCRIPTION FUFl.(kr) Ril Reactor Ve sci 925 RB Re etor licad 1.3 RB Plenum 2.1 Ril Pressuriur (includmg surge line) 0.5 RB OTSO *A* Tube Sheet 1.4 Ril OTSO *A" Tube llundle 1.7 Ril OTSO *A" Lower llead and J Legs 4.0 Ril "A* Ilot Leg 0.9 RD "A" Cold Legs 7.2 RB "A* Core flood Lme 0.6 Ril OTSO *II" Tube Sheet 36 0 Ril OTSO *D" Tube llundle 9.1 Ril OTSO "B" Lower liead and 11 egs 10.1 RD *D" llot leg 1.8 Ril *D" Cold legs 2.4 Ril *B" Core Flood Line 0.4 Ril Reactor Coolant Pumps 6.2 Ril Decay lleat Drop Line 1.$ Ril liasement Letdown Coolers 3.7' Ril 11asement Reactor Coolant Drain Tank 0.1 Ril Dasement RB Dasement and Sump 1.3 Ril Fuel Transfer Canal 18.9 RIl Core flood System 4.9 Ril incore Instrument Guide Tubes in 21 0 the "A" D nng Ril Upper Lndfittmg Storage Area 5.9 RD Tool Decontammation Facility 0.1 RB DWCS 3.7 Ril Defuelmg Tools 0.6 RD TRVFS 4.4 RD RD Drains 4.4 RD RCS Surface Films 4.6 TOTAL SNhlINVENTORY < 1086 l
- himimum Detectable Limit (h1DL) 4.3-6 UPDATE 2 - AUGUST 1997
I n (o) TAllLE 4.3 2 FINAL SNM INVENTORY llY LOCATION Auxiliary and Fuel llandling Buildings i OCATION DFSCRIPTION FITI,(kcl AX004 Sealinjection Valve Room 0.03 AX006 Make Up Pump Roorn IB 0.07' AX007 Make Up Pump Room I A 023' AX012 Audliary Building Sump Tank Room 0.10 AX015a/b Cleanup Filters 0.10' AX019 Waste Disposal Liquid Yalve Room 0 Ol* AX020 Reactor Coolant Bleed Tank Room til & IC 3.50 AX021 Reactor Coolant Illeed Tank Room l A 0.31 AX024 Auxiliary Buildmg Sump Filters 0.02 AX102 Ril Sump Pump Filter Room AX131 Miscellaneous Waste Tank Room 0.10 AX134 Misecllaneous Waste Tank Pump Room AX112 Seal Return Coolers and Filter Room 0.30' AXl14 MU & P Demineralizer Room l A 1.06 AXl15 MU & P Demineralizer Room - IB 0.13 AXl16 Make Up Tank Room 0.31 [m AX117 MU & P Filter Room 0.06 V) AX128 AX218 Instrument and Valve Room Concentrated Waste Storage Tank Room 0 01 0.01 AX501 RB Spray Pump 1 A 0.01 AX502 Ril Spray Pump - IB 0.01 AX503 DilR Cooler & Pump - I A 0.01 AX504 DilR Cooler & Pump - IB 0.01 F11001 MU Suetion Valve Room 0.46 F11002 Access Corridor F11004 Westinghouse Valve Room 0.16 Fil014 Annulus F11003a Make Up Discharge Valve Room 0.01 F11003b Make Up Discharge Valve Room 0.10 Fil101 MU & P Valve Room 0.32 Fill 09 Spent Fuel Pool "A* 3.80 Filll2 Annulus 0.01 Embedded Valves and Piping (MU System) 0.17 Embedded Valves and Piping (WDL System) 0.04 TOTAL SNM INVENTORY 11.46
'Mimmum Detectable Limit (MDL)
NOTE: All other locations contain less than 0.005 kg UO2 per area. O h 4.3-7 UPDATE 2- AUGUST 1997 l
l APPENDIX 4A O g DEFUEl.ING EOUIPMENT I
]
He major equipment and tools utihzed dunng defueling are described below.
- 1. Shielded rotatable work platform / canister positioning system (eps)
- 2. Defuelmg canisters
- 3. Defuehng vacuum system aithn systems
- 4. Core bore equip r,t
- 5. He autornated cutting equipment system (ACES)
- 6. long handled tools Q).:Shgjdndgla1 Ale Work Platform Once the Reactor llead and Plenum Assembly had been removed from the Reactor Vessel, the core was visible. In order to provide a working platform with the appropriate accommodations for the specialized defueling equiprnent and shieldmg for personnel performing the defueling operations, a totatable and shielded work platform was designcd, fabncated and installed on the top surface of the Reactor Vessel.
ne Stuelded Rotstable Work Platform is approximately 17 feet in diameter. The penmeter of the , j platform is a fabricated wide flange beam with roller assemblies mounted on the lower flange. He V roller assemblies mate with the support rail mounted on the support structure. A cable drive system provides the rotational drive for the platform. His platfonn supports sis inches of stainless steel shieldmg as well as some of the defueling tools and their reaction loads and the operators. nree transfer ports are provided to allow canisters to be installed and removed through the Shielded Work Platform. Two removable jib cranes are mounted on the Shielded Work Platform to aid the operators in the rnanipulation of the long-handled tools in the tool working slots. To avoid inadvertent movement of the Shielded Work Platform, a manual dise-type brake is attached to the senice platform of the shielded support structure. The skirt on the Shielded Work Platform serves as the disc. A cutaway view of the Shielded Work Platform is shown in Figure 4A 1. He shielded work platform and canister positioning system were left "as-is" for PDMS. (21 - Defueline Canisters Due to the post accident condition of the TMI 2 core, normal means of removal and shipment of the fuel were not possible. It was necessary to design and fabricate special canisters in which to contain and ship the core debris. nree types of canisters were required for the core debris; (1) filter, (2) knockout, and (3) fuel. The three types of canisters were required due to the various forms of debris, p 1 J 4A-1 UPDATE 2 - AUGUST 1997
which ranged from very small fuel fines to partial length fuel assemblics. Not only did these canisters serve as containers for removal and stupment of the core, they also serve as the long term storage containers, and by design, ensure subcriticahty of the contents. A canister consists of a circular pressure vessel, housing one of three types of intemals, depending on the function of the caruster, Except for the top closures, the outer shell is the sarne for all three types of canister design. He upper closure head design varies accordmg to the canister's function. The canister serves as a pressure vessel protecting against leakage of the canister contents as well as providmg structural support for the neutron absorbing materials. It is designed to withstand the pressures associated w1th expected conditions during shipping and storage and accident loads associated with postulated events intohing the handling or shipment of We canister. De fuel canister is a receptacle for large pieces of core debns. ne fuel canister consists of a cylindrical pressure vessel with a flat upper closure head Wrthin the outer shell, a full-length square shroud forms the internal cavity as shown in Figure 4A 2. This shroud is supported at the top by a bulkhead that ruates with the upper closure head. Both the shroud and core debris rest on a support plate that is welded to the shell. De support plate has impact plates attached to absorb canister drop loads and payload drop loads. He shroud assembly consists of a pair of concentric square stainless steel assemblies welded to cornpletely enclose four sheets of Doral, a neutron absorbing material. The shroud intemal dimensions are larger than the cross section of an undamaged fuel assembly. ne shroud extemal dimensions are slightly smaller than the inner diameter of the canister, thus prosidmg support at the shroud corners for lateral loads. De void area outside of the shroud is filled with a cement and glass bead mixture to the maxinmm extent practical to ehminate migration of the debris to an area outside of the shroud dunng a design basis accident. Le upper closure head is attached to the canister by eight equally spaced bolts. These bolts are designed to withstand the design pressure loads, handling loads, and postulated impact force due to shifting of the canister contents during in plant load drop or a shipping accident. He knockout canister, Figure 4A 3, was designed to be used as part of the vacuuming system with flow fittmgs which were capped or plugged after use The intemals module for the knockout canister is supported from a lower header welded to the outer shell. An array of four outer neutron absorber rods surround a central neutron absorber rod for enticality control. The four outer rods and the central absorber rod are filled with sintered D4C pellets. Lateral support for the neutron absorber rods and center assembly is provided by intennediate support plates. De filter canisters were designed to function in conjunction with either the Defueling Water Cleanup System or the Westinghouse Vacuum System to remove small debris particles from the water. The tilter assembly bundle that fits inside the ranister shell was designed to remove particulates down to 0.5 microns. Flow into and out of the filter canister is through two quick disconnect fitting. See Figure 4A 4. The intemal filter assembly bundle consists of a circular cluster of 17 filter elements, a drain lin and a neutron absorber assembly. He influent enters the upper plenum region, flows down past the support plate, through the filter media and down the filter element drain tube to the lower sump. The O 4A-2 UPDATE 2 - AUGUST 1997
l flow is from outside to inside with the particulate remaining around the outer perimew of the filter p clernent. De filtered water exists the canister through the drain line.
' All fuel, filter and knockout canisters that wrre utilized during the recoscry/defueling period were shipped offsite for storage. Ilowever, one canister vessel (outer shell), without any mternals, wu used to store three fuel assembly end fittings. His canister was placed into the canister storage rack located on the south end of the fuel transfer canal, widch is covered for PDMS. Three dummy canister test weight vessels are stored in the "A" spent fuel pool during PDMS.
)
l (3) Defuehng Yacuum S)Mrm - Airlift S3 sJrs )
%e vacuum system was designed to remove small sire fuel debris and fuel fmes from the debris bed and discharge them into an appropriate canister. It also was designed to be adapted to permit vacuuming of debris in the lower vessel head region. The system was composed of a pump, piping, vahmg, and other rniscellaneous components. It was located under, and supported from the Shielded Work Platform. It had a control console mounted on the auxiliary platform.
%e vacuum pickup nozzle was connected to a defueling canister by a flexible hose and was manually manipulated by a long handled tool supported from the Shielded Work Platform. The system was modular to permit remote installation and removal of the pump, load cell, vahing and piping sections.
%e airlift systems were devices designated to tramfer small fuel debris from the debris bcd to a fuel canister. Dese devices either discharged the debris directly to the canister or dumped the debris into a contamer which was then used to durnp the debris into the canister, ne airlift consists of a suction nozzje with air inlet, a separation chamber, and in some cases, a debris bucket. Depending on the unit, the debris bucket was removed from the airlift for dumping the debris into a fuel caruster or the debris bucket was integral with the separation chamber. He integral bucket was equipped with a discharge chute having a sliding door for dumping the debris into a fuel canister, lloses have been removed and the remaining hardware will remain "as is" during PDMS.
(4) Core lipJrEquipjDCut
!! was known that a large solidified mass existed below the loose core debris. Equipment was designed and fabricated which could drill into this mass and extract core stratification samples. The core tore equipment was subsequently modified to be fitted with a solid face bit to break up the solid material into smaller pieces The core bore equipment also was tised to cut some of the lower core support assembly structural elements. His equipment was mourWJ and supported from the rotatable platform. This equipment ha' been disassembled and is stored inside the R.D., on top of the "A&B" D ring missile shields (36T).
(5)-The Automated Cuttme Equinment System (ACFS) De ACES was installed and operated to cut the lower core support assembly stmetural elements to provide access to remove fuel from the core support assembly and the Reactor Vessel lower head. He system consisted of the required framework and mechanical elements to position and control a plasma torch underwater during structure cutting. These parts were located in the Reactor Vessel and (Ov) 4A-3 UPDATE 2 - AUGUST 1997
I were seniecd with electrical power and other senices delivered through umbilical cords from outside the vessel. %e plasnu torch was controlled and operated by a inodified commercial power supply. He movement of the torch wu powered and controlled by a special five ads control system This equipment was disassembled and is stored on the 347 clevation of the R D., east of the "B" D nng. Rd.done llaridi f d.hds ne Rotatable Shielded Work Platform is kcated approdmately 40 feet above the workmg arca in the core region his necessitates the use oflong handled tools which can be operated frorn the Shielded Rotatable Platform. The tools are operated through slots in the totatable platform. hiost tools are supported by an overhead crane that provides vertical and lateral motion. Several crancs are available for use, mcludmg the two jib cranes on the platform, the Reactor Building service crane and the polar crane. hiost of the long handled tools are stored in vanous tool storage racks inside the R.B. dunng PDh15. O O 4A 4 UPDATE 2 - AUGUST 1997
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.O i
s CIIAPTER 5 PDh1S RADIOLOGICAL CONDITIONS O O
- . . ~ . . - _ _ . .- - _ _
O V CHAPTER 5 PDMS RADIOLOGICAL CONTROLS TABLE OF CONTENTS SECTION TITLE PAGE 5.0 ' INTRODUCTION 5.0-1 5.1 GENERAL DECONTAMINATION ACTIVITIES 5. ' -1 5.1.1 GENERAL DECONTAMINATION OBJECTIVES . i-t 5.1.2 REACTOR BUILDING 5.1 2 5.1.2.1 Reactor Building Elevation 282'(Basement and Sump) 5.1 2 5.1.2.2 Reactor Building Elevations 305' and 34T 5.1 2 5.1.2.3 Reactor Building D Rings 5.1 2 5.1.2.4 Refueling Canal 5.1 2 O 51.3 AUXILIARY AND FUEL HANDI ING BUILDINGS 5.1 2 V 5.1.3.1 Auxiliary Building 5.1 2 5.1.3.2 Fuel Handling Building 5.1 2 5.1.4 OTHER BUILDINGS 5.1-3 5.1.4.1 Senice Building (Elevation 28l', 5.1 3 Tendon Access Gallery) 5.1.4.2 Senice Building (Elevation 305') 5.1-3 5.1.4.3 - Control Building Areas (East & West) 5.1 3 5.1.4.4 Turbine Building (Elevation 281') 5.1 3 5.1.4.5 : CACE Building 5.13 5.1.5 SYSTEMS 5.1-3 i 5.1.5.1 Reactor Coolant System 5.1 3 [ 5.1.5.2- Non-Reactor Coolant Systems 5.1 3
-(
\ i UPDATE 2 - AUGUST 1997
_. -~ .- - ,.
s CHAPTERS TABLE OF CONTENTS (Cont'd) SECTION' TITLE PAGE 5.2 SPECIFIC DECONTAMINATION ACTIVITIES 5.2-1 5.2.1 SPECIFIC DECONTAMINATION GOALS 5.2 1 5.2.2 GENERAL CilAPACTERIZATION OF 5.2-1 CONTAMINATIONS 4 5.2.3 DECONTAMINATION TECHNIQUES 5.2 2 5.3 PDMS RADIOLOGICAL SURVEY 5.3-1 't 5.3.1 RADIOLOGICAL ASSESSMENT 5.3-1 5.3.1.1 Pre-PDMS Radiological Survey Methodology 5.31 5.3.1.2 ' Post PDMS Radiological Survey Methodology 5.31 5.3.1.3 Remedial Decontammation Actisities 5,3-1 ; p 5.3.2 RADIOLOGICAL CONDITIONS AT BEGINNING OF PDMS 5.3-1 ( 5.3.2.1 Surface Contanunation at Beginning of PDMS 5.32 APPENDIX SA - Potential Reductions in Occupational Exposure SECTION TITLE PAGE S A.1
SUMMARY
AND INTRODUCTION 5A-1 SA. l.1
SUMMARY
$A-1 5 A. I.2 INTRODUCrlON SA-1 SA.2 BACKGROUND INFORMATION 5A-2 5A.2.1 RECOVERY TASKS SA-2 S A.2.1.1 Decontamination Task Force Report 5A-2 5A.2.2 FUTURE DECONTAMINATION TASKS 5A-3 5A.3 CRITERIA FOR PERFORMING THE EVALUATION SA-3
! ii UPDATE 2 - AUGUST 1997 i l l
APPENDIX 5A - Potential Reductions in Occupational Exposure (Cont'd) SECTION 11TLE PAGE 5A.4 RADIOLOGICAL CONDm0NS 5A-4 5A.4.1 CONDmCSS AT THE END OF PHASE Ill 5A-4 5A.4.2 CONDITIONS DURING PDMS SA-4 5A.4.3 CONDm0NS AT THE END OF PDMS 5A-4 5A.4.4 BASE CASE PLANT CONDm0NS 5A 5 5 A 4.5 ISOTOPIC CHARACTERIZATION SA 5 5A.5 OFF SITE DOSES DURING PDMS SA-6 4 5A.6 DOSE RATE ESTIMATION SA-6 S A 6.1 DECONTAMINATION TASKS SA-6
- 5A.6.1.1 Tasks Performed Robotically SA-7 5 A.6.1.2 Example of Estimating Dose Rates SA-7 ,
5A.6.2 RADIOACTIVE WASTE PROCESSING 5A-7 4 5A.7 ACTIVITIES EVALUATED- SA-8 5A.7.1 FUTURE DECONTAMINATION TASKS SA-8 5A.7.2 RADIOACTIVE WASTE MANAGEMENT TASKS SA-8 5A.7.3 ACTIVITIES DURING PDMS 5A-8 5 A.8 - - JOB-HOUR ESTIMATES SA 8 5A.8.1 SCOPE MODIFICATIONS SA-9
' 5A.8.1.1 - Immediate Additional Decontamination 5A 9 5 A.8.1.2 Final Decontammation As A Part of Decommissioning SA 10 SA.8.1.3 Example of Estimating Scope Modification SA-10 5 A.8.2 RADIATION AREA JOB-HOURS SA-10 -(
iii UPDATE 2 - AUGUST 1997
. _ _ _ - _ - _ _ . _ _ _ . _ . _ , _ . . _ . - . . _ - _ . . _ . . _ . . ~ . _ . _ _ . . . _
d APPENDIX 5A Potential Reductions in Occupational Esposure (Cont'd) SECTION TITLE PAGE 5 A.8.2.1 Example of Determining Radiation Area Job-hours 5A 11 S A.8.3 JOB-HOURS DURING PDMS SA 11 , 5A.8.4 RADIDACTIVE WASTE MANAGEMENT SA 11 5A.8.5 EXAMPLE OF JOB-HOUR ESTIMATION 5A 12 5A.9 PERSONNEL EXPOSURE 5A 12 5A.9.1 DECONTAMINATION TASKS 5A 13 5A.9.2 RADIDACTIVE WASTE MANAGEMENT 5A 13 J SA.9.3 PDMS TASKS 5A-13 5A.9.4 TOTAL PERSON REM 5A-14 5A.10 COMPARISON CASE YEAR PDMS SA 14 SA.ll CONCLUSION 5A 14 5A.12 REFERENCES 5A-15
- APPENDIX B -DECONTAMINATION ACTIVITIES 5B-1 Kelly Vacuumac $B 1 5B-1 Concrete Scabbling Strippable Coatings- 5B-1
, liigh Pressure Spray and Flushing 5B-1 i System Flushing 5B 1-i- Hands-on Decontamination 5B-2 e ROBOTIC DEVICES 5B-2 System In-Service Inspection (SISI) 5B-2 T' Remote Controlled Mobile Manipulator (RCMM) SB-2 iv UPDATE 2 - AUGUST 1997 1 1
a-a 4 l APPENDIX 5B (Cont'd) : SECTION TITLE g i Remote Reconnaissance Vehicle (RRV) SB 2 Remote Controlled Transport Vehicle (RCTV) 5B-3 [ Louie-2 5B 3 4 , WATER PROCESSING SB-3 i Submerged Demineralizer System 5B 3 EPICOR II 5B-4 Defueling Water Cleanup System (lon Exchanger-Portion) S B-4 4 i 3 4 i 4 4 i 4 v UPDATE 2- AUGUST 1997 .j
_. .= . . . . .. . . - . . . - . - . . - - . - - - . - . - . - . _ . . - . . . _ . ..- .- i 1 CHAPTER 5 ; TABLE OF CONTENTS (Cont'd) , y LIST OF TABLES ? TABLE NO. TITLE PAGE i' 5.1 1 BASELINE RADIOLOGICAL CRITERIA - REACTOR : 5.1 4 BUILDING 5.12 BASELINE RADIOLOGICAL CRITERIA AFHB 5.1 5 5.21 . SPECIFIC DECONTAMINATION GOALS - REACTOR 5.2 3 BUILDING ' 5.31 PDMS RADIOLOGICAL CONDITIONS REACTOR 5.3-3 BUILDING 53 2 PDMS RADIOLOGICAL CONDITIONS - AFHB 5.3-4 5.33 = PDMS RAblOLOGICAL CONDITIONS - OMER 5.3 13 i BUILDINGS ( 5.3-4 SURFACE CONTAMINATION - REACTOR BUILDING 5.3-14 5.35 SURFACE CONTAMINATION - AFHB 5.3-15 5.36 SURFACE CONTAMINATION - OTHER BUILDINGS 5.3 24 f SA 1 PHASE III ENDPOINT CRITERIA 5A 16 5A 2 END OF PHASE Ill AND BASE CASE DOSE RATES - 5A-17 REACTOR BUILDING
.5A 3 END OF PHASE III AND BASE CASE DOSE RATES - 5A-18 AUXILIARY AND FUEL HANDLING BUILDINGS -
SA-4 ESTIMATED Cs AND St ACTIVITIES IN THE 5A-19 REACTOR BUILDING SA 5 DOSE RATES AND ADJUSTED JOB-HOURS 5A-20 SA-6 ~ FINAL CLEANUP ACTIVITIES (TYPICAL) 5A-22 5A 7 RADIATION AREA JOB-HOURS 5 A , SA 8: ETN AND RWP HOURS 5A-25 ' Li.s _ ll . vi UPDATE 2 - AUGUST 1997
'l
7___._..__.___...._.___._._.._._.. I o i : I i i i CHAPTER 5 1 TABLE OECONTENTS (Co W'd) < LIST OF TABLES y
- TABLE NO. TITLE PAGE
, 5A 9
SUMMARY
TABLE FOR PERSON-REM EVALUATION SA 26 , l j - ANALYSIS i ' 5A 10 ESTIMATED TOTAL PERSON-REM FOR FINAL 5A 29 DECONTAMINATION OPTIONS 4 i i d d i 4 f I i-i 1 4 vii - UPDATE 2 - AUGUST 1997 L___-_._.__-__________ . . . - . . . , _ . . . - . . . . , . - - , _ . , , _ - . . _ . . . . . , , . . _
CHAPTER 5 PDMS RADIOLOGICAL CONDITIONS 5.0 - INTRODUCrlON One consequence of the March 19'!9 accidcat was wide spread radioactive contamination of the Reactor, Fuel Handling, and Auxiliary Buildings. Reactor Coolaat System water was released to the Reactor Building and overflowed to the Auxiliary and Fuel Handling Buildings. These areas required extraordinary decontamination efforts to achieve the cleanup program objectives. Contamination of areas outside the RB and AFHB were minor and limited. These areas outside the RB and AFHB were decontaminated a.xi either released for unrestricted use or configured such that the contamination is suitably contained. The objecth ; of the decontamination program were to remove and/or stabilize the contamination to reduce occupational exposure to workers and to prevent release of contamination to the environment during recovery and cleanup activities. In addition, a final decontanunation objective was to ensure that any remaining contamination was stable and suitably isolated for the PDMS period. The following discussion provides the major decontamination objectnes and techniques utilized at TMI 2, as well as the contamination levels remaining in the various areas of the plant. Also included (Appendir SA) are potential reductions in occupational exposure due to PDMS. O 4 i
-l
' 5.0 - 1 _ UPDATE 2 - AUGUST 1997
A / \ ')' 5.1 GENERAL DECONTAMINATION ACTIVITIES The decontammation of TMI 2 was accomplished in two phases. De initial phase improved the radiological conditions in the facility such that the dose rates were acceptably low to permit the necessary accident cleanup activities. He second phase was a decontamination program initiated for the purpose of systematically improving radiological conditions in the plant regardless of whether access was required for cleanup actisities. In order to establish goals for the TMI 2 decontammation program, decontanunation criteria were established for the RB, AFHB, and specific portions of other buildings, where relevant. The primary consideration in establishing the criteria for the initial phase was the anticipated need for personnel access during the cleanup and post cleanup timeframe. Table 5.1-1 presents the beeline Radiological Criteria for the RB and Table 5.1-2 provides the Baseline Radiological Cnteria for the AFHB The general decontammation criteria for piping systems, equipment, and components were based solely on their contribution to area dose rates. It should be emphasized that these criteria were intended as guidelines rather than absolute criteria. The guidelines were used in conjunction with common sense and good radiological practices. 5.1.1 GENERAL DECONTAMINATION OBJECTIVES ne general decontamination program objectives were assigned primarily by area. The general objectives establi:hed for each area were based on initial contamination levels, the need for personnel access, and the possibility of release of radioactivity to the emironmat. (oV) he facility was categorized into the following areas:
- n. Reactor Building
- 1. Reactor Building elevation 282'
- 2. Reactor Building elevations 305' and 347'
- 3. Reactor Building D-Rings
- 4. Refueling Canal
- b. Auxiliary and Fuel Handling Buildings
- 1. Auxiliary Buildmg Cubicles (total of 99)
- 2. Fuel Ha.xiling Bui' g Cubicles (total of 37)
- c. Other Buildings
- 1. Senice Building (elevation 28 l', Tendon Access Gallery)
- 2. Senice Building (elevation 305")
- 3. Control Building Areas (East & West)
- 4. Turbine Building (elevation 28l')
- 5. CACE Building
- d. Systems
- 1. Reactor Coolant System
,, 2. Non-Reactor Coolant Systems He general decontammation objective for cach area is outlined below.
(U) 5.1-1 UPDATE 2 - AUGUST 1997
5.1.2 REACTOR BUILDING O 5.1.2.1 Reactor Building Elevation 282'(Basement and Sump) The general decontamination objectives in the Reactor Building basement were to enhance the ability to characterize the basement with respect to residual fuel and to promote the long-term stability of contamination during the PDMS period Re specific program objectives included remosing the sludge from all accessible areas in the Reactor Building basement and conduct a high pressure flush of walls, floors, and overheads. 5.1.2.2 Reactor Building Elevations 305' and 347' he original general decontammation objective on elevations 305' and 347' of the Reactor Building was to reduce dose rate levels low enough to support continued personnel occupation during defueling operations and other associated actisities. These dose rate goals were achieved to allow for defueling operations. However, post-defuelmg draindown of CFT 1 A has increased the dose rates for elevations 305' and 347'such that the PDMS general area dose rate goal had to be increased to 100 mR/hr. 5.1.2.3 Reactor Building D-Rings The general decontamination objectives within the upper portion of the D-rings were to support efforts to locate, characterize, and remove fuel from the pressurizer, ste.un generator, and reactor coolant piping and to evaluate the feasibility of RC pump removal. 5.1.2.4 Refueling Canal ne original general decontanunation objective in the refueling canal was to reduce dose rates sufficiently to tupport construction activities for defueling equipment installation. These dose rate goals were achieved for installation of defueling equipment and actual RV defueling. However, post-defueling pumpdown of the RV and the refueling canal has increr. sed the dose rates in the refucling canal such that the PDMS general area dose rate goal had to be increased to 100 mR/hr. 5.1.3 AUXILIARY AND FUEL HANDLING BUILDINGS 5.1.3.1 Auxiliary Building The Auxiliary Building was categorized into 99 separate areas or cubicles, which accounted for essentially all of the walls, floors, and overhead areas in the building. He general decontanunation objective for the contammated cubicles was to support ongoing cleanup actisities. The Baseline Radiological Cntetis referenced in Table 5.1-2 were applied to each Auxiliary Building cubicle as a guideline for the decontanunation efforts. 5.1.3.2 FuelHandling Building He Fuel Handling Building was categorized into 37 separate areas or cubicles which accounted for essentially all of the walls, floors, and onrhead areas in the building He general 5.1 - 2 UPDATE 2 - AUGUST 1997
4 r . decontamination objectives for the contammated areas within the Fuel Handling Building weie established to suppott ongoing cleanup activiti es and to stabilize locahzed contanunation. As in the Auxiliary Building, tie Baseline Radiological Criteria referenced in Table 5.1 2 were applied to each Fuel Handling Buildmg cubicle for use as guideline for the programmatic decontammation effort. 5.1.4 OTHER BUILDINGS 5.1.4.1 Senice Building (Elevation 281', Tendon Access Gallery) De general decontamination objective in the Senice Building was to identif fhot spots and remose or shield contammation. One objective was the partial removal and stabilization of the contammated cork 3 seal material from a building construction joint. 5.1.4.2 Senice Building (Elevation 305') The general decontammation objective in Jie Senice Building (elevation 305') was to identify hot spots and remove or shield contammation. 5.1.4.3 Control Building Areas (East & West) The general decontanunation objectim in this portion of the Control Building Areas was to identify hot spots and remove or shield contamination. 5.1.4.4 Turbine Buildmg (Elevation 28l') The general decontamination objective in this portion of the Turbine Building (elevation 281') was to identify hot spots and remove or shield contamination. 5.1,4.5 CACE Building ne general decontamination objective in the CACE Building was to identify hot spots and remove or shield contamination. 5.1.5 SYSTEMS 5.1.5.1 Reactor Coolant System ne general decontammation objective for the Reactor Coolant System was to remove residual fuel and other contamination to the maximum extent practical. 5.1.5.2 Non-Reactor Coolant Systems The general decontamination objective for the non-reactor coolant systems was to reduce radiation dose , rates low enough to support general area dose rate goals. The systems decontammation program was integrated with the systematic cubicle and area decontammation programs. C '. t t 5.1 - 3 UPDATE 2- AUGUST 1997 l l 1
TABLE S.1 1 O BASELINE RADIOLOGICAL CRITERIA - REACTOR BUILDING GENERAL AREA SURFACE CONTAMINATION DOSE RATE (dom /100 cm*) (mR/hr) bel.OW 7' ABOVE 7' infrequent (Quarterly) Elevation 305' to 34T <100 <50,000 NOTE 3 Elevation 34T and Above < 30 <50,000 NOTE 3 Refuchng Canal <600 <50,000 NOTE 3 Top of D-Rings <100 <50,000 NOTE 3 Access Not Required D-Ring Interior, El. 349' and Above "AS IS" "AS IS" NOTE 3 Basement (El. 282') "AS IS" "AS IS' NOTE 3 O NOTES:
- 1. Rese criteria refer to general access areas and should not be used as maxirnum leu's allowed (i.e.,
hot spots could have higher dose rates).
- 2. Areas of the Reactor Building not listed in this table are left "as is* for PDMS.
- 3. Since access is not required above 7', and the overheads are generally not the major contributors to the general area dose rate, these areas may be left without decontammation.
l l l WDATE 2 - AUGUST 1997
7 k TABLE 5.1-2 BASELINE RADIOLOGICAL CRITERIA AFHB - GENERAL AREA MAXIMUM SURFACE DOSE RATE HOT SPOT CONTAMINATION ACCESS FREOUENCY (mR/hr) (mR/hr) (dom /100 cm*)
- Routine - <2.5 10 NOTE 1 j
(40 hr/wk access) Frequent - < 50 100 NOTE 1 (daily access) I Occasional- < 500 1000 NOTE 1 , (weekly access) Infrequent - < 1000 2000 NOTE 1 (monthly access or less) P J q
- NOTES
2
, 1. Smearable contammation (dpm/100 cm ) will be either:
- a. <1000 below 7/<10,000 in overheads (generally applied to corridors and access ways) or
- b. <50,000 below 7'/<50,000 in overheads (generally applied to cubicles)
.1
/
{N 5.1 - 5 UPDATE 2 - AUGUST 1997
( V 5.2 SPECIFIC DECONTAMINATION ACTI\TTIES The prunary decontamination activities at TMI 2 were in the AFHB and the Reactor Buildmg. Some decontamination tasks were also required in other areas such as the Turbine and Senice Buildings. However, the extent of work required in these areas was relatively minor compared to the AFHB and Reactor Building work scope. 5.2.1 SPECIFIC DECONTAMINATION GOALS Specific decontamination goals were developed by applying the baseline decontanunation criteria introduced in Section 5.1 to specific areas within the facility. Note that these specific decontanunation goals apply to general area access and do not represent the maxtmum levels allowed (i.e., hot spots could have higher dose rates and contammation levels). Categorization of the facility into specific areas was done within the context of the general decontammation program objectives as discussed in Section 5.1.1. In the case of the Reactor Building, the physical layout of the building and the anticipated personnel access requirements favored the use of the general enteria as specific goals. The Auxiliary and Fuel Handling Buildings and the specific portions of other buildings are configured in such a way as to favor applying the general criteria to much smaller areas, designated as cubicles, to derive the specific decontamination goals. He specific decontamination goals for the Reactor Building are presented in Table 5.3 1. Table 5.3-2 presents specific decomammation goals for the Auxiliary and Fuel Handling Buildings and Table 5.3 3 encompasses other buildmgs. i O j 5.2.2 GENERAL CHARACTERIZATION OF CONTAMINATION V Rere were various conditions of radiological contamination at TMI-2 after the accident. The majonty of contammation can be characterized as a layer ofloose particulate covering soluble contammants that had been absorbed into the coatings and the surface layers of concrete and dried. Several areas, such as the Reactor Building basement, presented particularly difficult decontammation challenges. Originally, it was anticipated that water flushing would remove the majority of surface contammation. However, subsequent to the early decontammation efforts, decontammated surfaces often became recontaminated to the earlier contanunation levels. In order to identify the sources of recontamination, survey techniques were developed to determine if the recontamination was from outside sources or from other mechanisms. It was determined that much of the recontamination resulted from soluble contammants leaching from the surface materials into which they had been absorbed during and following the accident. Te:hiques were developed to physically remove coatings from floors or to remove a layer of concrete from bol eca M and non-coated surfaces. After water flushing removed the loose surface contammation on areas where large open surfaces were contammated, scabblers were used to reduce contammation levels to permit cleanup actisities with as little personnel exposure as possible. The floor contammation levels in most areas of the AFHB were reduced to those typical of pre-accident conditions by this technique. In the Reactor Building, radiation levels in frequently accessed areas were reduced by 85% by scabbling accessible areas and shieldmg finite sources. In the AFHB cubicles, it was necessary to flush piping in some areas to remove contained contammants t T before scabbling w2s performed. The logic was to conduct a gross flush of an area using remote nozzles. A U 5.2 - 1 UPDATE 2 - AUGUST 1997
survey was then performed. Ifless exposure was required to perform a system flush of piping followed by scabbling. the piping flush was initially performed. Some exceptions to this logic were encountered because flow points through piping were totally blocked or a very localized area contained a large source of radioactive material. In these cases, the source of contammation was physically removed. Approximately 17 Ci of contammation remains in the AFIIB floor drains and approximately 600 Ci remains in the AFIIB sumps with over 99% contained in the Auxiliary Building smnp. The AFHB sumps are in senice. Most floor drams in the AFilB remain open and unaltered. Some drains were altered to have either a ball float valve instahed or to be plugged. This was done for those drains that had become a source of contamination in their cubicle. 5.23 DECONTAMINATION TECIINIQUES The major decontammation techniques utilized during the cleanup period are described in Appendix 5B. O 5.2 - 2 UPDATE 2 - AUGUST 1997
I D 5.3 PDMS RADIOLOGICAL SURVEY nis section describes the radiological conditions at the TMI 2 facility upon entering PDMS. These con &tions are expressed in terms of general area dose rate, loose surface contammation, and general isotopic distribution. 5.3.1 RADIOLOGICAL ASSESSMENT Upon completion of cleanup activities (includmg decontammation) in a given area or cubicle, the area was isolated to prohibit uncontrolled access. Deactivated systems traversing the area or cubicle were drained, vented, and isolated. De subject area was, at that point, configured for long-term momtored storage and available for a fmal PDMS radiological assessment. This assessment was performed utilizing radiological surveys (in this case, radiation, contanunation and air activity surveys performed by radiological controls technicians) as a basis for determuung whether the established decontamination program endpoints were achieved as well as to dor ument the radiological con &tions which existed upon entering PDMS. If at the conclusion of the radiolnical assessment it was determined that satisfactory radiological conditions were not achieved, additional de.ontamination efforts were undertaken or exceptions to the goals were taken. 5.3.1.1 Pre-PDMS Radiological Suney Methodology Radiation contanunation and air activity sun eys were routinely performed during the course of the cleanup program in support of work activities. Rese surveys were performed in accordance with regulatory and industry standards and practices to verify and document ra6ation and contammation levels for use in controlling personnel exposure. These surveys were then evaluated as to whether or not they supported the V( q f conclusion that decontammation endpoints had been achieved in those instances where existing suneys were judged unsuitable for substantiating decontammation endpoints, additional surveys were conducted. 5.3.1.2 Post-PDMS Radiological Sunty Methodology During the PDMS period, radiological conditions within the facility will be monitored through sampling and periodic suncillance. Surveillance activities for the Reactor Building consist of radiation suntys in conjunction with planned Reactor Building inspections. He purpose for conducting these surveys is to provide assurance that conditions are stable or to provide early indication of any changing conditions which may require corrective action. The Radiological Survey Plan for PDMS is described in Section 7.2 4.2. 5.3.1.3 Remedial Decontammation Actisities in the event that changing conditions are indicated, an evaluation will be performed as to the need for, and form of, corrective action to be taken. In general, areas will be assessed on a case-by-case basis with the decidmg factors being the area's i.npact on personnel exposure and the possibility of a release to the emironment. The Urut 2 PDMS organization will be staffed to provide the capability to take corrective action or call upon additional resources necessary to take corrective action. 5.3.2 RADIOLOGICAL CONDITIONS AT BEGINNING OF PDMS Table 5.3 1 lists the specific radiological goals for the TMI 2 Reactor Building and the corresponding radiological conditions as of the most current radiological surveys existing in September 1992. The RB 7 (V I 5.3-1 UPDATE 2 - AUGUST 1997
radiological conditions listed in Table 5.3 1 reflect routxied-oft, average PDMS survey data for the entire O cubicle / area in question. Rese data were compiled in the manner described in Section 5.3.1.1. Table 5 i a lists the equivalent information for the AFHB as of November 1993. Table 5.3 3 provides a summary of the radiological conditions for the balance-of plant areas not covered by Tables 5.31 and 5.3 2. 5.3.2.1 Surface Contamination At Beginning Of PDMS To establish a baseline at the beginning of PDMS, the radioactivity present as surface contamination m various areas of the facility has been evaluated. This information serves as an initial reference for the evaluation of any future activities in the respective areas, in order to appraise the radioactivity present as loose surface contamination upon entry into PDMS, an analytical model was constructed utilizing available loose surface contammation data, generalized waste stream isotopic distributions and estimates of surface area. This information is formatted in a marmer similar to the general area dose rate and loose surface contamination data presented in Tables 5.3-1, 5.3 2, and 5.3 3. Only surface contamination was considered; fixed contamination or contamination intemal to piping systems or equipment was omitted. The generalized waste streams or distribution of prmeipal isotopes are referenced on each of the tables. Table 5.3 4 lists the data obtained from the analytical model described above for the TMI-2 Reactor Building Table 5.3-5 hsts the equivalent data for the AFHB and Table 5.3-6 provides a similar sununary for the balance of plant areas. All of the calculations of the quantities of curies listed are based on the specific decontammation goals given on Tables 5.3-2 and 5.3 3. 5.3-2 UPDATE 2 - AUGUST 1997
O 3 O TABLE S.3-1 PDMS RADIOLOGICAL CONDITIONS - REACTOR BUILDING AREA DESCRIPTION SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE GENERAL AREA SURFACE DOSE RATE CONTAMINATION DOSE RATE CONTAMINATION (mR/hr) (dom /100 cm*) (mR/hr) (dom /IDO cm*]
<100 <50,000 150 2,000,000 Elevation 305' to 347
<30 <50,000 50 710,000 Elevation 347 and Above
<100 <50,000 120 670,000 Refueling Canal D-Ring Interior, EL 349'and Above "AS IS" "AS IS" 300 280,000 l "A" D-Ring "AS IS" "AS IS" 200 220,000 "B" D-Ring
<100 <50,000 40 270,000 Top of D-Rings
("A" D-ring) 50 190,000 ("B" D-ring) 13asement, El. 282' "AS IS" "AS IS" 56,000 " Note I
- The radiological conditions in this table reflect rounded-off, average PDMS survey data.
" This is the decay-corrected dose rate taken by ROVER (TPB 85-3 Rev. O, February 1985 )
NOTES:
- 1. This area is inaccessible; no meaningful data exists.
5.3-3 UPDATE 2 - AUGUST 1997
TAELE 5.3-2 PDMS RADIOLOGICAL CONDITIONS- AFIIB SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAT11 NATION GENERAL AREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/ Overheads DOSE RATE <7'/ OVERHEADS NUMBER AREA DESCRIPTION (mR/hr) (dom /100 cm2) (mR/hr) (dom //100 cm')
AXu01 RB Emerg. Cooling Booster <2.5 <l,000/<10,000 1.5 59tV2,500 Pumps Area [WGP-1 <2.5 <l,000/<10,000 2.4 81,000/6,500 Shielded Enclosure] AX002 Access Corridor <2.5 <l,000/<10,000 1.5 110/l,300 AX002a N2 Piping System <2.5 -1,6t'X1/N/A 1.8 110/N/A AX003 Access Area <2.5 <l,000/<10,000 0.8 2,900/2,300 AX004 Scal Injection Valve Rm <1000 <50,000/<50,0N) 120 68,000/750,000 AX005 MUP Pump 1C Rm <500 <50,000/<50,000 8 40,000/30,000 AX006 MUP Pump iB Rm <500 <50,000/<50,000 60 88,000/29,000 AX007 MUP Pump I A Rm <500 (50,000/<50,000 40 9,200/215/00 AX003 Spent Resin StorTank IB Rm "AS IS" "AS IS* 710 960,000/3,400,000 AX009 Spent Resin Stor Tank i A Rm "AS IS" " ASIS
- 1,700 3,000,000/6,0(X),000 AX010 Spent Resin Transfer Pump Rm " ASIS" "AS IS" 180 1,400,000/5,100,000 AX011 AB Sump Tank Pump / Valve Rm <50 <5,C%/<$0,000 8 3,200/4,900 l AX012 Aux Bldg Sump & Tank Rm <50 <5,000/<50,000 560
! 390,000/68,tKX) l AX0I3 Evap Cond Tanks, Pumps <500 <l.tXX1/<lo,000 $ 130/140 Demins Rm AXO14 RC Evaporator Rm <500 <50,000/<50,000 I8 20,000/8,100 l AX015a Cleanup Filters Rm <500 <50,000/<50,000 I10 10,000/8,700
*The radiological conditions in this tabic reflect rounded-off, average PDMS survey data. N/A - Not Applicable 5.3-4 UPD. - AUGUST 1997
[ TABLb /2 (Cent'd) 'd PDMS RADIOLOGICAL CONDITIONS- AFilB SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENEP.AL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION CURICI .E DOSE RATE <7'/ Overheads DOSE RATE . <7*/ OVERHEADS NUMBER AREA DESCRIPTION (mR/hr) (dem/100 cm2) (mR/hr) (dem//100 cm*) '.
AX015b Cleanup After Filters Rm <500 <50,000/<50,000 38 21,000/13,000 , AX016 Cleanup Demineralizer 2A Rm <500 <50,000/<50,000 38 21,000/13,000 AXO17 Cleanup Demineralizer 2B Rm <500 <50,000/<50,000 110 10/100/8,700 l AX018 Waste Transfer Pumps Rm <500 <50,000/<50,000 10 17,000/14,000 AX019 Waste Disposal Liquid <500 <50,000/<50,000 19 7,800/5,700 Valve Rm AX020 RC Bleed flo! dup Tanks <500 <50,000/<50,000 160 520,000/200,000 IBAICRm i
- AX021 RC Bleed floidup Tank i A Rm <500 <50,000/<50,000 18 1,800/12,000 AXO22 North Stairwell <2.5 < l.000/N/A 0.3 440/N/A p AX023 Elevator Pit and <10 <SJ,000/N/A 14 25,000/N/A .
AX024 Aux Bldg Sump Filters Rm <500 <50,000/<50,000 15 156,400/11,000 AX025 Area Between Service, <500 <l.000/<lo,000 3.5 1,000/310 Control, & RB AX026 SealInjection Filters Rm <500 <50,000/<50,000 12 9,000/1,100 AX027 South Stairwell <2.5 <l.000/N/A 0.2 480/N/A AX101 Radwaste Disp. Cntrl Panet Area <2.5 <1,000/< 10,000 0.2 90/110. . t AX102 RB Sump Pumps Filters Rm <1000 <50,000/<50.tXM) 47 9.300/4,200 l
- AX103 Motor Control Center 2-1IEli Rm <2.5 <l 000/<10,000 0.2 470/480 53-5 UPDATE 2 - AUGUST 1997
TA: LE 5.3-2 (Cont'd) PDMS CADIOLOCICAL CONZlTIONS - AFID SPECIFIC DECONTAMINATION COALS PDMS RADIOLOGICAL CONDITIONS
- GENERALAREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/Oterheads DOSE RATE OVERIIEADS NUMBER AREA DESCRIPTION (mR!hr) (dom!100 cm2) (mR/hr) (dom!!100 cm*)
AX104 Motor Control Center 2-21EB Rm <2.5 <1,000/<10,000 0.2 470/470 AX105 Substation 2-1IE Rm <2.5 <l.000/<10,000 0.2 48W530 AX106 Substation 2-21E Rm <2.5 < l .000/<10,000 0.2 480/530 AX107 Motor Control Center 2-1IEA Rm <2.5 < l,000/< 10,000 O_2 460/530 AX108 Motor Control Center 2-21EA Rm <2.5 < l .000/< 10,000 0.2 480/500 AX109 Nuclear Services Coolers <2.5 <l.000/<10,000 0.2 100/130
& Pumps Area AXl10 Intermediate Coolers Area <2.5 <1,(X)0/<lo,000 0.2 100/130 AX11I Intenned Cooling Pumps <50 <1,tX)o/<10,000 0.7 440/410
& Filters Room AXX 12 Seal Return Coolers & <1000 <50,(XX)/<50,000 99 350,000/38,000 Filter Room AX113 Waste Gas Analyzer Rm <$0 <50,000/<50,000 19 22,000/6,400 AXX 14 MUP Demineralizer I A Rm *AS IS" "AS IS* 73,000 9,R00/34,0tX)
AX115 MUP Demineralizer IB Rm " ASIS" "AS IS* 68,000 31,0(X1/2so,000 AX116 Makeup Tank Rm (500 <50,000/<50,000 60 310.000/23,000 AX117 MUP Filters Rm <1000 "AS IS" 940 330,000,000/2,400 AXl18 Spent Fuel Coolers and Pumps Area <2.5 <l .000/< 10.000 1.I I,000/3,000 AXI19 Spent Fuel Demineralizer Rm <2.5 <1J100/<lo,twXX) .4 480/330 AX120 Spent Fuel Filters Rm <2.5 <l,000/<10,000 0.6 360/1,000 5.3-6 1 li 2 - AUGUST 1997
TABli D). ( MCoat'Q i PDMS RADIOLOCL_/ CONDITIONS - AFH3 \ SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDI110NS* GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION - CUBICLE DOSE RATE <7'/Oterheads DOSE RATE <7'/OVERilEADS NUMBER A.Nf1DESCRIPTidN (mR/hr) (dom /100 cm2) (mR/hr) (doent/100 cm*) AX121 Inside Elevator Cab <2.5 <l.000/N/A 03 250/N/A AX122 North Stairwell <2.5 <l 000/N/A 02 470/N/A + AX123 Access Area (includes AX-136) <2.5 <1,000/<10,000 0.2 160/140 AX124 Concent. Liq. V/ar.e Pump Rm <500 <50,000/<50,000 33 3.300/1,800 AX125 Waste Gas Decay Tank IB Rm <300 <50,000/<50,000 0.2 1,00C/1,000 AX126 Waste Gas Filter Rm <500 <50,000/<50,000 0.2 100/160
'AX127 Waste Gas Decay Tank I A Rm <500 <50,000/<50,000 0.6 6,400/690 AX128 Valm & Instrument Rm <500 <50,000/<50,000 2.7 1,000/1,300 AX129 Deborating Demin. IB Rm <500 <50,000/<50,000 0.3 1,000/600 AX130 Deborating Demin. I A Rm <500 <50,000/<50,000 0.5 540/520 AX131 Misc. Waste 1Ioldup Tank Rm <50 <5,000/<50,000 120 1,900/6,500 AX132 Unit I and Unit 2 Corridor <2.5 <l,000/<10,000 0.2 100/100 AX133 Sonth Staintell <2.5 <l,000/N/A 0.2 500/N/A AX134 Misc. Waste Tank Pumps Rm <50 <50,000/<50,000 13 13,000/45,000 AX135 Radwaste Disposal Control Pnts <2.5 <!,0(X)/<l0,000 0.2 130/120 AX201 North Stairwell <2.5 <1,000/N/A 0.2 450/N/A j AX202 Elevator Shaft <2.5 <1,000/<10,000 0.2 450/480 AX203 4160V Switchgear 2-lE Rm <2.5 <I,000/<10,000 0.2 480/460 5.3-7 UPDATE 2- AUGUST 1997
TADLE 9.3-2 (Cont'd) PD31S RADIOLOGICAL CONDITIONS- AFilB SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/Onrheads DOSE RATE <7'/OVERIIEAD NUMBER AREA DESCRIPTION (mR/hri (dom /100 cm2) (mR/hr) (dom //100 cm')
AX204 4160V Switchgear 2-2E Rm <2.5 <, 'KXV<lo,000 0.2 4RO/4RO AX205 RB Purge Air Sup. and ify <2.5 <l.000/<lo,000 0.7 100/130 CntrlExh Area AX206 RB Purge Air Exhaust Unit B <50 N/A 10 200,000/N/A AX207 RB Purge Air Exhaust Unit A <50 N/A 13 200,000/N/A AX208 Aux Bldg Exhaust Unit B <50 N/A 0.4 3,900/N/A AX209 Aux Bldg Exhaust Unit A <50 N/A 0.7 10,000/N/A AX210 Fuct flandling Bldg Exhaust Unit B <50 N/A 0.9 12,000/N/A AX211 Fuct flandling Bldg Exhaust Unit A <50 N/A U.3 7,200/N/A AX212 Decay Ht Surge Tk & Substation <2.5 <I,0(XV<l o,000 0.2 100/90 Area AX213 Unit Substations & Access Arca <2.5 <l,000/< l0,000 0.2 130/120 AX214 Decon Facility [ Internal Area of Decon <2.5 < t,0txV<lo.000 0.3 150/100 Facility Tanks] <2.5 <l.000/N/A 04 4,400/N/A AX215 Fuelllandling Bldg Supply <2. 5 N/A 0.2 450/N/A AX216 Aux Bldg Supply Unit <2.5 N/A 0.2 450/N/A AX217 Access Area <2.5 <!,000/< 10,000 02 I20070 5.34 1 ~E 2 - AUGUST 1997
f) (j Q TABig (Cent *Q PDMS RADIOLOGICAL CONDITIONS- AFHB SPECIFIC DECONTAMINATION GOAI.S PDMS RADIOLOGICAL COMDITIONS* GENERAL AREA SURFACE CONTAMINATION - GENERAL AREA SqFACE CONTA.MINATION CUBICLE DOSE RATE <7'l Overheads DOSE RATE <7'lOVERH FADS NUMBER AREA DESCRIPTION ' (mR/hr) (dom /190 cm21 (mR/br) (dome //100 cm'l
' AX218 Concentrated Waste Stor. Tank Rm <500 <50,000/<30,000 15 1,900/1,000 AX219 inst. Racks & Atmospheric Monitor <2.5 <I,000/<10,000 0.3 390/5,900 Arca AX220 Caustic Liquids Mixing Area <500 <l,000/<10,000 1.4 440/360 AX221 Caustic Liquids Mixing Area <500 <l,000/<10,000 0.8 450/RRO Corridor AX222 South Stairwell <2.5 <l 000/N/A 0.2 100/N/A AX223 Air flandling Units General Area <2.5 <5,000/<10,000 0.8 490/450 ALOI Elevator Machine Rm - <2.5 <l.000/<10,000 02 100/170 AX302 North Stairwell <2.5 <l,000/<10,000 0.2 480/4w AX303 Elevator and Stairwell Access <2.5 <l,000/<10,000 0.2 510/510 AX304 . Aux. Bldg. Exhaust Fan #8 <2.5 <l.000/<10,000 0.6 750/510 A.X305 FuelIlandling Bldg. Exh. Fan #10 <2.5 <1,000/<10,000 'O2 650/390 4
AX401 Roof <2.5 <1,000/N/A 0.2 90/N/A AX402 Cooling Water Surge Tanks Rm <500 <50,000/<50,000 0.2 110/230 - AX403 Damper Rm <500 <50,000/<50,000 0.2 120/130 AX501 RB Spray Pump 1 A Rm <25 <5,000/<50,000 17 370,000/2,100,000 AX502 RB Spray Pump IB Rm <25 "AS IS" 31 110,000/540,000 t 5.3-9 UPDATE 2 - AUGUST 1997
TAILE 5.3-2 (Cont'd) P;MS RADIOLOCICAL CONZlTIONS - AFIIJ SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/ Overheads DOSE RATE <7'/OVERilEADS NUMBER AREA DESCRIPTION (mR/hr) (dem/100 cm2) (mR/hr) (dom //100 cm*)
AX503 Decay lleat Removal Cooler & <25 <50,000/<50,000 11 43,000/230,000 Pump 1 A Rm AX504 Decay licat Removal Cooler & <2.5 <50,000/<50,000 6.1 15,000/89,000 Pump IB Rm Fii001 Makeup Suction Vaht Rm <500 "AS IS' 19 70,000/89,000 F11002 Access Corridor <2.5 <1,000/<10,000 1.5 1,000/1,500 Fl!003a Makeup Discharge Vaht Rm <1000 <50,000/<100,000 69 I40,000/77,000 FII003b Makeup Discharge Valve Rm <1000 <50,000/< 100,000 220 510,000/I80,000 FII004 Westinghouse Vaht Rm <500 <50,000/<50,000 50 38,000/I 100,000 Fl1005 Mini Decay IIcat Vault <500 <50.000/< 54,000 2.4 2,700/I,500 F11006 Decay iIcat Service Coolers Area <500 < l,000/< 10,000 6.4 1,100/1,400 FII007 Neutrl and Reci Boric Acid <500 < l,000/< 10,000 0.8 ItX1/460 Access Area Fi1008 Neuti. .zer Tanks Pumps Rm <500 <50,000/<50,000 180 22,000/4,700 FIIOO9 Neutralizer Tanks Rm <500 <50,000/<50,000 150 21,000/8,300 FIIO10 Reclaimed Boric Acid Tank Rm <500 <50,000/<50,000 4.3 2,400/6,300 FII011 Reclaimed Boric Acid Pump Rm <500 <50,000/<50,000 9.I 14,000/20,000 FII012 Neutralizer Tanks Filters Rm <500 <50,000/<50,000 31 2,800/1,200 F11013 Oil Drum Storage Area <500 <I,000/<10,000 0.2 100/100 5 3-10 t E 2 - AUGUST 1997
m
/-,'s 1,
TARti (
)'(Cent'O .( ~~
s
\- PDMS RADIOLOGdO[ CONDITIONS- AFH3 ~ '
SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/ Overheads DOSE RATE <7'lOVERHEADS NUMBER AREA DESCRIPTION (mR/hr) (dem/100 cm2) (mR/br) (dem//100 cm*)
FI1014 Annulus <500 <50,000/<50,000 110 35,000n.500 ' Fil10I . MUP Valve Rm <500 <50,000/<50,000 200 850,000/14,000,000 Fil102 East Corridor <2.5 <l,000/<10,000 1.1 200/31,000 , [ Chemistry Sample Shielded <2.5 <!,000/N/A 4 . 380,000/N/A
. Storage Cage]
Fil103 Sample Rm <50 <50,000/<50,000 1.2 4,000/I,600 n"0/ Weu Corridor <2.5 <I,000/<10,000 0.2 120/100 F11ILi IVodel':oom <2.5 <1,000/<lo,000 0.2 100/100
'"~orometer Shielded <2.5 <l.000/N/A 21 22,000/N/A Enclosure] .
Fil106 Monitor Tanks & Sa.nple Sink <2.5 <l 000/<10,000 0.7 330/110 Area Fil107 Trash Compactor Area <2.5 <l,000/<10,000 0.2 100/130 FII108 Truck Bay <2.5 <l.000/<10,000 0.2 70/70 l Fil109 Spent Fuel Pool A "AS IS" "AS IS*/N/A 230 55,000,000/N/A l (Under fuel pool cover) i I Fill 10 SDS Spent Fuel Pool B <2.5 <l,000/ N/A 6.9 180,000/N/A (Under Fuel Pool Cover) Fili 11 Fuel Cask Storage <1000 <1,000/N/A 03 150,000/N/A
- l. (Under FucI Pool Cover) i Filll2 Annulus <100 . <50,000/<50,000 19 3,500/340 Fil201 East Corridor <2.5 <l,000/<10,000 1.0 420/9,400 5.3-1 I UPDATE 2 - AUGUST 1997
TACLE 5.3-2 (Cont'C) PDMS RADIOLOGICALCONDITTONS- AFilB SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SURFACE CONTAMINATION CUBICLE DOSE RATE <7'/ Overheads DOSE RATE <7'/OVERIIEADS NUMBER AREA DESCRIPTION (mR/hr) (dom /100 cm2) (mR/hr) (dem!/100 cm*)
FII202 West Corridor <2.5 <1,000/<10,000 0.2 4R0/490 FIl203 Surge Tank Area <500 <50,000/<50,000 28 1,000/ Inaccessible Fil204 Standby Pressure Control <300 <l.000/<10,000 0.2 1,000/1,000 Area Fi1205 Annulus <100 <50,000/<50,000 8.7 700/8,500 R1301 Upper Spent Fuel Pool A <2.5 < t,000'< 10,000 3.9 240/300 Area (Above Fuel Pool Cover) Fil102 SDS Operating Arca <2.5 <l,000/<10,000 1.2 480/400 FII301 Upper Standby Pressure <2.5 < t,000/<10,000 0.2 300/160 Control Area FH304 Annulus <500 <50,000/<50,000 0.6 2,200/2,200 Fil305 Spent Fuel Pool Access <2.5 <!,000/<10,000 1.3 390/900 Area 5.3-12 1 'E 2 - AUGUST 1997
- p. ,q
, (3
/
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\ l Thj5.3_3 7
(V) PDMS RADIOLOGICAL CONDITIONS - OT11P.R BUILDINGS l SPECIFIC DECONTAMINATION GOALS PDMS RADIOLOGICAL CONDITIONS
- GENERAL AREA SURFACE CONTAMINATION GENERAL AREA SLPJACE CONTAMINATION CURICLE DOSE RAF <T/0wrheads DOSE RATE <TG' 2ir." ADS (dom /Ino em2) (mR%r) #4r ,7*00 cm*1 NUMBER AREA DESCRIPTION (mRT SB000 Senice Building El. 281' <2.5 <l.000'<10.000 0.4 # !310 M-20 Arca <23 <l.000/<10pm 0.3 110/I13 5B002 M-20 Area Sump <2.5 < l.000!N/A 0.4 1tX W N/A SB002 Senice Building EI. 30T <2.5 < l.000/< 10.000 0.2 1107380 SB100
<2.5 <lpW<10.000 0.2 1.300/4pm
[RB Containment Centrol Cubicle Secondary Clem Lab] <2.5 <l.000/<10,000 0.2 1,700/140 SB500 Tendon Access Gallery <23 <l .000/<10.000 0.4 110/110 f TB000 Turbine Building EI. 28l' <2.5 <l.000/<10.000 ol 1C3/i00 pal 98 CACE Building <2.5 <1.000/<10.000 0t ItW100 RA101 PWST Ilouse <2.5 <l.000/<10.000 0.2 120/110 lP%3T Samp] <2.5 <l.000,H/A 0.2 120sN/A RAIO4 BWST Area <2.5 <I,000/N/A 0.3 90/N/A (1) T1e radiological conditions in t'ais table reflect ramded-off, average PDMS sur=cy data. (2) Docs not include surface 5.ontamination on the cork scam. 5.3-13 UPDAlli 2 - AUGUST 1997
O TA111,0 5.3-4 SURFACE CONTAMINATION - REACTOR llUILDING AREA DESCRIPTION PRISCIPI.E ISOTOPES S HIFS Elevations 305' to 347 Cs 137 9.7 E 1 St 90 1.9 E 1 Elevation 347 and Above Cs137 6.7 E 1 St 90 3.3 E 1 Refueling Canal Cs 137 2.8 E 2 St 90 2.2 E 2 D Ring Interior Cs 137 3.2 E 2 Elevation 349'and Above St 90 2.0 E 3 Basement, Elevation 282' Cs 137 6.5 E+2 St 90 5.9 E+2 0 3.3 14 c , _ _ e 11,,7 8
O - t , s i f i , i
- TABLE 5.35 i i l
) SURFACE CONTAMINATION- AFHB
- CUBICLE
!. NUMBER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIES (**) 1 1 1 AX001 RB Emerg. Coolmg Booster Pumps Area C 1.32E-3 ! f AX002 Access Corr 6r B 5.45E-5 AX002a . Na Piping System C 6.67E-6 , 'AX003 Access Arca C 935E-4 AX004 SealInjection Valve Rm B 933E-3 i AX005 MUP Pump IC Rm B 436E-3 t h AX006 MUP Pump 1B Rm A 934E-3 i 8 - ! AX007 MUP Pump IA Rm B 1.00E-3 i 1 AX008 Spent Resin Storage Tank IB Rm B 130E-1 I j- AX009 Spent Resin StorageTank I A Rm B 234E-I j AX010 Seit Resin Transfer Pump Rm B 6 74E-2 ! 1 AX011 Aux Bidg Sump Tank Pumps and Valve Rm B I .18E-4 : i
- AX012 Aux Bldg Sump and Tank Rm B 3.97E-2 [
!- 'AX013 Evap Cond Tanks. Pumps ard Dcmins Rm B 231E-5 L 4 i AX014 RC Evaporator Rm A 2.27E-3 ' j AX015a Cicanup Filters Rm A 3.73E-4 i 4 f
- 53-15 UPDATE 2 - AUGUST 1997
, i 1 i
TABLE 5_3-5 (Conrd) SURFACE CONTAMINATION - AFIIB CUBICLE NUMBER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) . CURIES (") AX015b Cicanup After Filters Rm A 7.71 E-4 AX016 Cicanup Dcmineralizer 2A Rm A 1.23E-3 AX017 Cleanup Dcmineralizer 2B Rm A 6.03 E-4 AX018 Waste Transfer Pumps Rm B 1.06E-3 AX019 Waste Disposal Liquid Vahr Rm A 7.W E-4 AX020 RC Biced Iloidup Tanks liland IC Rm A 3 05E-AX021 RC Bleed floidup Tank 1 A Rm B 7.97E-4 AX022 North Stairwril B 2.80E-5 AX023 Elevator Pit and Associated Equipment B l .01 E-3 AX024 Aux Bidg Sump Filters Rm B . 78E-4 AX025 Area Between Service, Control, and RB B I .73 E-4 AX026 Seal Injection Fdters Rm C 1.40E-4 AX027 South Stairwc!! B 1.71 E-5 AXiOI Radwaste Disposal Control Panel Area B 1.96E-5 AX102 RB Sump Pumps Filters Rm B 2.6 t E-4 AX103 Motor Control Center 2-11EB Rm C 2.RRE-5 5 3-16 UPDATE 2 - AUGUST 1997 O O O
, . _ _. .. . _ _ -. . . __ _ ._=
, u/
a [. TABLE 5.3-5 (Cont'd) SURFACE CONTAMINATION- AFHB , CUBKI.E I. NUMBER . AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIES (") , A%104 Motor Control Center 2-21EB Rm B 3.03E-5 AX105 Substatxm 2-IIE Rm B 7.10E-5 AX106 Substation 2-21E Rm B 3.04E-5 AX107 Motor Control Center 2-IlEA Rm B 8.94E-5
- j. AX108 Motor Controt Center 2-21EA Rm A 6.46E-5 AX109 Nucicar Services Coolers and Pumps Arca B 3.42E-5
- AX110 Intermediate Coolers Area C 3.54E-5 l AX11I Intermed Cooling Pumps and Filters Rm B 7.16E-5 AXl12 ' Seal Return Coolers and Filter Rm B 2.43 E-2 AXII3 Waste Gas Analyzer Rm B 2.97E-3 AX114 MUP L. w. lizer I A Rm B 5.99E-4 AX115 MUP Dcmineralizer iB Rm B 2.7EE-3 AX116 MakeupTank Rm A 2.15E-2 AXl17 MUP Filters Rm C 2.5NE+ 1 i AX118 Spent Fuel Coolers and Pumpt Area C 2.89E-4 1
5.3-17 UPDATE 2- AUGUST 1997
i TABLE S.3-5 (Cont'd) SURFACE CONTAMINATION- AFIIB CUBICLE NUMBER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIES (") AX119 Spent Fuel Dcmineralizer Rm B 2.5iE-5 AX12(, Spent Fuci Filters Rm A 9.33E-6 AX121 Inside Elevator Cab B 1.14E-5 A f122 North Stairwell B 2.95E-5 AX123 Access Area (mcludes AX-136 Ilot Tool Room) B 9.45E-5 AX124 Concentrated Liquid Waste Pump Rm B 1 # E-4 AX125 Waste Gas Decay Tank IB Rm B 1.92E-4 AX126 Waste Gas Filter Rm B 4_75E-6 AX127 Waste Gas DecayTank I A Rm B 1.17E-3 AX128 Valve and Instrument Rm B R 02E-5 AX129 Deborating Demineralizer iB Rm B 5 80E-5 AX130 Deboratmg D mineralizer IA Pan B 3.21E-5 AX131 Miscellaneous Waste lloidup Tank Rm B 2.64E-4 AXI32 Corridor Betmen Unit I and Unit 2 B l 47E-4 AX133 South Stairwc!! B 2.41 E-5 AX134 Miscellaneous Waste Tank Pumps Rm A 1.3RE-3 5 3-18 UPDATE 2 - AUGUST 1097 O 9 9
4 3 s s i r- , TABLE 5.3-5 (Cent'd) . } SURFACE CONTAMINATION- AFHB [
. t CUBICLE !
- . NUMitER AREA DESCRIPTION PRINCIPAL ISOTOPES (*)' CURIES (")'
!l l AX135 Radwaste Disposal Control Pancis B 4.61E-6 ,
- AX201 North Stairwell B 2.54E-5 i i.
AX202 Elevator Shaft B 2.16E-5 : 1
. AX203 4160V Switchgear 2-lE Rm B 1.05E-4 !
, AX204 4160V Switchgear2-2E Rm B I.09E-4
- AX205 RB Purge Air Sup. and Hy Ctrl Exh Area A 3.95E-5 [
AX206 RC Purge Air Exhaust Unit B B 1.41E-2
. AX207 kB Purgc Air Exhaust Umt A B I.56E-2 ;
) AX208 Aux Bids Exhaust Unit B B 2.42E-4 i AX209 Aux Bldg Exhaust Unit A B 6.24E-4 j AX210 Fuci Handling Bldg Exhaust Unit B B 6.ME-4 ] j AX21I Fuelllandling Bldg Exhaust Unit A B 3.56E-4 ! i
- j. ; AX212 Decay Heat Surge H and Substation Arca B 633E-5 i AX213 Unit Substations and Access Arca C 8.. l lE-5 ;
AX214 Decon Facility C 1.68E-4 j l AX215 Fuci Handling Bldg Supply Unit C 2.80E-5 1 i 4 s 5 3-19 UPDATE 2- AUGUST 1997 1 t 4 4
TABLE 5.3-5 (Cont'd) SURFACE CONTA511 NATION - AFIIB CUBICLE NU\fRER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIEL(**) AX216 Aux Bldg Supply Unit B 33(E-5 AX217 Access Area B 6.94E-5 AX218 Concentrated Waste Storage Tank Rm B l .44E-4 AX219 Inst Racks and Atmospheric Monitor Area B 639E-5 AX220 Caustic Liquids Mixing Area B 4.2SE-5 M 221 Caustic Liquids Mixing Area Corridor B l .12E-4 AX222 South Stairm!! B 4 89E-6 AX223 Air 11andling Units General Area C 3.79E-4 AX301 Elevator Machine Rm C 5.67E4 AX302 North Stairwed B 2.40E-5 AX303 Elevator and Stairull Access C 4.83 E-5 AX304 Auxiliary Building Exhaust Fan #8 C 2.22 E-5 AX305 Fuel llandling Building Exhaust Fan #10 A 1.94E-5 AX401 Roof A 1.54 E-4 AX402 Cooling Water Surge Tanks Rm C 1.55E-5 AX403 Damper Rm B 1.26E-5 5 3-20 UPDATE 2 - AUGUST 1997 9 O O
m k TABLE S.3-5 (Cent'd) SURFACE CONTAMINATION- AFHB ! t CUBICLE NUMBER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIES (") I AX5'11 RB Spray Pump 1 A Rm A 2.98E-2 AX502 RB Spray Pump 1B Rm B 3.56E-3 ft AX503 - Decay licat Remov Cooler and Pump I A Rm .A 3.08E-3 AX504 Decay Ileat Remov Cooler and Pump IB Rm A 2.67E-3 FII001 Makeup Suction Valve Rm C 1.57E-2 ' RID 02 Access Corridor C 2.17E-4 F11003a Makeup Discharge Valve Rm B 3.45 E-3 + , F11003b Makeup Discharge Valve Rm B 1.75E-2 FII004 Westinghouse Valve Rm C 6.66E-3 ; ' ~ FIIDOS Mini Decay IIcat Vault B 3.61 E-5 . F110 % Decay licat Service Coolers Area B 4.4RE-4 FII007 Neutri and Rect Boric Acid Access Area B 430E-5 r
- Fil008 NeutralizerTanks Pumps Rm B 2.17E-3
! FII009 NeutralizerTanks Rm B 2.94E-3 [ i t F11010 Reclaimed Boric Acid Tank Rm A I .97E-4 ! i ! l i. j i 5.3-21 UPDATE 2 - AUGUST 1997 i i ! [
TABLE S.3-5 (Cont'd) SURFACE CONTAMINATION- AFIIB CUBICLE NUMBER AREA DESCRIPTION PRINCIPAL ISOTOPES (*) CURIES (") FII011 Reclaimed Boric Acid Pump Rm A 9.16E-4 Fii012 Neutra!izer Tanks Filters Rm B 4.29E-5 Fl!013 Oil Drum Storage Arca B 4.49E-6 F11014 Annulus A 5.22E-3 F11101 MUP Valve Rm B I.14E-1 Fil102 East Corridor B 5.67E-4 FIIIO3 Sample Rm B 2.40E-4 Fil104 West Corridor C 3.20E-5
~
Fil105 Model Rm B 4.38E-4 Fil106 Monitor Tanks and Sample Sink Area C 7.03E-5 Fil107 Trash Compactor Area B 5.62E-6 Fil108 Truck Bay A 5.46E-5 Fil109 Spent FucI Pool A ' C 1.35E+2 FIII10 SDS Spent Fuct Pool ' C 4.62E-2 Fill 1I Fuct Cask Storage ' C l .14E-2 Fill 12 Anmalus B 4.39E-4 FII201 East Corridor B 1.51 E-4 Fil202 West Corridor B 1.07E-4 (*) .A metal cover with an access door was placed owr these arcas to prevent spread of contamination. 5.3-22 U 2 - AUGUST 1997
m y\ n 1 ( ) (v' ) & ~J TABLE S 3-5 (Cont'd) l SURFACE CONTAMINATION- AFIIB CUBICLE NUMBER AREA DESCRil' TION PRINCIPAL ISOTOPES (*) CURIES (") B 4.35 E-5 FII203 SurgeTank Area C 2.36E-4 FII204 Standby Pressure Control Area B 7.24E-5 FII205 Annulus C 5.30E-5 FII301 Upper Spent Fuel Pool A Area B I .53E-4 FII302 SDS Operating Area Upper Standby Pressure Control Area C 1.92E-4 l FII303 B 6.05E-4 Fil304 Annulus C 2.65 E-4 FII305 Spent Fuel Pool Access Area NOTES: (*) The principal isotopes and their relative distnbution are dermed below. The Sr-90 value sw.w4s the sum of the Sr-90 and Y-90 isotopes which are in equilibrium; the Cs-137 value represents the sum of the Cs-137 and Ba-137m isotopes which are in equihbnum. The "A", ~B", and "C" categories relate to normal, makeup, and defueling waste streams, revectively. Only those isotopes important from an offsite dose perspective are included. A B C Sr-90 0.08 Sr-90 0.29 Sr 00 0.63 Cs-137 0.92 Cs-137 0.71 Cs-137 0.28 Pu-238 4.43E-6 Pu-238 167E-5 Pu-238 4 25E-4 Pu-239 5.39E-5 Pu-239 2.04E-4 Pu-239 5 18E-3 Pu-240 1.43E-5 Pu-240 5.41 E-5 Pu-240 1.37E-3 Pu-24 I 4.86E-4 Pu-24I I 84E-3 Pu-24I 0.04 Am-241 1.56E-5 Am-141 5.92E-5 Am-141 1.50E-3 Pm-I47 0.04 (**) Ecsc are calculated values based on the specific decontamination values given on Table 5.3-2. 5.3-23 UPDATE 2- AUGUST 1997
TABII5.34 CUBICLE SURFACE CONTAMINATION - OTilER BUILDINGS NU%fBER AREA DFSCRIPTION CU RIESt **) PRINCIPAL TSOTOPES(*) SB000 Scrsice Building El. 221' A I.3 tE4 SB002 M-20 Area A I.16E4 SD002 M 20 Arca Semp A I (ME-5 SB100 Semce Building El. 305' A 7.23 E-5 [RB Containment Cor: trol Cubicle] B 5.72E-5 (Secondary Clem Lab] B 3 8 tE-4 SB500 Tendon Access Gallcry A 1.22E4 TB000 Turbine Building El. 2R1* A 591E4 PA108 CACE Building B 2.69E-5 RA101 PWST Pump flouse A 2.72E-5 [P%3T Sumpi A 2.53E-6 RA104 B%TT Arca A 2.92E-5 NOTES (*) The principal isotopes and tleir relathe distribution are defined below. The Sr-Whalue represcrds the sum ofIhe St-90 and Y-90 botors which .re in equihbrium, the Cs-137 value represents the sum of t!e Cs-137 and Ba-137m isetepes ahich are in equilibrmm. Tlr *A* and "B" categorb relate lo normal and make up waste streams, respecthcly Only those isorges important from an ofTsite dose persgttnt are included. A B Sr-90 0.08 Sr-90 0.29 Cs-137 0 92 Cs-137 0 71 Pu-23M 4 43E-6 Pu-238 167E-5 Pu-239 5.39E-5 Pu-239 2.04E-4 Pu-240 1.43E-5 151-240 5 4IE-5 Pu-241 4 86E-4 Pu-241 184E-3 Am-24: 1.56E-5 Am-141 5.92E-5 ('*) These are calculated values based on t!e spxific decontamination values ghtn on Tabic 5.3-3. 5.3-24 UPDATE 2 - AUGUST 1997 9 9 9
< l APPENDIX 5A l POTENTIAL REDUCTIONS IN OCCUPATIONAL EXPOSURE DUE TO POST DEFUELING MONITORED STORAGE 4
[ NOTE: ne following is a historical treatise that discusses the results of the Decontammation Task Force Report (DTFR)(Reference 2) conducted in 1985. De DTTR prosides an evaluation of the reduction in occupational exposure attributed to PDMS. The actual radiological conditions existing m December 1993, i.e., at the time of entry into PDMS, as given in Section 5, have a negligible impact on , the conclusions reached by this study. This Appendix will remain in its current condition to maintain j that historical perspective. No further attempt will be made to be consistent with the rest of the SAR as I l it is revised.] 5A.I
SUMMARY
ANDINTRODUCTION - 5A.I.]
SUMMARY
A comprehensive evaluation of the person-rem associated with additional required decontamination of the TMI 2 plant aAct completion of the " Cleanup Program" (i e., Phase 111 Endpoint) has determined that defening this decontamination for a period of 30 years will result in a potential occupational exposure savings in the range of 4,500 to 9,800 person rem. In calculating this savings, person tem resulting from decontammation tasks, radioactive waste processing tasks, and exposuret due to tasks performed during PDMS were estimated. Table SA 10 contains a summary of the estunated range of person rem for both 7 immediate additional decontammation and fmal decontammation as a part of decommissioning, assummg a i 30 year period of PDMS. 5A.I.2 INTRODUCTION Following the completion of the prerequisites identified in Section 1.1.2.1 of the PDMS SAR, the TMI 2 plant will enter PDMS. To enter this mode, the TMI 2 plant must be in a safe, stable condition so that it does not pose a risk to the health or safety of the public. GPU Nuclear has determined that by deferring any remaining decontanunation until after PDMS, a significant savings in the occupational radiation exposures will be realized. He difference in occupational exposure as a result of performing these tasks after a period of PDMS (final decontamination as a part of decommissioning) instead of at the end of Phase 111 (immediate additional decontamination) is based on: o Reduction in radiation dose rates due to the natural decay of radioactive materials remaining in the plant; o Advances in remote technology directly applicable to fmal cleanup actisities; o Advances in chernical decontanunation methods; and o A longer development penod to plan, engineer and, in some cases, perform further decontammation activities, i nis report quantifies the potential savings in occupational exposure which might be realized due te delaying the fmal decontamination until af er the PDMS period. He evaluation performed considers two cases:
"immediate additional decontammation" (i.e., immediately after Phase 111) and "fmal decontammation as a part of decommissioning" after PDMS; the delay is assumed to be approximately 30 years. Neither case calculates person rem exposures for normal decommissioning activities. 'nds report describes the plant corxhtions assumed to exist both before and sfler decontammation actisities. Dese conditions form the 5, A - 1 UPDATE 2 AUGUST 1997'
basis for the dose rates used in the person t em evaluati,c. De basis for the job-hour estimates used in the analysis are also provided in this report. TVs study i' considered a reasonable estimate of the potential person rem savings associated with PDMS ba# .4 available information. In addition to the two cases listed above, person-rem are also evaluated assuming final decontanunation as a part of decommissioning after a PDMS period of 20 years. This case is included to facilitate comparison with the person rem stated in Supplement 3 of the Programmati: Emironmental impact Statement (Reference 1), which assumes a PDMS duration of 20 years followed by 4 years of fmal decontammation. It is not the plan of GPU Nuclear to have a PDMS period of 20 years nor has GPU Nuclear specified a duration for completmg the decontamination activities, except that PDMS shali not continue beyond the time of decommissioning TMI 1. His study should not be interpreted to imply e. commitment by GPU Nuclear to employ specific decontamination techniques, not to achieve specific endpoint radiological conditions during fmal decontammation. 5A.2 DACKGROUNDINFORMATION 5A.2.1 RECOVERY TASKS Decontamination and dose reduction aethities have been conducted continuously during the TMI 2 cleanup period. Major tasks that have been performed are: o Decontamination Task Force Report (Reference 2) o Gross Decot < anination Experiment (Reference 12) o Dose Reduction Working Group Tasks o Auxiliary and Fuel liandling Duildings (AFIIB) decontanunation o Reactor Duilding (RB) decontamination in general, these tasks have been geared to ensunng that the recovery tasks, including stabilizing the plant and defueling, were performed in a radiological emironment that ensured occupational exposures were ALARA. During the cleanup period, the priority tasks were those associated with remosing the nuclear fuel from the plant. Dese tasks were important as they reduced the risk to public health and safey. SA 2.1.1 Decontamination Task Force Report ne Decontamination Task Force Report (DTFR) was prepared by a jomt group of TMI 2 organization representatives. It is the most comprehensive, as well as the most current, miev of the effort required to decontaminate TMI 2. The report was prepared by a task force formed in 1985 to evaluate the problems and aethities associated with achieving Phase 111 radiological completion criteria by mid 1988. He objective of the Decontanunation Task Force was to arrive at a consensus technical approach to each of the major areas ofdecontamination work identified by the Task Force. De ten major areas evaluated by the Task Force were: o Remote equipment development o Sludge transfer and disposal o D-nng dose reduction and decontamination
- 5. A - 2 UPDATE 2 AUGUST 1997
m o Reactor Building basement recovery o AFIID surface decontamination o Non-RCS systems decontamination o Reactor Building ilVAC modifications o Reactor Building Phase 111 surface decontamination o Reactor Coolant System decontamination o Reactor Buildmg Phase III decontammation waste management ne Task Force l,ased its evaluations, technical approaches and schedules on available technical plans. During the evaluation, several critical areas were identified including remote technology development and Reactor Buildmg basement cleanup. Dese remain entical areas as the end of Phase 111 approaches. SA.2.2 FUTURE DECONTAMINATION TASKS Tasks performed for unmediate decontammation (Post Phase ill decontamination) or for fmal p decontamination as a part of decommissioning (Post PDMS decontanunation) differ from cleanup tasks in that the priority will have changed from removing fuel in order to reduce the risk to the health and safety of i the public, to rernoving radioactivity from the plant in order to provide a radiologically improved emironment for future plant work. As the maximum annual oft-site dose from routine releases during PDMS is projected to be 0.02 mrem /yr (PDMS S AR Table 8.1-5), the PDMS configuration presents neghgible risk to the health and safety of the public (see Section SA.5). Future work, thercfore, will be planned to maintain worker radiation exposures ALARA. SA.3 CRITERIA FOR PERFORMING TIIE EVALUATION To perform the person rem study in a thorough and efficient manner, the following cnteria were established for the evaluation: o Task evaluations should reflect GPU Nuclear's current plans. o Person-tem estimates should be developed for the two cases such that the calculated difference (i.e., the savings)is representative. C k))- v
- 5. A - 3 UPDATE 2 AUGUST 1997
o The expected plant conditions at the end of Phase Ill should be projected, based on cunent progress and on going cleanup actisities. o No time dependence is assumed to account for the actual schedule for perfonning the work. o The decontamination techniques to be employed are assumed to be those emisioned by the DTFR. o hejob-hours spent in radiation areas should be estimated from histoncal data for totaljob-hours and time spent in radiation areas. 5A.4 RADIOLOGICAL CONDITIONS SA.4.1 CONDITIONS AT11!E END OF PilASE Ill Table SA 1 summanzes the Phase !!! endpoint dose rate entena. Some of these enteria have already been met or exceeded Chapter 5 of the PDhtS SAR desenbes the rasological con 6tions at the end of Phase !!!, which are the conations that will be in effect at the beginning of PDhfS. In general, the plant can be categorized into three areas: o Auxiliary and Fuel llandliig Duildings. Some cubicles are locked because of high radiation levels, however, general area dose rates throughout the buildmgs are relatively low and permit personnel access. hiany areas, includmg halhvays and access corridors, allow essentially unlmuted access. Surface contamination, particularly in the overhead areas, is relatively high in Th112. liowever, the surface contanunation present does not normally impact the general area dose rate (see Table SA 3 and Table 5.3 2 of the PDhtS S AR); o Reactor Duilding Entry and Operating Levels. General area dose rates allow access to most areas, and pernut reasonably long stay times (see Table SA 2 and Table 5.3 1 of the PDh1S S AR) Based on available data, dose rates on the 305' elevation are generally less than 150 mrerWhr. On the 34T elevation dose rates are corally less than 50 mrem /hr; and o Reactor Duildmg Basement and D-Rings. Radiological con &tions in these areas severely limit accessibihty and stay times. He average dose rates in the basement are in excess of 10 R/hr. The "D" D-ring dose rates average approximately 1 R/hr and those in the "A" D ring approximately 500 rnierdhr, These dose rates are based on surveys and other available documentation (e g., References 3 & 4). SA.4.2 CONDITIONS DURING PDMS It is planned that PDMS will go into effect at the completion of Phase III, therefore the radiological con &tions in the plant at the beginning of PDMS will be the same as those at the end of Phase Ill. Dunng PDMS the dose rates in the plant will decrease due to natural decay of the radionuclides. SA.4.3 CONDITIONS AT TIIE END OF PDMS The dose rates for fmal decontamination as part of decommissioning will be lower than those for imme6 ate additional decontammation due to the decay of the radionuclides, After a 30-year period of PDMS, dose rates were assumed to be half of those existing at the end of Phase 111. His accounts for the decay of 5.A-4 UPDATE 2 AUGUST 1997
l domirtant dose rate contributor (i.e., cesium 137), which has a half life of 30 years. ( 5A 4 4 HASE CASE Pl. ANT CONDITIONS he dose rate enteria assumed as endpoint Boats for the fmal decontamination actnities are the dose rates l postulated for a typical pressurized water reactor dunna shutdown. His condition is refened to as the " base case" plant combtion. Radiological data frorn six Babcock & Wilcox plants: Oconec Units 1,2 and 3, Arkansas Nuclear One Unit 1, Rancho Seco and TMI Unit 1, were used to establish the base case plant condition (Reference 5) TMI l dose rates were obtained from TMI l surveys, and data for the other plants were obtained from the llrown Boven Reactor (IlllR) GMilli Report titled " Evaluation of the Dose Rate Data of Various Nuclear Power Plants with ha:.. Nuclear Steam Supply Systems"(Reference 6). His report lists both the TM1 1 survey data and the common mean of the radiation levels at the six plants. The common mean from all six Dabcock & Wilcox plants is used when it is available. Ilowever, for most areas in the AFilB the data in the BBR Report lacked sufficient detail, therefore, the TMI l survey data fomis the base case criteria for these areas. Tables $A 2 and $A 3 list the base case dose rates for the Reactor 1 11uilding and AFilB respectively, with the Phase 111 endpoint dose rates listed for cornpanson. It can be noted, from these tables, that: o Dose rates in the AFilB, including conidors and cubicles, are, in general, comparable in both cases. o Conditions in the T'dl.2 Reactor 11uildmg are not comparable to the base case plant. Dose rates throughout the Reactor Buildmg, especially in the D rings and basement, are much higher at TMI 2. T'N Hus, it can be concluded tlut future decontamination tasks will be concentrated in the Reactor lluilding, (%./ ) with a smaller cfTort required in the AFilB.
$A 4.5 ISOTOPlc CllARACTERIZATION Radioactive source terms expected to remain at the TMl 2 plant are discussed in Chapter 8 of the PDMS SAR. The predominant dose contnbutors are the gamma emitters. This is due to the penetrating nature of gamma radiation, as compared to beta radiation which is non penetrating and easily shielded Beta dose rates are nonnally not the controlling radiological concem for plant personnel. Radionuclides which decay by way of beta or alpha particle emissions are prinarily a concern for internal uptake from inhalation of airborne radioactivity. Experience during the recovery period has shown that occupational internal exposures are small compared to the external A
U
- 5. A - 5 UPDATE 2 AUGUST 1997
l gamrna aposures. Therefore the analysis to assess person rern savings associated with PDMS has focused on ducct ganuna occupational esposures. The ganuna enutter tlat will be the predominant dose contributor 9 i l l is Cs.137, which has a 30 year half hfe. Thus, dose rates are expected to decrease to $0 percent of their i origmal values during the 30 year period of PDMS assurned in this evaluation
'the areas of highest radioactisity are the basement and the D rings. Tasks required to decontaminate these areas are expected to consume a significant number of person rem The major contributor m the basement is i the block wall wiuch has a current Cs 137 activity of approximately 19,000 curies The total radioseth1ty in the D rmgs is estmtated to include approximately 17,000 cunes of Cs 137. TWe SA-4 lists the cesiurn and strontium actmtics estirnated for these areas.
5A.$ OPF SITE DOSES DURING PDMS Table 8.13 of the PDMS SAR hsts the estinuted maximum off site doses from both routme and accidental releases daring PDMS. The annual doses from routine releases are based on airbome and liquid releases frorn the Reactor fluildmg. The maximum doses, which are from the airbome source, are 2 x 10' rnremlyr to the bone and I x 10' mrem /yr total body. Annual doses from contarnination in the AFlH3 are expected to be negligible, when compared to those from the Reactor Buildmg, due to the lower radioactivity levels in the AFini for both the liquid and airborne p"athways. The estimated aruiual 4off site doses from airborne adioactivity from the AFill) are 1 x 10 mrem'yr to the bone and 3 x 10 nuemlyr total body (Reference 7) The rout'ne release annual off site doses are within the PDMS adnunistrative linuts listed in Table 8.1-4 of the PDMS SAR and, therefore, their reduction will not be the focus of future work. 5A 6 DOSE RATE ESTIMATION Dose rates estmtated for immediate additional decontamination are based on current data and the dose rate criteria for the end of Phase 111. The dose rates for performing the fmal decontamination as a part of decommissiorung are assumed to be 50 percent of those for immediate additional decontamination. This results from the decay of Cs 137, the predominant dose contnbutor, over the assumed PDMS period of 30 years. 5A 61 DECONTAMINATION TASKS The dose rates used in the person-rem analysis, which are listed in Table SA 5, are estimated considenng the followmg: o Phase III endpomts (see Table SA 1), o Average dose rates for 1987, from Exposure Tracking Nurnbers (Table SA 8 lists the applicable E1Ks); o Planned Robotic use; o Decrease in dose rates as decontamination work proceeds; and o Ttme spent in radiation areas with lower dose rates than the area being decontaminated. No dose rates are listed for health physics support as previous experience shows that 20 percent of the total 5.A 6 UPDATE 2 AUGUST 1997
perwerem expecded is attributable to health physics. He person-rem for health physics support in this evaluation will, therefore, be bued on this criterion. i SA 6.1.1 Tuks Performed Robotically Due to the high general area radiation dose rates present in all areas of the Reactor Building buement and in some areas of the D rings, it is assumed that many of the tasks in these areas will require remote or scrni remote operations. Remote tasks to be performed in the basement include removal of the block wall, general cleanup acthities and cubicle cleanup. De use of robotics for cleanup of the D rings wi!! be limited , by physical accessibihty. It is assumed that all robotic operations will be remotely controlled from an operations center outside the Reactor Building. Although robotic operations will be performed remotely, ; personnel exposures will be incurred for robotic support tasks, including robot deployment and retrieval, robotic maintenance, repair and decontanunation.
! .To estimate the personnel exposure associated with robotic tasks, radiation work permit (RWP) hours and person rem data for previous robotic activities, which are tracked on an exposure tracking number (ETN) data base, were used. As a significant amount of work utilizing robotics was completed in 1987, data for this year were used to develop an average exposure estimate. Based on data obtained from the ETN data base, the average exposure rate associated with robotic activities was 64 mrem'hr, Tuks that have been accomplished usir ; robots include basement charucterization, including sideo and ;
radiological examinations; buement fic >r and wall concrete bore samples; basement wall and overhead gross flushing; and basement floor desl dging. ]( 5A.6.1.2 Example of Estimating Dose Rates As an example of how the dose rates used in this evaluation are estimated, the approach used for estimating the basement dose rates is considered representative. The estimated average dose rate for immediate additional decontammation of the basement, excluding the block wall, is 100 mrenVhr. His is based on: o Current dose rates for placement and mcintenarre tasks associated with the use of robots - 64 mrenVhr (from ETN C91K001); , t o Limited personnel access into the basement after dose rates decrease to approximately 10 R/hr (Reference 8); o increased time spent in the basement as the dose rates approach the base case criteria of approximately 20 mrem /hr; and o Time spent in areas oflower dose rates. It is auumed that most of the decontamination will be performed robotically until the dose rates have docrened sufficiently to allow fairly long personnel entries into the basement.
$A.6.2 RADIOACTIVE WASTE PROCESSING The dose rate for radioactive waste processing is based on the dose per cubic foot of waste processed during 1987, adjusted to account far the fact that waste from highly radioactive areas, such as the Reactor Building p basement, will be processed. His waste has higher activities than those in areas of the Reactor Building or AFHB from which waste is presently processed. The dose rate for final decontamination as a part of 5.A-7 l'PDATE 2 AUGUST 1997
- .- -, . . .- - . - . . . - - - . _ -. . - ~.~ - - -.
decornmissioning is assumed to be half the dose rate for immediate additional decontanunation because of the 30 year half life of Cs 137. It is assumed that the volume of radioactive waste that will be processed is in the range of 120,000 183,000 cubic feet This volume is within 20 percent of the total volume shipped through 1987. 5A.7 ACTMTIES EVALUA1ED Re activities that are considered in this person-rem analysis include final decontamination tasks, radioactive waste rnanagement tasks associated with the decontamination, and actiuties dunng PDhtS. 5A.7.1 FUTURE DECONTAhi! NATION TASKS De " future decontanunation tasks" are defined as those tasks which are required to improve radiological conditions from the Phase !!! endpoint criteria to the base case level. The actual work required to accomplish the fmal cleanup, listed in Table SA-6, was estimated based on the following sources: o Decontanunation Task Force Report o Decontamination Approaches o Decontamination Technical Specifications' o Thil 2 Reactor 13uilding Blockwall Report and Task Force Report (References 9 and 10) nese future tasks will also include processing the radwaste produced by the decontamination 5A 7.2 RADIDACTIVE WASTE MANAGEMENT TASKS Radioactive waste will be generated dunng the decontammation of TMI 2. Radwaste tasks will include removing this waste from the Reactor i uilding or AFIID; processing the waste (e g , compacting), and shipping it to the appropnate disposal facility. 5 A.7.3 ACTIVITIES DURING PDMS Work performed in radiation areas during PDMS includes maintenance in th Reactor Buildmg and in the AFilD, surveys and inspections in these facilities and radioactive waste processing. 5A.8 J0D-110UR ESTIMATES Re rnost comprehensive study pertaming to the work scope and approach for decontammation actisities is the DTFR which defmes decontamination tasks necessary to achieve the Phase 111 endpoint criteria, along with the associated job-hours required to complete these tasks. The information presented in Reference 2, categorized by Reactor Buildmg and AFliB actinties, assumes 3 years to complete the decontammation tasks. He DTFR is used as the basis for the job-hour estimates for this present evaluation as many of the tasks The Decontamination Technical Specifications are not associated with the operating license. They are reports discussing decontanunstion techniques. 5.A - 8 UPDATE 2 AUGUST 1997
Q specified in the report will also be required for future decontamination. Ilowever, the total number of hours required to reach the base case enteria from cunent or Phase 111 endpoint conditions will differ from those listed in the DTFR, and no time dependence is assumed for reducing the dose rates to the base case criteria. l In addition, the hours developed in the report were irutially used for budget purposes and hence were billable hours (i c., total hours committed) rather than an estimate of the actual hours spent in a ra&ation field. Herefore, the job-hour estimates in the DTFR were modified for this evaluation. This modification considers: o ne difference in scope between the DTFR estimates and the Post Phase til required aditional decontanunation; and o ne fraction of the totaljob-hours that will be spent in radiation areas. Table SA 7 lists the estimated job-hours that will be spent in radiation fields, using the categories in the DTFR, with the DTFR totaljob-hours included for comparison. He radiation areajob-hours are given for three cases: 1) irnmediate additional decontamination following Phase ill: 2) fmal decontanunation as a part of decomtnissioning following 20 years of PDMS (the Reference 1 base case), and 3) fmal decontarmnation as a part of decommissioning following 30 years of PDMS (the base case assumed in this SAR) These categories were expanded for this evaluation, as shown in Table SA 5. Hejob hours assumed for each of these categories are also listed. Previous history has shown that person rem attributable to health physics support is a percentage of total person rem, therefore nojob-hours are listed for this task. 5A.81 SCOPE MODIFICATIONS O Identified modifications to the work scope are used to estimate the change in total job-hours based on a change in the desired endpoint of the tasks or a change in the decontammation program starting point. 5A.8.1.1 Immediate Additional Decontammation in all areas of the Reactor Buildmg the Phase ill endpoint dose rates are higher than the base case plant dose rates, therefore the scope is increased to account for the lower dose rate goals of the final decontamination activities. Modification is also made to tasks that require fewer job-hours because the task is partially completed or a reassessment of the job-hours required has been made. In general, most areas of the AFIIB are close to the base case enteria. Therefore, the originaljob-hour estunate is reduced to show that only a fraction of the work outhned by the Task Force will remain at the end of Phase 111. Generally, the modifications tojob-hours are determmed as followr o The job-hours increase if the base case is lower than the Phase 111 endpoints assumed in the DTFR and decontamination has not reduced dose rates significantly, it also increases if the current Phase ill endpoint is higher than the endpoint assumed in the DTFR and decontammation has not reduced it to below the endpoint. An example of an increase in scope is the D-rings where the general area dose rate criteria has been decreased, but, because access is difficult, little decontammation has been done in the lower sections of the "B" D-ring. In addition, the average dose rates in the "A" D nng are also considerably above the dose rate enteria, o ne job-hours are unchanged if the work already done has reduced the present dose rate to a level where an equivalent amount of work is required to attain the base case criteria as that originally assumed to reach the Phase Ill endpoints. For exarnple, it is assumed that the scope for desludging ( 5.A-9 UPDATE 2 AUGUST 1997 l
previously unaccessed areas of the Reactor Building basement is the same as for the Phase 111 destudging tasks. o ne job-hours are decreased if decorvamination has reduced the dose rates to close to the base case enteria, or to a point where less wo.-k is expected to obtain these criteria. A decrease in scope is seen primarily in the AFIIB where n est areas, except for some cubicles, are at or close to the base case dose rates. Every task in the AlllB is assumed to require fewer job-hours than listed in the DTFR. In addition to these general assumptions for job-hour modifications the following should be noted. o The DTFR includcd job-hours for possible modification to the RB HVAC systems to establish isolation zones to prevent re contanunation. It has been determined, based on work experience, that modification to the IWAC system will not be required. o Robotics. The job-hours in the DTFR included robotic development. This task will not involve time in a radiation area, therefore the hours are not included in this evaluation 5A.8.1.2 Final Decontamination As A Part of Decommissioning ne required job-hours for each task are generally assurned to be lower for decontamination deferred until after PDMS, his accounts for the decay that will occur during PDMS. thus reducing the job hours required to meet the base case dose rates. 5A.8.1.3 Example of Estimating Scope Modification As for the previous dose rate cumple (Section 5A.6.1.2), Reactor Building basement recoveryjob-hours will be used to demonstrate the job-hour modifications, both in scope and time spent in radiation areas. For immediate additional decontamination, the job-hours required to complete this work are increased by a factor of thrcc (3) as compared to the DTFR. His increase is based on: o A decrease in the endpoint criteria. The interim radiological endpoint criteria for the DTFR were i R/hr general area in the basement, and 5 R/hr general area in the cubicles. These dose rate criteria have been reduced to 20 mr/hr general area and 1 R/hr general area in the cubicles, respectively, for the base case plant; o De assumption that no appreciable dou rate reduction has occurred in the basement since the DTFR was written. For fmal decontamination as a part of decommissioning the job-hour requirement is also assumed to increase, based on these factors. This increase takes into account the decrease in dose rates in the basement during the 30 years assumed for PDMS, but also recognizes the fact that the dose rates will still be very high at the start of decontamination. 5A.8.2 RADIATION AREA J0D HOURS Re job hours developed in the DTFR e e total, billable hours (i.e., they include time spent on engineering tasks such as design; adnunistrative tasks, such as record keeping; and planning and preparation that are spent away from radiation areas). He time spent by radiation workers for dressing out and other worker activities is also compensated for in the estimate of the time spent in radiation areas. Herefore, a factor 5.A - 10 UPDATE 2 AUGUST 1997
l needs to be s,eplied to the total hours developed in the DTFR to account for these tasks which are not performed la 4adiation areas. To determine thejob-hours spent in radiation areas, the totaljob-hours listed in the DTFR were modified by factors based on actual billed hours and RWP hours (i.e., job-hours in radiation areas) for the decontamiration group workers in 1987. These RWP hours include the hours spent on health physics support. Based on presious expenence, it is assumed that the job hours that apply to health physics support equal 20 percent of those expended on decontarmnation tasks. Table 5A 8 brts the ETNs, used for tracking RWPs, and the total RWP hours that were used in determining these rnodification factors. These factors wete not developed for indhidual tasks but by t3pe of task, as listed below: RB AFlfB ' Robotics 0.15 N/A Dxon 0.20 0.50 Chaincternation 0.20 0.30 Systems 0.10 0.10 ne Robotics radiation arca factor (RAF) was used for tasks that are expected to require extensive use of robots. For tasks where Imtited robot use is expected, the RAF for decontammation is assumed De RAF for systems work is assumed to be the same for both buildmgs. Dese assumed radiation area factors also assume that as dose rates decrease, allowing longer periods to be spent in radiation area, the percentage of to'J job-hours spent in radiation areas will increase 5A.8.2.1 Exarnple of Determining Radiation Area Job hours 4 ( i Although both robotics and personnel access will be used in decontaminatmg the basement, the radiation area factor for robotic work is assumed for this task. The radiation area factor for robotic work is based on the following: o Totaljob-hours for robotics tasks, in 1987, was 4990, o Total hours spent in radiation areas for robotic work in 1987 was 820 hours (ETN C91K001); o Radiation area hours that apply to llealth Physics Support are 20% of the hours for decontamination tasks. Based on the above data a radiation area factor of 0.15 was derived for robotic tasks: (820/1.2 rad areas hrs)/4990 total hours = 0.15) 5A.8.3 JOB-110URS DURING PDMS It will be necessary for personnel to spend some time in the Reactor Building and AFilB for maintenance tasks and surveys. He job-hours for the period of PDMS are not estimated by task but, rather, are based on estimated annual job-hour requirements. 5A.8.4 RAD 10 ACTIVE WASTE MANAGEMENT Job-hour estimates have not been made for radioactive waste maragement tasks. The person rem for these
.g tasks are based on the total volume of waste that it is estimated will be processed.
p 5.A - 11 UPDATE 2 AUGUST 1997
5A 8.5 EXAMPLE OF J0D IlOUR ESTIMATION The decontarnination aethities for which person rem are estimated 6ffer from task desenptions in the DTTR. For trutance, the totaljob-hours for Reactor Buildmg Recovery are assumed to include the hours spent on Eeneral decontamination in the basement, decontamination of the basement cubicles, and removal of the block wall. This example focuses on the hours spent in radiation areas dunng decontamination of the basement cubicles. The total number of job-hours that are estimated for the basement recovery tasks is: Ra6ation Areas Job-hours = DTFRjob-hours ' scope modification
- raaation area factor
= 114610
- 3.00
- 0.15
= 51575 hours it is estimated that half of these hours will be spent on basement cubicle decontamination, therefore, this activity is estimated to require 25,788 job-hours in ra6ation areas.
5 A.9 PERSONNEL EXPOSURE Personnel exposures were detennined for two cases:
- 1. Irnmediate additional decontamination (i e., at the end of Phase 111); and
- 2. Fmal decontamination as a part of decommissioning (i c., after PDMS)
The parameters described in the earlier sections were used in computmg a range of estimated person rem for final decontamination of the TM12 plant, radwaste management and tasks that will be performed during PDMS. t O ! 5.A-12 UPDATE 2 AUGUST 1997 l l
I i n g 5A.9.1 DECONTAMINATIONTASKS
~
De person rem estimated for the decontammation tasks listed in Table SA $ are the products of the dose rates and radiation area job hours also listed in that table. As desenbed in Section 5A.6.1, the dose rates for each task, for immediate additional decontamination, are based on the dose rate criteria for the start and fmish of each task ne dose rates for fmal decontamination as a part of decornmissioning are assumed to be . half of the dose rates for immediate additional decontammation, to account for radionuclide decay. i ! In determining person rem, the job hours spent in radiation areas are required in this evaluation these hours i are developed from the total, billable job hours developed in the DTFR for decontamination tasks required to meet Phase Ill enteria, as described in Section $A.8. %e radiation areajob-hours are applied to several categories of tasks, such as Reactor Building cubicle cleanup, which are listed in Table SA 5. The ;
- person rem are estimated for each of these categories, as listed in Table SA 9. Ilowever, there are many
. aethities associated with each of these task categories. De array of activities considered in developing the I total person rem estimate for decontamination of TMI 2 are listed in Table $A 6. Table SA-6 gives a general idea of the extensive amount of work required to complete future decontammation activities.
j llistorical data from similar studies show that person tem estimates for forecasted tasks tend to be a factor of 2 high when compared to the actual person rem received in performing the tasks (Reference 11). To account for these uncertainties the person rem for cach task can be expressed as a range. ne lower bound of the :ange is, therefore, assumed to be half of the calculated person rem dose. In addition. to allow for uncertainty, the upper bound is assumed to be 10 percent higher than the calculated number. The person rem for health physics support is 20% of the total person rem for the other decontamination tasks. ,
$A.9.2 RAD 10 ACTIVE WASTE MANAGEMENT The total person rem estimated for radioactive waste management associated with the decontammation tasks is 360 550 person rem if the decontamination is performed at the end of Phase !!! and 180 280 4 person-rem if additional required decontamination is deferred for a 30 year period of PDMS. This is based on the assumption that the volume of waste processed will be in the assumed range of 120,000 to 183,000 cubic feet, which is within 20% of the total radioactive waste processed to date. The estimated person rem savings from radwaste processing is 180 280 person rem, which accounts for less than 5 percent of the total savings.
5A.9.3 PDMS TASKS
- it is estimated that the occupational exposure during PDMS will be in the range of 230 490 person rem.
This dose is less than 10% of the total estimated person rem from fmal decontamination as a part of decommissioning. Tasks to be performed during PDMS, along with the estimated person rem are: Maintenance & Radwaste Processing 190-410 Surwys and Inspections 10 - 80 Total 230 490 O S. A .13 UPDATE 2 AUGUST 1997 i
,-,ym,w ,- ,--yw.-.,, .'.r.
c.,.e --.m.. , ..._..,,.m.'.. -,,mm,,,-,,m... .,-.mm.,-.m_,..~, , , _ - . , ._.,_.,_--,-.-__.---,--+.-we,-...e
5A.9.4 TOTAL PERSON REh! ne estunated person rem for the two cases considered in this evaluation are listed by category in Table SA 10. He total person rem are: Immediate Additional Decontamination 7,160 15,550 person-rem Fmal Decontamination as a part of of Decommissioning 2,710 5,770 person rern ne difference between these ranges (i.e., approximately 4,500 9,800 person tem) is the estunated savmss when further decontamination is deferred until after a 30 year period of PDhiS. 5A.10 C0hiPARISON CASE 20 YEAR PDh!S ne person rem savings listed in the PElS (Reference 1)is based on a PDh15 duration of 20 years ne estimated person-rem savmgs, assuming 20 years of PDhtS, has been determined for companson with the PElS numbers. He esttmated total person rem for this case are 4,100 - 8,800 perron rem, with a resultant savmgs in the range of 3,100 6,800 person rem. His person-rem estimate for fmal decontandnation following PDhtS was determined using the methods desenbed in this Appendix with a 20-year decay factor applied to the dose rates. 5A 11 CONCLUSION Tabic 5A 10 lists the estimated person rem for immediate additioral decontamination and for fmal decontamination as a part of decommissioning assuming a 30-year period for PDhtS. %c estimated occupational person-rem savmgs from deferring further decontamination of the Th112 plant until after a PDhtS period of 30 years is in the range of 4500 - 9800 person-rem. Person tem estimated for activities during PDh1S are snull in comparison to the total savings. The range of estimated savings has been detennined based on present knowledge of tasks that will be performed during the decontamination, and assuming that the decay of radionuclides danng PDhtS will be at least that of Cs 137, the predorninant dose contnbutor, it should be noted that Cs 137, with a half hfe of 30 years, is one of the longer-hved isotopes. It is, therefore, conceivable that the dose rates at the end of PDhtS will have decreased by more than 50% Robotic development during the period of PDh1S could also decrease the person-tem from defeued decontamination due to increased efficiency and reliability of the robots, i e , less job-hours required in radiation areas. His person rem evaluation demonstrates that defernng decontammation of Th112 beyond that required to achieve the postulated PDhtS conditions for a period of 30 years will result in an occupational person-tem savings in the range of 4,500 9,800 person-tem; the shorter period of 20 years postulated in Reference I will provide estinuted savings in the range of 3,100 6,800 person rem. As has been shown, the risk to the general public is negligible during PDhtS (References 7 and 13). Therefore, this large estimated person-rem savings clearly wanants deferral of further decontamination beyond that currently included in the TMl 2 Cleanup Program for the period of PDhtS. O
- 5. A - 14 UPDATE 2 AUGUST 1997
( 5A.12 REFERENCES
- 1. NUREG-0683, Supplement No. 3, " Programmatic Emironmental Impact Statement...." August 1989
- 2. " Decontamination Task Force Report," by P. R. Bengel, et al, December 18,1985
- 3. 'lechnical Bulletin 85 1,
- Reactor Building General Area Radiation Survey Maps,"
Rev. 3, February 3,1988
- 4. Technical Bulletin 8610, ""B" Steam Generator TLD Characterization," Rev. O, February 18, 1986
- 5. GPU Nuclear memorandum 9240-88 4372 from J. E. Tarpinian to D. W. Turner, " Comparison of TMI 2 with other B&W Plants," dated February 2,1988
- 6. BBR GMBil Report Number $95 C01 A (82),
- Evaluation of the Dose Rate Data of Various Nuclear Power Plants with B&W Nuclear Steam Supply Systems"
- 7. GPU Nuclear rnemorandum 6615 90-0188, from S. E. Acker to E. D. Schrull, "PDMS SAR Ch 8 Dose Calculations," dated November 5,1990
- 8. Technical Plan TPOfrMI 188, "TMI 2 Cleanup Program Post Defueling Monitored Storage,"
Revision 0, January 1987
- 9. " Disposition of the Reactor Building Blockwall,* by P. R. Bengel, et al, June 19,1987 (Attachment to GPU Nuclear Memorandum 4440-87 048)
- 10. " Task Force Report Reactor Building Basement Decontamination," November 1987
- 11. GPU Nuclear memorandum 9240-88-4521, from D. J. Merchant to J. E. Tarpinian, " Review of 1986 Collectise Dose Goal," dated May 9,1988
- 12. TP0/TMI-009, " Gross Decontamination Experiment Report," Bechtel National, Inc., September 1982
- 13. GPU Nuclear memorandum 6615 92160, from S. Acker to E. Schrull, " Dose Calculation Results per anemo C312 921045, PDMS SAR Rev.16," dated October 27,1992.
{d
- 5. A - 15 UPDATE 2 AUGUST 1997
TAllLE 5A.g PilASE 111 END POINT CRITERIA GENERAL AREA DOSE RATE AHf A DESCRIPTION R4ir EE ACTOR DUILQEiQ. ' Refueling Canal <0.10 Elevation 347' and above (except D. Ring) <0.03 Elevation 305' to 347' <0'.0 Dasement(El 282) "As is* AUX 11,1ARY BUILDEiO' Corridors <0 0025 Other Areas <0.0$ IELilANDLING Dull, ult {G' Corridors <0 0025 Other Areas <0 05 QIIIERJ1UllJ21SGS Turbine Building <0 0025 Chemical Cleanup Duilding <0 0025 (except EPICOR 11 pump area .a be left operable) Service Building Containment <0.0025 Tank Arca SQ1T1
- 1. Conditions pertain to general area; excludes " hot spots" (e g., basement block wall) and those that are locked high radiation areas (e g , seal injection valve room).
- 5. A - 16 UPDATE 2 AUGUST 1997 O
TABLE SA 2 END OF PHASE III AND BASE CASE DOSE RATES REACTOR BUILDING DOSE RATES (mrem /br) 1 OCATION TMi2 Baw fm Top of Rea: tor Vessel 5 - 60 55 Support structure Reactor Vessel Head 24 - 30 155 Stand Area Pressurizer Top 400 - 600 25 65 OTSO Sides 600 - 3,000 10 35 OTSG Top 80 200 10 35 00 - >2,000 281' Reactor Buildirig 10,000 50,000 <.2 18 D Ring Walkways 25- 80 8 Incore Table 120 150 4 Equiptrentllatch Area 80 - 200 4 n 5.A - 17 UPDATE 2 AUGUST 1997 l
1 l TABLE 5A-3 END OF P11ASE Ill AND DASE CASE DOSE RATES AUXILIARY AND FUEL IIANDLING BUILDINGS DOSE RATES (mrem /hr) LQCATION TMI-2 Base Case General Areas Walkways (non cubicle areas) 281' <0.2 20 0.2 18 305' <0.2 <0.2 328' <0.2 <0.2 347 0.5 30 0.2 5 Other Areas Make up Pump Areas 20-80 ( A) 26 (high pressure injection) <500 (B) ND (C) Concentrated Waste Storage Tank Area 10 22 90 Spent Resin Tank Area ND (B) 0.2 1 (A) 3800 Miscellaneous Waste Tank <50 65 Reactor Coolant Evaporator Room <50 2.2 6 0 Spent Fuel Pool Cooler 0.21.0 0.2.5 Waste Transfer Pump Room <60 0.6 1.0 Reclaim Boric Acid Tank <0.6 15-26 Reclaim Bonc Acid Pump <100 8-12 Neutralizer Pump Room 5 36 18-40 Decayllcat Vaults 5-50 3-18 (both) Reactor Building Spray Pumps 415 (A) 0.2-2.5 (both) 50-150 (B) Auxiliary Sump ND 10-25 Sealinjection Valve Room ND 5 20 Makeup Valve Areas 2 500 12 100 ND no data
- 5. A - 18 UPDATE 2 AUGUST 1997
. . _ . _ _ - - - ~ . _ - . _ _ - - -
j t L ! TABLE 5A-4 l
- ESTIMATED Cs AND Sr ACTIVITIES IN THE REACTOR BUILDING t-
! t ! Cs 137 Sr.90 SOURCE (Curies) (Curies) ! { 1 l D-Ring 'A' 1,660 80 . . - D-%g 'B' 15,000 750
- . Block wall 19,000 750 l Previously Subrnerged. 7,000 300
( Floors and Walls - , j Sediment 460 450 a l l l l-l 'N_QTE: These activities are from Chapter 8 of the PDMS SAR ! L i f 1-i i I. A-f .. t r s 1: i i
- i i
! 5.A - 19 UPDATE 2 AUGUST 1997 l ) 1 4- -
._-___..__:.:.__._.___.--._.......___-....,.,. _._._,.,.,_,n_._:.._._.._.,
i TABLE 5A 5 DOSE RATES AND ADJUSTED JOB-IlOURS POST-PilASE 111 POST PDMS INDEX DECONTAMINATION DOSE RATES DOSE RATES NUMBER TASKS (mrrm/hr) JOH-IlOliRJ (mrem /hr) JOH-IIOliRS 1.0 Reactor Buildmg 1.1 Preparations / Support Activities 1.1.1 Characterization 30 1,901 15 1,721 1.1.2 Ventilation Control - 0 - O and Area Isolation 1.1.3 11ealth Physics Support' - 0 - 0 1.1.4 Engineccing Support 30 2,810 15 3,810 1.2 BaLment General Cleanup 100 26,740 50 21,009
; 1.3 Basement Cubicle h Cleanup 100 25,788 50 17,192 1.4 Basement Block Wall Removal 2
- 8,,96
< - 5,731 1.5 D-Ring Dose Reduction 150 9,400 75 4,700 1.6 D-Ring Final Decontamination 100 14,848 50 11,136 1.7 Dome and Polar Crane Decon 50 762 25 381 1.8 El. 347'-0" DecoWDose Reduction 30 4,700 15 2,350 1.9 El. 347'-0" Final Cleanup 20 37,303 10 18,651 1.10 Et, 305'-0" Decon
/ Dose Reduction 50 4,700 25 2,350 1.1I El. 305' 0" Final Cleanup 30 38,064 15 19,032 1.12 Systems Decontamination 1.12.1 Reactor Coolant Sys, 50 267 25 267 1.12.2 Non-RCS Systems 50 L1,2 9 25 ? 392 Subtotal 179,268 110,720 5.A - 20 UPDATE 2 AUGUST 1997
l ( TABLE SA 5 (Cont'd) DOSE RATES AND ADJUSTED JOR-HOURS POST PHASE IU POST PDMS INDEX DECONTAMINATION DOSE RATES DOSE RATES NUMBFR .16EES (arem/br) JOB-HOURS (mrem %r) JOB-HOURS 20 Auxiliary and Fuel Handling Buildings 2.1 , Preparations / Support Activities 2.1.1 Characterization 10 1,092 5 767 2,1.2 Health Physics Support' - - - - 2.1.3 . Engineering 10 139 5 28 Support 2.2 AFHB Decon/ Dose Reduction 10 19.682 5 8.065 Subtotal 20,913 8,861 Total 200,181 119,581 NOTES
- 1. Nojob-hours or dose rates are listed for health physics. Person-tem for health physics support is assumed to be 20%
of total person-rem for other tasks,
- 2. Person rem for removal of the blockwall were developed in Reference 9, therefore, no dose rates urre assumed in this evaluation.
i
' k- 5.A - 21 UPDATE 2 AUGUST 1997
TABLE SA-6 O FINAL CLEANUP ACTIVITIES (TYPICAL) REACTOR BUILDING 1.1 Preparations / Support Activities 1.4 Basement Block Wall Removal Characterization Remove liand Rail j Ventilation Control and Area Isol. Remove Elevator Access Platform Health Physics Support Demolish Elevator Work Platform Engineermg Suppon Core Bore Concrete Beneath Access Remove Structural Steel 1.2 Basement General Cleanup Remove Seismic Screen Core Drill Concrete Block Walls Desludge Floors Cut / Remove Steel Be.ms Gross Water Flush Remove Access Ladder High Pressure Water Flush Decontammate Elevator Pit Remove Instrument Racks Fill Pit With Concrete Remove Pumps, Heat Exchangers Dismantle Stairs Concrete pad removal Piping, Cable, Conduit Removal 1.5 D-Ring Dose Reduction Scarify Concrete Wall Scarify Floors Low Pressure Flush All Surfaces Coating Removal From Steel Surfaces Limited Removal of Mirror Insul. Coating Removal From Contmnt. Liner Lower Pressure Flush Exposur: Surf. Remove Hoses, General Trash High and Ultrahigh Pressure Flush Remove Ventilation Ducts Limited Coatings Removal Clean Floor Drains Placement of Temporary Shielding histall Waste Handling System Flush /Mose Basement Sludge 1.3 Basement Cubicle Cleanup 1.6 D-Ring Final Decon Remove Door, Cages to Cubicles Complete Insulation Removal Concrete Removal for Robot Access Low Pressure Flush All Surfaces Desludge Sump liigh and Ultrahigh Pressure Flush Flush, Scarify Sump Removal of Contanunated Coatings Chemical Clean Letdan Coolers Clean Platforms, Structural Steel Remove Letdown Coolers Decontanunate RC Pumps Gross Water Flush All Cubicles Scarify Inside D-Ring Walls High Pressure Flush All Cubicles Scarify All Bathtub Ring Areas Remove Piping, Cables, Conduit Paint Removal From Steel Surfaces Scarify Floors of All Cubicles Sandblast RCS Component Surfaces ScarifyWallsin AllCubicles Decontammate/ Remove HVAC Ducts Miscellaneous Coatings Removal 5.A - 22 UPDATE 2 AUGUST 1997
. ~ -. - . _ - - - -. .- . _ . - - - .- .. -
TABLE 5A-6 (Cont'd) O FINAL CLEANUP ACTIVITIES REACTOR HUILDING 1.7 Dome and Polar Crane Decon 1.10 El. 305'-0" Decon/ Dose Redaction Scarify Crane Rail Support Gross Fiush of Overheads and Walls Remove Crane Motors / Equipment Limited Equipment Removal Remove Paint From Crane Placement of Temporary Shielding Remove Dome Coating Above LOCA Ducts Remove Dome Coating Other Areas 1.11 El. 305'-0" Final Cleanup 1.8 El. 347-0" Decort Dose Reduction
/
Aggressive Decon Overheads, Walls Remove Concrete Pads Low Pressure Water Flush Decontammate/ Remove Equipment . Limited Equipment Removal Clean Drains Placement ofTemporary Shielding Decontammate/ Scarify Floors Remove Steel Coatings 1.9 El. 347-0" Final Cleanup Decon/ Remove Defueling Equipment Remove Air Cooler Motors and Fans Aggressive Decon Overheads, Walls Scarify Walls Around Open Stainvell Remove Concrete Pads Remove LCSA Pieces From CFr Decontammation/ Remove Equipment Decon/ Remove Core Flood Lines p Clean Drains Decontammate/ Scarify Floors Remove Overheads Around Open Stair Remove Steel Coatings 1.12 Systems Decontammation Decon/ Remove Defueling Equipment Remove LOCA Ducts Reactor Coolant System Scabble Missile Shields Non-RCS Systems Decontammate Head Cut Up and Remove Plenum AUXILIARY AND FUEL HANDLING BUILDINGS 2.1 Preparations / Support Activities 2.2 AFHB Decon/ Dose Reduction Characterization Decon Spent Resin Storage Tanks Health Physics Support Decon Makeup Demin Cubicle Engineering Support Decon Cleanup Demin Cubicle Decon Misc. Waste Holdup Tank Room Decon SealInjection Valve Room Decon Other AFHB Areas Systems Decon v
- 5. A - 23 UPDATE 2 AUGUST 1997
TABLE SA 7 O' RADIATION AREA JOB-HOURS ; DTFR' TOTAL RADIATION AREA JOB-HOURS' TASK DESCRIPTION JOB-HOURS Post-PDMS Post-Phase III 20-Year PDMS 30-Year PDMS 1 RB Characterization 6,400 1,280 1,280 1,280 2 RB Visual Survey 21,060 421 421 421 3 Sludge & Core Bore Visual 13,330 200 20 20 4 RB Decon/ Dose Reduction 24,320 4,864 3,405 2,432 { 5 RB Decon Maintenance 69,680 13,936 9,755 6,968 l 6 Destudge RB Basement 63,650 9,548 9,548 9,548 1 7 Decon D Rings Above 282' 37,120 14,848 1.' 992 11,136 2 42,979 34,383 8 RB Basement Recovery 114,610 51,575 l 9 Establish HVAC Control 104,250 0 0 0 l 10 Decon 347' and above 190,320 38,064 26,645 19,032 11 Decon 305' and above 190,320 38,064 26,645 19,032 q 12 Robotics 22,880 0 0 0 13 Non RCS Systems Decon 119,600 2,392 2,392 2,392
]
14 RCS System Decon 13,360 267 267 267 15 Site Engineering 76,190 3,810 3,810 3,810 SUBTOTAL 1,067,090 179,268 140,158 110,720 : .,me 45,760 686 686 686 e! 1 AFHB Charactenzation 2 Decon Mamtenance 109,200 2,730 1,638 546 3 Tech Spec 106,080 0 0 0 Decontamination 4 AFHB Cubicle Decon 291,200 14,560 10,192 7,280 5 Systems Characterization 27,040 406 243 81 6 Non RCS Systems Decon 239,200 2,392 1,196 239 7 Site Engmeering 55,640 139 28 28 SUBTOTAL 874,120 20,913 13,9&4 8,861 TOTAL 1,941,210 200,181 154,141 119,581 EQIES 1 DTFR - Decontanunation Task Force Report 2 Post-Phase III - immediate additional decontammation Post PDMS - Final Decontamination as a part of decommissioning
- 5. A - 24 UPDATE 2 AUGUST 1997
( TABLE SA-8 ETN AND RWP HOURS ETN RWP HOURS Reactor Building Roboucs C91K001 819.25 Characterization C90K001 1,213.63 Decontamination C30C001 0.0 C30D001 10.80 C30E001 0.0 C30D002 498.02 K20H015 170.07 D47H004' 1,322.93 Auxiliary & Fuel Handling Building Decontamination J30C011 397.12 130C012 2,218 47 J30C013 1,491.62 J30C070 148.18 J30J103 3,353.57 130C082 307.93 J30C083 18.78 J25H001 57,482.77 J30J040 538.65 Characterization Systems 190H001 664.32 E30C001 1,664.37
- Not all the RWPs included in this ETN were for decontamination work. The hours listed are only for decontamination tasks.
\ 5.A - 25 UPDATE 2 AUGUST 1997
TA BLE SA-9
SUMMARY
TABLE FOR PERSON-REM EVALUATION ANALYSIS POST-PilASE 111 INDEX DECONTAMINATION DO5E RATES POST-PDMS NUMBER TASKS (mrem /h r) (JO B-IIO URS) PERSON-REM' JOB-IIOURS PERSON-REM 1.0 Reactor Building 1.1 - Preparations / Support Activities 1.1.1 Characterization 30 1.901 30-60 1,721 10-30 1.1.2 Ventilation Control & Area Iso. - 0 0-0 0 0-0 1.1.3 llealth Physics Support' - 1,110 - 2,450 370-R20 1.1.4 Engineering Support 30 3,810 60-130 3,810 30-60 1.2 Basement General Cleanup 100 26,740 1,340-2,940 21,009 530-1,160 1.3 Basement Cubicle Cleanup 100 23,788 1,290-2,840 17,192 430-950 1.4 Basement Block Wall Removal - 8.5 % 180-400 5,731 100-210 1.5 D-Ring Dose Reduction 150 9.400 710-1,550 4,700 180-390 1.6 D-Ring Final Decontamination 100 14,848 740-1,630 11,136 280-610 1.7 Dome and Polar Crane Decon 50 761 20-40 381 0-10 1.8 El. 347'-0* Decon/ Dose Reduction 30 4,700 70-160 2,350 20-40 1.9 El. 347*-0" Final Cleanup 20 37,303 370-820 1R,651 90-210 1.10 El. 305'-0" Decon/ Dose Reduction 30 4,2 !?n-? m 2.350 30-60 5.A - 26 UPDATE 2 AUGUST 1997 O O O
.s m
TABLE SA-9 (Cent *d)
SUMMARY
TABLE FOR PERSON-REM EVALUATION ANALYSIS POST-PHASE Ill INDEX DECONTAMINATION DOSE RATES POST-PDMS ' NUMBER TASKS (mrem /hr) (JOR-HOURS) PERSON-REM' JOB-HOURS PERSON-REM ~ 1.11 El. 305'-0* Final Cleanup 30 38,064 570-1,260 19,032 ' 140-3 to - 1.12 Systems Decontamination 10-20 1 1.12.1 Reactor Coolant System 50 267 10-20 267 0-10 , 1.12.2 Non-RCS Systems 50 2,392 60-130 2,392 30-70 Subtotal 179,268 6.680 - 14,690 110.720 2,240 -4,940 2.0 Auxiliary & Fuel lland'ing Buildings i 2.1 Preparations / Support Activities -
'2.1.1 Characterization 10 1,092 10-10 767 Neg. ,
i 2.1.2 IIcalth Physics Support' 20-50 0-10
'{
- 2.1.3 Engineering Support - 10 139 Neg. 28 Neg.
1 2.2 AFIID Decon/ Dose Reduction 19,682 100-220 R,065 20-40 i i Subtotal 20,913 130-280 8,861 20 , Total 200,181 6,810-14,970 119,581 2,260
- i. .
i i t
' 5.A - 27 UPDATE 2 AUGUST 1997 l
l
l l TABLE 5A-9 (Cont'd)
SUMMARY
TAllLE FOR PERSON-REM EVALUATION ANALYSIS NOTLS
- 1. Dose rates for Post PDMS are 50% of those listed for Post Phase Ill.
- 2. Person-rem for health physics support are assumed to be 20% of total person-rem.
- 3. Person-rem for block wall removal are from Reference 9.
- 4. The range of person-rem is assumed to be from 50% below to 10% above the product of dose rate and job-hours columns from Table SA-9.
5 Neg. - Less than 5 person-rem. O O
- 5. A - 28 UPDATE 2 AUGUST 1997
4 TABLE 5A 10
'%)~ ESTIMATED TOTAL PERSON REM FOR FINAL DECONTAMINATION OPTIONS POST PHASE lil' POST PDMS*
. TASK Efrson-rem) (Person-rem) -
During PDMS N/A 230 - 490
, Cleanup 6,800 - 15,000 2,300 - 5,000 Radwaste Handling . 360 - 550 180 280 Total 7,160 - 15,550 2,710 - 5,770 PDMS Person rem Savings' 4,500 - 9,800
'The savings is defined to be the difference in total person rem required for deferred and immediate decontammation.
O NOTES
- 1. Immediate additional decontammation.
- 2. Final decontarmnation as a part of decommissioning (30-year PDMS).
i f S.A - 29 UPDATE 2 AUGUST 1997 L______________.__._.._..___ _ _ . - _ . - _
p DECONTAMINATION ACTIVITIES - The prunary techniques used to decontanunate the TMI-2 facility are listed below and discussed as appropriate.
- a. Kelly Vaccumac b.Scabbling c.Strippable Coatings d.High Pressure Spray
- c. System Flushing
- f. Hands-on Decontammatica Kelly Wecumac During the decontanunation program, steam cleaning machines, called Kelly Vaccumac, were procured.
Contanunation was removed from surfaces by a steam cleaning wand, then vacuumed much like conventional steam cleaning unit used on carpetmg. Concrete Scabbbng s A concrete scabbler was developed at TMI-2. This machine removed a layer of concrete surface from the floors of many contaminated areas of the plant and vacuumed the residue. Scabbled surfaces were then coated or painted to prevent recontammation. Strippable Coadags Strippable coatings made of carious polymer substances were applied to surfaces. When subsequently removed, gross particulate contammation adhered and also was removed. Hich Pressure Sorav and Flushing Sewral different types of high pressure spray and flush operations were used to decontaminate the plant areas. Rese techniques varied water temperature and pressure; however, in each case, water was the only medium used to remove loose contamination from walls and surfaces in the various areas of the plant. System Flushing Processed borated water was circulated through most contammated systems to remove coranunants All of the systems wre then dramed to the extent practical and isolated unless required for PDMS activities.
/ \
5.B - 1 UPDATE 2 - AUGUST 1997
llands-On Decontaminaligo Hands on decontammation is a mechanical tech"ique that was generally used on floors and walls to remove loosely held contamination. Typically, rags, absorbent cloths, brushes, and abrasive pads were utilized for this decontammation technique. In addition, mechanically driven hand brushes and floor brushes were utilized where required Grit or abrasives were added, as necessary, to aid in surface cutting or removal of a portion of the surface if the contanunation was more tightly fixed. Occasionally, approved chemical decontamination agents were used to enhance the effectiveness of hands-on techniques. However, no major chemical decontammation activities were conducted. ROBOTIC DEVICES In several cases, the decontammation equipment was mounted on remotely controlled robots and used to treat surfaces that were inaccessible to personnel nese robots were used in the Reactor Buildmg basement and AFilB cubicles. Each robot is described below. System In-Senice Inspection (SISD In 1982 a device was developed to perform remote samplmg and radiation surveys in the makeup purification demineralizer cubicles. This device was a small track, tethered vehicle that contained several CCTV cameras, a radiation detector, and a small manipulator arm for sample collection. Determmation of the levels of smearable contamination of the floor surface was the extent of the sample collection intended. Remote Controlled hiobile Maninulator (RCMhD in December 1982 a RChiM device, designated " Fred" was purchased. This device was tethered, six-wheeled vehicle with a simple manipulator. He RChiM was used to remotely flush areas in the plant to muumize the requirement for personnel exposure in high radiation areas for extended periods of time. The acquisition of this device essentially was the fint use of remote equipment to perform clean up tasks as a means of reducing occupational radiation exposure. He first use of the RCMM was in the Auxiliary Buildmg for flushing the "B" and "C" reactor coolant bleed tank cubicles and makeup pump IB room. A high pressure water nozzle mounted on the end of the marupulator arm was used to flush the rooms. A support camera with pend and tile capabilities was used to pronde additional viewing capability. Remote Recontamnce Vehicle (RRV) Dunng the first part of 1984 a Remote Reconnaissance Vehicle, designated " ROVER," was delivered to TMI. The RRV was a six wheeled, tethered mobile vehicle. It contained CCTV cameras foniewing in the forward and rear directions. He tethered playout and takeup was controlled by the RRV operator. A space on the front of the RRV allowed attachment of devices to perform remote activities. During imtial use of the RRV, a radiation detector on an x-y positioner allowed surveys and vertical radiation profiles to be performed. Later the RRV was equipped with flushing, core boring, concrete removal and sludge removal tools. 5.B - 2 UPDATE 2 - AUGUST 1997
i l i Remote Controlled Transport Vehicle (RCTV) g In 1984 a RCTV was loaned to GPU Nuclear Corporation by the Department of Energy. It contained a manipulator arm mounted to a telescoping column. Power and controls to the RCTV are transmitted through a tether that follows the movement of the transporter. Lessons leamed during the implementation of the presiously used remote vehicles helped make the first use of the RCTV a success. He first effort with the RCTV was to perform a radiation survey and measure the radiation profile of the "A" snd "B" makeup demineralizer tanks. le91t2 GPU Nuclear designed and fabricated a six -bled, skid-steered remote vehicle named Louie 2, primarily ' to remove the highly contaminated grout pad from the floor of the seal injection valve room, he device was configured with a three-piston scabbler and a vacuum shrood to remove the scabbled debris. The robots developed and used at TM1-2 performed a variety of recovery tasks including data acquisition, dose reduction, and waste management. Aroughout the entire cleanup period, new equipmem and techniques were developed as necessary to meet the challenges presented by the cleanup task, in some cases, the most efficient and cost effective means of mitigating the contanunation and mmunizing exposure was to remove the contaminated equipment involved. It was then packaged and shipped to the appropriate commercial radioactive waste disposal facility. O WASTE PROCESSING Several processing systems were developed for use in dispositioning both the accident water and the water necessary to accomplish the TMI-2 cleanup operations. Primarily, water use was restricted to the original accident generated water, which was cleaned and reused throughaut the cleanup program. The pnmary water processing systems developed were the Submerged Demineralizer System, EPICOR II, and the Defuelmg Water Cleanup System. Suhmerced Demineralizer Svstem The Submerged Demineralizer System (SDS) was a primary water processing system utilizing shielded, expendable vessels. The system was located in the Unit 2 Spent Fuel Pool 'B' and shielded by the pool water. He processing media were zeolites and it was most effective in removing soluble cesium and strontium. At the completion of TMI 2 waste management activities, the system was deactisated, the expendable vessels were disposed of at a commercial radioactive waste burial site or shipped to a DOE research facility, and the shielding water was disposed of as part of the accident generated water disposal program. V 5.B - 3 UPDATE 2 - AUGUST 1997
EPICOR 11 EPICOR 11 was configured as a primary processing system utilizing three disposable demmeralizer vessels (liners). The first position lmer utilized zeolites as a processing media and served to remove soluble cesium and strontium. The second and third liners utilized organic resins as the process media and served primarily as polishers. Defuehne Water Cleanup System (lon-Exchancer Portion) The Defueling Water Cleanup System (DWCS) was a primary processing systern. It utilized a disposable liigh Integrity Contamer (HIC) with zeolites as a processing media. It was effective in remosing soluble cesium and strontium. Like the SDS,it was deactivated at the ennclusion of the TMI 2 waste management actisities. O 5.B - 4 UPDATE 2 - AUGUST 1997
1 l l i 1 4 i i 1 CHAPTER 6 DEACITVATED SYSTEMS AND FACILITIES 1 l
- CHAPTER 6
DEACTIVATED SYSTEMS AND FACILITIES TABLE OF CONTENTS
6.0 INTRODUCTION
6.0-1 6.1 DEACTIVATED FACILITIES 6,1-1 6.1.1 DELETED 6.1-1 6.1.2 DELETED 6.1-1 6.1.3 CONTAINMENT AIR CONTROL ENVELOPE 6.1-1 6.1.4 CIRCULATING WATER PUMP HOUSE 6.1-1 6.1.5 CIRCULATING WATER CHLORINATOR HOUSE 6.1-1 6.1.6 NATURAL DRAFT COOLING TOWERS 6.1-2 6.L7 MECHANICAL DRAFT COOLING TOWER 6.1-2 DELETED 6.1-2 6.1.8 6.1.9 TENDON ACCESS GALLERY 6.1 2 b] f 6.1.10 RIVER WATER AND FIRE PUMP HOUSE 6.1 2 6.1.11 BWST PIPE CHASE 6.1-3 6.1.12 CONTROL BUILDING (M 20) AREA EAST 6.1-3 6.1,13 CONTROL BUILDING (M-20) AREA WEST 6.1-3 6.1.14 DELETED 6.1-3 6.2 DEACTIVATED PASSIVE SYSTEMS 6.2- 1 6.2.1 MAIN AND REHEAT STEAM SYSTEM 6.2-1 6.2.1.1 System Design 6.2-1 6.2.1.2 PDMS Function 6.2-1 i UPDATE 2 - AUGUST 1997
-p) t v
)
_ _ . . _ . . , . _ _ . . _ _ _ . . - _ _._ _ . ~ . . _ . _ _ . . . . _ . .._ _ _ _ . _ _ . _ _ CHAPTER 6 , TABLE OF CONTENTS (Cont'd) SECTION IITLE . PAGE s
; 6.2.2 _ PRIMARY NtTLEAR PLANT HYDROGEN SUPPLY 6.2-1 SYSTEM 6.2.2.1 System Design 6.2-1 6.2.2.2 PDMS Function 6.2-2 6.2.3 FUEL HANDLING AND STORAGE SYSTEM 6.2-2 I 6.2.3.1 System Design 6.2-2 6.2.3.2 PDMS Function 6.2-2
^ 6.2.4 - STANDBY REACTOR COOLANT PRESSURE 6.2-2 CONTROL SYSTEM 4 6.2.4.1 System Design 6.2 2 6.2.4.2 PDMS Function 6.2-2 2 i t 6.2.5 MINI DECAY HEAT REMOVAL SYSTEM 6.23-6.2.5.1 System Design 6.2-3 6.2.5.2 PDMS Function 6.2 3 6.2.6 SPENT FUEL COOLING SYSTEM 6.2-3 i 6.2.6.1 System Design 6.2-3 i 6.2.6.2 PDMS Function 6.2-3 6.2.7 ' REACTOR COOLANT MAKEUP AND PURIFICATION - 6.2-4 SYSTEM 6.2.7.1 System Design- 6.2-4 6.2.7.2-- PDMS Function - 6.2-4 6.2.8 - DECAY HEAT REMOVAL SYSTEM 6.2-5 6.2.8I System Design- 6.2-5
. 6.2.8.2 ' PDMS Function 6.2-5 ii UPDATE 2 - AUGUST 1997
_ - _ . . . ~ . _ _ _ . _ . . . . _ _ . . _ . _ _ _ _ . _ _ _ _ _ _ . l CHAPTER 6 TA.BLE OF CONTENTS (Cont'd) e 6.2.9 REACTOR BUILDING LEAK RATE TEST SYSTEM 6.2-5 6.2.9.1 System Design 6.2-5 6.2.9.2 PDMS Function 6.2-6 6.2.10 SERVICE AIR SYSTEM 6.2-6 e 6.2.10.1 System Design 6.2-6 6.2.10.2- PDMS Function 6.2-6 6.2.11 CHEMICAL ADDITION SYSTEM 6.2-6 6.2.11.1 System design 6.2-6 6.2.11.2 PDMS Function 6.2-6 6.2.12 REACTOR BUILDING EMERGENCY SPRAY SYSTEM 6.2-6 4 ( 6.2.12.1 System Design 6.2-6 6.2.12.2 PDMS Function 6.2-7 4 6.2.13 NUCLEAR SAMPLING SYSTEM 6.2-7 6.2.13.1 System Design 6.2-7 6.2.13.2 PDMS Function 6.27 6.2.14 NUC. PLANT AND RADWASTE NITROGEN SYSTEMS 6.2-7 _6.2.14.1 System Design 6.2-7 6.2.14.2 PDMS Function - 6.2-7 6.2.15- DECAY HEAT CLOSED COOLING WATER SYSTEM 6.28 6.2.15.1 System Design 6.2-8 6.2.15.2 PDMS Function-' 6.2-8
. [UY iii UPDATE 2 - AUGUST 1997
l CHAPTER 6 TABLE OF CONTENTS (Cont'd) SECTION- TITLE EAGI 6.2.16 RADWASTE DISPOSAL, WASTE DISPOS AL GAS SYSTEM 6.2-8 , 6.2.16.1 System Design 6.2 8 6.2.16.2 PDMS Function -6.2-8 6.2.17 REACTOR BUILDING EhERGENCY COOLING WATER 6.2-8 SYSTEM 6.2.17.1 System Design 6.2-8 6.2.17.2 PDMS Function 6.2-9 6.2.18 INTERhEDIATE CLOSED COOLING WATER SYSTEM 6.2-9 6.2.18.1 System Design 6.2-9 6 2.18.2 PDMS Function 6.2-9 ( 6.2.19 FUEL POOL WASTE STORAGE SYSTEM 6.29 6.2.19.1 System Design 6.2-9
-6.2,19,2 PDMS Function 6.2 10 6.2.20 TEMPORARY NUCLEAR SAMPLING SYSTEM 6.2-10
- 6.2.20.1 System Design 6.2-10 6.2.20.2 PDMS Function 6.2-10 6.2.21 OTSG CHEMICAL CLEANING SYSTEM 6.2-10 6.2.21.1 System Design 6.2-10 6.2.21.2 PDMS Function 6.2-10 2
6.2.22 CORE FLOODING SYSTEM 6.2-11 iv UPDATE 2 - AUGUST 1997 l l hlD
CHAPTER 6 TABLE OF CONTENTS (Cont'd) b Q SECTION ~ IIILE PAGE
.6.2.22.1 System Design 6.2 11 6.2.22.2 PDMS Function 6.2.11 6.2.23 FEEDWATER SYSTEM ~ 6.2- 11 6.2.23.1 System Design 6.2-11 6.2.23.2 ~ PDMS Function 6.2-11 6.2.24 NUC. SERVICES CLOSED COOLING WATER SYSTEM 6.2-11 6.2.24.1 System Design 6.2 11 6.2.24.2 PDMS Function 6.2-12 6.2.25 INDROGEN RECOMBINER 6.2-12 6.2.25.1 System Design 6.2 12 6.2.25.2 PDMS Function 6.2-12 6.2.26 EMERGENCY FEEDWATER SYSTEM 6.2 12 6.2.26.1 System Design 6.2-12 6.2.26.2 PDMS Function 6.2-10 6.2.27 REACTOR COOLANT SYSTEM 6.2-12 6.2.27.1 System Design 6.2-12 6.2.27.2 PDMS Function 6.2-13 6.2.28 SOLID RADWASTE DISPOSAL SYSTEM 6 .2-13
'6.2.28.1 System Design 6.2-13 6.2.28.2 PDMS Function -6.2-13 i
6.2.29 REACTOR FEEDWATER HEAT SYSTEM - 6.2-13 v UPDATE 2 - AUGUST 1997 l l ? - , . .. . - . , _ - - - . - _ . --
s CHAPTER 6 TABLE OF CONTENTS (Cont'd) SECTION TITLE PAGE 6.2.29.1 System Design 6.2-13 6.2.29.2 PDMS Funetion 6.2-13 6.2.30 STEAhi GENERATOR SECONDARY VENTS & DRAINS 6.2-14 SYSTEM 6.2.30.1 System Design 6.2-14 6.2.30.2 PDMS Function 6.2 14 6.2.31 DEWATERING SYSTEM 6.2 14 6.2.31.1 System Design 6.2-14 6.2.31.2 PDMS Function 6.2-14 ( ( 6.2.32 FUEL TRANSFER CANAL FILL AND DRAIN SYSTEM 6.2-14 6.2.32.1 System Design 6.2 14 6.2.32.2 PDMS Function 6.2-15 6.2.33 DELETED 6.2-15 6.2.33.1 DELETED 6.2-15 6.2.33.2 DELETED 6.2-15 6.2.34 DELETED 6.2-15 6.2.35 DELETED 6.2-15 6.2.35.1 DELETED 6.2-15 6.2.35.2 DELETED 6.2 15 vi UPDATE 2 - AUGUST 1997 v
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~ CHAPTER 6
( TABLE OF CONTENTS (Cont'd) x SECTION I[Ild PAGE 6.2.36 SUBMERGED DEMINERALIZER SYSTEM 6.2-15 6.2.36.1 System Design 6.2-15 6.2.36.2 PDMS Function 6.2-15 6.2.37 DELETED 6.2-15 6.2.38 OTSG RECIRCULATING SYSTEM 6.2-15 6.2.38,1 System Design 6.2-15 l 6.2.38.2 PDMS Function 6.2-16 6.2.39 DECONTAMINATION SERVICE AIR SYSTEM 6.2-16 6.2.39.1 System Design 6.2-16 6.2.39.2 PDMS Function 6.2-16 O V 6.2.40 DECON PROCESSED WATER SYSTEM 6.2-16 6.2.40.1 System Design 6.2-16 6.2.40.2 PDMS Function 6.2-16 6.2.41 SLUDGE TRANSFER AND PROCESSING SYSTEM 6.2-16 6.2.41.1 System Design 6.2-16 6.2,41.2 PDMS Function 6.2-17 6.2.42 DEFUELING WATER CLEANUP SYSTEM 6.2-17 6.2.42.1 System Design 6.2-17 6.2.42.2 PDMS Function 6.2-17 6.2.43 PLASMA ARC NITROGEN SYSTEM 6.2-17 vii UPDATE 2 - AUGUST 1997
CHAPTER 6 TABLE OF CONTENTS (Cont'd) h..- V SECFION .IIILE ZAfd; 6.2.43.1 System Design 6.2-17 6.2.43.2 PDMS Function 6.2-17 6.2.44 Reactor Building Sump Level System 6.2-17 6.2.44.1 System Design 6.2-17 6.2.44.2 PDMS Function 6.2 18 6.2.45 Reactor Building Air Sampling System 6.2 18 6.2.45.1 ' System Design 6.2-18 6.2.45.2 PDMS Function 6.2-18 6.2.46 Waste Disposal Liquid, Reactor Coolant System 6.2-18 6.2.46.1 System Design 6.2-18 O 6.2.46.2 PDMS Function 6.2-18 V 6.2.47 Secondary Plant Sampling System 6.2-18 6.2.47.1 System Design 6.2-18 6.2.47.2 PDMS Function 6.2-18 6.2.48 Makeup Water Treatment and Condensate Polishing System 6.2-19 6.2.48,1 System Design 6.2-19 6.2.48.2 PDMS Function 6.2-19 6.3 DEACTIVATED SYSTEMS 6.3 1 6.3.1 AUXILIARY STEAM SYSTEM 6.31 6.3.2 -- DELETED 6.3-1 viii UPDATE 2 - AUGUST 1997 l O ! U .
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CHAPTER 6 O G TABLE OF CONTENTS (Cont'd) {ECTION TITLE .PAGE 6.3.3 DELETED 6.3-1 6.3.4 DELETED 6.3-1 6.3.5 DELETED 6.3 1 6.3.6 DELETED 6.3-1 6.3.7 DELETED 6.3 1 6.3.8 INSTRUhfENT AIR SYSTEM 6.3-1 6.3.9 DELETED 6.3-1 6.3.10 DELETED 6.3-1 6.3.11 DELETED 6.3-1 6.3.12 DELETED 6.3-1 6 3.13 DEMINERALIZED SERVICE WATER SYSTEM 6.3-1 6.3.14 DELETED 6.3-2 6.3.15 NUCLEAR SERVICES RIVER WATER SYSTEM 6.3-2 6 3.16 REACTOR BUILDING NORMAL COOLING WATER 6.3-2 SYSTEM 6.3.17 REACTOR BUILDING PENETRATIONS FORCED AIR 6.3-2 COOLING SYSTEM 6.3.18 DELETED 6.3-2 6.3.19 CIRCULATING WATER SYSTEM 6.3-2 6.3.20 DELETED 6.3-2 6.3.21 ENVIRONMENTAL BARRIER SYSTEM 6.3-2 ix UPDATE 2 - AUGUST 1997
CHAPTER 6 TABLE OF CONTENTS (Cont'd) O o SEC110N TITLE PAGE 6.3.22 PENETRATION PRESSURIZATION SYSTEM 6.3-3 6.3.23 GLAND STEAM SYSTEM 6.3-3 6.3.24 DELETED 6.3-3 6.3.25 DELETED 6.3-3 6.3.26 DELETED 6.3-3 6.3.27 DELETED 6.3-3 6.3.28 DELETED 6.3-3 6.3.29 DELETED 6.3-3 6.3.30 DELETED 6.3-3 6.3.31 DELETED 6.3-3 ( 6.3.32 DELETED 6.3-3 6.3.33 DELETED 6.3-3 6.3.34 DELETED 6.3-3 6.3.35 DELETED 6.3-3 6.3.36 DELETED 6.3-3 6.3.37 DELETED 6.3-3 6.3.38 DELETED 6.3-3 6.3.39 DELETED 6.3-3 6.3.40 TEMPORARY NUCLEAR SERVICES CLOSED 6.3-3 COOLING SYSTEM 6.3.41 DELETED 6.3-3 6.3.42 DELETED 6.3-4 ( x- UPDATE 2- AUGUST 1997
CilAPTER 6 TAHLE OF CONTENT!, (Cont'd) SECTION TITLE Ed.91 6.3.43 DELETED 6.3-4 6.3.44 DELETED 6.3-4 6.3.45 DELETED 6.3-4 6.3.46 DELETED 6.3-4 6.3.47 DELETED 6.3-4 6.3.48 DELETED 6.3-4 6.3.49 DELETED 6.3-4 6.3.50 EARTilQUAKE DETECTION SYSTEh1 6.3-4 6.3.51 REACTOR COOLANT PUhiPS hiOTOR OIL DIU.!N 6.3-4 SYSTEhi 6 3.52 DELETED 6.3-45 6.3.53 DELETED 6.3 5 6.3.54 DELETED 6.3 5 6.3.55 DELETED 6.3 5 6.3.56 POLAR CRANE 6.3 5 6.4 DELETED 6.4-1 6.5 SYSTEhi REFERENCES 6.5-1 xi UPDATE 2 - AUGUST 1997
4 i { CHAPTER 6 il l TABLE OF CONTENTS (Cont'd) . l ! . LIST OF TABLES i 4 j TABLE NO. IIILE i FAGE : 1 ! 6.11 DEACTIVATED FACILITIES 6.1-4 i i i ) i l 6.2 1 DEACTIVATED PASSIVE SYSTEMS 6.2 20 j 6.31. DEACTIVATED SYSTEMS 6.36 l i , ! 6.5 SYSTEM REFERENCES 6.51 l l ! 2 2 . 'I
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CilAPTER 6 p i ) V' DEACTIVATED SYSTEMS AND FACILITIES
6.0 INTRODUCTION
This chapter describes those systems and facilities which have been deactivated for PDhtS. There ne two categories of deactivated system a and facilities: 1) descurated systems and facilities with passive PDhtS functiord and 2) deactivated systems and facilities. Le first category consists of those systems or facilities which have been deactivated but provide the passive function (s) during PDhtS o contamination control and/or containment isoL: tion. No effort will be expended to maintain the design fimetional capability of these systems and facilities. However, the passive function (s) of the affected systems or facilities will be maintained throughout PDMS to provide reasonable assurance that TMI 2 can be maintained in the PDhtS condition with no nsk to the health and safety of the public. He passive function of containment isolation will be maintained as required in the PDMS technical specifications section 3.1.1.1, primary containment isolation. Tbc passive function of contamination control will be maintained by adherence to the requirements of 1000 PLN-4010.01 (GPUN Corporation Padiation Protection Plan) and 1000 PLN 7200.04 (PDMS Quahty Assurance Plan). The second category consists of those systems and facihties which are deactivated because they senc no active or passive function during PDMS. No maintenance is required and no attempts will be made to
/ 'g preserve or maintain these systems and facilities.
NJ' Tables 6.1 1, 6.2 1, and 6.3 1 provide a listing of those facilities and systems which will be deactivated durmg PDMS. %ese tables also provide other relevant status information for the listed facilities and systems. Equipment, components, and parts may be removed from systems and facilities designated as deactivated, and used for other purposes, provided their removal does not adversely affect the PDMS function of the system (s) or facilities involved, his may include complete system dismantlement, cornponent removal for use elsewhere, and possible conversion of portions of systems for other uses. Similarly, equipment, components, and parts may be removed from systems and facilities designated as deactivated with PDMS passive functions prosided, the passive function is not compromised or prosided, decontamination activities have negated the need for the passive function of contamination control. A detailed original design description of deactivated facilities and systems may be found in the TMI 2 FSAR and/or system operating descriptirn books. 6.0-1 UPDATE 2 - AUGUST 1997
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6.1 DEACTIVATED FACILITIES l Table 6.1 1 provides a listing of deactivated facihties for TMI 2. Also listed are the status ofinternal contamination and relevant remarks regarding the fmal la.wp of the facility. He following sections ' l address the facihty desenption and the PDMS function of the facility. 6.1.1 DELETED 6.1.2 DELETED 6.1.3 CONTAINhtENT AIR CONTROL ENVELOPE The CACE was installed during the cleanup period and provided space to mobilize equipment and materials needed to support the in-contauunent activities through defueling Location of the CACE at the equipment hatch allows equipment and matenals to be moved into and out of the Containment Building with a minimum of 6fficulty through the equipment hatch airlock doors. The CACE sen ed as an aid in the control of the spread of contanunation and airbome radioactivity during those tmies when the airlock doors were opened. During PDhtS, this facility serves no active function and as such is designated deactivated. Certain types of RD entries require the CACE to be in a closed condition; thus, the CACE senes a passise (~'T funcson. Q 6.1.4 CIRCULATING WATER PUh!P 110USE The Circulating Water Pumn 11ouse contains the six circulating water pumps and their controls which provide a flowpath from the circulating water fiume to the main condenser. During PDh;S, this facility serves no active or passive function and as such is designated deactivated. The contents of this facility are being dismantled. 6.1.5 CIRCULATING WATER Cl!LORINATOR 1100SE ne Circulating Water Chlorinator llouse contained the chlonne evaporator, chlorinator aJ ejector and storage tanks in ad& tion to the sulfuric acid storage tank This facility supported the che-ical treatment of the Circulating Water System. All equipment and hazardous materials (chlotate, sulfuric acid)lmv been rernoved. During PDhtS, this facility sen es no active or passive function and as such is designated deactivated. 6.1 1 UPDATE 2 . AUGUST 1997 O i V
6.1.6 NATURAL DRAIT COOLING TOWERS Two cooling towers were pro ided as a heat sink for the plant circulating water system Dunng PDMS, ther.c towers se ve no active or passive function and as such are designated deactivated Therefore, the combustible matenal from this facility has been removed The aircraft warmng lights will remain opetational. 6.1.7 MECilANICAL DRAFT COOLING TOWER The original design of the Mechanical Draft Cooling Tower was to remove the heat added by the Senice Cooling Water Systems before it was eturned to the nyer. This facility has been dismantled and the combun.ble material has been removed, only the concrete basin ani parnp house remain. 6.1.8 DELETED 6.1.9 TENDON ACCESS GALLERY The Tendon Access Gallery provida! access for initial positioning and tighterung of the Tendon Post Tensioning System during construction The Tendon Post Tensioring System is grouted and as such does not require periodic torquing During PDMS, this facility serves the passive function of contamination control and hou',ing an opettlional sump pump 6.1.10 RIVER WATER PUMP 110USE The River Water Pump flouse proiided a structure for intake water supplied to the vanous nuclear and senice water systems and the Unit 2 diesel fire pump (FS P 1) in the adjacent Fire Pump flouse structure. Dunne PDMS, these facilities and the systems within these facihties serve no active or passive function and as such are designated deactivated The contents of this facility are being distr.antled and the combustivle matenal removed Following this dismantilement, the 312' elevation of the River Water Pump llouse is being convened to serve as a facility for the storage and testing of various non-radiological robotic equipment and mock ups. Use of this area for this purpose has no effect on the deactivated status of this facility as related to PDMS. 6.1 2 UPDATE 2 - AbGUST 1997
i Ch
's] 61.11 DWST PIPE CHASE The llorated Water Storage Tank (BWST) Pipe Chase is an underground tunnel extendmg frorn the BWST into the Auxiliary Building on the east side. It enclosed piping for the Decay lleat Removal and Duilding Spray Systerns. During PDMS, this area serves the passive function of contanunation control.
6.1.12 CONTROL DUILDING (M 20) AREA EAST 1hc Control fluildmg Area is the plant area below elevation 30$' between the Turbine, Reactor, and Senice Buildings. The cast portion is separated from die west by a bamer wall and houses the motor-driven Emergency Steam Generator reed Pumps. This area prosides access to the Tendon Access Gallery on the east side and tho Control and Senice Duildings from the Turbine Building. During PDMS, this area serves the passive functions of housing containment isolatiora and contamination control. It also contains a portion of the cork scam monitoring system. The contents of this facility are being dismantled. 6.1.13 CONTROL DUILDING (M 20) AREA WEST 1he Control Duilding Area west portion houses the tmbine<lriven Emergency Steam Generator Feed Pump, Main Steam Isolations, Relief and Atmospheric Vent Valves, the Control Building Area Sump, and Unit Substations 2 34 and 2 44. This area also provides access to the Tendon Access Gallery on the west side. O During PDMS, this area serves the passive functions of housing containment isolations and contamination contro!. Ilowever, one sump pump and a portion of the cork seam monitoring system will remain operational The contents of this facility are being dismantled. 6.1.14 DELETED 6.1-3 UPDATE 2 - AUGUST 1997 V
TABLE 6.1-1 DEACTIVATED FACILITIES PDMS INTERNAL FACILITY DESCRIPTION FUNCTION CONTAMINATION ISOLATION REMARKS DELETED CACE PASSIVE YES YES Residual fixed contamination in some floor arcas Circulating Water Pump flouse NONE NO NO Circulating Water Chlorinator NONE NO NO Ilouse Natural Dran Cooling Towers NONE NO NO Aircran warning lights are operational Mechanical Dran Cooling Tower NONE N/A N/A Dismantled. DELETED Temion Access Gallery Sump PASSIVE YES YES Residual contamination in sump; one pump operational. DELETED 6.1-4 UPDATE 2 - AUGUST 1997 O O O
TABLE 6.1-1 (Cont'd) DEACTIVATED FACILITIES PDMS INTERNAL CONTAINMENT FACILITY DESCRIPTION FUNCTION CONTAMINATION ISOLATION REMARKS
- BWST Pipe Chase PASSIVE YES NO Residual contamination in pipe chase and Decaylicat Removal piping.
Control Building (M-20) Area East PASSIVE YES YES Residual contannnaten in access trunk to tende access gallery. Cork Scam Monitoring Sym Control Bu;1 ding (M-20) Arca West PASSIVE YES YES Residual contamination in main steam " lines and control building area sump; one sump pump operational Cork Scam Monitoring Sy m DELETED i 4 1 0 6.1-5 UPDATE 2- AUGUST 1997 i i 4
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6.2 DEACTIVATED PASSIVE SYSTEMS [ "Ris section describes those systems which have been deactivated but provide some passive function i durmg PDMS. No effort will be expended to maintain the design function of these systems. He identified passive function or functions of the afTected systerns will be maintained through the PDMS configuration of the TMI 2 facility. Equiprnent, components, and parts may be removed from these systems provided the passive function is not compromised or provided decontamination and dismantlement activities have negated the need for the passive function. Removal may involve dismantlement, .:.ap recovery or convasion of components or systems to other uses in support of dismantlement or in support of Unit I activities. Table 6 2 1 provides a listing of deactivated passive systems for the TMI-2 facility. Also listed are the system code, status of containment isolation, whether the system has interral contamination. and relevant remarks regardmg the fmal layup of the systein. Each of the following sections addresses the original design function of the system and its PDMS passive function or functions. Additional reference information is listed in Section 6.5. 6.2.1 MAIN AND REllEAT STEAM SYSTEM 6.2.1.1 System Design ne main steam piping was originally designed to deliver main steam from the steam generators to the high pressure turbine. It was also designed to provide main steam to the steam generator feed pump turbines, emergency steam generator feed pump turbine, second stage reheaters of the moisture separator icheaters, turbine bypass vahrs, and the turbine gland seal system. Ac pnmary function of the Reheat Steam Piping System was to deliver reheat steam from upstream of xj the high pressure turbine to the moisture separator-reheaters to reheat the high pressure turbine exhaust steam It also provided scheat steam to the steam generator feed pump turbines. The Main and Reheat Steam piping system and components are being dismantled. 6.2.1.2 PDMS Function This Main and Reheat Steam System provides no active function dunng PDMS. The passive functions provided by this system during PDMS are Containment isolation and contammation control. He Containment isolation function is provided by maintaining valves MS VI A/B, MS V2A/B, MS V4A/B, MS Vil A/B, MS-V7A/B, MS V15A/B, MS V50A/B, MS V51 A/B, MS V224, MS V225, MS V226, and MS V227 in the closed position. 6 2.2 PRIMARY NUCLEAR PLANT HYDROGEN SUPPLY SYSTEM 6.2.2.1 System Design The Nuclear Plant flydrogen Supply System was designed to store nuclear grade hydrogen and supply it at a reduced pressure to the make-up tank in the Aunliary Buildmg. Hydrogen was used in the pnman coolant to reduce the concentration of free oxygen. 6.2 - 1 UPDATE 2 - AUGUST 1997 p G
6 2.2 2 PDhtS Function This sy stem provides no active function during PDhtS. The passive function prosided by this system during PDhtS is contamination control. All hydrogen bottles have been removed. 6.2.3 FUEL llANDLING AND STORAGE SYSTEh! 6.2.31 System Design Ihc Fuel liandling and Storage System was designed to provide the capabihty of receivmg new fuel assemb;ies, storing new and spent fuel assemblics, delivenng new fuel to the reactor core, removing spent fuel from the reactor cose, rearranging fuel within the core and transferring spent fuel from the site. His systern was nubfied and utilized during the cleanup period to receive fuel canisters and transfer them from the Reactor Vessel to the Spent Fuel Pool for eventual transfer off site. 613.2 PDhtS Function This syrtem provides no active function during PDhtS. He passive functions provided by this system during PDhiS are Containment isolation and contamination control. He Containment isolation function is provided by blind flanges instalied downsticam of valves Fil VI A/B with test valves Fil-VIC/D on the flanges in the closed position Contamination within Spent Fuel Pools (SFP) "A" and "B" is controlled by a sheet metal cover installed above each SFP. The covers consist of steci roof decking supported by structural steel beams stmnc across each SFP. Included in each cover is a filter to contain the spread of contamination from the SFP and an access port to allow personnel access to the inside of the SFP. 6.24 STANDBY REACTOR COOLANT PRESSURE CONTROL SYSTEh! 6.2 41 System Design ne Standby Reactor Coolant Pressure Control System was designed and installed during the TMI 2 cleanup period for maintaining RCS volume and pressure control. 6 2.4.2 PDhtS Function his system provides no active function during PDh15. The passive function provided by this system during PDNtS is contamination control. 6.2-2 UPDATE 2- AUGUST 1997 O
6.2.5 MINI DECAY HEAT REMOVAL SYSTEM 6.2.5.1 System Design l De Mini Decay Heat Removal (MDHR) System wu designed and installed during the TMI 2 cleanup j period. The MDHR systeht was designeo to remove heat from the Neactor Coolant System by ! forced circulation through the core, provide a means for sampling the Reactor coolant System and l control ambket temperature and airborne contamination levels in the pump and best eschanger j vauk. 6.2.5.2 PDMS Function ! nis system provides no active function during PDMS. ne passive function provided by this system during PDMS is contamination control. This system was never actually used durms the cleanup period, i 6.2.6 SPENT FUEL COOLING SYSTEM , 6.2.6.1 System Design ; i ne Spent Fuel Cooling System was designed to remove the decay heat generated by the spent fuel
- stored within the fuel pools and purify the water in the Fuel Transfer Canal and the fuel storage pools.
Other functions of the system were as follows:
- a. Filled and drained the Fuel Transfer Canal.
4
- b. Purified the water in the Fuel Transfer Canal during refueling operations. 1
- c. Purified the water in the Borated Water Storage Tank after a refueling. ,
e i
- d. Drained and filled the storage pools when required.
- e. Purified the water in the RCS while in the decay heat removal mode of operation.
- f. Recovery flow path for DWCS FTC 'A' SFP Clean Up System.
6.2.6.2 PDMS Function : his system prmides no active function during PDMS. He passive functions provided by this system dunng PDMS are Containment isolation and contammation control. He Containment isolation , function is provided by maintaining valve SF-V105 in the closed position. t I 6.2-3 UPDATE 2 - AUGUST 1997
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1 1 l 6.2,7 REACTOR COOLANT MAKEUP AND PURIFICATION SYSTEM j 6.2.7.1 System Design The Makeup and Purification System was designed to provide a means for controlling the reactor coolant imentor) durinE reactor power operations as well as maintaining the water quality and chemistry of the coolant mthin presenbed specifications. The system also served to accomplish the following:
- a. Provided seal injection water to the Reactor Coolant Pumps to establish a pnmary coolant pressure boundary and to supply pump coohng water,
- b. Provided a means of venting radioactive and flanunable gases from the RCS.
- c. Added rnakeup water to the core flooding tanks.
- d. Served a safety features function by injecting high pressure water into the RCS in the event of a LOCA.
- e. Provided an indication of failed fuel 6 2.7.2 PDMS Function This system provides no active function during PDMS, The passive functions provided by this system during PDMS are Containment isolation and contamination control. The Containnent isolation function is provided by rnaintaining valves MU Vl6A/II/C/D, MU V366, MU V368, MU Y376, MU V377, MU Vl8, MU V379, MU V31$, MU V316, MU V380, MU V381, MU V382, MU-V383A/11/C/D, and MU V384A/B/C/D in the closed position.
6.2 - 4 UPDATE 2 - AUGUST 1997 9
l 6.2.8 DECAY IIEAT REMOVAL SYSTEM l 6.2.8.1 S 3stem Design he Decay lleat Removal system was designed to proside the following functions:
- a. Remove core decay heat after the reactor coolant had reached the minimum temperature possible with condensate and feedwater cooling (250'F) lleat was removed from both the core and pressurizer.
b Fill, recirculate, punfy (via the spent fuel systern), and drain the Fuel Transfer Canal for refuelmg.
- c. Minimize the consequences of a loss of coolant accident in the following manner:
- 1. By injecting borated water and sodium hydroxide solutions into the core at a low reactor pressure.
- 2. Ily providmg long term cooling after a LOCA by recirculating water from the Reactor Building sump to the core.
- 3. By supplying the suction of the high pr ssure injection makeup pumps for long-term cooling after a LOCA.
c
- 4. By supplying the buildmg spray pumps with water from the BWST, the sodium hydroxide tanks, or from the Reactor Building sump-d= Circulating the contents of the BWST tank for midng and sampimg.
e Serve as a Recovery flow path for borated water to the defueling water clean up system 6.2.8.2 PDMS Function This system provides no active function during PDMS. He passive functions provided by this system dunng PDMS are Containment isolation, contamination control, and RB sump level and pumpdown flow path. The Containment isolation function is provided by maintaining valves DH V3, Dil V4A'B, Dil V6A/B, Dil Vl87, Dil V205, and Dil V225 in the closed position 6.2.9 REACTOR BUILDING LEAK RATE TEST SYSTEM 6.2.9.1 System Design he Reactor Buildmg leak rate test system was designed to verify the leak tightness of the Containment. (O) 6.2 5 UPDATE 2 - AUGUST 1997
l 6.2.9.2 PDMS Function his sptem serves no active function during PDMS. He passive function provided by this system during PDMS is Containment isolation which is ensured by installed blind flanges on penetrations R 571 A and R S?)D a1x! a welded cap on R-571B. 6.2.10 SERVICE AIR SYSTEM l l 6.2.10.1 System Design l l ne Semcc Air was piped to quick-disconnect hose coanections throughout the plant and was used by I the dernincratizer systems for resin mixing and transport and other pneumatic equipment. Air for these ! purposes was oil free, but not specially dned. Breathing air could be taken from the Semce Air System at any of the outlets if the proper adapters were used. 6.2.10.2 PDMS Function his system provides no active function during PDMS. However, portions of the Semce Air System are utilized by the Compressed Air Systern (see section 7.2.6 4). He passive function provided by this system during PDMS is Containment isolation. He Containment isolation function is prosided by maintaining valve SA V20 in the closed position. 6 2,11 CllEMICAL ADDITION SYSTEM 6 2.11.1 System design ne Chemical Addition System was designed to perfonn the follomng functions:
- a. Dissolve chemicals in demineralized water, b Store dissolved chemicals.
- c. Provide positive manual control of the transfer of chemicals to the appiopnate system in the required quantity and concentration.
- d. Provide a means of sampling all chemical concentrations.
6.2.11.2 PDMS Function his system provides no active function during PDMS. He passive function provided by this system during PDMS is contamination control. 6 2.12 REACTOR BUILDING EMERGENCY SPRAY SYSTEM 6 2.12.1 System Design The Reactor Building Emergency Spray was designed to cool the Reactor Building atmosphere following a RCS piping rupture, thereby effecting a pessure reduction within the buildmg, and consequently, miniminns the potential leakage of radioactivity from the building to the site and surrounding areas. He spray system also functioned to remove radioiodine from the Reactor Building atmosphere by chemical reaction and to wash suspended particulate radioactivity out of the Reactor Buildm; atmosphere. 6.2 - 6 UPDATE 2 - AUGUST 1997
1 I 6.2.12.2 PDhtS Function f'\ nis system provides no active function during PDhtS. However, RB pressure monitoring is provided
'd via DS Vl46,147, and 149 for RB pressure indication and a 0.5 psig Hi RB pressure alarm A portion of the R.B. spray system piping will be utilized as a fhw path for R.B. sump draining if required. The passive functions provided by this system during PDhis are Containment isolation and contamination control. De Containment isolation function is provided by maintaining valves BS V148, BS VIA/B BS V130A/B in the closed position.
6.2.13 NUCLEAR SAhiPLING SYSTEh! 6.2.13.1 System Design ne Nuclear Sampling System was designed to provide the capabihty to obtain representative liquid ard gas samples from nuclear systems. 6.2.13.2 PDhtS Function his system provides no active function during PDhtS. Ac passive functions provided by this system during PDhtS are Containment isolation and conta:nination control. The Containment isolation function is provided by maintaining valves CA V8, CA V9, and CA Y10 in the closed position. 6.2.14 NUCLEAR PLANT AND RADWASTE NfrROGEN SYSTEhtS 6.2.14.1 System Design O The Nuclear Plant and Radwaste Nitrogen System was designed to store and supply nuclear grade
's nitrogen at various pressures to several systems in the Reactor and Auxiliary Buildings and storage tanks in the yard.
6.2.14.2 PDhtS Function his system provides no active function during PDhtS. The passive functions provided by this system dunng PDhtS are Containment isolation and contxnination control. The Containment isolation function is provided by maintaining valve Nht V52 in the closed position. All nitrogen bottles have been removed. s [ T
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6.2-7 UPDATE 2 - AUGUST 1997
6.2.15 DECAY llEAT CLOSED COOLING WATER SYSTEh! 6.2.15.1 System Design The Decay IIcat Closed Cw!mg Water System (DilCCW) was designed to cool the Decay lleat Removal Cmlers, the Reactor Cmlant Drain Tank Leakage Cwlers, Decay lleat Removal Pump and hiators, and the Decay lleat Closed Cooling Water Pump hiotor. Tne DliCCW system provided a barrier between the reactor emlant and the Nuclear Scruces River Water Sy stem to prevent the release of radioactisity to the emironment. During the recovery period a portion was used as a return flow path for the DWCS FTC/"A" S F. Pool Clean Up System. 6.2.15.2 PDhtS Function This system provides no active function during PDhtS. The passive functions prouded by this system during PDhtS are Containment isolation and contamination control. The Contaiment isolation function is prmided by maintaining valves DC-V103, DC VI15, and DC V137 in the closed position. 6.2.16 RADWASTE DISPOSAL, WASTE DISPOSAL GAS SYSTEh! 6 2.16.1 System Design 1he Radwaste Disposal, Waste Disposal Gas System was designed to provide a means to collect potentially radioactive gas from components and tanks in the plant, compress and dehver this gas to the waste gas decay tanks, store the gas for decay, and recycle or release the gas through the unit vent at a controlled rate within the linuts of10 CFR 20. During the recovery period a portion of the system was utilized as a flow path for the DWCS Reactor Vessel Clean Up System 6 2.16.2 PDhtS l unction This system provides no active function during PDhtS. The passive functions provided by this system during PDhtS ate containment isolation and contanunation control. The containment isolation function is provided by maintaining WDG V 199 in the closed position. Contamination within the system is controlled by the installed air filters for the contaminated portions of the system. 6.2.17 REACTOR HUILDING EhiERGENCY COOLING WATER SYSTEh! 6.2.17.1 System Design 11e Reactor Buildmg Emergency Cmling Water System us designed to provide cmbng water to the Reactor Buildmg cooling units in the event of a LOCA. 6.2 - 8 UPDATE 2 - AUGUST 1997 l
6 2.17.2 PDMS Function
- p} This system provides no actne function during PDhlS. %e passive function provided by this system
(/ during PDMS is Containment isolation. We Containment isolation function is provided by maintaming valves RR V5A/B/C, RR V6C/D/E, RR VI1 A/B'C/D/E, RR Y25 A/B/C/D/E, RR V28 A/B/C/D'E, RR V75 A/II/C/IL'E, RR V86, RR V88, IWV90, RR V92, RR Y94, RR Y96, and IWV98 in the closed position. 6.2.18 INTERMEDIATE CLOSED COOLING WATER SYSTEM 6.2.18.1 System Design he hitermediate Closed Cooling Water System was designed to provide coohnE water to the following equipment located mside the Reactor Building. a ne Makeup and Purification Synem Letdown Coolers, MU C 1A and IB.
- b. He mechanical seal area and cooling jacket for each of the four Reactor Coolant Pumps.
- c. He stator coil cooler for each of the 69 control rod drive mechanisms.
- d. He Steam Generators llot Dain Cooler.
6218.2 PDMS Function
/O nis system prmides no active function during PDMS. The passive functions provided by this system
\ danng PDMS are Containment isolation and contamination control. The Containment isolation function is providoc' by maintaining valves IC V3, IC V4, IC V5, and IC-V207 in the closed position 6.2.19 FUEL POOL WASTE STORAGE SYSTEM 6.2.19.1 System Design his system was designed and installed during the TMI 2 clear.up penod it provided a temporary storage facility for radioactive liquid wastes from the Reactor Building Sump and the Miscellaneous Waste lloldup Tank without contammating the Fuel Storage Pool. It also provided the capability to transfer liquid wastes from one set of stomge tanks to another or to the Auuliary Buildmg Emergency Cleanup System. Most of this system has been removed. The only remaining portion is WG P 1 (flow path from RB sump) and its associated piping. This portion was never used during the cleanup period.
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(") 6.2-9 UPDATE 2 - AUGUST 1997
6219.2 PDMS Tunction T1us system provides no active function during PDMS. The passive function provided by this system dunng PDMS are containtnent isolation and contamination control. The containment isolation function is prosided by main'aining SWS I'V 1 in the closed position with the penetration blind flange installed 6 2.20 TEMPORARY NUCLEAR SAMPLING SYSTEM 6220.1 System Design This system was designed and mstalled during the TMI 2 cleanup penod The temporary nuclear sampling system prosided representative liquid and gas samples from selected points containing post accident waste. 62.202 PDMS Tunction "Dtis system prosides no active furiction dunng PDMS. The passive function prosided by this system dunng PDMS is contamination control. 6.2.21 OTSO CilEMICAL CLEANING SYSTEM 6 2.21.1 System Design The OTSO Chemical Cleaning System was designed to chemically ternove corrosion products and contarnmants from selected portions of the Condensate, Teedwater, Emergency Feedwater and Main Steam S) stems prior to operation, and from the Steam Ocnerators as needed during the life of the plant. The Feedwater, Condensate, and Main Steam Chemical Cleaning System used temporary piping and cquiptnent connected to the cdsting system piping. During the recovery period a portion of the chemical cleaning piping was utilized for support systems, Decon Senice Air, Temporary Decon Water and the Fue! Transfer Canal Fill and Drains System. 6 2.21.2 PDMS Function This system prosides no active function dunng PDMS. The passive functions prosided by this system danng PDMS are Containment isolation and contammation control. The Contamment isolation functmn is provided by maintainmg valves in the Decon Senice Air System, the fuel trantfer canal fill and drains system and the Temporary Decon Water System DSA V-004, DSA V 006, PW WO69, PW. V-099, TDW V 001 and TDW V 003 in the closed position. 6.2 - 10 UPDATE 2 - AUGUST 1997
6.2.22 CORE FLOODING SYSTEht i ] J 6.2.22.1 System Design
'Ibe Core Fkolmg System was designed to flood the Reactor Vessel core with borated water in the event of a major loss of coolant accident.
During the accovery period a portion of the Core Fkaling System was utilized as a sainpling flow path for the DWCS Reactor Yessel Clean-up System. 6.2.22.2 PDh15 Function
'Ihis system provides no active function during PDN15. The passive functions prmided by this system during PDh!S are Containment isolation and contamination control. The Contamment isolation function is prmided by maintaining valves CF VI 14 A/B, CF V144, CF V145, CF-Y146, and CF V129A'D in the closed position.
CFT 1 A is being used for storage of LCSA pieces that were cut up and removed from the Reactor Vessel, this tank is vented to the atmosphere via an installed filter. 6.2.23 FEEDWATER SYSTEh! 6.2 23.1 System Design The Feedwater System was designed to supply feedwater to the steam generators in addition. the system raised the temperature of the feedwater before entry into the steam generators. 6.2.23 2 PDhtS Function This system prmides no active function dunng PDh!S. The passive functions provided by this systern during PDhtS are Containment isolation and contamination control. The Contamment isolation function is provided by maintaining valves FW Vl7A/D, F\WVl9A/B, FW V35 A/B, and IYV68A/D, in the closed position. 6.2.24 NUCLEAR SERVICES CLOSED COOLING WATER SYSTEh! 6 2.24.1 System Design
'this system was designed to supply normal and emergency cooling water through various plant equipment and system heat exchangers to remove heat generated by operation The heat was subsequently transferred to the Nuclear Senices River Water System.
!d' 6.2 - 11 UPDATE 2 - AUGUST 1997
6.2.24.2 PDMS Function his system prmides no active function during PDMS. The passive function provided by this system during PDMS is Containment isolation ne Containment isolation function is prmided by maintaining valves NS V72, NS V81, and NS V210 in the closed position 6 "I IIYDROGEN RECOMDINER 6 2.25.1 System Design flydrogen accumulation in the Containment following a LOCA was controlled by the post-LOCA hydrogen recombiner with a purge to atmosphere through an appropriate filter train provided as a backup. After the accident, the two post LOCA hydrogen recombiners were transferred to TMI Urut 1. Only portions of the piping sy stem presently remain. 6 2.25.2 PDMS Function his sptem prmides no active function during PDMS. He passive function provided by tlus system dunng PDMS is Contairunent isolation ne Containment isolation function is provided by traintairung valves All V90A/H, All V120A/B, AII Vl49, and All V151 in the closed position 6 2.26 EMERGENCY FEEDWATER SYSTEM 6 2.26.1 System Design ne Ernergency Feedwater Systern was designed to provide a limited amount of feedwater to the steam generators on loss of all reactor coolant pumps or feedwater purnps. 6.2.26.2 PDMS Function his system provides no active function during PDMS. He passive functions provided by this system during PDMS are Containment isolation and contamination control. He Containment isolation function ;s provided by maintaining valves EF V12A/H, EF-V33A/B, and EF V36 in the closed position. 6.2.27 REACTOR COOLANT SYSTEM 6.2 27.1 System Design The RCS consists of the Reactor Vessel, two 5ertical Once nrough Steam Generators, four shaft scaled reactor coolant pumps, an electncally heated pressurizer and interconnecting piping. The system, located entirely within the Containment, is arranged in two heat transport loops, each with two reactor coolant pun;ps and one steam generator. The reactor coolant was transported through piping connecting the Reactor Vessel to the steam generators and flowed downward through the steam generator tubes, transferring heat to the stearn and water on the shell side of the steam generator. The reactor coolant was returned to the reactor through two lines, each contaming a reactor coolant pump. 6.2 -12 UPDATE 2 - AUGUST 1997
6.2.27.2 PDhtS Function his system prmides no active function during PDhtS. De passive function provided by this system during PDMS is contamination control. Contamination within the system is controlled by the installation of a contamination barrier over the open Reactor Vessel. His barrier consists oflead shieldmg installed above the shielded work platform to reduce radiation levels emanating from the Reactor Vessel and a moisture / contammation barrier. In addition, the RCS has been drained to the extent practical; less than 10 gallons of water remain in the Reactor Vessel and less San 100 gallons remain in each OTSO (meluding blegs). Boron impregnated glass shards were also added to the Reactor Vessel (Ref. Section 4 3.5). 6.2.28 SOLID RADWASTE DISPOSAL SYSTEh! 6.2.28.1 System Design ne Solid Radwaste Disposal System was designed to store and transfer resm, radioactive liquid wastes, and concentrated boric acid for processing, packaging, and subsequent transportation to a disposal site. During the recovery period portions of the Solid Radwaste Disposal Sy: tem were used in the sludge aansfer and processing system Refer to Section 6.2. 6.2.28.2 PDhtS Function his system presides no active function dunng PDhtS The passive fanction provided by this system dunng PDhtS is contamination control. 6.2.29 REACTOR TEEDWATER llEAT SYSTEhi 6.2.29.1 System Design ne Reactor Feedwater licat System was designed and installed to provide a method to heat and add chemicals to the RCS water. He system was rendered obsolete before its construction was completed. The pump, heaters, and chemical addition equipment for the system have been removed. Only miscellaneous piping and valves remain. 6 2.29.2 PDhiS Function his system presides no active function during PDhtS. He passive function provided by this system dunng PDhtS is contamination control. A) (V 6.2 - 13 UPDATE 2 - AUGUST 1997
6.2.30 STEAM GENERATOR SECONDARY VENTS & DRAINS SYSTEh! 6.2.30.1 System Design ne Steam Generator Secondary Side Vents and Drains System was designed to dram the water from the secondary side of the steam generators. In addition, it provided a flow path for nitrogen to purEc all non-condensable gases from the OTSGs. It also prosided the flow paths for sampting and chemical cleaning of the OTSGs. 6.2.30 2 PDMS Function This system provides no active function dunng PDMS. He passive function presided by this system during PDMS is Contamment isolation ne Containment isolation function is provided by maintaining valves SV Vl8 and SV V55 in the closed position. 6.2.31 DEWATEIUNG SYSTEM 6.2.31.1 System Design ne Dewatering System was designed and installed during the TMI 2 cleanup period. This system removed and filtered the water from submerged defueling canisters and provided a transfer padi to the Defueling Water Cleanup System for processing. The Dewatermg System also provided the cover gas for canister shipping. The water was removed from the defueling canisters to reduce the weight of the canisters for shipping and prevent the precious metal catalysts from being submerged. A cover of argon was provided to reduce water intrusion when the canister was in the water, reduce air intrusion when the canister was out of the water, and reduce the pyrophoricity potendal of the debris within the canister. 6.2 31.2 PDMS Function his system provides no active function during PDMS. He passise function provided by this system dunng PDMS is contamination control. Contammation within the system is controlled by a cover over the "A" rpent fuel pool. 6.2 32 FUEL TRANSFER CANAL FILL AND DRAIN SYSTEM 6.2.32.1 System Design This system was designed and installed during the TMI 2 cleanup period it provided a means of suppl3ing borated water to the Reactor Duildmg from the borated water storage tank through a spent ft.el cooling pump (for high flow) or a sandpiper pump (for low flow) for filhng the Fuel Transfer Canal and the Internals Indexing Fiuure. In addition, this synem was used to distnbute processed water in the Reactor Duildmg. This system was installed, in part, because the normal Fuel Transfer Canal fill capabihty had been lost due to an inaccessible closed valve in the Reactor Buildmg. 6.2--14 LTDATE 2 - AUGUST 1997
I l 6.2.32.2 PDMS Function
\ Ris syrtern prmides no active function during PDMS. Le passive functions provided by this system during PDMS are Containment isolation and contarmnation control. ne Containtnent isolation function is provided by maintaining valves PW V069 and PW V099 in the closed position. Rese same valves provide the isolation function for the processed water storage and distnbution system.
6.2.33 DELETED 6.2.33 1 DELETED 6.2.33.2 DELETED 6.2.34 DELETED 6 2.35 DELETED 6.2.36 SUBMERGED DEMINERALIZER SYSTEM 6 2.36.1 System Design ihe SDS was a temporary liquid radwaste processing system designed to reconcentrate the fission products contained in the waters in the Reactor Buildmg sump, the RCS and liquid waste systems and reduce the fission products to levels acceptable for fmal treatment through the EPICOR 11 system. He SDS was installed during the TMI 2 cleanup period. f
\ 6 2.36.2 PDMS Function His system provides no active function during PDMS. He passive function provided by this system during PDMS is contamination control. The SDS system and "B" spent fuel pool were drained of water and the "B" spent fuel pool was covered.
6.2.37 DELETED 6.2.38 OTSG RECIRCULATING SYSTEM 6 2.38.1 System Design ne OTSG Recirculating System was designed and installed during the TMI 2 cleanup period. The primary function was to: ! 1. Remove radioactivity from steam generator RC H-1B (OTSG "B," due to a tube leak), and
- 2. Chemically treat water in both steam generators (RC H 1 A,1B) for wet-Qup condition.
6.2 - 15 UPDATE 2 - AUGUST 1997 U
6 2.38.2 PDhtS Function This system provides no ac;ive function during PDhtS. The passive functions prouded by this system are Contamment isolation and contamination control. The Containment isolation function is presided by inaintaimng vah es GR VI A, iB, GR V7A and 78 in the closed position. 6 2.39 DECONTAhtlNATION SERVICE AIR SYSTEh! 6.2.39.1 System Design The function of the Decontamination Senice Air Sy stem was 'o supply a source of air during the clean-up period (in addition to the existing senice air system) to:
- a. The Urut 2 Reactor Buildmg for support of work actinties
- b. The Urut 2 AFilB for operation of decontamination equipment.
6.2.39.2 PDh1S Function 1his system proudes no active function during PDhtS, lhe passive function provided by this system is Containment isolation which is provided by maintaining valves DSA V004 and DSA V006 in the closed position. 6.2 40 DECON PROCESSED WATER SYSTEh! 6.2.40.1 System Design The Decon Processed Water System ns designed during the cleanup period to proude a sufficient source of flush water for the sludge transfer and processing systerns and for decontananation in the AFilB and RR The NLil pump, piping, and valves and the Elliott pump were included in this system. l~wo centrifugal pumps, DPW P 12A and 128, a distribution header, and a valve manifold supplied the previously installed hose network. 6.2.40.2 PDhtS Function This system proudes no active function during PDh1S. The passive functmns provided by this system are Containment isolation and contamination control. Contaitunent isolation is provided by maintaming s alves DW V28, TT)W-V001, and TDW V003 in the closed position The NLB pump, the Elliott pump and hose distribution system have been removed. 6.2,41 SLUDGE TRANSFER AND PROCESSING SYSTEhi 6.2.41.1 System Design The Sludge Transfer and Processing System was designed during the cleanup period to proside for removal, processing, and solidification of sediment in the Reactor Building basement, Auxiliary Building Sump, and selected AFl!B tanks The system modified the Spent Resin Storage Tanks and portions of the Waste Disposal Solid system piping. 6.2 -16 UPDATE 2 - AUGUST 1997
l 6.2.41.2 PDhtS Function
~h This system provides no active function dunng PDhtS. The passive functions provided by this system
[V are Contatnrnent isolation and contammation control Containment isolation is provided by maintainmg valves WDS FV612 and WDS IT614 in the closed position. 6.2 42 DEFUELING WATER CLEANUP SYSTEh! 6.2 42.1 System Design he DWCS is a system which was designed dunng the cleanup penod and installed to clean two bodies of water.
- 1) %e fuel Transfer Canal / Spent fuel Pool Cleanup System was a temporary liquid processing system which was designed to process water contained in the SFP and/or the FTC.
- 2) ne Reactor Vessel Cleanup System was. temporary liquid processing system which was designed to process water contamed in the Reactor Vessel.
6.2 42.2 PDhtS Function nis system provides no active function dunng PDhtS. The passive functions provided by this system are Contamment ise'ation and contamination control. ne Containment isolation function is prosided by maintaining valves SF V105, CF-Vl45, CF V114B, CF V129B, DC V137, DC V103, Dil Y205, Dil Vl87, DWC-V038, DWC V040B, DWC V037, DWC V040A, WDG-Vl99, WDL V1092, O DWC V316, and DWC V318 in the closed position He above valves mclude other system designators V considered as part of the DWCS. 6 2,43 PLASH 1A ARC NITROGEN SYSTEh! 6.2 43.1 System Design ne Plasma Arc Nitrogen System was designed during the defuelu's period to proside a source of quenching gas during plasma are cuttmg operations of the Reactor Vessel internals. 6.2 43.2 PDhtS Function This system prmides no active function durmg PDh1S. The passive function provided by tius system is Contamment Isolation which is provided by maintainmg valves PAN V-005, 017 and 019 in the closed position, 6.2.44 REACTOR BUILDING SUh1P (LEVEL hiEASUREh1ENT) 6 2.44.1 System Design The Reactor Building Sump Level System was designed and installed during the clean-up period to measure the 8' of water which accumulated in the Reactor Building Basement. It utilized an instrument air purge (bubbler) and measured the back pressure of the air which corresponded to a height of water in the R.B. Basement.
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(m) 6 2 -17 UPDATE 2 - AUGUST 1997
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6.2.44.2 PDhtS Function his system provides no active function during PDMS. He passive functions provided by this system are contamination control and Containment isolation which is provided by maintaining valves RDS. IV.1009,1011,1013 and 1014 in the closed position. 6.2.45 REACTOR BUILDING AIR SAh!PLING SYSTEh! 6.2451 System Design ne Reactor Building Air Sampling System was designed to provide the capability of obtaining representative R B. air samples during normal and emergency conditions and Radiation monitoring of the ventilation exhaust air during purge operations. 6.2.45.2 PDMS Function his system provides no active function during PDMS. He passive function provided by this systern are contamination control and Containtncnt Isolation which is prosided by mamtaining valves All V. 143,145,168,169,170,171,227 and 230 in the closed position. For PDMS a new R.B. air sampimg system was designed and installed through R.D. personnel airlock A2. 6.2.46 WASTE DISPOSAL LIQUID,IGACTOR COOLANT 6.2.46.1 System Design ne Waste Disposal Liquid, Reactor Coolant System was designed ta receive and transfer reactor coolant from inside the Reactor Buildmg to locations outside the R.B. for storage and'or processing for reuse or disposal. 6.2.46.2 PDMS Function The system provides no active function dunng PDMS. %c passive functions provided by this system are Containment Isolation and Contamination Control. He Containment isolation function is prosided by maintaining vah es WDL.V.1092 and WDL.V.1125 in the closed position. 6.2.47 SECONDARY PLANT SAMPLING SYSTEM , 6.2.47.1 System Design The Secondary Plant Sampling System morutored the concentration ofimpunties and chemical additives in the water and steam which recirculated in systems in the secondary plant. During the clean-up period the sampling system was used to rnonitor OTSO levels and provide a flow path tur draining the OTSO's. 6.2.47.2 PDMS Function his system provided no active function for PDMS. %c passive function prosided by this system is contamination control. 6.2 - 18 UPDATE 2 - AUGUST 1997
- . - .. ..- - - _ - . _ - - _ _ - - . . . -.-. -. - . . . . - - - - - ~ - . -
6.2.48 MAKEUP WATER TREATMENT AND CONDENSATE POLISHING SYSTEM 6.2.48.1 Symem Dwign The makeup water treatment systern processed Susquehanna Rher water and pronded high purity dmuneralised water to the Domineralized Senice Water System, it also supplied the condensate pelishing symem which reduced the level of suspended and dissoh'od impunties in the condensate and , foodwater synems to acceptable levels. This systems is being dismantled. 6.2.48.2- PDMS Function This system provides no active function for PDMS. De passive function provided by tids system is , contamination control. ! f f 4 i I l Y 4 P 4 s a h 6,2 -l 9 UPDATE 2 - AUGUST 1997 n-,S e,c v m w wwN-,.--,m-w.,,--e-,.-, mrv.,- mmer--,,~,ses-e~w-e.,re~ - v&~-n--- ,o-v , - - vw.+.- ~ ~~..-u,-~n--,---- -,,--,ew------- ,- .e---w, ann.a w a , a r --
TABLE 6.2-1 DEACTIVATED PASSIVE SYSTDIS SYSTD1 CONTAINMENT INTERNAL SYSTEM DESCRIPTION CODE ISOLATION CONTAMINATION REMARKS Main and Rcheat Steam MS YES YES Primary Nuclear Pla n Ilpfrogen Supply 11Y NO YES All bottles ofgas removed from plant. FuelIIandling and Storage FIi YES YES Standby Pressure Control System SPC NO YES Mini Decay iIcat MDil NO YES Spent Fuel Cooling SF YE5 YES RC Makeup & Purification MU YES YES Decay IIcat Removal Dil YES YES See Defueling Water Ocan-up System RB Leak Rate Test LR YES NO Service Air SA YES NO Portions utilized by compressed air system. Chemical Addition CA NO YES See nucicar sampimg system 6.2 - 20 UPDATE 2 - AUGUST 1997 O O 9
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( l 1 t *) TABLE 6.2-1.(Cont'd) DEACTIVATED PASSIVE SYSTE3IS SYSTEh1 CONTAINMENT INTERNAL SYSTEM DESCRIPTION CODE ISOLATION COf ; AMINATION REMARKS Reactor Building Emergency Spray BS YES YES RB pressure signals are provided for RB pressure indication and Alarm. j l Nuclear Sampling System SN YES YES CA-VR, CA-V9, CA-V10 are considered t part ofnuclear sampling system. Nuclear Plant and Radwaste Nitrogen NM YES YES Decay lleat Closed Cooling DC YES YES See Defueling Water Clean-up System Waste Disposal Gas WDG NO YES See Defueling Water Clean-up System Reactor Bldg. Emergency Cooling Wtr RR YES NO Intermediate Closed Cooling IC YES YES j Fuct Pool Waste Storage WG YES YES SWS-FV-1 Containment Isolation Temporary Nuclear Sampling System SNS NO YES Capability to sample MWIIT maintained. OTSG Chemical Cleaning SGC YES YES Core Flooding CF YES YES Sec Defueling Water Cican-up System Feedwater FW YES YES Nuclear Services Closed Cooling Water NS YES NO 6.2 - 21 UPDATE 2 - AUGUST 1997 _ _ = _ _ _ __.
TABLE 6.2-1 (Cont'd) , DEACTIVATED PASSIVE SYSTEMS SYSTEM CONTAINMENT INTERNAL SYSTEM DESCRIITION CODE ISO 14 ATION CONTAMINATION REMARKS Ilydrogen Recombine- IIR YES YES Emergency Feedwater EF YES YES Reactor Coolant System RC NO YES Weste Disposal- Solid WDS NO YES WDS-FV-612 and 614 are considered part of Sludge Transfer System. RC FeedwaterIIcat RCF NO YES SG Secondary Vent & Drain SV YES YES Dewatering System DS NO YES Fuel Transfer Canal Fill & Drain FCC YES YES Fuct Transfer System FII YES YES DELETED Submerged Demineralizer System SDS NO YES 6.2 - 22 UPDATE 2 - AUGUST 1997 O O O
s 7s ..,,
%Y TABLE 6.2-1 (Cont'd)
DEACTIVATED PASSIVE SYSTEMS SYSTEM CONTAINMENT INTERNAL SYSTEM DESCRIPTION CODE ISOLATION CONTAMINATION REMARKS DELETED
' OTSG Recirculation .GR YES YES Decon Service Air DSA YES NO Decon Process Water DPW YES YES DW-V28 is considered part ofdecon Ic xessed water system.
Studge Transfer and Processing System STS YES YES Modified SRSTand WDS piping. Defueling Water Cleanup System DWC YES YES Plasma Arc Nitrogen PAN YES NO Reactor Building Sump (Level) RBS YES YES Reactor Building Air Sampling System All YES YFE Waste Disposal Liquid, Reactor Coolant - WDL YES YES See Defueling Water Clean-up System i Sceandary Plant Sampling SS NO YES Water Treatment and Condensate WT NO YES Polishing . 6.2 - 23 UPDATE 2 - AUGUST 1997 m _ ___- _
6.3 DEACTIVATED SYSTEMS h Q The section provides a description of those systems which are deactivated because they serve no active or passive function during PDMS. Table 6.3 1 lists the deactivated systems for the TMl 2 facility and their PDMS status. All deactivated systems require no maintenance and no attempts will be made to presen c no-maintain these systems. Equipment, components, and parts may be removed from these systems and facilities, and used for other purposes provided their removal does not adversely affect any PDMS function. Each of these systems is described in the following sections. 6.3.1 AUXILIARY STEAM SYSTEM he Auxiliary Ste tm System was designed to supply process steam to the following equipment:
- a. Reactor Coolant Evaporator
- b. DELETED
- c. Turbine Gland Seal Steam System
- d. Turbine Driven Emergency Steam Generator Feed Pump
- c. DELETED
- f. Unit 1/ Unit 2 Aux Steam Cross-Connect Piping.
6.3.2 DELETED 6.3.3 DELETED 6.3.4 DELETED O 6.3.5 DELETED h 6.3.6 DELETED 6.3.7 DELETED
-6.3.8 INSTRUMENT AIR SYSTEM The Instrument Air System was used throughout the plant for nuclear and non-nuclear instrumentation and controls, and for pneumatic devices where oil and moisture-free air was required. Portions of the instrument air system are incorporated within the compressed air supply system. For additional information see Section 7.2.6.4.
6.3.9 DELETED 6.3.10 DELETED 6.3.11 DELETED 6.3.12 DELETED 6.3.13 DEMINERALIZED SERVICE WATER SYSTEM The Demineraliacd Service Water System was designed to receiw demmeralized water from the Deminerahzmg System. The water was processed for oxygen removal, stored under a nitrogen blanket, and y distributed to senice as needed. He distribution headers supplied the Reactor Buildmg, Turbine Building, ' 6.3-1 UPDATE 2 - AUGUST 1997
Control and Senice Buildmg, the Auxiliary and Fuel Handling Building, and the Diesel Generator Building. During the clean-up period, a portion of the D.W. Sy stem was utdized in the Decon Process Water System (See Section 6.2.40). I he Radwaste Pumps Seal Water System, a subsystem of the Demineralized Senice Water System, provided seal water to pumps that handled contanunated fluid in either nonnal or abnormal operations. During PDMS, tlus water is supplied by the Domestic Water System. 6 3.14 DELETED 6.3.15 NUCLEAR SERVICES RIVER WATER SYSTEM The Nuclear Senices River Water System (NSRW) supphed cling water for all nuclear related and fuel handimg requirements and various HVAC senices. He system took suction from river water and retumed the water to the river via the Mechanical Draft Cooling Tower. The NSRW equipment in the Screenhouse has been removed. 6.3.16 REACTOR BUILDING NORMAL COOLING WATER SYSTEM ne Reactor Building Normal Cooling Water System was designed to circulate treated water through the Reactor Building air cooling units and to remove the heat thus transferred to the water by passing it through an evaporative coolet. 6.3.17 REACTOR BUILDING PENETRATIONS FORCED AIR COOLING SYSTEM The Reactor Buildmg Penetrations Forced Air Cooling System provided cool air to the feedwater and r steam penetrations in the Reactor Buildmg to maintain the temperature of the concrete surrounding the penetrations within allowable limits. 6.3.18 DELETED 6.3.19 CIRCULATING WATER SYSTEM The Circulating Water System was designed to provide the cooling water requirement for the main condenser. In addition, the system prmided cooling water to the condenser vacuum pump coolers, deicing water supply to the River Water Pump House, and watet to the chlonne ejector. 6.3.20 DELETED 6.3.21 ENVIRONMENTAL BARRIER SYSTEM The Emironmental Barrier System was designed to provide added assurance that containment atmosphere leakage to the emvironment, through the isolation valves of selected process lines, would be as low as practical, his system provided a barrier to emironmental leakage in addition to the two barriers which were normally supplied by means of the inside and outside Containment isolatica valves. O 6.3-2 UPDNT 2 - AUGUST 1997
1 ! 6.3.22 PENETRATION PRESSURIZATION SYSTEM
/q (j) The Penettstion Pressurization System was designed to reduce any leakag of containment atmosphere through penetration sleeve cavities by means of a positive nitrogen gas prec.sure in the enclosed air spaces incorporated in the penetration design.
6.3.23 GLAND STEAM SYSTEM The Turbine Gland Scaling System scaled the turbine shaft and rotor to ensum that steam was contained within the turbine casings and to prevent air inleakage. The system provided taese functions automatically by controlhng the gland steam pressure to an adequate sealing pressure under all conditions of turbine operation. 6.3.24 DELETED 6.3.25 DELETED 6.3.26 DELETED 6.3.27 DELETED 6.3.28 DELETED 6.3.29 DELETED 6.3.30 DELETED O V 6.3.31 DELETED 6.3.32 DELETED 6.3.33 DELETED 6.3.34 DELETED 6.3.35 DELETED 6.3.36 DELETED 6.3.37 DELETED 6.3.38 DELETED 6.3.39 DELETED 6.3.40 TEMPORARY NUCLEAR SERVICES CLOSED COOLING SYSTEM This system was designed and installed during the TMI-2 cleanup period. It provided cooling for the mini decay heat removal heat exchangers. A blank plate was installed in the inlet to spent fuel pool heat exchanger I A to divert sufficient flow through the Temporary Nuclear Services Closed Cooling System. ( ) 6.3.41 DELETED 6.3-3 UPDATE 2 - AUGUST 1997
l 6.3.42 DELETED 6.3.43 DELETED 6.3.44 DELETED 6.3.45 DELETED 6.3.46 DELETED 6.3.47 DELETED 6.3.48 DELETED 6.3.49 DELETED 6.3.50 EARTHQUAKE DETECTION SYSTEM He SMA 3 Strong Motion Accelerograph System manufactured by Kinemetrics, Inc. was installed in the basement of the Reactor Building annulus to measure the strength of seismic events. he SMA-3 was a multi-channel strong motion accelerograph. It featured central recording on magnetic tape cassettes with remote accelerometer and starter packages. The system remained in standby condition and would only activate if an earthquake caused the starter to actuate the recording circuits and tape transports. 6.3.51 REACTOR COOLANT PUMPS MOTOR OIL DRAIN SYSTEM Ris system was used to drain oil leakage from the RCP motors. Oil was diverted from the upper beanns oil splash shielding and lower bearing oil splash shielding to oil shield drain tanks located in the Reactor Building basement. Each of the four (4) reactor coolant pumps has the capacity to contain as much as 138 gallons of oil. More than 40% of that oil was removed prior to PDMS. Not all of the oil was removed because the dose rates severely limit access to the area of the pumps and oil collection / drain systems. liowever, the oil collection tanks meet NFPA-30 requirements and are equipped with flame arrestors. Therefore, the residual oil in the tanks does not present a fire hazard. O 6.3-1 UPDATE 2 - AUGUST 1997
6.3.52 DELETED O 6.3.53 DELETED 6.3.54 DELETED-6.3.55 DELETED 6.3.56 POLAR CRANE The Reactor Building Polar Crane is an electrically powered, pendant operated rotary bridge crane with a single trolley. The crane was originally designed and constructed with a capacity of 500 tons and an auxiliary hoist capacity of 25 tons. During the post accident clean-up period, the crane was extensively refurbished with the refurbishment including deactivation of the cab controls and installation of pendant , operating controls and a new festoon cable. It was subjected to a load test to recertify it e a main hoist capacity of 170 tons and an auxiliary hoist capacity of 25 tons in support of reactor vesse; acad service platform removal. In October of 1990, the crane manufacturer notified owners that structural defects had been found in a similar crane and recommended certain modifications to preclude failures in ali crane of this type. In lieu of performing the modification', GPUN elected, with the manufacturer's approval, to downgrade the crane to a main hoist capacity of 88 tons and an auxiliary hoist capacity of 9 tons. This was sufficient capacity to satisfy the lifting and handling needs for the remainder of the defueling program. In March of 1994, the crane manufacturer again notified owners of additional defects identified in a similar crane and recommended further remedial modifications. GPUN deactivated the polar crane. During the O PDMS period repairs and modifications will be evaluated and the crane may be refurbished to restore it to a lifting capability sufficient to support PDMS related work and future dismantlement and decommissioning needs. 4
.( _~
6.3-5 UPDATE 2 - AUGUST 1997
TABLE 6.3-1 DEACTIVATED SYSTIBIS SYSTEM PD11S CONTAINMENT INTERNAL SYSIT,M DESCRIITION CODE FUNCTION ISOLATION CONTAMINATION REMARKS Auxiliary Steam AS NONE NO NO Instrument Air IA NONE NO NO Portions of the Inst. Air System will be used in the Compressed Air System Demin Water DW NONE NO NO DW-V28 transferred to decon processed water (DPW) system. Nuclear Services River Water NR NONE NO NO Reactor Building Normal Cooling RB NONE NO NO Penetration Cooling PC NONE NO NO Cire. Water CW NONE NO NO Environmental Barrier EU NONE NO NO Penetration Pressurization PP NONE NO NO Gland Steam GS NONE NO NO 6.3-6 UPDATE 2- ST 1997
i s TABLE 6.3-1 (cont'd) DEACTIVATED SYSTEh1S SYSTEh! PDMS CONTAINhfENT INTERNAL SYSTEM DESCRIITION CODE FUNCTION ISOLATION CONTAAIINATION REMARKS i Temp. Nuc. Sve. Closed TNS NONE NO NO Cooling l Earthquake Detection ED NONE NO N/A RC Pump Motor Oil Drains RO NONE NO NO l Polar Crane N/A NONE NO YES 1 l l 6.3-7 UPDATE 2 - AUGUST 1997
6.4 DELETED i
, . e 6.4 - 1 UPDATE 2 - AUGUST 1997
6.5 SYSTEM REFERENCES Provided below is a list of reference documents that provide additionalinformation. Unless otherwise specified, drawing numbers refer to Burns and Roe Flow Diagrams. REFERENCE DOCUMENT NO. DOCUMENT DESCRIPTION l , 6.51 GPU Dwg. 302-2700, Reactor Building Isolation Sheets 1,2 & 3 Flow Diagram, Unit 2 6.5-2 DELETED 6.53 Doc. No.15737-2-J 16-001 Bechtel Instr. Index 6.5-4 Work Orders 2555 and 3475 Burns & Roe Instr. List 6.5 5 Dws No 302 2219 Radiation Monitoring, Station Vent i 6.5-6 DELETED 6.5-7 DELETED 6.58 &chtel Dwg. Demineralized Senice Water No, 3 5737-2-M74 I'WO1 P&lD 6.5-9 DELETED e 6.5 10 DELETED 4 Y
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT NO. DOCUMENT DESCRIPTION 6.5 11 Bechtel Dwg. P&ID Contamment Air Control No. 2-M74-CDW01 Envelope HVAC System 6.5 12 Bechtel Dwg Piping and Instrument Diagram No. 2-M74-DW01 (P&lD). Demineralized Service Water l System l 6.5-13 Bechtel Dwg. No. 2-M74 DWC01 Defueling Water Clean-Up Reactor Vessel f Clean-Up System j l 6.5 14 Bechtel Dwg. No. 2-M74-DWC02 Defueling Water Clean-Up Fuel Transfer Canal / Spent Fuel Pool Clean Up System I 6.5 15 Bechtel Dwg. No. 2-M74 DWC03 Defueling Water Clean-Up Auxiliary Systems 6.5 16 Bechtel Dwg. No. 2 M74-PW01 Processed Water Storage and Recycle System I 6.5 17 Bechte! Dwg. No. 2 M74 RBC01 Reactor Building Chilled Water System 6.5 18 Bechtel Dwg. No. 2 M74-SDS01 Piping and Instmment Diagram (P&lD
- SDS Feed and Monitor Tank System 6.5 19 Bechtel Dwg. No. 2-M75 FCC01 Schematic Flow Diagram - Fuel Transfer Canal Fill System 6.5 20 Bechtel Dwg. No. 2-POA-1303 General Arrangement Plenum Removal Reactor Building 6.5-21 Dechtel Dwg. No. 2 POA-6401 General Arrangement Fuel Handling Building Plan El 347'-6" 6.5-22 Dwg. No. 2002 Main & Reheat Steam and Gland Seal Steam Systems 6.5 23 Dwg. No 2003 Bleed Steam System 6.5 - 2 UPDATE 2 - AUGUST 1997 9
SYSTEM REFERENCES (Cont'd) REFERENCE - DOCUMENT NO. ~ DOCUMENT DESCRIPTION 6.5 24 Dwg. No. 2004 Auxiliary Steam System 6.5 25 Dwg. No. 2005 Feedwater and Condensate System , .6.5 26 Dwg. No. 2006 Make-Up Water Treatment and Condensate Polishing 6.5 27 Dwg. No. 2007 Demineralized Senice Water System 6.5 28 Dwg. No. 2008 Feedwater Heater Vents, Reliefs and Misc Drains System 6.5-29 Dwg. No. 2009 Feedwater Heater Drains System 6.5 30 Dwg. No. 2010 Condenser Air Extraction System 6.5-31 Dwg. No. 2011 Turbine Lube Oil Purification and Transfer System 6.5 32 Dwg. No. 2012 Instrument and Senice Air System 6.5-33 Dws. No. 2013 Domestic Water . s 6.5 34 Dwg. No. 2014 Senice Air System 4 6.5-35 Dwg. No. 2015 Turbine Plant Sampling System -
- Secondary Plant 4
6.5-36 Dwg. No. 2018 Secondary Senices Closed Cooling Water 6.5-37 Dwg. No. 2021 Circulating and Secondary Senices River Water System 6.5-38 Dwg. No. 2023 Circulating and Secondary Senices River Water System 4 k O 6.5-3 UPDATE 2 - AUGUST 1997
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT NO. DOCUMENT DESCRIPTION 1 6.5-39 Dwg. No. 2024 Reactor Coolant, Make-Up and ! Purification System i 1 1 6.5-40 Dwg. No. 2025 Chemical Addition 6.5-41 Dwg. No. 2026 Spent Fuel Coolmg and Decay Heat i System I 6.5-42 Dwg No. 2027 Radwaste Disposal Reactor Coolant ! I Liquid 6.5-43 Dwg. No. 2028 Radwaste Disposal Gas 6.5-44 Dwg. No. 2029 Intermediate Closed Cooling Water System 6.545 Dwg. No. 2030 Nuclear's Senices Closed Cooling Water System 6.5-46 Dwg. No. 2031 Nuclear's Samp!mg System 6.5-47 Dwg. No. 2033 Nuclear Senices Rher Water System 6.5-48 Dwg. No. 2034 Reactor Buildmg Emergency Spray an Core Floodmg 6.5-49 Dwg. No. 2035 Decay Heat Closed Cooling Water System 6.5-50 Dwg. No. 2036 Nitrogen for Nuclear and Radwaste Systems
+
6.5 51 Dwg. No. 2037 Fire Protection System 6.5-52 Dwg. No. 2038 Diesel Fuel Emergency Diesel Generator 6.5-53 Dwg. No. 2039 Raduste Disposal- Solid 6.5-4 UPDATE 2 - AUGUST 1997
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT DOCUMENT DESCRIPTION
- 6.5 54 Dwg. No. 2040 Heat and Ventilation, Turbine and Control Building Areas 6.5-55 Dwg. No. 2041 Reactor Building Ventilation, and Purge System 6.5 56 Dwg. No. 2042 Auxiliary Building Heating & Ventilation System 6.5-57 Dwg. No. 2044 Heating Ventilation and Air Condition (Control Building, Cable, Battery and Switchgear Rooms 6.5-58 Dwg. No. 2045 Radwaste Disposal Miscellaneous Liquid 6.5-59 Dwg, No 2046 Reactor Building Normal Cooling Water 6.5-60 Dwg. No. 2047 Heating and Ventilation, Cin River Water Pump House ano House 6.5-61 Dwg. No. 2049 Heating and Ventilation, Emergency Diesel Generator Building
- 6.5-62 Dwg. No. 2076 Piping Specialty List 6.5-63 Dwg. No. 2219 Heating and Ventilation, Buildmg Air intake and Exhaust 6.5-64 Dwg. No. 2343 FuelHandling Building Heating &
Ventilation System 6.5-65 Dwg. No. 2385 Senice Building Heating & Ventilation System 6,5 66 Dws. No. 2391 Senice & Control Building - Domestic 1 Water. Sanitary Waste & Contanunatioa Drains
! 6.5-5 UPDATE _2 - AUGUST 1997
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT DOCUMENT DESCRIPTION l 6.5.67 Dwg No. 2397 Reactor Building Penetrations Isolation Valve Seal Water System 6.5-68 Dwg. No. 2414 Steam Generator Secondary Side Vent and j Drains 6 5-69 Dwg No.2440 Main, Bleed & Auxiliary Steam System & Traps 6.5-70 Dwg No 2475 Heating and Ventilation, Coagulator Buildmg, Circulating Water Chlorinator House, and Mechanical Draft Cooling Tower Pump House 6.5-71 Dwg. No. 2492 Radwaste Pumps Seal Water 6.5 72 Dwg. Fo. 2496 Sump Sump Discharge 6573 Dwg. No. 2497 Reactor Building Penetration Forced Air Cooling System 6.5-74 Dwg. No. 2517 Reactor Building Leak Rate Test 6.5 75 Dwg. No. 2524 Feedwater IIcater Nitrogen Blanketing 6.5-76 Dwg. No. 2532 H: and CO2 Supply Systems (Secondary Plant) 6.5-77 Dwg. No. 2551 Lube Oil System - Emergency Diesel Generator 6.5-78 Dwg. No. 2552 Startmg Air System, Emergency Diesel Generator 6579 Dwg. No. 2596 Flow Diagram, Hydraulic and Pneumatic Fuel Transfer System 6.5-80 Dwg. No. 2601 Reactor Coolant Pumps Seals 6.5 - 6 UPDATE 2 - AUGUST 1997
_ ._ __ . _ . . . _ _ _ . _ . . _ . . ._._._._._._____m.. _ _ . _ _ _ _ _ . _ _ . . _ _ .
'I SYSTEM REFERENCES (Cont'd)
REFERENCE -DOCUMENT DOCUMENT DESCRIPTION 6.5 81- - Dwg. No. 2606 OTSG Chemical Cleaning System .
- 6.s82 ' Dwg. No. 2626 Lab & Penetration Pressurization Gas ,
Systems and Hydrogen for Make-Up Tank - 4 6.5-83 Dwg. No. 2632 Radwaste Disposal Reactor Coolant 4 1.eakage Recovery
- 6.5-84 Dwg. No. 2633 Flow Diagram Oil Splash Shield Drain Piping for R. C. Pump Motors 6.5-85 Dwg. No. 2634 Glarx! Steam Seal System Turbine Building .
6.5-86 Dwg. No. 2636 Emironmental Barrier System 6.5-87 Dwg. No. 2668 SSCCW and Secondary Plant Sampling . System, TB & CBA 6.5-88 GPUN Dwg. No. 2E 3510-1024 P&ID, Reactor Building Sump Recirculation System
-6.5 89 GPUN Dwg. No. 2E 950-02 001 SDS Plan View
, 6.5-90 GPUN Dwg. No. 2R 950-21001 P&lD Composite Submerged Demmeralizer System l 6.5 91 Dwe. No. 3004, Sheet 1 4160V Switchgear One Line Diagram 6.5-92 Dag. No. 3005, Sheet 3 480V Unit Substations One Line Diagram
- 6.5 93
- Dwg. No. 3011 Safety Features Actuation System Block
. Diagram 4 6.5-94 Dws. No. 3024-106 Block Diagram Misc. Seismic Monitoring System 4 l . 6.5-7 UPDATE 2 - AUGUST 1997
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT DOCUMENT DESCRIPTION 6.5-95 Dwg. No. 3073 Diesel Generators Sheets 57,59, and 60 6.5-96 Dwg. No. 3091 Safety Features Actuation System Index 6.5-97 Dwg. No. 4148 Reactor Building Containment Wall Penetrations Schedule 6.5-98 Dwg. No. 4479037 Steam Turbine Piping-Steam Drain & Gland Diagram 6.5-99 Dwg. No. 614F177-5 (B&R File No. 01-00-0504) Seal Oil Diagram - Turbine Generator 6.5 100 Dwg. No. 614F177-6 Seal Oil System (B&R File No.1 + 0504) 6.5-101 Dwg. No. 7213843 (B&R File No. 01-00-0210) E. H. Fluid STS & Lube Diagram 6.5 102 Bechtel Dwg. Piping and instrument No.15737-2-M74-DSO! Diagram - Dewatering System 6.5-103 TM1-1 Dwg. No. C-302-05 Unit 1 Auxiliary Steam System 6.5 104 TMI l Dwg. No. C 302-051 Auxiliary Steam System of TMI Unit No.1 Auxiliary Boilers 6.5-105 TMl 1 Dwg. No. C-302-162 TMI-l Plant Filtered HA to and from TM1 #2 Plant 6.5-106 TMI-l Dwg. No. C 302-163 The 1,000,000 Gallon Demineralized Water Storage Tank 6.5-107 TMI-l Dwg. No. C-302 301 The Unit 1 Plant Generator Gas & Vents 6.5 - 8 UPDATE 2 - AUGUST 1997 O
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT DOCUMZST DESCRIPTION
- 6.5-108 TMl 1 Dwg. No. C-302-671 Unit 1 - Sampling; Liquid and Gas 6.5 109 TMI-l Dwg. No. C 302 692 Liquid Waste Disposal System for Unit 2
6.5 110 TMl-1 Dwg. No. C 302-84 Unit 1 - Auxiliary Building Heating & Ventilation 6.5 111 TMI l Dwg. No. D 60372 Gland Seal & Ejector System 6.5 112 Dwg. No. E-032 Misc. Power Panel Schedules, SDS 6.5 113 Dwg. No. E 302-191 Unit No.1 OTSG Chemkal Cleaning System ~
- 6.5-114 TMl-1 Dwg. No. E 302196 Unit No.1 OTSG Chemical Cleaning System 6.5-115 TMI l Dwg. No. E-302-231 Fire Service Water 6.5 !!6 Dwg. No, M006 Auxiliary Building Emergency Liquid Clean-Up 6.5 117 Dwg. No. M011 Flow Diagram Condenser Air Extraction Filtration System 6.5 118 Dwg. No M012 Alternate Condensate System 6.5 119 Dwg. No. M013 HVAC Chemicci Cleaning and Health Physics Building 6.5-120 Dwg. No. M014 Fuel Pool Waste Storage System 6,5-121 Dwg. No. M015 Auxiliary Building Emergency Liquid Clean-Up System Sampling 6.5-122 Dwg. No. M016 Flow Diagram, Temporary Auxiliary &
Fuel Handling Building HVAC O O 6.5-9 UPDATE 2 - AUGUST 1997
SYSTEM REFERENCES (Cont'd) REFERENCE DOCUMENT DOCUMEhT DESCRIPTION 6.5-123 Dwg. No, M021 Long Term OTSG "B" Cooling 6.5 124 Dwg. No. M022 Standby Reactor Coolant System Pressure Control System 6.5 125 Dwg. No. M041 " Temporary" Nuclear Services Closed Cooling Water System 6.5 126 Dwg. No. M043 Mini Decay Heat Removal System 6.5-127 Dwg. No. M044 & MG45 Temporary Nuclear Sampling System, SNS 6.5-128 Dwg. No, M208 Chemical Cleaning & HeeJth Physics Building Fire Protection O O 6.5 - 10 UPDATE 2 - AUGUST 1997
.e - ,-4.----A-_ -,Adas.me al.4-4-a.el,.ia & A E D.r emm_.,..A4.4.% 44 me.A as A-a.m- h-.--4m--We-.A.*-*seM*a-fJ-.aae*v4aMA - m= e e ..As a
) i CHAPTER 7 OPERATIONAL SYSTEMS ' 1 AND FACILITIES , I l I i i a d 5 'd 4
CHAPTER 7 ( OPERATIONAL SYSTEMS AND FACILITIES TABLE OF CONTENTS. SECTION I[III PAGE NO.- ~
7.0 INTRODUCTION
7.0 1 7.1 OPERATIONAL FACILITIES 7.1 1 i , 7.1.1 CONTAINMENT (REACTOR BUILDING) 7.1 1 7.1.1.1 PDMS Function 7.1-1 7.1.1.2 Containment Structure 7.1-1 i 7.1.1.3 Containment Functional Design 7.12 7.1.1.4 Facility Description - 7.1-2 7.1.1.5 Evaluation 7.1-2 7.1.2 AUXILIARY BUILDING 7.1 3
- 7.1.2.1 PDMS Function 7.1-3 7.1.2.2 Facility Description 7.1-3 7.1.2.3 Evaluation 7.1 3 7.1.3 FUEL HANDLING BUILDING 7.1 3 7.1.3.1 PDMS Function 7.1-3 7.1.3.2 Facility Description 7.1 3 7.1.3.3 Evaluation 7.1-4 -
7.1.4 ' FLOOD PROTECTION 7.1-4 7.1.4.1 PDMS Function 7.1-4 7.1.4.2 Facility Description 7.14-7.1.4.3 Evaluation . 7.1-5 (
\
i UPDATE 2 - AUGUST 1997
1 [ CIIAPTER 7 ( TABLE OF CONTENTS (Cont'd) SECTION TITLE PAGE NO. 7.1.5 AIR INTAKE TUNNEL 7.1-5 7.1,5.1 PDMS Function 7.1-5 7.1.5.2 Facility Description 7.1-6 7.1.5.3 Evaluation 7.1-6 7.1.6 UNIT 1/ UNIT 2 CORRIDOR 7.1-6 7.1.6.1 PDMS Function 7.1-6 7.1.6.2 Facility Description 7.1-6 7.1.6.3 Evaluation 7.!-7 7.1.7 CONTROL AND SERVICE BUILDINGS 7.1-7
'l.7.1
. PDMS Function 7.1-7 7.1.7.2 Facility Description 7.1-7 c 7.1.7.3 Evaluation 7.1-7 Turbine Building 7.1 8 k 7.1.8 7.1.8.1 PDMS Function 7.1-8 7.1.8.2 Facility Description 7.1-8 7.1.8.3 Evaluation 7.1-8 7.2 OPERATIONAL SYSTEMS 7.2-1 7.2.1 CONTAINMENT SYSTEMS 7.2-1 7.2.1.1 Containment Isolation 7.2-1 7.2.1.2 Containment Atmospheric Breather 7.2-2 7.2.1.21 PDMS Function 7.2-2 7.2.1.2.2 System Description 7.2-2 7.2.1.2.3 Evaluation 7.2-3 7.2.1.3 Contamment Ventilation and Purge 7.2-6 7.2.1.3.1 PDMS Function 7.2-6 7.2.1.3.2 System Description 7.2-6 7.2.1.3.3 Evaluation 7.2-6 7.2.1.4 Containment Airlocks a'd Equipment Hatch 7.2-6 ii UPDATE 2 - AUGUST 1997
/] CHAPTER 7 TABLE OF CONTENTS (Cont'd)
SECTION *[lILE PAGE NO. 7.2.1.4.1 PDMS Function 7.26 i .7.2.1.4.2 System Description 7.2-7 7.2,1,4.3- Evaluation 7.27 7.2.2 FIRE PROTECTION, SERVICE, AND SUPPRESSION 7.2-7 7.2.2.1. PDMS Function 7.2-8 7.2.2.2 System Description 7.2-8 7.2.2.3 Evaluation 7.2 10 7.2.3 RADIOACTIVE WASTE hMNAGEMENT 7.2-11 7.2.3.1 Radioactive Waste Miscellaneous Liquids 7.2-11 ! 7.2.3.1.1 PDMS Function 7.2-11 7.2.3.1.2 System Description 7.2-11 7.2.3.1.3 Evaluation 7.2 12 7.2.3.2 Sump Pump Discharge and Miscellaneous Sumps System 7.2 13 7.2.3.2.1 PDMS Function 7.2-13
\ 7.2.3.2.2 System Description 7.2 13 7.2.3.2.3 Evaluation 7.2 13 7.2.4 RADIATION MONITORING 7.2-14 7.2.4.1 PDMS Function 7.2-14 7.2.4.2 Radiological Surveys 1.2-14 7.2.4.2.1 AFHB Radiological Surveys 7.2-14 4
7.2.4.2.2 Contamment Radiological Surveys 7.2-14 7.2.4.3 Efiluent Monitoring 7.2-15 7.2.4.3.1 AFHB Airborne Evaluation 7.2 16 7.2.4.4 General Radiological Monitoring 7.2-17 7.2.4.5 Evaluation 7.2-17 7.2.5 ELECTRICAL SYSTEMS 7.2-17 7.2.5.1 PDMS Electrical Distribution Sysem 7.2-17 7.2.5.1.1 PDMS Function 7.2-17 O iii UPDATE 2 - AUGUST 1997
)
. CHAPTER 7 TABLE OF CONTENTS (Cont'd) -
SECTIONi TITLE fAGE NO.
'7.2.5.1.2- System Description -7.2-17 7.2.5.1.3 Evaluation 7.2 18 7.2.5.2 Normal and Emergency Lighting 7.2-19 7.2.5.2.1 PDMS Function 7.2 19 7.2.5.2.2 System Description 7.2 19 7.2.5.2.3- Evaluation 7.2 20 7.2.5.3 Communications System 7.2 20 7.2.5.3.1 PDMS Function 7.2-20
- 7.2.5.3.2 System Description 7.2 20 7.2.5.3.3 Evaluation 7.2-21 7.2.6 PDMS SUPPORT SYSTEMS 7.2-21 ,
-7.2.6.1 Auxiliary Building Ventilation System 7.2-21 7.2.6.1.1 PDMS Function 7.2-21
\- 7.2.6.1.2 System Description 7.2-21 7.2.6.1.3 Evaluation 7.2-22 7.2.6.2 Fuel Handling Building Ventilation System 7.2-22 7.2.6.2.1 PDMS Function 7.2 22 7.2.6.2.2 System Description 7.2 22 7.2.6.2.3 Evaluation 7.2 22 7.2.6.3 Air intake Tunnel Ventilation System 7.2-23 7.2.6.4 Compressed Air Supply System 7.2-23 7.2.6.4.1 - PDMS Function 7.2-23 7.2.6.4.2: System Description 7.2-23 7.2.6.4.3 Evaluation 7.2 24 7.2.6.5 Building Inteakage Waterproofing System 7.2-24 7.2.6.5.1 PDMS Functum 7.2 24
'7.2.6.5.2. System Description - 7.2-24
. ~\ iv UPDATE 2 - AUGUST 1997.
. . . _ - . ~ . . . . - . - . - - . - . . . - _ . . . . - . - . . . - . - - . . . - _ - . - . -
CHAPTER 7 L . !
- TABLE OF CONTENTS (Cont'd)
SECTION TITLE PAGE NO.- . 7.2.6.53 Evaluation 7.2 25 7.2.6.6 Sewers 7.2 25 7.2.6.6.1 PDMS Function 7.2 25 e 7.2.6.6.2 System Description 7.2-25 i i 7.2.6.63 Evaluation 7.2 25 7.2.6.7 Domestic Water System 7.2-25 7.2.6.7.1 PDMS Function 7.2-25 7.2.6.7.2 System Description 7.2 25 .
- . 7.2.6.7.3 Evaluation 7.2-26 7.2.6.8 Control Room Ventilation System 7.2-26
- 7.2.6.8.1 PDMS Function 7.2 26 7.2.6.8.2 System Description 7.2 26
[ 7.2.6.8.3 Evaluation 7.2-26 i 7.2.6.9 Cable Room Ventilation System 7.2-27 7.2.6.9.1 PDMS Function 7.2-27 3 I J
- \
v UPDATE 2 - AUGUST 1997 4
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0 4 ( CHAPTER 7 TABLE OF CONTENTS (Cont'd) SECTION TITLE PAGE NO. + 7.2.6.9.2 System Description 7.2 27' r 7.2.6.9.3 Evaluation 7.2 27 i 7.2.6.10- Service Building Ventilation System 7.2 27 7.2.6.10.1' PDMS Function 7.2 27 7.2.6.10.2 System Description 7.2 27 7.2.6.10.3 Evaluation 7.2 28 7.2.6.11 PDMS Alarm Monitoring System 7.2 28 l 7.2.6.11.1 PDMS Function 7.2 28 - r 7.2.6.11.2 System Description 7.2 28 7.2.6.11.3 Evaluation 7.2 29 l 7.2.6.12 Control Building - Mechanical Equipment Room 7.2-29 4 Ventilation System } . 7.2.6.12.1 PDMS Function 7.2 29 ! 7.2.6.12.2 . System Description 7.2-29 7.2.6.12.3 Evaluation 7.2-29 7.2.6.13 Control Building Area Ventilation System - 7.2-30 7.2.6.13.1 PDMS Function 7.2-30 7.2.6.13.2 - System Description 7.2-30 7.2.6.13.3 - Evaluation 7.2-30 7.3 - SYSTEM REFERENCES 7.3-1 Y d
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CilAPTER 7 O- TAllLE or CONTENTS (Cont'd) 1,lST Ol' TAlli,ES TAlli,E I)lll 1%GE 7.1 1 OPERATIONAL FACILITIES 7.1 9 7.21 OPERATIONAL SYSTEh!S 7.2 31 7.2 2 CONTAINMENT IS01.ATION TABLE 7.2 34 7.2 3 OPERATIONAL SUMP SYSTEMS FOR PDMS 7.2 47 73-4 DELETED O O vii UPDATE 2 - AUGUST 1997
I CllAPTER 7 TABLE OF CONTENTS (Cont'd) ,, LIST OF FIGURES FIGURE TITI,E PAGE No. 7.2 1 CONTAINMENT ATMOSPl!ERIC BREATIIER "MOST 7.2 51 PORTABLE" PATlIWAY MODEL 7.2 2 R/DIATION SURVEY LOCATIONS; R.B. ELE'- ATION 347' . 7.2 $2 7.2 3 RADIATION SURVEY LOCATIONS; R.B. ELEVATION 305' 7.2 53 O O viii UPDATE 2 - AUGUST 1997 I
1 p Cl! APTER 7 OPERATIONAL SYSTEMS AND FACILITIES
7.0 INTRODUCTION
This depter dcacribes those systems and facilities which will be maintained in an operational condition I during PDMS. Generally, those facilities which are maintained operational are those buildmgs or areas ' I that contain operational systems or partially operational systems. An operational facihty may include both operational systems (see Section 7.2) and deactivated systems (see Chapter 6). Operational facilitics and systems serve several functions within the scope of PDMS activities, includmg support of site operations, maintenance activities, and surveillance activities. Tables 7.1 1 and 7.2 1 provide a hsting of those facilities and systems which will be maintained in an operational condition during the PDMS period. These tables also provide other relevant information conceming the status of the listed facilities and systems. The followmg systems and facilities discussed in Chapter 7 provide reasonable assurance that TMl-2 can be me.intained in the PDMS condition with no risk to the health and safety of the public: 1) the Containment structure, 2) the Containment Ventilation and Purge System; 3) the Containment Atmospheric Dreather; 4) the Fire Detection, Service, and Suppression System; 5) the Audliary Building Ventilation System; 6) the Fuel 11andlmg Building Ventilation System; 7) the associated support and momtoring systems; and 8) the unit's flood protection capabilities. (3 b o) V 7.0 1 UPDATE 2 - AUGUST 1997
p 7.1 OPERATIONAL FACILITIES Thil 2 facilities required to be operational during PDhtS are desenbed in this section. Facilities are required to be operational to support operational systems within those facilities and/or to isolate internal contamination from the emironment. Table 7.1 1 provides a listing of operetional facilities for Thil 2 during PDhtS. Each internally contarninated facihty is identified along with any relevant remarks regardmg the fmal layup of the facility. Each of the following sections addresses the PDhtS function of the facility, the facihty description, and applicable evaluations. Additional reference information is listed in Section 7 3. 7.1.1 CONTAINhiENT(REACTOR BUILDING) 7.1.1.1 PDhtS Function The primary function of the Containment during PDhtS is as a contamination bamer. He Containment provides shielding of the emironment from the contained radiation it also provides the means to assure that any efiluents from the Containment will be controlled, filtered, and monitored. The Containment was designed to withstand approximately 60 psi ofinternal pressure, airplane crashes, a safe shutdown-carthquake, tornados, floods and other natural phenomena. He Containment was also designed with the capabihty to isolate any radioactive materials produced as a result of accidents or other ur plar'ned events O Q Although modifications were made to several of the Containment penetrations, (See Table 3.7 1) the structural capabilities of the Containtnent were not significantly diminished by the accident or any of the cleanup actisities and are expected to be retained through the PDhtS period. The Containment design and isolation capabilities relied upon during PDhtS are desenbed in the following sections. 7.1.1.2 Contaiament Structure he Containment is a reinforced concrete structure cornposed of cylindrical walls with a flat foundation mat and a dome roof and lined with a carbon steel liner. He structure provides biological shielding for normal and unanticipated conditions. The steelliner enclues the equipment and systems which remain inside the Containment and ensures that an acceptable upper limit ofleakage of radioactive material will not be exceeded under the worst unanticipated es ent. He foundation slaS is reinforced with corwentional carbon steel reinforcing. The cylindrical walls are prestressed with a grouted tendon post tensioning system in the vertical and horizontal directions. The dome roofis prestressed utilinns a three-way grouted tendon post tensiomng system. The inside sudace of the Reactor Building is lined with carbon steel to ensure a L high degree ofleak tightness. The thickness of the liner plate is 3/8 in. for the cylinder,1/2 in for the dome and 1/4 in for the base. O U] 7.1 1 UPDATE 2 - AUGUST 1997
he foumiation(
- t ars on the bedtock and is 1I feet 6 inches thick with nn additional 2 foot thick concrete slab atxt , we liner plate. He cylindrical portion has an it side diameter of 130 feet, wall thickness of 4 feet, and a hc;ght of 157 feet from the top of the foundation slab to the spring ime. He roofis a shallow dame which has a large ra6us of 110 feet and a transition radius of 20 feet 6 inches.
7.1.1.3 Containment functional Design During PDh1S, the Contairunent senes primanly as a contamination barrier and provides shieldmg from the radiation due to cornained contanunation. All efIluents will be controlled, filtered, and monitored he functional requirements for the Reactor Buildmg during PDhis are listed below;
- a. He Containment pressure will be main;ained at equihbrium with atmospheric pressure by utihr.ing a passive ventilation systern (see Containment Atmospheric Breather, Section 7.2.1.2) via the Auxihary 13uilding.
- b. Containment isolation will be maintained by a single passive barrier either inside or outside of Containment on cach Containment penetration Active isolation capability is not required for PDhtS except for the Containment Atmospherie Breather and the RH Purge Containment valves. Various passive means, or their equivalent, are accepcble for piping systems and inciude locked closed valves, closed and deactivated rernote manual valves, closed and deactivated automatic valves, and blind flanges.
- c. hiomtoring of effluent releases will be provided by existing and/or ad&tional morutoring equipment as designated in Section 7.2.4.
7.1.1.4 Facility Description Systems within the Containment not required to be maintained in an operational condition during PDhtS have been deactivated. ne electric power circuits m the Contamment have been deenergized except for those necessary for PDh!S monitoring, inspection, and surveillance equipment, and other PDh1S support acqiirements. Pnor to each inspection inside the Containment, circuits will be energized to provide lighting and power for required equipment. 7.1.1.5 Evaluation he Containment was originally designed to withstand airplane crashes, seismic events, tomados, floods, and other natural phenomena. Although there were modifications made during the cleanup period to several of the piping penetrations, these modifications were performed so that the structural integnty of the Containment has been rnaintained Neither the accident nor any activity dunng the cleanup period has significantly degraded any of the structural capabilities of the Containment, herefore, the Containment is structurally capable of withstandmg the original design basis events (except internal pressurization) during the PDhtS period without further analysis. He intemal pressure of the Contaimnent during PDhtS is controlled by a passive breather system (see Section 7.2.1.2). His system will maintain the Reactor Building in equilibriutt xith atmospheric pressure (via the Auxiliary Building) at all times during its use, in addition, a rar.ge of postulated events has been imestigated (see Chapter 8) and none of these events could result in any significant pressurization of the Containment. Herefore, enn with the reduced pressure capabilities due to the 7.1 2 UPDATE 2 - AUGUST 1997
existence of mcxhfied penetrations, the Containment is capable of performing its intended function of contamination isolation tiuoughout the range of normal and postulated unanticipated events.
'I he Containment will remain isolated during PDhtS. The Containment Atmospheric lireather and the RB Purge isolation valves will close on a liigh RIl pressure,if in operation.
7.1.2 AUXILIARY BUILDING i 7.12.1 PDh!S Function De Auxiliary Builing will serve prtmarily to support operation of the liquid radwaste, Auxiliary Builing sump, ventilation, and effluent monitonng systems required for PDh15 activities. 7.12 2 Facility Description ne Audliary Building shares a common wall with the Fuel 11andling Buildmg on the west side and has a vertical air intake shan attached to the east wall. He Auxiliary Building is rectangular in plan with three main floors of slab beam and flat slab construction. At the east exterior wall, a large door opening is located at grade level. His door opening ia not protected from an aircran impact loading or external missiles (see Section 3.5). The Auxiliary Building is accessible from the Senice Bui! ding, the Fuel llandling Building, and the Unit 1 - Unit 2 corridor. During PDh1S, the Auxiliary Building Ventilation System and filters will be maintained in an operational con & tion and operated as required. The auxiliary sump, auxiliary sump tank, and associatal level indication will remain operational as well as the 480/277 VAC power to lighting, and sump level inacation circuits. hiost loads of 480 VAC and above have been deenergized at the switchgear and/or motor control centers, llowever, selected loads (e g , welding receptacles, heaters, pump rootors, and fan motors) will remain energized and available tbr use, as needed. The Auxiliary Builing will be accessible for periodie surveillance entries and other limited actisities. 7.1.2.3 Evaluation System operations and activities in the Auxiliary Buildmg during PDh1S are at a reduced level, thereby substantially reducing the potential for spread of contamination. Auxiliary Building sump and liquid radwaste systems are operational to collect and process any liquids in the building to minimize uncontrolled accumulation ofliquids during PDhtS. 7.1.3 FUEL llANDLING BUILDING 7.1.3.1 PDhtS Function During PDh1S, the Fuel 11andling Building is not required for storage of new or spent fuel. Ilowever, it may be utilized for the temporary staging of site-generated radwaste or other appropriate uses. 7.1.3.2 Facility Description ne Fuel liandling Building shares a common wall on the cast side with the Auxiliary Building and a common truck bay with the Unit 1 Fuelliandling Building on the north end. One bridge crane, common [m
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to both buildings, was provided for fuel handling; no separating wall exists above the operating floor, te., elevation 347'-6" ne Containment is located on the south side of this building Two staiusc steel lined, reinforced concrete fuel storage pools are located in the buildmg. During PDhtS, the Tuct flandling Buildmg Venti'ation System and filters will be maintah.ed in an operational condition and will be operated as required for elevations below 347-6" The operating floor (el. 347-6) area is ventih ted by the Thfi 1 ventilation system. Electric distribution will remain configured to power low voltage (120/208 VAC) lighting loads and fire l 1 detectors. l All fuel canisters have been removed from the spent fuel pools and shipped oft-site. Doth fuel pool stnictures will remaia intact. He SDS has been deactivated. The Fuel Transfer Tubes have been isolated Access to the fuel pool area from Thfi 2 will be appropriately controlled to prevent unauthorized access to the TM1 1 fuel pool area which is classified as a vital area of Thil 1. He FIIB truck bay will be accessible from and under opetutional control of TMI 1. 7.1.3.3 Evaluation he Fuelllandlmg 13uildmg configuration for PDMS mirunuzes sources of contanunation; therefore, the potential for spread of contamination is very low. 7.1.4 FLOOD PROTECTION 7.1.4.1 PDMS Function The existmg unit flood protection capabilities will be maintained for PDMS and are based on a maximum water elevation of 311.0 ft. under flood conditions. The probable maximum flood (PMF) for the Susqueharma River at liarrisburg was established by the Army Corps of Engineers as 1,600,000 cfs. He water surface profiles routed downstream to the site results in a PMF of 1,625,000 cfs, which corresponds to a site elevation of 308.7 ft ne water surface elevation at the tip of Three Mile Island is 304.0 ft and 303.0 ft at the intake structure for the design flood. At these locations for the PMF, the calculated surface elevations are 310 0 ft and 309.0 ft., respectively. See Section 2 4.3. 7.1.4.2 Facihty Description Although station grade, at 304.0 ft., is above the water surface profile, dikes are prosided around the site to protect the station from wave action for the design flood. The top elevation of the protective dike at the tip of nree Mile Island is 310 0 ft., which provides a freeboard of 7.1-4 UPDATE 2 - AUGUST 1997
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^N approximately six feet above the design flood at that location. ne dikes along both sides of the island descend uniformly from ele.ation 310.0 ft. to elevation 305.0 ft., winch is sufficient to protect the entire site for the design flood. A dike with a top elevation of 304.0 ft. extends across the southem end of the site.
Structures are presided with complete protection at the exterior fsces rather than a: tempting to protect individual equipment or systems ne waterstops between adjacent building walls and mats were designed to be capable of withstandmg a maximum water head of 45 ft which is in excess of the inaumum head associated with the flood level. He exterior slidmg doors and flood panels are provided with watertight seals. Specific design features of these structures are:
- a. Containment - here are no external openings in the Containment below the 305 ft elevation,
- b. Fuelllandhng Building - There are no external openings in the Urdt 2 Fuel Itmdling Buildmg that require flood protection. The railroad door in the Urut 1 portion of the Fuelliandling Buildmg is designed to be watertight.
- c. Control Building - Flood panels are provided for all ground level exterior entrances.
- d. Auxiliary Euildmg - A flood panel is prmided for the east roll up door entrance.
- c. Air intake The openmgs in the Air Intake Tunnel are kcated higher than the probable maximum flood level except for a water tight hatch located at ground level, southeast of the Q
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V f. General - Doors and entrances (not flood protected) to the Control Buildmg Area are either watertight or are provided with flood panels. All openings that are potential leak paths (e.g., ducts, pipes, conduits, cable trays) are sealed. 7.1.4.3 Evaluation in addson to specific building flood protection provisions, the entire site is protected by an early warmng system provided by the Federal State River Forecast Center and a dike with a top elevation of 310.0 ft. The probable maximum flood is calculated to reach elevation 308.5 ft. on the wet side and 308 ft on the east side of the site. P.-refore, systems and facilities seguired to support PDMS actisities are protected from flooding. 7,1.5 AIR INTAKE TUNNEL 7.1.5.1 PDMS Function During PDM S, the Air intake Tunnel provides a pathway for screened air to the following operational plant ventilating systems:
- a. Reactor Building Ventilation
- b. Auxiliary Buildmg Ventilation
- c. Fuel llandling Building Ventilation
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- d. Control Building Ventilation.
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- c. Senice Building Ventilation
- f. Control Building Area Ventilatico ne Air intake Tunnel protects th:se plant ventilating systems from airbome debris, flood water, and l l
fire. 7.1.5.2 Facility Description he Air intake Tunnel consists of a cylindrical intile tower with screens and baflies, a 100,000 gallon sump, and an underground tunnel leading to the plant ventilating systems. He tunnel floor drains to the sump he tunnel leads to a vertical air intake shan which branches out into the individual supply ducts for the plant ventilating systems. He sump will be pumped out via a temporary pump, when required. 7.1.5.3 Evaluation The Air intake Tunnel is maintained during PDMS to provide an air supply pathway for operational plant ventilating systems. The structure is designed to protect the Air intake System against projectiles and floodmg. ne openings in the tower are above the probable maximum flood level, and the hafIled intake and screen prevent projectiles from entering the intake. De Air Intake Tunnel, by design, also helps prevent the spread of fire into plant vent]ating systems. Combustible material has been removed from the Air intake Tunnel. 7.1.6 UNIT 1/ UNIT 2 CORRIDOR 71.6.1 PDMS Function During PDMS, the Unit 1/ Unit 2 corridor serves as an operational facility to proside:
- a. Ileated weather enclosure for various operational system piping such as domestic water, Unit I discharge to 1%TS and the Unit 1 Processed Water Storage Transfer System
- b. Access to the Auxiliary Building from the cast outside yard through rollup security door 10.
- c. Interconnectmg corridor beturen Unit I aad Unit 2.
7.1.6.2 Facihty Description Re Unit 1/ Unit 2 corridor is a heated passageway nmning north to south adjacent to the east side of the Turbine, Seniec and Control, and Auxiliary Buildings. It is a steel frame structure with metal siding over a concrete base floor, with a partial block wall up to the windows to the outside east yard. The roof consists of built up layers of felt and asphalt. 7.1-6 UPDATE 2 - AUGUST 1997
1 I o 7.1.6.3 Evaluation The Unit 1/ Unit 2 corridor provides sufficient heating during PDMS to prevent freczmg of the enclosed operational system piping. All contaminated equipment has been removed or adequately isolated to nuninut.e the spread of contanunation. 7.1.7 CONTROL AND SERVICE DUILDINGS 7.1.7.1 PDMS Function ne Control Building houses the Unit 2 Control Room, a relay room, two deactivated inverter battery rooms, a cable spreadmg room, and a mechanical equipment room. Although the PDMS Alarm Morutonng System directs instrument alarm outputs to the Unit 1 Control Room, the Unit 2 Control Roorn annunciators / panel indications will be relied upon to provide specific information should the need arise. The Senice Building houses the operational Compressed Air System compressors, receiver tanks, and associated piping. In addition, the Senice Building provides access to the Reactor Building, the Auxiliary Duildmg, the Control Building and the air intake tunnel. He ControVSenice Building sump with one sump pump will remain in senice. 7.1.7.2 Facility Description ne Control and Senice Buildings are separated by a common wall. He Control and Senice Dutidmgs Q)
! are rectangular buildmgs with a common foundation rnat. He floors of the Control Building are supported by interior walls. A peripheral gap of 4 in. between edges of the floors and the inside face of the exterior walls has been provided to create a structural separation between the exterier structure and the interior structure He purpose of this separation was to protect vital and sensitive Control Room equipment from dmamic aircraft impact loadmg to which the exterior walls and the roof could be subjeued Door openings and other penetrations in exterior willis of the Control Buildmg that are susceptible to aircraft loadmg have been shielded by reinforced concrete shield walls.
Dunng PDMS, the Control Dailding Ventilation Systems (i.e., Control Room IIVAC, Mechanical Equiprnent Room HVAC, and Cable Room IIVAC) and the Senice Duilding Ventilation System will be imintamed in an operational condition and will be operated as required. Additionally, electrical distribution will remain configured to power low voltage (120/208 VAC) lighting loads, fire detectors, communications and panel annunciators. 7.1.7.3 Evaluation ne Control and Scnice Buildings configuration for PDMS muumizes sources of contammation, therefore, the potential for spread of contammation is very low. U] 7,1-7 UPDATE 2 - AUGUST 1997
7.1.8 TURDINE BUILDINO 7.1.8.1 PDhtS Function The Turbine Building will be utilized to house operational support systems and deactivated passive systems during PDhiS. Operational :ystem.= located in the Turbine Building include:
- a. The main 13.2KV power feed disconnects,480!!20VAC distnbution systems and a 125VDC distnbution systern.
- b. Sump pump and sump discharge lines from Th11 Unit 2.
- c. DELETED d DELETED
- e. Sewage system for the temporary personnel access facihty.
- f. Domestic water gnd heat trace
- g. Turbine Building crane and clevator.
Deactivated passive functions in the Turbine Buildtng are Reactor Building containtnent isolations and contamination isolations. 7.1.8.2 Facility Description The Turbine Buildtng is rectangular in plan with three main floors of slab-beam construction. It houses the main turbine and condenser, a portion of the steam supply system and the condensate and feedwater systems. During PD51S, the Turbine musiliary and support systems will remain deactivated. The Turbine Building does not have flood protection. Equipment, componems and parts may be removed from deactivated systems provided their remova! does not adversely affect the PDhis function of the system (s) involved in the Turbine Buildmg. 7.1.8.3 Evaluation The Turbine Building Configuration for PDhtS minindres sources of contamination, therefore the petential of spread of contamination is very low. 7.1 8 UPDATE 2 - AUGUST 1997
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- w. G G TABLE 7.1-1 OPERATIONAL ' 7 LITIES CONTAINMENT INTERNAL FACILITY DESCRIPTION ISOLATION CONTAMINATION REMARKS CONTAINMENT YES YES OPERATIONAL ONLY TO TIIE EXTENT NECESSARY TO (REACTOR BUILDING)
- SUPPORT OPERATING SYSTEMS.
AUXILIARY BUILDING YES YES OPERATIONAL ONLY TO T1IE EXTENT NECESSARY TO SUPPORT OPERATING SYSTEMS. FUEL IIANDLING BUILDING YES YES OPERATIONAL ONLY TO THE EXTENT NECESSARY TO SUPPORT OPERATING SYSTEMS. FLOOD PROTECTION NO NO MANUALLY INSTALLED FLOOD PROTECTION DOORS WILL BE AVAILABLE FOR WATER INGRESS PROTECTION. AIR INTAKE TUNNEL NO NO OPERATIONAL ONLY TO llIE EXTENT NECESSARY TO SUPPORT OPERATING SYSTEMS. i UNIT 1/ UNIT 2 CORRIDOR NO YES OPERATIONAL ONLY TO T11E EXTENT NECESSARY TO SUPPORT OPERATING SYSTEMS. CONTROL AND SERVICE NO YES OPERATIONAL ONLY TO T1IE EXTENT NECESSARY TO BUILDINGS SUPPORT OPERATING SYSTEMS. TURBINE BUILDING YES YES OPERATIONAL ONLY TO TIIE EX IENT NECESSARY TO SUPPORT OPERATING SYSTEMS. 7.1- 9 UPDATE 2- AUGUST 1997
7.2 OPERATIONAL SYSTEMS ne following sections describe the sy stems that will be maintained in an operational condition during PDMS. Table 7.2 1 provides a listing of those systems. Also listed are the system code, status of containment isolation function, internal contamination, and any relevant remarks regarding operation in the PDMS configuration. Each of the following sections addresses the PDMS function of the specific system, system description, and an evaluation. A list of references providing more detailed information is included in Section 7.3. As required by the PDMS Quality Assurance Plan, procedures are in-place to control the operations, surveillance, testing, modifications, and maintenance of the operational systems. 7.2.1 CONTAINMENT SYSTEMS ne following systerns or portions of systems provide the necessary measures to ensure that the Contamment matntains its pnmary function as a contanunation bamer throughout PDMS:
- a. Portions of systems consisting of Containtnent penetration isolations
- b. Containment Ventilation and Purge System
- c. Contsinment Atmospheric Breather
- d. Contamment Airlocks and Equipment Hatch.
7.2.1.1 Containment isolation \ An operating nuclear power plant requires doubP barrier isolation capabilities so that no single, credible failure or malfunction of an active component can result in leakage due to loss of Containment isolation, ne installed double barriers in piping systems usually include vanous types of valves located inside and outside of the Reactor Building penetration. Due to the non-operating and defueled condition of TMI.2 during PDMS, there are no piping systems which penetrate the Contamment that require double barrier active isolation capability, nerefore, except for the Containment Atmospheric Breather and RB pressure indicating piping, all piping penetrations will be isolated by a single, passive barrier either inside or outside of Containment. locked closed valves, closed and deactivated remote manual valves, closed and deactivated automatic vahrs, and blind flanges are examples of acceptable means of Containment isolation during PDMS. Table 7.2 2 describes the means ofisolation for each Containment penetration. Any system required to be operated during PDMS which necessitates the opening of a Containment isolation feature shall be controlled by procedures which delineate the requirements for opening and reestablishing isolation of the penetration. Several Containment penetrations were modified during the cleanup period to preside capabilitics necessary for cleanup operations. The piping penetratinn moditications installed during the cleanup period were designed to withstand a maximum of 5 psig of pressure, (Reference 7.3 12). All other - piping penetrations that have not been modified are presumed to have a retained capability at or near their original design pressure. Although Table 7.2-2 provides a listing of specific means of Containment isolation, any equivalent V 7.2-1 UPDATE 2 - AUGUST 1997
means of isolation is acceptable and will not constitute a change to the Containtnent isolation capability. For cumple, a vah c may bc removed and replaced by a blind flange and result m no change in the Containment isolation capability. 1 7.2.1.2 Containment Atmospheric Breather 1 7.2.1.2.1 PDMS Function he Containment Atmospheric fireather has la added to the Containment to provide passive pressure control of the Containment relative to ambient atmospheric pressure via the Auxiliary Building and to , establish a "most probable pathway" through which the Containment will " breathe." The Containment Atmospheric Breather is designed to provide a specific pathway through which the Containment atmosphere can aspirate to maintain pressure equihbriuri with the emironment external te the i Containment. He Breather is designed so that the Containment atmosphere will preferentially aspirate ! through it to the Auuliary Building rather than through other potential Containment leak paths. In l addition to assuring that the Containment structure will not exper:ence significant pressure differential, positive or negative, which could threaten the structural capability of the contammation boundaries prmid-d by the Containment, the pathway assures that efiluents from the Containment to the emironment will pass through the filtered Dreather pathway. De Breather is a passive system which requires only periodic inspection of the IIEPA filter and maintenance of the system isolation valve. Besides the isolation valve, there are no active parts which can fail and cause the Breather to become inoperative. He Containment Atmospheric Breather model is shown on Figure 7.21. He Breather is assured to be the "most probable" patimay because the size of the Breather is very large compared to other potential leak paths. He analysis wiuch demonstrates that the Containment Atmospheric Breather is the most probable patimay is given in Section 7.2.1.2.3. Finally, the unfiltered leak rate test required by PDMS T echnical Specifications 3/4.1.1.2 ensures that the flow through the breather remains greater than 100 times the flow through the other unfiltered leak paths. 7.2.1.2.2 System Description ne Breather is a passive system consisting of a 6 in. diameter duct with A IIEPA filter and a welded plate installed downstream of the IIEPA that holds four sample filter paper frames. Each frame holds a set of two sample filter papers that can be remosed for radionuclide analysis. De PDMS configuration is shown on GPUN Drawing 302 2041 De Hydrogen Control System line is used as the Containment Atmospheric Breather. Filter position All F 33 contains a 24" x 24" x 11 %" IIEPA. There is an isolation valve between l l 7.2-2 UPDATE 2 - AUGUST 1997
Contairunent and the Breather IIEPA filter that will automatically close upon receipt of a Containment pressure increase of % psi. The purpose of this isolation is to protect the Breather liEPA filter in the event of a significant fire in the Reactor Buildmg.' he Breather is operated in the following modes:
. Passive Breathing All V 3A, AII V 52, All V 153, Ali V 154 and All V 25 are open anc' All V-4 A and Ali-%120A are closed. A filter housing door downstream of All F 33 is opened in this configuration, the Reactor Building is allowed to naturally aspirate via a l{ EPA filtered patimay to the Auxiliary Building which, in turn, either naturally aspirates or is ventilated to the etnironment through yet another set ofliEPA filters. In the event ofloss of air or loss of power, All %52 will fait closed.
. DOP Testing DOP testing of the llEPA filter is performed without the sample filter paper frames in place. ANSI N5101980, Testing of Nuclear Air-Cleaning Systems will proside guidance in the performance of DOP testing of the llEPA filter.
Prior to operation of the RB Purge System, the RD Breather will be isolated. In this configuration, valve All %120A, All V153, All V154, All V 25 and All V 52 will be closed and valves All V 3A anc All V-4A will be open. Prmisions have been made to allow semi annualsample filter paper removal and assay and reinstallation or replacement of the llEPA filter. 7.2.1.2.3 Evaluation his section demonstrates that the Containment Atmospheric Dreather is the "most probable pathwa)" f by which the Containment can discharge air to (or intake air from) the emironment. It is presumed that the Contamment Atmospheric Breather can be deemed the "most probable pathway" if the mass - flowrate through the breather system m response to an atmospheric pressure change is orders of magnitude larger than the mass flowtates through all other pathways m response to the same pressure changes. He mass flowrate through the Breather in response to a pressure differential can be calculated from its flow resistance. Sunitarly, the mass flowTate through "all other paths" can be calculated from the flow resistance for "all other paths " He flow resistance for "all other paths" can be calculated from the rate at which the pressure in the Containment attempted to achieve atmospheric equilibriu n when the Containment was scaled and pressurized. For the purpose of calculating the flowrates, the Containment is visualized as shown in Figure 7.2 1. The various known or potential leaks have been lumped together as an " equivalent" leak. He Contairunent Atmospheric Dreather has been modeled as a 30 ft straight length of 6-inch diameter pipe with one llEPA filter. ( 5 psig. The maximum overpressure in Containment from the postulated worst case fire is estimated to be approximately (A
' 7.2-3 UPDATE 2- AUGUST 1997
The flow tiuough the Breather vent and the " equivalent" leak will te calculated ustng the extended Bernoulli equation. For the Breather; (P./p) + (%/2g.) = G'./p) + [f(1/d)] (%/2g.) + (DP/p) (1) where: P = the pressure in the Containment p = density of Containment air
% = the (negligible) velocity in the Cont 2inment
- g. = gravitational constant = 32.17 Lbm-ft/1.bf sec' P. = the ambient pressure I = pipe length d = pipe diameter f = friction factor for flow through the 30 ft pipe
% = velocity in the Breather vent pipe DPr = pressure drop across the !! EPA filter Since the breather system is designed to allow the Containment to respond to small changes in atmospheric pressure, the pressure differences will be small and the flow in the 6 inch diameter pipe will l assumed to be latninar. In that case, f = 64/Re = 64/(d%p/ ), where p is the absolute viscosity of The velocity in the pipe can be wntten in terms of the rnass flowrate in the pipe:
% = (4md(p nd')) (2) where:
m, = the mass flow rate in the Urcather vent pipe d = the diameter of the Breather vent pipe ne pressure drop across the IIEPA filter can also be wntten in terms of the mass flowrate m the pipe. (DP/p) = (lbp') (3) where: K is the rated pressure drop across the filter of 1 inch of water at 1000 CFM which equals 0.312 lbf sec/ft'. (Note that this assumes the llEPA filter pressure drop is linear with flow. His is conservative since the actual pressure drop will be less at the expected lower than 1000 CFM flow rate which in tum would allow raore flow through the Breather.) 7.2-4 UPDATE 2 - AUGUST 1997
Substitutmg into equabon (1): l b = (P. P.) / [(128 /(n p g ))(1/d') + (K/p)] (4) For the *cquivalent" leak, the result is the same escept tlut the tenn for the llEPA filter is absent. rm = (P. P.)/ [(128 /(n p p))(l'd')) (5) In this case, the quantity (1/d') must be detemuned. To fmd an equivalert value of(t/d') for the leaks, data ftorn the leak test of the Containment were used in the test, the proportional leak rate was takulated as 0.0852% per day when the Containment was held at 70.6 psia. Since the pressure in the Containment is proportiona to the air mass in the Containment, the proportional leak rate is the leak path mass flowrate, mi, divided by the air mass in the Containment at the time of the rneasurement.
.0852Wday = mi/M = mi/ (pV) where:
M = the Contamment air mass V = the Containment free volume = 2.1E6 ft' , Converting the leak rate from % per day to inverse seconds, and combining with equation (5): (0 000852/(24 x 3600)) = (P. . P.) / [(p V)(128 /(p ng ))(1!d')] (6) As a result: h (t/d')a = (P. - P.) / [(128V /np) (0.000852 / (24 x 3600))] Which, in turn, leads to the determination that the equivalent value of(1/d') for the leak paths is: 4 (1/d')a = 2.61E10 ft [lf the length of the leak path is on the order of the Containment wall thickness (i.e.,4 ft), the total leak diameter would be 0.042 inches.] He ratio ofleak flow to liteather vent flow can then be wntten as' mi / m, = [(128 Ing,)(lld') + K) / [(128 Ing,)(l'd')a] (7) with: (1/db = 480 ff' K = 0.312 lbf sec/ft' His gives the ratio of mass flow rates as: (rni / m,) = (0.00717 + 0,312) / 3.89E$ = .000001 Therefore, the Containment Atmospherie Breather elcarly is "the most probable pathway" p 7.25 UPDATE 2 - AUGUST 1997
As stated in Section 7.2.1.2.2, there is a welded plate installed downstream of the !! EPA filter that holds i four sample filter paper frames; each frame holds s set of two filter papers. The air flow into and out of the contatnrnent via the Hrcather also passes through each set of two sample filters. (For the purposes of this discussion, the sample filter papers closest to the Dreather llEPA will be referred to as the No I filters and the sample filter papers farthest from the Dreather !! EPA and closest to the Auxiliary lluildmg atmosphere as the No. 2 Filters ) ne Dreather llEpA filters the air leaving the Containment into the Auxiliary Building. Filter No. I collects the matenal that may pass through the llEPA filter. Filter No. 2 filters and samples the air cenne back into the Contamment from the Auxiliary Building All four of the No. I Filters are rernoved semi annually and one is assayed for radioactisity. If any aetnity is found on the filter, it will be assumed that, for the samphng time penod, a hke amount of actisity was released from the Contaminent into the Auxiliary Buildmg (i c., an assurned eniciency of 509 his is a very conservative approach since the sample filter papers used have a collection efliciency of grcater than 90*b Using this rnethodology, any activity assumed to be released is captured on the filter and will be assmed to have been released over the six month time period. Since the filter deposition is cumulative, this method provides determinative (but not real time) rnorutoring to verify that efHuents through the Hrcather are wittun the calculated values m Chapter 8. Due to the extremely low releases calculated for PDhtS, the sample filter paper is deemed adequate for detennitung the releases anticipated during PDhtS. 7.2.1.3 Containment Ventilation and Purge 7.2.1.3.1 PDhtS Function During PDhtS, the Containment Ventilation and Purge System ensures that uncontrolled atmospheric migration of radioactive conteunation will not ercate a hazard to either the public or site perronnel. 7.2.1.3.2 System Description The Containment Ventilation and Purge System will be maintained in an operational condition to support actisities in the Contairunent (e g., surveillance entnes, matntenance) during PDNIS. Testing to ensure operabihty of the Contairunent Ventilation and Purge includes llEPA filter pressure drop, eshaust flow rate, DOP testing (guidance provided by ANSI N5101980), and visual inspection of the filter train. He Containment Ventilation and Purge Systern consists of a single operational Containment purge exhaust unit, make-up air supply associated ductworh, dampers, and filters De purge exhaust umt (maximum flow 25,000 efm) draws air from the D nngs through HEPA filters, and discharges to the station vent. ne PDhtS contiguration is shown on GPUN Drawing 302 2041. 721.33 Evaluation Operation of the Containment Vent and Purge System prosides fresh air to the Containment while prosiding a filtered, monitored exhaust path. Atmospheric radiation monitoring, as described in Section 7.2.4, provides for monitoring of airbome releases from the system by using momtors located in the exhaust duct and in the station sent. nis ensures that releases from the Containment to the emironment are minimal. 7.21.4 Containment Airlocks and Equipment Hatch 7.2.1,4.1 PDhtS Function 7.2-6 UPDATE 2 - AUGUST 1997
The airlock doors will be used during PDMS for Containment ingress and egress. The airlocks are O Q designed as a double dxr system with one of the doors always closed during routine entry into the Containment Under these conditions, the Containment remains isolated (an enclosed volume) at all times aad the amount of Containment air which could be released to or from the Containtnent is limited to the volume inside of an airlock assembly. 7.2.1.42 System Description the Equipment flatch is heated in the southwest quadrant of the Containment. It is a 24 ft. 8 inch diameter,20 5 ton cover for a 23 ft diameter penetration in the Containment wall. Its design purpose is to accommodate the movement of large objects into and out of the Containment. A removable personnel airlock assembly (airlock 1) is incorporated into the Equipment llatch. The airlock has a 9 ft. outside diarneter, a 12 ft 6 in. length, and weighs approximately 15 tons. Both the Equipment flatch and the airlock are double gasketed with the Equipment flatch bolted to a steel flange on the Containment wall. A separate personnel airlock assembly (airlock 2) is located in the southeast quadrant of the Contamment. Its use is intended primarily for personnel access and is permanently mounted in the Containment wall. Both personnel airlock assemblics are manually operated and require no electric power to open or close. 7.2.1.4.3 Evaluation There are situations when it is necessary to open both doors of an airlock assembly simultaneously, as is the case of movement of a long piece of equipment into or out of the Containment. The relevant issues associated with opening both airlock doors simultaneously during PDMS include: (
- a. Release of radioactivity during maintenance or surveillance activities, transient conditions, or other natural or man-made events,
- b. Assurance that the airlock doors can be closed.
Typically, both airlock doors would be open only for the period of time necessary to complete the relevant aethity and the Reactor Buildmg Purge System would be operating, thereby minimizing any effluent. Natural phenomena, such as floods and high winds, are not considered to pose a safety concern because there is adequate warning to ensure that the airlock doors could be closed prior to a significant rHease. From a seismic standpoint, the airlocks serve no structural function, and therefore, will not affeu the seismic integrity of the Containment. 7.2.2 FIRE PROTECTION, SERVICE, AND SUPPRESSION Fire Protection is provided during PDMS to minimize the ;ersal for a release of radioactive material due to a fire f . A contaminated area, protect systems ma#,ained opCrational during PDMS, minimize the liabihty and property risk from potential fires, and mimmize any potential tisk to the operating unit on the same site. O) U 7.27 UPDATE 2 - AUGUST 1997
i 1 nese objectives are achieved by minimizing the potential for a fire by strict control of combustible l taaterials and ignition sources and by presiding a system of detection and suppression suitable to ) deal with any potential fire. 7.2.2.1 PDh1S Function I ne overall fire protection objectives are achieved by providmg a system of fire protection features designed to ensure the following prmuu y functional requirements:
- a. Fire detection shall be provided to the extent that any credible fire will be detected
- b. Portable fire extinguishers shall be provided in areas of the facility, as necessary, to provide adequate fire suppression capability.
- c. Presence of flammable and'or combustible liquids and materials shall be nummired to the maximum extent practical.
7.2.2.2 Systern Description ne original Thil 2 system of fire protection has been modified to address the functional requirements for fire protection for PDhtS. He Fire Detection, Service, and Suppression System as configured for PDMS are shown on GPUN Drawing 302 231, Sht.1. De measures required to provide the necessary degree of fire protection are desenbed below.
- a. He yard fire main will be maintained pressurized using the station fire pumps in Unit I with the altitud tank as a backup water souret
- b. A heat sensitve wire fire detection system has been installed which provides detection capability in the Reactor, Auxiliary, Fuel llandling, Control, Senice and Turbine lluildings. The detectors are divided into sis (6) rones with the fire and trouble alarms transmitted to the control panel on the TMI Unit 1 Turbine Deck that alarms in the Thfl.1 Processing Center. The detector will actuate when its temperature reaches 150 to 165 degrees Fahrenheit. The t installation of this system supports the plan to remove the original plant high voltage detection system.
The operable portion of the fire detection and alarm systems will be tested annually by perienning channel functional tests and supenised circuit tests. For non-supenised circuits between the local panels and the remote monitoring station-in the Thti 1 Control Room, testing will be perfonned every 31 days to demonstrate operability. I c. Equipment-related fire detectors, installed on various components within the plant
- to monitor a specific hazard and automatically trip the associated fire suppression j system, have been deactivated along with the related fire suppression sy stem.
l 7.2-8 UPDATE 2 - AUGUST 1997
4
- d. De lialon systems' protecting the Air intake Tunnel have been deactivated by l
- removing the lialon cylinders and deenergizing the ultrasiolet arxl pressure ;
detectors. The llalon system protecting the relay room has been deactivated by removing the cylinders. De fire detection system will remain operational to I monitor these areas. 4 c. Portable fire extinguishers and self contained breathing apparatus are staged with ! cmergency response crew equipmer,t. Additional portable fire extinguishers are
- located throughout the plant as needed to support work activities.
- f. Transient combustibles inside the Containment and the AFilB have been removed to the maximum extent practical.
- g. The oil has been drained from the main turbine and lube oil reservoir, feedwater i pump turbines, emergency feedwater pump turbine, emergency feedwater pumps, condensate pumps, condensate booster pumps, and the hydrogen seal oil unit.
- h. The charcoal filters have been removed from all IWAC systems in Unit 2.
- i. The 12 in, fire service loop, which runs through the Diesel Generator ,
Building, AFilR, Control Building area and TurW.ne Building (east and west), i has been cut and capped off. The Diesel Generator Building has been turned over to TMI 1. Fire Service Water System standpipes have been configured to the East and West side of the Turbine Building which permits connection of the ** cal fire hydrants to the 331' elevation of the Turbine Deck by way of ! staged 0.9 hoses. This will allow responsive action by the Station Fire : y Brigade and or local Fire Depauments, r t J. All portions of the Fire Protection bystem located inside buildings in areas where the fire harard risk is small have been deactivated.
- k. The station fire brigade is fully trained to assure that the personnel are familiar with system configurations, plant layout, and the procedures in Unit 2.
- 1. The Fire Protection Program and housekeeping inspections and their frequency are addressed in plant procedures,
'The air intake tunnel halon system was removed because the probability of an airplane crash in the vicinity >
of the air intake tunnel was estimated to be less than 2E.7/ year and because of the presence of heat activated
- detectors that will shutdown the ventilation system W close the venti fan dampers upon detection of a fire. i 7.29 UPDATE 2 - AUGUST 1997 . , .ms, . , . - - , . - . . , , ~ , . , . - . . . ... . _. _.......-_.,__,-._ -. .._ .~ , - . _ . . . - . _ . . . _ , . . . - . ,
7.2.2.3 Evaluation i ne scope of fire protection has been reduced for areas in which systems have been deactivated and cornbustibles have been sigmficantly reducud, so that the corresponding fire hazards have been miminized. Deluge systems in the airhntake tuttnel, the Auxiliary, the Turbine, and Control Buildings have been deactivated for PDMS. Rese are no deluge systerns in the Containment. Detection desices proside contacts for st:penisory indication that each dnice is operatiorn! and, in the event of detector actuation, indicates the location where the wire detector actuated the alarm. The station fire brigade is under the supenisory control of Unit 1. Upon detection of a fire in Unit 2, the ststion fire brigade will respond to the specific location in Urut 2. His response in accordance with ongoing station fire brigade training and procedures will ensure mitigation of a fire in Urut 2 dunng PDMS. He fire protection and suppression systerns are configured to provide adequate capability to eninguish any potential fire during PDMS. O 7.2-10 UPDATE 2 - / < JST 1997 l
l s 7.2.3 RAD 10 ACTIVE WASTE MANAGEMENT Liquid radwaste management systems that are operational during PDMS are the Radioactive Waste - Miscellaneous Liquids System and the Sump Pump Discharge and Miscellaneous Sumps System. Major portions of these two systems are operational during PDMS to prevent localized flooding and to pro ide proper disposal of effluents. 7.2.31 Radioactive Waste . Misce!!aneous Liquids 7.2.3.1.1 PDMS Function
+
During the PDMS period, portions of the WDL system remain operational. This status prosides assurance that significant quantities ofliquid wastes will not accumulate in an uncontrolled manner in the Auxiliary and Containment Buildings. Liquid radwaste in these buildings may result from either rainwater inleakage or PDMS activities, ne WDL system achieves its objective by meeting the following criteria.
- a. Existing sumps in the Auxiliary and Containment Buildmgs will be morutored and pumped, as required.
- b. Liquid storage capabihties have been provided for accumulation until sufficient quantities are available for batch processing, as necessary.
f' c. Existing Unit I and Unit 2 WDL system tie ins have been restored to prmide ( capabihty to process Unit 2 low level 1. quid radwastes. 7.2.3.1.2 System Description The WDL system is designed to tr;ccive liquids from the Auxiliary and Fuel llandling Buildings, t assorted equipment in these buildings, and from the Containment sump. This system has the capability to retain waste liquids for radioactive decay, sampimg, filtering, or transfemng liquids 4 to TMI l for processing and/or dispoal. He PDMS configuration is shown on GPUN Drawing 302 2045. Liquid waste sourccs from the AMIB and Containment include tank drains, vents, Alter drains, flush line drains, and floor drains. Rain water inleakage into the ATHB also accumulates in the Auxiliary Building sump. Some floor drains have the potential of spreading contamination from their drain lines back into the room. Selected drains may have a ball float valve device or plug installed to preclude tbc spread of contamination, ne WDL components that will be utilized during PDMS include:
- a. MWHT (WDL T 2) A 19,518 gallon capacity, stainless steel, horizontal tank provioes the waste water feed for transfer / processing. It collects water from the operable sumps, which are preferentially lined up to it, ne tank hr.s redundant inlet fihers (WDL F 8A & B) to partially clean up the influent.
V "'2-11
. UPDATE 2 - AUGUST 1997
i
- b. Sumps ne Containment Basement and Auxiliary Building sumps are lined with stainless steel. Level indication will be maintained for the Auxiliary Buildmg Sump and Containment Basement sumps.
- c. Sump pumps Le Auxiliary Building sump has dual pumps that operate manually as needed he Reactor Build.ng Spray and DilR sumps (four total) have a single pump that operates manually as needed Capacities of the pumps are as follows: Auuliary Buildmg sump purnps 100 gpm; Reactor Building Spray and DliR sump pumps 50 gpm. nese pumps will be maintained operational and placed in a manual control mode.
d ABST (WDL T 5)- A 3,085 gallon capacity, stainless steel, honzontal tank and its associated equipment will be maintained operational as a backup in case the MWiiT is not available for extended periods. Selected WDL tanks (e g , the MWitT and ABST) have cartridge type liEPA filters installed on selected opening (s) to protect against airbome releases from these tanks. Deactivated WDL tanks are vented to the waste gas disposal header which has a cartridge type liEPA filter installed to protect against airborne scleases. Dunng PDMS, the Auxiliary Building Ventilation System is not required for waste wat:r transfer operations in the Auxiliary Buildmg. nem f ransfer operations do not pose an airbome problem since any airbome coritaminants would be released from the water prior to entenng the MWET or the ABST. Contaminants stirTed up m the MWiiT or ABST will be contained by the liEPA filters installed on these WDL tanks. He Containment Sump and Sump Pumps are in an undetermined condition. Because the existing plant sump pumps WDL P-2A and 2B have not been refurbished, an altemate flow path will be utilized to drain down the RB basement sump. A tie-in to the WDL system has been provided . outside of penetration R 593. His tie-in nms from the decay heat temoval and buildmg spray pump suction header isolation valves to the "A" RB spray pump room sump via BS P I A drain valves. 7.2.3.1.3 Evaluation Because a majority of plant systems are deactivated, have been drained, and placed in a layup condition, there are a limited number of activities that can generate liquid waste during PDMS. Liquid waste in the remaining systems and accumulated water inleakage will be adequately handled by periodic batch processing to TMI l using the operational portions of the %T)L system with TMI l cross-ties and discharged via approved pathways. ni: ensures minimum exposure to plant personnel and numnuzes releues to the emironment in accordance with 10CFR20 and $0. l 7.2-12 UPDATE 2 - AUGUST 1997
7.2.3.2 Sump Pump Discharge and hiiscellaneous Sumps System 7.2.3.2.1 PDMS Function nere are a number of sumps in TMI 2 that will be maintained in an operational con & tion during PDMS. He various sumps and their locations are listed in Table 7.2-3. Maintaining the various building sumps operational assures that water buildup does not cause adverse localized floodmg. These sumps will contain water that is either clean or slightly radioactive. Clean water is presently routed to the Industrial Waste Treatment System (1%TS) Ra&oactive water will be processed and &scharged via approved pathways; slightly radioactive water will be pumped to the IWTS and released in accordance with 10 CFR 20,10 CFR $0 and NPDES regulations. The discharge from the IMTS is morutored for ra&ation in accordance with the ODCM. 7.2.3.2.2 System Description The design 2 of the various sumps are delmeated in the applicable documents referenced m Section 7.3. He PDMS configuration is shown on GPUN Drawing 302-2496. He sumps have the capabihty of being pumped automatically with the pumps controlled by float switches; however, they will normally be operated in the manual mode with a lugh level alarm that annunciates in the control room and the PDMS Alarm Monitoring System. Sump level is monitored by level detectors located in the respective sumps The exceptions are the Circulating Water Chlonnator Huiling. Circu'ating Water Pump ilouse, and the Air intake Tunnel Norma sumps which will employ portable sump pumps to pump down the sumps as necessary. Water from the floor drains that enters these sumps is tenerally not contanunated, although sumps within the Turbine Buildmg, Control Buildmg Area, Control and Service Building, and Tendon Access Gallery have recirculation and grab sample lines to pemut sampimg for radioactisity. 7.2.3.2.3 Evaluation In general, the functional requirements of each sump and sump pump have been determined on an indnidual basis. Monitoring oflevel in the various sumps by remote means and/or visual inspections ensures that accumulated leakage is transferred for processing in a timely manner. Sampling will be used to quantify radioactive content and ensure proper waste stream processing. Herefore, operation of the sump pump ischarge system casuret liquid waste streams generated dunng PDMS are adequately transferred for ultimate pmcessing and do not adversely affect the PDMS plant conditions. 7.2-13 UPDATE 2 - AUGUST 1997
l 7.2.4 RADIATION h10NITORING 7.2.4.1 PDhtS Function During PDhtS, the radiation monitonng requirements for the facility are primarily those associated with assuring the stability of the radiological conditions in the facility and effluent monitoring. He off site dose calculations for normal time periods and unanticipated events (see Chapter 8) are based on assurned and measured radiological con &tions associated with the various areas of the facihty, in order to assure that the off site dose calculations for the various events remain bounding, the radiological con &tions must be periodically monitored to assure they remain within acceptable bounds. In ad6 tion, all efiluents must be monitored to assure all off-site releases are within acceptable bounds, as well as to meet regulatory requirements for effluent reporting. [3 roader radiological conditions monitonng will be conducted throughout the facility to assure compliance with good radiological conditions practices and 10 CFR 20. Rese radiological monitoring actisities are required to support other PDhtS activities such as visual inspections, preventive maintenance or other routine tasks. 7.2.4.2 Radiological Suntys 7.2.4.21 AFIIB Radiological Suneys Radiological surveys will be conducted on a periodic basis to monitor radiological conditions in the Auxiliary and Fuel 1-landling Buildmgs. These radiological surveys will be conducted quarterly and will consist of air sampling, loose surface contarmnation, and radiation does rate suntys. In adation, TLDs may be placed in fixed locations and changed out periodically to monitor dose rates over a long-term period. Radiological survey results will be reviewed and evaluated for trends to prmide early detection of deteriorating radiological conditions. 7.2.4.2.2 Containment Radiological Surveys Periodic Containment radiological surveys are required to provide information regarding the stability of the radiological conditions inside the Containment. As stated in Section 7.2.4.1, this infomsation is necessary to perioically validate the off site releases as calculated in Chapter 8. Radiological suntys just outside the containment airlock doors will be conducted quarterly, as expressed in Regulatory Guide 1,86 Position 3.C. Radiological surveys inside containment will be conducted semi annually, as a minimum, at the approximate locations shown on Figures 7.2 11 and 7.2 12. hionthly radiological sun eys in the Containment were performed atter the RB was placed in its PDh15 condition in order to develop an adequate data base. These surveys consisted ofloose surface contamination and radiation dose rates at all survey locations and at least one air sample inside the contamment. De semi-annual surveys will collect data from the same lo:ations. In addition, TLDs nuy be placed in fixed locations and changed out periodically to monitor dose rates o er a long-term period. These surveys will be reviewed and evaluated for any indicated trends. His will either provide assurance that contamination conditions inside the Containment are stable or will provide early indication of any changing con &tions which may require corrective action. 7.2-14 UPDATE 2 - AUGUST 1997 O
l 7.2.4.3 EfIluent Monitoring Airbome efDuents will be monitored during active and passive ventiladon of the Containment. Periodic operation of the Reactor Building Purge may be necessary danng personnel entries. During Reactor Buildtng Ventilation System operation, the station venulation stack monitor, llP R 219 or llP R-219A, will provide real time monitoring of releam ne Reactor Building effluent monitoring system is shown on GPUN Drawing 302 2219. During periods when the Containment ventilation systems are not operating, airborne effluents frorn the Containment will continue to be monitored as discussed in Section 7.2.1.2.3. ne Containment is passively vented to the Auxiliary Buildmg through a breather pathway which will be filtered using a llEPA filtration system. Dunng periods of Reactor Building Purge operation the brt.ather pathway will be isolated. During passive ventilation the purge exhaust will be isolated and the Containment wid be vented through the breather IEPA filter. On a semi annual basis, a sample filter paper installed downstream of the IIEP A filter will be assayed for its radioactivity content to evaluate any release to the emironment dunng periods ofinactisity. ne Containment Atmospheric Breather System allows pressure equahzation between the Containment and the emironment via the Auxiliary Building. For this reason there is no motive force to cause contammation to leak out other than througn the Containment Atmospheric Breather. Herefore, it is not anticipated that releases to the emironment will occur through pathways other than the breather. Operating procedures for Containment isolation and operation of airlock doors will to used during PDMS to ensure isolation is maintained in the AFHB, negligible airbome efIluent discharges are anticipated during normal events. This conclusion is based on three factors: 1) de facto isolation of the AFHB,2) prevention of airbome events within the AFHB, and 3) the periodic monitoring of radiological conditions. Access to the AFHB is limited and controlled by site procedures. he ventilation system exhaust, as shown in GPUN Drawings 302 2042 houses a prefilter for large particulates, and two banks of DOP tested HEPA filters in series With the AFilB essentially isolated, there is no motive force to generate sigruficant airbome contamination levels, and any airborne contamination that might develop is filtered by the ventilation system exhaust pathway prior to release to the Station Vent. Prevention of airbome contamination within the AFHB results, in part, from the lessened level and frequency of plant system operations and reduced access and actisities of plant personnel. In addition, intemally contaminated systems inside the AFHB are drained ofliquids, and isolated by closing the respective boundary valves. Spent fuel pool "A/B" were sealed since they contained loose contammation suHicient to pose a contamination spreading concem 7.2-15 UPDATE 2 - AUGUST 1997
i nroughout PDMS, an ongoing radiological surveillance program will monitor radiological conditions within the AFliB. By means of various suneys, as described in Section 7.2.4.2, potential degradation of radiological conditions will be identified in order for appropriate remedial actions to be taken A special monitoring program of AFIIB airbome levels, see Section 7 2.4.3.1, was conducted for a one year period prior to PDMS, and was continued for a minimum of one year after implementation of PDMS. He information gathered during these evaluations constitutes an extensive data base that provides additional assurance that AFIIB airborne emuent releases will be insigni6 ant in nature. In addition to the special monitoring prograrn, whenever the AFIIB ventilation systems are operated during PDMS, the llEPA filtered exhaust is also monitored by the real tune sampling of the Station Vent Monitor, thus assuring a controlled, monitored emuent release. Considering that the AFlIB has orders of magnitude less contarmnation than the Contamment the airbome emuent controls described above are sumcient for assuring airborne emuent releases from the AFilB during nonnal events will be insignificant. A certain amount ofinleakage into sumps is anticipated during PDMS and penodic discharges will be necessary, initial samples will be taken and analyzed to quantify radioactive efDuents. All radioactive liquid discharges will be via an approved pathway which will provide dilution and monitoring capabilities. 7.2.4.3.1 AFilB Airbome Evaluation A special morutoring program was designed to evaluate particulate airbome concentrations in the AFIIB prior to, and after, entry into PDMS. He purpose of the evaluation was to determine the airbome levels in the AFlIB during steady state conditions. The most representative sample point of unfiltered AFiiB airborne particulates was directly upstream of the plant ventilation system IIEPA filter banks. Installed plant momtoring capabilities existed at both of these locations, i c., llP R 221 A for the Fuel llandling Building and IIP R 222 for the Auxiliary Buildmg. % moving filter paper mechanisms of these plant monitors was disabled, creating fixed filter sample points. %e filter papers at these plant monitors were periodically changed out, and air sample results were reported quarterly. The special monitoring program was temporanly suspended whenever plant activities in the AF11B were expected to generate significant airbome levels. He special AFIIB airbome evaluation concluded that the average particulate airbome generation rates of the Auxiliary and Fuelllandling Buildings were significantly below the 2.4 E-4 uci/sec acceptance criterion. He acceptance enterion was less than 1% of the TMI 2 Recovery Technical Specification for release of rate particulates with halflives greater than eight days. This corresponds to less than 1% of 0.024 uci/sec when averaged over any calendar quarter. 7.2 16 UPDATE 2 - AUGUST 1997
0 ( 7.2.4.4 General T.adiological Monitoring it is anticipated that the routine radiological surves will only be performed in areas requiring access for visual inspection, preventive maintenance, or other routine tasks. "High radiation,"
"high contamination," and scaled areas will not norraally be accessed for routine sun'eys unless access is required for some other purpose. Radiological support of work during PDMS will be conducted in accordance with Radiological Controls procedures and good radiological work practices.
7.2.4.5 Evaluation ne radiological effluent and monitoring programs described above address the principal radiological cencerns for PDMS. These programs assure the radiological conditions in the facility are moaitored and any significant deteriorating conditions will be identified in a timely period and appropriate correction action taken. Also, both liquid and gaseous effluents are monitored to
- .ssure all radioactive releases are within acceptable bounds. These monitoring programs, in cmjunction with general radiological controls activities, assure that the radiological aspects of PDMS e : appropriately addressed.
7.2.5 ELEC'lRICAL SYSTEMS During PDMS, sarious plant systems will be required to remain operational to support the monitoring, protection, and surveillance activities associated with PDMS. Some systems require T { s continuous operation while others require only intermittent operation. Due to the need for electncal power support for these activities the PDMS Electrical Distributic, System will be maintained 4 operational and remain energized during PDMS. } 7.2.5.1 PDMS Electrical Distribution System 7.2.5.1.1 PDMS Function During PDMS, the TMI-2 Electrical Distribution System will be maintained operational and energized to provide reliable power mmces for the PDMS support systems and their associated controls and instrumentation. Power vil also be available for area lighting, receptacles, heating and ventilation to support PDMS sarveillance activities. In some instances, systems utilized for PDMS surveillance activities may require nergization from local control stations prior to commencing the surveillance actisity. 7.2.5.1.2 System Description ne TMI 2 Electrical Distribution System is powered from a 13.2 KV offsite power source. The 13.2 KV/480 VAC transformers, in Unit Substations 2-31,2-32,2-35,2-45 and 2-37, proside 480 VAC power to lo:ations in the Turbine, Senice and Auxiliary Buildings. Unit Substation 2-31 provides 480 VAC to bus Unit Substation 2-22E in the Control Building. All of the PDMS electrical loads are consolidated on these six buses. (D)
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V 7.2-17 UPDATE 2 - AUGUST 1997
Unit Substations 2 31 and 2 32 provide 480 VAC to five (5) Motor Control Centers (MCCs), MCC 2 3111, MCC 2 33A, niCC 2 31A, MCC 2-32A and MCC 2-42C. These contain combination motor starters using molded case circuit breakers and magnetic contactors. He low voltage 120/208-volt AC distribution system supplies control, instrumentation, and power loads requiring unregulated 120/208-volt AC power. It consists of distribution panelboards, branch breakers, and transformers located in and powered from 480-volt MCCs through 480CO-volt dry-type transformers. A 125 volt rectifier provides Dr oower to a single distribution panel. The rectiner is normally fed from Unit Substation 2 22E. In me esent of a power loss, an automatic transfer switch will provide backup power from the Unit i Station Blackout bus. All PDMS DC control have been consolidated on this panel. He vital 120-volt AC system consists of distribution panels,2 12R and 2-22R, fed from regulated transformers. Tney receive power from Unit Substation 2-22E through 480/120-volt step-down-transformers. An automatic transfer switch provides backup power to panel 2-12R from the Unit i Ststion Blackout bus in the event primary power is lost. The regulated 120-volt AC power system supplies control and instrumentation loads as well as power for communication and annunciators. Single line diagrams of the Unit 2 AC distribution system are shown on GPUN Drawings 206-201, 206-202,206-203,206-204,3009,3010,3010,5016,3017 Shts.1,2 and 3, E021, E025, El16, 2E21-011 and 2E21-012. 7.2.5.1.3 Evaluation ne Electrical Distribution System has been modified to meet the requirements of PDMS. Due to the deactivation of the reactor and its associated support systms, Class IE emergency diesel backed power systems are no longer required, la support of this, the emergency diesel generators have been turned over to TMI-l and the Engineered Safety Feature buses no ionser have connection capability to the emergency diesel generator buses 2DG 1 and 2DG-2. The Engineered Safety Feature buses will no longer be considered Class !E. All non-PDMS support systems and components have been deactivated and isolated from the power distribution system. Admmistrative controls lunt been developed and are in place to govem the use of PDMS st pport systems and prevent unauthorized use of deactivated systems Load consolidation has been performed in order to reduce the number of energized circuits, which reduces plant maintenance and surstillance activities, thereby enhancing overall plant safety. DC power required during PDMS is supplied through a rectifier. The Electrical Distribution System, as modified for PDMS, will provide sufficient reliable electrical power to support all PDMS activities with enhanced overall plant and personnel safety. In the event that all electrical power is lost, actions will be taken expeditiously to restore power. In the unlikely event that power cannot be restored within 14 days, a report will be submitted to the NRC within 30 days detailing the plans and schedule to restore power. 7.2-18 UPDATE 2 - AUGUST 1997
} g
) 7.2.5.2 Normal and Emergency Lighting 7.2.5.2.1 PDMS Function TMI Unit 2 is provided with normal lighting systems using mercury-vapor, fluorescent and incandescent lununaries. These systems provide illumination for PDMS support actisities and for personnel safety. All lighting not required for security and monitoring actisities will be tumed off.
Lighting will be energized as needed for maintenance actisities. Installed emergency lighting wul be maintained during PDMS. One-half of the normal lighting originally designed and installed is available throughout TMl 2 except in the RB. Ne ul lighting within the RB is pro ided by strings of lights installed on the 305' and 347' elevations. The lighting is adequate to support PDMS inspection and test activities without additional illummation from permanently installed buildMg lighting Eight-hour portable emergency lighting will be carried by ensrgency personnel crews entering the buildings. This lighting will be staged with emergency response crew equipment. Routine entry crews will carry flashlights. 7.2.5.2.2 System Description The PDMS lighting system is powered from normal AC power sources; an exception to this is the RB lighting system discussed below. This system utilizes three types ofluminaries: mercury-vapor, fluorescent and incandescent. The mercury vapor lummaries are powered from 480/277-volt systems directly from the 480-volt unit substations or from 480-volt motor control Q centers. The fluorescent and incandescent lummaries are powered from 208/120-volt systems atilizing 30 KVA step-down transformers which are supplied from the 480-volt sources. In
\.V/ general, the mercury-vapor lumiraries are used in high ceiling areas, the fluorescent lununaries in almost all other areas, and the incandescent lummaries where emironmental conditions require their use. Exit signs are powered from the normal lighting system; with backup battery for these signs from the emergency lights.
Emergency lighting consists of sealed beam lamps powered by batteries which initiate operation upon loss of the normal lighting system. This lighting is provided to ensure safe egress for personnel. Additional exit information will be provided by postings. The RB normai lighting system consists oflights on the 305' and 347' elevations fed from Portable Power Distribution Centers (PPDC) or " power buggies." These power supplies were o iginally installed in the RB to support defueling activities. Two power buggies are located on Ge 305' elevation and two are located on the 347' elevation. The power feed is from either USS ;-35 or USS 2-45 and is configured such that the two power buggies on each e.evation are er.c gtzed from different sources, i.e., on each elevauon, one-half of the lighting is fed from one source a .d the other halfis fed from the other source. In the event one source of power is lost during an ete.ry, adequate Ughting would remain to assist in the safe evacuation of personnel.
,m 7.2-19 UPDATE 2 - AUGUST 1997 l
l
7.2.5.2.3 Evaluation The majority of the existing lighting sy stems remains operational during PDMS. Sufficient lighting capability is provided for anticipated support activities If further needs anse, temporary lighting will be added for specific PDMS activities. 7.2.5.3 Communications System 7.2.5.3.1 PDMS Function The TMI-2 Communications System during PDMS will provide normal communication channels throughout Unit I and Unit 2. In addition, the Communications System will provide the capability to announce alarms and alert personnel to radiation and fire hazards. 7.2.5.3.2 System Description Portions of the original system have been retained for PDMS as follows:
- a. Normal Page - Party System This system 's powered from a separate 120-volt, single-phase AC power bus.
The system is compatible with TMI Unit I and was merged with the TMI Unit I system through a merge-isolate switching arrangement in the control room to provide normal communication channels throughout TMI Units I and 2 during PDMS. The system consists of handsets, amplifiers, loudspeakers, evacuation tone generator, isolating transformer, and the necessary special equipment to preside a paging channel and three party line channels.
- b. Badio-Antenna Ssstem This system consists of antennas located at strategic points within the TMI 2 PDMS Buildings to ensure full coverage for radio communications. This system is the back-up system for a loss of the normal page-party system.
- c. Commercial Telechone Systes This system's trunk lines are leased from the Bell Atlantic Company. The handsets and switching equipment are maintained by GPU Senice Corporation personnel. This system provides links with all on site as well as off-site locations.
7.2-20 UPDATE 2 - AUGUST 1997
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Q 7.2.5.3.3 Evaluation
- The communication system will remain in an operational condition during PDMS to provide the following capabilities:
- a. Communications throughout Unit I and Unit 2.
- b. Communication for identification of fire, injury, and flood. Alarms are generated from the TMl 1 Control Room.
- c. Communication for evacuation of normally unoccupied areas. Some of the areas identified may be used for storage of equipment and thus require occasional ingress and egress.
7.2.6 PDMS SUPPORT SYSTEMS The operational systems discussed in this section provide the necessary measures to support PDMS activities. Ahhough they do not directly ensure protectiw functions, their operation is necessary to carry out anticipated operasca, inspection, surveillance, ad maintenance activities through PDMS. 7.2.6.1 Auxiliary Building Ventilation System 7.2.6.1.I PDMS Function O) The Auxiliary Building Ventilation System will be maintained in an operational condition to support PDMS activities. When in operation, this system perfonns the following functions:
- a. Provides fresh, filtered, heated air in sufficient quantity to maintain room temperatures compatible for personnel and equipment.
- b. Minimizes the spread of contammation by providing air flow from clean areas to potentially contammatJ areas and to the exhaust.
4 c. Filters exhaust air. The system will also operate to proside frxze protection, as necessary, for liquid systems inside the Auxiliary Building. 7.2.6.1.2 System Description The Auxiliary Building Ventilation System is a forced-flow heating and ventilating system consisting of supply and exhaust subsystems, with exhaust HEPA filter train', which prosides 8 A local differential pressure indicator is installed across each HEPA filter. These indicators are routinely checked on a monthly surveillance when the ventilation system b in senice. During PDMS, the ventilation system may be out of senice for extended periods of time. No checks will be performed on the HEPA filters when the sentilation system is shutdown. The surveillance of the HEPA filters will N: resumed when the system is returned
,q to senice, l I V 7.2-21 UPDATE 2 - AUGUST 1997
i once-through ventilation with no recirculation. The discharge dampers of the s pply and exhaust fans are closed when the ventilation system is not opertting. He PDMS configuration is shown on GPUN Drawing 302-2M2. 7.2.6.1.3 Evaluation During PDMS, Auxiliary Building ventilation and air handling equipment proside a filtered pathway during system operation to meet industrial and radiological requirements. Sources of contammation have been mmmuzed (e g., fuel removed, fuel pool drained, layup of deactivated systems); therefore, spread of potential contammation during PDMS has been greatly reduced. 7.2.6.2 Fuel Handling Buildmg Ventilation System 7.2.6.2.1 PDMS Function , Fuel Handling Building Ventilation System will be maintained in an operational condition to support PDMS activities. When in operation, this system performs the followmg functions:
- a. Provides fresh, fdtered, heated air in sufficient quantity to maintain room temperatures suitable for personnel and equipment.
- b. Minimizes the spread of contanunation by providing air flow from clean areas te potentially contammated areas, and then to the exhaust.
- c. Filters exhaust air.
- d. Maintains the lower elevations (328',301', and 281') of the Fuel Handling Building separate f.om the operating deck, which is maintained at a slightly negative pressure by the Unit I ventilation system.
The system wdi also operate to provide freeze protection, as necessary, for liquid systems inside the Fuel Handimg Bailding. 7.2.6.2.2 System Description ne Fuel Handling Building Ventilation System is a forced flow heating and ventilating system consisting of supply and exhaust subsystems, with exhaust HEPA filter train, which proside once-through ventilation with no recirculation. He operating deck and Fuel Handling Building truck bay are separated from the remainder of the Fuel Handling Building and are ventilated by the Unit I ventilation system. The PDMS configuration is shown on GPUN Drawing 302-2343. 7.2.6.2.3 Evaluation During PDMS, FHB ventilation and air handling equipment provide a filtered pathway during system operation to meet industrial and radiological requirements. Sources of contammation have been minimimi (e.g., fuel removed, fuel pool drained, layup of deactivated syste:ns, covers installed on spent fuel pools); therefore, spread of potential contammation during PDMS has been 7.2-22 UPDATE 2 - AUGUST 1997
l ? C'\ V greatly reduced. 7.2.6.3 Air intake Tunnel Ventilation System he Air Intake Tunnel will be maintained only as a supply pathway for screened air to plant ventilating systems during operation. It consists of a cylindrical intake tower with screens and baf!!cs, a 100,000 gallon sump, and an underground tunnel leading to the plant ventilation systems. He PDMS configuration is shown on GPUN Drawing 302-2219. During PDMS, the Air Intake Tunnel provides a supply pathway for ventilation systems operation to meet industrial and radiological requirements. 7.2.6.4 Compressed Air Supply System 7.2.6.4.1 PDMS Function Portica of the original plant Instrument and Senice A4 Systems will be utilized during PDMS to proside cornpressed air to operational pneumatic devices in the following systems:
- a. Waste Disposal- Liquid
- b. Auxiliary Building Ventilation System
- c. Fuel Handling Building Ventilation System v d. Control Building Ventilation System
- c. Service Building Ventilation S3 stem
- f. RB Purge System
- g. RB Breather System 7.2.6.4.2 System Description The Compressed Air Supply System consists of two air-cooled air compressors, receivers, and the piping and valves required to distribute compressed air to operational pneumatic desices. He major components, piping, and valves of the original plant Instrument /Senice Air Systems have been incorporated as part of the Compressed Air Supply System. Two i i
\d 7.2 23 UPDATE 2 - AUGUST 1997
i air cooled air compressors are used to supply air to the modified system in place of the original wrter-cooled compressors. He primary air compressor, SA P4, is located in the Senice Building 280' elevation with a backup air compressor, SA-P 2, also located in the Senice Building 280' elevation. He Compressed Air System will be operated continuously to support operations. The PDMS configuratica is shown on GPUN Drawing 302-2014, Sht. 3. 7.2.6.4.3 Evaluation he Compressed Air Supply System primarily utilizes the portions of the original plant Instrument /Senice Air System which are required to store and distribute air to pneumatic devices which are operational during PDMS. Since cooling water will not be available during PDMS to cool air compressors, air-cooled air compressors have been used. 7.2.6.5 Building Inleakage Waterproofmg System 7.2.6.5.1 PDMS Function During PDMS, the TMI-2 building waterproofmg systema serve to direct roof rainwate into the site stormwater dramage system and prevent groundwater from entering buildings throt ,b joints, penetrations, and cracks. 7.2.6.5.2 System Description The plant waterproofing systems consist of:
- a. Building roofmg systems
- b. Basement waterproofmg from groundwater.
- c. A cork scam monitoring system (see Section 1.1.2.2.4).
The building roofs, except for the Auxiliary and Reactor Buildings, are a built up system of asphalt, felts, and insulation on both concrete and steel decks. Ibinwater is directed sia roof slope to roof drams which carry the rainwater to the site stormwater dramage piping. All runoffis collected in a retention basin which can be monitored prior to discharge into the Susquehanna River. All basement walls are poured concrete. To prevent groundwater inleakage, the following were performed;
- a. All penetrations through basement walls were sealed.
- b. Expansionjoints between building foundations were sesed with waterstops, cork filler, and epoxy scalant.
- c. Constructionjoints were keyed to deter water seepage through them.
7.2-24 UPDATE 2 - AUGUST 1997
l ) (p) 7.2.6.5.3 Evaluation in preparation for PDMS, various building seams, link seals, and major cracks have been repaired to the extent practical to muumize expected inleakage from storms and high groundwater levels. The inleakage rates and flowpaths experienced to date do not atTect plant equipment required for PDMS. Additionally, the Sump Pump Discharge and WDL system are operational to transfer accumulated water to mmmuze potential spread of contammation due to localized flooding. 7.2.6.6 Sewers 7.2,6.6.1 PDMS Function ne basic function of the sewage collection system is to transport sewage from TMN structures to the Sewage Treatment Plant. The PDMS confimration is shown on GPUN Drawing 302151.
'7.2.6.6.2 System Description Sewage from the temporary personnel access facility (TPAF) in the Turbine Building and several outbuildings is routed to the Sewage Treatment Plant (STP) which serves both TMI l and TMI 2.
He major operational portion of the Sewer System is underground gravity flow piping that prmides for the transport of sewage from the Unit 2 support facilities to the STP. p I f 7.2.6.6.3 Evaluation
\J ne Sewage Treatment Plant will process sewage from the TPAF and several outbuildingt The majority of TMI 2 sewage piping is underground below the frost line. He original plant sanitary waste / sewage system is deactivated. 1 7.2.6.7 Domestic Water System 7.2.6.7.1 PDMS Function During PDMS, portions of the existing domestic water systern will remain operational to pro ide domestic water senices required during PDMS.
7.2.6.7.2 System Description The domestic water system is maintained as a modified operational system. Unit 2 is supplied with domestic water from Unit I which is then distributed to Unit 2 support facilities. Domestic water is provided to the radwaste seal water unit in the Auxiliary Building, to the TPAF in the Turbine Building, and to several outbuildings. He PDMS configuration is shown on GPUN Drawings 302-158 Sht. 4.
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'V 7.2-25 UPDATE 2 - AUGUST 1997
7.2.6.7.3 Evaluation Since personnel access into the plant will be infrequent, only one source of domestic water is required in the Turbine Building. The Auxiliary Building header supplies water to trailers west of the unit one SB0 Diesel Generator and domestic water to the seal water unit. Unit I and Unit 2 support facilities will remain operational, therefore, domestic water will ccatinue to be supplied to them. 7.2.6,8 Control Room Ventilation Systra 7.2.6.8.1 PDMS Function The Control Room Ventilation System will be maintained in an operational condition to support PDMS activities. This system provides fresh, filtered, heated or cooled air in sufficient quantity to support personnel occupancy and equipment protection. 7.2.6.8.2 System Description ne Control Room Ventilation System consists of one supply fan (AH-C-16B) runnmg in a forced ventilation mode during normal year round conditions. The supply fan will pnmarily recirculate the control room air as it is heate& cooled. A small amount of fresh air (outside air) will be force supplied by bypass booster fan (AH-C-l(X). Exhaust fan (AH-E-35) will return control room air to the suction of supply fan (AH-C-168). A small amount of the control room air will be l " exhausted" out of this recirc mode, primarily by exfiltration dampers in the control room and sia l the kitchen and toilet fans. His provides for a small amount of air change per day. 1 Heating is controlled by a thermistor located in the control room exhaust duct. It prosides signals to a programmable contro!!ct in the control room chiller at either of the setpoints that activates two steps ofinstalled duct heaters on demand. As conditions dictate, this same thermistor prosides signals at either of the programmable controller setpoints to give cooling sia the 10 ton control room chiller. Cooling can also be called for by a humidistat located in the supply air duct work near the supply fan (AH-C-16B). If the conditions in the control room or the nukeup air cause the humidity to go above either of the two setpoints, the corresponding cooling circuit in the 10 ton chiller w411 operate to condense out tLis moisture. Neither cooling or heating functions will operate unless supply fan (AH C-16B) is runnmg and satisfying a flow switch in the supply air duct. Additional outside air can be prosided by performing special operations when economizer mode is desired or the chiller malfunctions and additional cool outside air is desired. 7.2.6.8.3 Evaluation During PDMS, Control Room ventilation and air handling equipment prosides a filtered pathway for active operation to meet industrial and radiological requirements. The Control Room Ventilation System is maintained operational for the maintenance and surveillance entries into the TMI-2 Control Room and in response to off-normal conditions. 7.2-26 UPDATE 2 - AUGUST 1997 O
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. (v ) 7.2.6.9 Cable Room Ventilation 3ptem 7.2.6.9.1 PDh1S Function The Cable Room Ventilation System will be maintained in an operational condition to support PDhiS activities. When in operation, this system prosides fresh, filtered, heated air in sufficient quantity to maintain room temperatures suitable for personnel and equipment. 7.2,6.9.2 System Description The Cable Room Ventilation System is a forced flow heating and ventilation system consisting of a supply and exhaust-retum subsystem which provides ventilation with partial recirculation. When the ventilation system is not operating, a damper in the bypass duct will open, allowing free passage of air in the exhaust-return duct system. 7.2.6.9.3 Evaluation During PDhtS, Cable Room ventilation and air handling equipment provide a filtered pathway during system opention to meet mdustrial requirements and provide the appropriate emironment s for instrumentation and annunciator equipment. 7.2.6.10 Senice Building Ventilation System O (d 7.2.6.10.1 PDhtS Function The Senice Building Ventilation System will be maintained in an operational condition to support PDhis activities. When in operation, this system performs the following functions:
- a. Provides fresh, filtered, heated air in sufficient quantity to maintain room temperatures suitable for personnel and equipment.
- b. hinunuzes the spread of contanunation by providing air flow from clean areas to potentially contaminated areas, and then to the exhaust.
- c. Filters exhaust air.
7.2.6.10.2 System Description He Senice Building Ventilation System is a forced flow heating and ventilation system consisting of supply and exhaust subsystems. Exhaust HEPA filter trains, which proside once-through ventilatio::. with partial recirculation of clean areas.
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7.2-27 UPDATE 2 - AUGUST 1997 l
7.2.6.10.3 Evaluation During PDMS, Senice Building ventilation and air handling equipment proside a filtered pathway during system operation to meet industrial and radiological requirements. This system is maintained operational for paonnel ingress and egress to the Reactor Building, Auxiliary Buildmg, and Unit 2 Control Room, for maintenance and sun eillance entries into :he Senice Building, and prosides ventilation for the CAS backup compressor. 7.2.6 I1 PDMS Alarm Monitonng System 7.2.6.11.1 PDMS Function ne function of the plant computer alarm system is to notify plant operations personnel of an abnormal plant conation which requires op rator action to correct or which represents a threat to plant, personnel or equipment safety. The PDMS Alarm Monitoring System provides the means to remotely monitor select TMI 2 alarms and TMI-2 station vent monitor signals in the TMI-l Control Room via the TMI l plant computer. As required by the GPUN Emergency Plan, the PDMS Alarm Monitoring System is designed such that if the remote monitonng of the alarms in Unit I becomes inoperable, the TMI 2 Control Room alarms and station vent monitor signals can be monitored from the annunciators and other recorders / equipment in the TMl 2 Control Room. The alarms and functions to be monitored are listed in Operating Procedure 1105-22, Response To PDMS alarms. (Ref. 7.3-13). 7.2.6.11.2 System Desenption ne plant computer uses four types of alarm information display systems - alarm CRTs, alarm displays on a Utihty CRT, alarm summanes on a Utility CRT and an alarm pnnter. The modifications that were necessary to facilitate installation / operation of the PDMS Alarm Monitonng System were as follows:
- 1. A fiber optics cable link was installed between the TMI-l computer system in the OSF Buildi:.g and the TMI 2 multiplexer unit located in the TMI-2 Control Room.
- 2. A multiplexer unit was installed in the Unit 2 Control Room to interface with all required signals from the field (i.e., sensors or annunciators) or the Unit 2 Control Room annunciators. He multiplexer performs the necessary signal processing to comert the 6gital and analog signals to a light signal which is transmitted back to the TMI l computer via the fiber optics cable link.
- 3. He required digital alann inputs and analog signals were interconnected to the multiplexer unit.
- 4. The multiplexer receives 120VAC power from a 480/120VAC regulated transformer.
His transformer receives 480VAC power from one of two sources. Normally it will be fed from the TM1-2 480VAC system or, as a backup, it can Le fed from one of TM1-l's 480VAC B.O.P. power systems.
- 5. A Mini Uninterruptible Power Supply (UPS) provides backup power to the multiplexer in 7.2-28 UPDATE 2 - AUGUST 1997
_ . _ _ _ . _ ~ . _ _ ~ _. .._._ . _ ___._
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case of failure of the normal 120 VAC power source. The UPS will provide this backup i-Q power for a nummum of one hour. _6. Computer software was developed and tested to generate the required audible annunciator signals and the displays on the existing CRTs in the TMI l Control Room. 7.2.6. I 1.3 Evaluation The PDMS Alarm Monitoring System does not meet NFPA requirements with respect to supervised circuits. However, the intent of the NFPA standards is met by the use of a multiplexer / communication link trouble alarm, which notifies the Unit I computer and operators of a failure in the multiplexer or fiber optics cable. In addition, in the event of a communications problem, the system software will suppress all other TMI 2 alarm conditions to avoid overloading the Unit 1 operators with spurious alarms that normally would result &om such a failure. 7.2.6.12 Control Building-Mechanical Equipment Room Ventilation System 7.2.6.12.1 PDMS Function Part of the Control Building-Mechanical Equipment Room Ventilation System will be maintained in an operational condition to support PDMS. The modified synem provides fresh, filtered, heated air in sufficient quantity to maintain room temperatures suitable for personnel and equipment. 7.2.6.12.2 System Description ne Control Building-Mechanical Equipment Room Ventilation System is a forced flow heating ventilation system consisting of a supply and exhaust subsystem which prosides once-through l ventilation with no recirculation. 7.2.6.12 3 Evaluation During PDMS, the Control Building-Mechanical Equipment Room Ventilation and air handling equipment prosides a filtered pathway during system operation to meet industrial and radiological requirements. He Control Building Mechanical Equipment Room Ventilation Syctem is maintamed operational for maintenance and surveillance entries into the mechanical equipment room and in response to off-normal conditions. During the winter it prosides pipe freeze protection.. .' M 7.2-29 UPDATE 2 - AUGUST 1997 4
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7.2.6.13 Control Building Area Ventilation System 7.2.6.13.1 PDMS Function The Control Building Area Ventilation System will be maintained in an operational condition to support PDMS. When in operation, the system prosides fresh, filtered, heated air in sufEcient quantity to maintain room temperatures suitable for personnel and equipment. 7.2.6.13.2 System Description The Control Building Area Ventilation System is a forced flow heating and ventilation system consisting of a supply fan, air filter, and duct heater umt, which provides heated (when required) fresh air into the Control Building Area and Tendon Access Gallery, 7.2.6.13.3 Evaluation During PDMS, the Control Building Area Ventilation and air handling equipment prosides an active filtered pathway that meets industrial requirements. The Control Building Area Ventilation System is maintained operational for maintenance and surveillance entries into the Control Building Areafrendon Access Gallery and to provide pipe freeze protection. O
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L) TABLE 7.2-1 b--, OPERATIONAL SYSTEMS t ~ SYSTEM CONTAINMENT INTERNAL F ESCRIPTION SVS. CODE ' ISOLATION CONTAMINATION REMARKS
' CONTAINMENT All YES YES PASSIVE SYSTEM, PERIODIC INSPECTION OF IIEPA FILTERS ,
ATMOSPHERIC & ASSAYOFSAMPLEFILTERPAPERS BREATHER
' CONTAINMENT AlI .YES YES OPERATED IN PURGE MODE TO SUPPORT CONTAINMENT
- VENTILATION & ENTRIES PURGE CONTAINMENT AIRLOCKS RBA YES N/A AIRLOCKS FOR PERSONNT1/ EQUIPMENT ACCESS;
& EQUIPMENT EQUIPMENT HATCil WILL REMAIN IN PLACE liATCH FIRE PROTECTION FP N/A N/A ZONE FIRE DETECTION SHALL BE OPERATIONAL +
THROUGHOtJT OPERATIONAL PLANT AREAS t FIRE SERVICE FS YES N/A PROVISIONS II AVE BEEN M ADE AND EQUIPMENT liAS BEEN STAGED TO IJTILIZE LOOP IIYDRANTS ' FIRE SUPPRESSION FS N/A N/A PORTABLE FIRE SUPPRESSION EQUIPMENT ARE STAGED EQUIPMENT WITH EMERGENCY RESPONSE CREW EQUIPMENT j WASTE DISPOSAL WDL YES YES NECESSARY EQUIPMENT / TANKS TO PROCESS WATER WILL BE MAIN- , - LIQUID (MISC) TAINED OPERATIONAL ONLY TIIE BUILDING SUMP PUMPS TIIE MISC. - WASTEIlOLDUPTANK(WDL-T-2),111E ABST(WDL-T-5) AND INTERCONNECTING PIPE SilALL REMAIN OPERATION AL FOR WATER REMOVAL FUNCTIONS. SUMPFUMP SD N/A YES FACILITIES ARE SEALED TO LIMIT EXTERIOR WATER INGRESS. ' ! DISCIIARGE & PERIODIC SUMP PUMP OPERATIONS WILL PREVENT SUMP ACCUMU-MISCELLANEOUS LATION OF DRAINAGE AND INADVERTENT INLEAKAGE. 7.2-31 UPDATE 2 - AUGUST 1997
TABLE 7.2-1 (Cont'd) OPERATIONAL SYSTEMS YSTEM CONTAINMENT INTERNAL DESCRIPTION SYS. CODE ISOLATION CONTAMINATION REMARKS RADIATION llP N/A YES RADIATION MONITORS AND ALARMS REMAIN IN OPERATION MONITORING AS DEEMED NECESSARY BY RADCON. SELECTED llP MnNITORING AND SURVEY PROGRAMS ARE ALSO CONTINUED ELECTRIC EE N/A N/A ELECTRICAL EQUIPMENT W111CII SUPPORTS OPERABLE SYSTEMS DISTRIBITilON AND FACILITIES SilALL REMAIN OPERATIONAL. AREA LIGIITING SIIALL BE AVAILABLE T11ROUGIIOLTT THE PLANT LIGIITING & EL N/A N/A EMERGENCY LIGilTING COMMUNICATIONS COM N/A N/A COMMUNICATIONS WILL BE OPERATIONAL TO TIIE EXTENT NECESSARY FOR PDMS ACTIVITIES AUXILIARY BUILDING All N/A YES VENTILATION WILL BE OPERATIONAL TO TIIE EXTENT VENTILATION NECESSARY FOR PDMS ACTIVITIES FUEL HANDIMG All N/A YES VENTILATION WILL BE OPERATIONAL TO THE EXTENT BUILDING VENTILATION NECESSARY FOR PDMS ACTIVITIES AIR INTAKE TUNNEL All NO NO MAINTAINED ONLY AS A SUPPLY PATilWAY VENTILATION COMPRESSED AIR IA/SA NO NO AIR-COOLED AIR COMPRESSORS USE PORTIONS OF SUPPLY INSTRUMENT AND SERVI, 'i lR SYSTEMS. BUILDING INLEAKAGE CS NO YFS WATERPROOFING WILL MINIMI7,E IN-LEAKAGE TO Tile EXTENT TIIAT ANY INLEAKAGE CAN DE ADEQUATELY llANDLED BY PERIODIC TRANSFER TO BUILDING SUMPS 7.2-32 UPDATE 2 - ST 1997
. __ . - .. ~ . . _ _ _ _ _ _
O O 3 TABLE 7,2-1 (Cent *d) OPERATIONAL SYSTEMS SYSTEM CONTAINMENT INTERNAL DESCRIPTION SYS. CODE . ISOLATION CONTAMINATION REMARKS SEWERS SE N/A NO OPERATIONAL FOR OLTTBUILDINGS AND TEMPORARY PERSONNEL ACCESS FACILITY ONLY. DEACTIVATED FOR IN-PLANT PORTIONS : DOMESTIC WATER 'DO N/A NO OPERATIONAL ONLY TO THE EXTENT REQUIRED TO SUPPORT PDMS ACTIVITIFS CONTROL BUILDING ' All N/A NO VENTILK: JON WILL BE OPERKDONAL TO THE EXTENT VENTILATION NE'TSSARY FOR PDMS ACTIVITIES. CONSISTS OF THREE OPERATIONAL SUBSYSTEMS; CONTROL ROOM, CABLE ROOM AND HEC 1IANI".AL EQUIPMENT ROOM VENTILAT10N SYSTEMS SERVICE BUILDING AH N/A YES VENTILATION WILL BE OPERATIONAL TO Tile EXTENT i VENTILATION NECESSARY FOR PDMS ACTIVITIES; INCLUPES, CONTROL BUILDING AREA VENT PDMS ALARM N/A N/A MONITORING SYSTEM , i s b 7.2-33 UPDATE 2 - AUGUST 1997 l
TABLE 7.2 2 CONTAINMENT ISOLATION TABLE l Line Operational Size Isolation Penetration Service System System (inches) Valves Status R-524 Fuel Transfer Canal SF NO 10 SF-VIOS Manual-Locked Fill Line Closed (LC.) R-525 Decay IIcat Coolant Dil NO 12 Dil-V3 De-energized Letdown 1/2 Dil-V225 Manual-LC. R-526 Steam Generator "A" CA NO 1/2 CA-V8 De-energized-L.C. Sample Line R-527 Core Flood Tank CF NO I CF-Vl44 De-energized-LC. Bleed & Sample R-528 Steam Generator "B" CA NO 1/2 CA-V9 De-energized-L.C. Sample Line R-529 Reactor Coolant WDL NO 4 W DL-V125 De-energized-LC. Drain Pump Discharge R-530 Steam Generator SV NO 2 SV-V55 De-energized-L.C. Side Vent & Drain R-531 Decay lleat CCW for DC NO 8 DC-VI15 De-energized-LC. RC Leak Recovery System R-532 FuelTransfer Tube Fil NO I Fil-VID Manual-LC. 39 Blind Flange Installed G 7.2-34 UPDATSAUGUST 1997 G ,
.g .. .
73- } L . _- / }
-q TABLE 7.2-2 (Cont'd)
CONTAINMENT ISOLATION TABLE Line Operational Sire Isolation , Penetration ' Service Svsten Systene (inches) Valves Ststas R-533 FuelTransferTube Fil NO 1 FH-VIC Manual-LC. 39 Blind Flange Installed R-535 ' Demineralized Water DW NO 3 DW-V28 Manual-LC. R-536 Plasma Are Nitrogen PAN NO 2 PAN-V5 Manual-LC. 3/4 PAN-Vl7 Manual-LC. 1 PAN-Vl9 Manual-LC. R-537 Nitrogen & Fill CF NO I CF-V145 Manual-LC. to Core Flood Tank 1 CF-VI14B Manuai-L.C.- 1/2 CF-V129B Manual-LC. l R-538 Pressurizer Steam SN- NO 1/2 CA-V10 De-energized-LC.
& Water Space Sample Line l
R-539 DWCS DC NO 8 DC-V103 De-energized-LC. Isolation 1/2 DC-V137 Manual-L.C. ! R-54i Letdown Line to MU NO 2 I/2 MU-V376 De-energized-LC. Purification Demin. R-542 ' DWCS Borated Dil NO 3 DII-V187 Manual-LC. Water Flush i Dil-V205 Manual-LC. -[ R-543 Reactor Building NM NO 1 NM-V52 Air Disabled-LC. I Nitrogen lleader i 7.2-35 UPDATE 2 - AUGUST 1997
TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operational Size Isolation Penetration Service System System Onches) Valves Status R-544 Nitrogen & Fill to CF NO I CF-V146 Manual-L.C. Core Flood Tank I/2 CF-V129A Manual-L.C. I CF-Vi14A Manual-L.C. R-545-A Building Spray BS NO 1 BS-V146 Operational - open for RB pressure indication Pressure Sensing R-545-B DWCS Samp!c DWC NO 3/4 DWC-V038 Manual-L.C. Isolation 3/4 DWC-V040B Manual-LC. R-545-C DWCS Sample GWC NO 3/4 DWC-V037 Manual-L.C. Isolation 3/4 DWC-VD40A Manual-LC. R-545-D Reactor MU NO 2 MU-V377 Desurgized-LC. Coolant Pump Seal Water Return R-546 Pressurizer, RC, WDG NO 4 WDG-V199 De-energized-LC. OTSG & Core Flood Tank Vents R-547 Reacto WDL NO 4 WDL-Vil26 De-energized-L.C. Building Sump Pump (DEACTIVATED Discharge PORTION) R-548 Fire Protection FS YES 4 FS-V639 Manus LC. R-549 Reactor Building All YES 16 All-VIB Operational Inlet Purge 4 All-V908 Manual-LC. Line 1/2 All-V149 Manual-LC. 7.2-36 UPDAT 2 AUGUST 1997
(J v J TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line - Operational Size Isoistion Penetratica ' Service System System (inches) Valves States R-550 ' ' Reactor Building All YM 36 All-VIA Operational ~ All-V90A
' Inlet Purge Line 4 Manual-L.C.
1/2 All-V151 Manual-LC.
. R-551 Reactor Building All YES 36 ' AII-V4A Operational
- Outlet Purge Line 10 All-V52 Operational 4 All-V120A Manual-LC.
1/2 All-V153 Operational ' R-552 Reactor Building All YES 36 All-V4B Operational
' Outlet Purge Line 10 All-V7 Air Disabled-LC.
10 All-V81 Air Disabled-L.C. 4 All-VI20B Manual-1..C. R-553 DWCS to RB DWC NO 2 WDL-V1092 De-energized-LC. Isolation R-554-A Instrument Air, Purge IA NO 1/2 All-V213 Manual-LC. < 1/2 All-V214 Manual-L.C. 1/2 All-V221 Manual-L.C. R-554-B Air Sample Supply All NO 1/2 AlI-V169 Manual-L.C. ,
- (Radiation Detection) 1/2 AII-V230 Manual-L.C.
R-554-C Building Spray BS NO I BS-V147 Operational- open for RB pressure indication Pressure Sensing , R-554-D Instrument Air DWC NO 3/4 DWC-V316 Manual-L.C. to DWCS - 3/4 DWC-V318 Manual-LC. 1/2 Pipe Cap Installed 7.2-37 UPDATE 2 - AUGUST 1997
TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operational Size Isolation Penetration Service System System (inches) Valves Status R-555-A Air Sample Supply AII NO I All-V143 Manual-LC. (Radiation Detection) I/2 All-V168 Manual-LC. R-555-B Air Sampic Return AH NO I Al1-V145 Manual-L.C. (Radiation Detection) 1/2 AII-V171 Manual-LC. R-555-C PCI isolated Ground PCI NO I Blind Flange Installed Cable R-555-D Air Sample Retum All NO 1/2 All-V170 Manual-LC. l (Radiation Detection) 1/2 AH-V227 Manual-LC, R-557 To RC Pump Oil NS NO 8 NS-V72 Air Disabled-L.C.
& Motor Coolers 1/2 NS-V210 Manual-L.C.
j R-558 From RC Pump NS NO 8 NS-VH1 Air Disabled-LC. Oil & Motor Coolers R-559 Intennediate IC NO 3 IC-V5 Air Disabled-LC. l Cooling to Control i Rod Drive Mechanisms R-561 Ifigh Pressure TDW NO I TDW-V001 Manual-LC. Water i TDW-V003 Manual-L.C. 10 Blind Flange Installed i R-561 Decon Service DSA NO I DSA-V004 Manual-LC. Air 3/4 DSA-V006 Manual-LC. 9 7 2-38 UPDAl AUGUST 1997
.n.
~\
Q,I ] G' - l TABLE 7.2-2 (Cont'd)
. CONTAINMENT ISOLATION TABLE Line .
Operational Sire Isolation
, Penetration Service System System (inches) Valves States R-562-A Instrument Air IA' NO 1/2 All-V215 Manual-L.C.
I . Supply 1/2 All-V216 . Manual-L.C. 1/2 All-V223 Manual-LC. R-562-B Pressure AII NO 1/2 All-V147 Manual-LC. Transfer Fans 1 Pipe Cap Installed i R-562-C ' . Building Spray BS NO 1 BS-V148 Manual-LC Pressure Sensing R-562-D ' RB Sludge WDS NO I WDS-FV612 Manual-LC. Transfer NO 1 WDS-FV614 Manual-LC. I R-563 Intermediate IC NO 6 IC-V4 Air Disabled-LC. Cooling System 1/2 IC-V207 Manual-LC. R-565 Processed Water PW NO 3 PW-V69 Manual-LC. Supply PW to Reactor i PW-V99 Manual-L.C. Building 4 Flange Installed i R-566 Service Air SA NO 2 1/2 SA-V20 Air Disabled-LC. R-567 Intermediate IC NO 6 IC-V3 Air Disabled-L.C. Cooling System R-569 Sec. System SV NO 3 SV-V18 L.C. Flush & Drain R-570 Ifigh Pressure MU NO 2 1/2 MU-V16A De-energized-LC. Injection Line 1/2 MU-V315 Manual-LC. 7.2-39 UPDATE 2- AUGUST 1997
~ -
TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operational Size Isolation Penetration Service System System (inches) Valves Status R-571-A Integrated Leak LR NO ! Blind-Flange Installed Rate Test R-571-B Integrated Leak LR NO 1 Welded Cap installed Rate Test R-571-C Building Spray BS NO I BS-V149 Operational - op= for RB high pressure alarm signal Pressure Sensing R-571-D - Integrated Leak LR NO 1 Blind-Flange Installed Rate Test R-572 liigh Pressure MU NO 1/2 MU-V316 Manual-LC. Injection Line 2 1/2 M U-V16B De-energized-LC.
& Makeup 2 1/2 MUU-V18 Air Disabled-L.C.
R-573 Reactor Coolant MU NO 3/4 MU-V379 Manual-L.C. Pump Seal Water Supply 3/4 MU-V383A Manual-LC. 3/4 MU-V384A Manual-LC. R-574 Reactor Coolant MU NO 3/4 MU-380 Manual-L.C. Pump Seal Water 3/4 MU-V383B Manual-LC. Supply 3/4 MU-V384B Manual-L.C. R-575 Reactor Coolant MU NO 3/4 MU-V381 Manual-LC. Pump Seal Water 3/4 MU-V383C Manual-L.C. Supply 3/4 MU-V384C Manual-LC. R-576 Reactor Coolant MU NO 3/4 MU-V382 Manual-LC. Pump Seal Water 3/4 MU-V383D Manual-LC. Supply 3/4 MU-V384D Manual-LC. 7.2-40 UPDAT AUGUST 1997
Q
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TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operational Sire Isolation , Penetration Service Systein System (inehes) Valves Status i R-577 Reactor Building Air RR NO 8 P.R-VSA De,..w. pad-LC. . Coolers 1 RF-V28A Manual-L.C. 1/2 R4-V86 Manual-LC. R 578 Reactor Building Ali RR NO 6 RR-VI1 A Air Disabled-LC. Coolers 1/2 RR-V75A Manual-L.C. 6 RR-V25A Air Disabled-LC. R-579 Reactor Building Air RR NO 8 RR-V5B De-energized-LC. Coo!:rs 1 RR-V28B Manual-LC. 1/2 RR-V83 TAanual-LC. R-580 Reactor Building Air RR NO 8 RR-V5C De-energized-LC. Coolers 8 RR-V6C De-energized-LC. I RR-V28C Manual-L C. 1/2 RR-V90 Manual-LC. 1/2 RR-V92 Manual-L.C. R-581 Reactor Building Air RR NO 6 RR-VIIC Air Disabled-L.C. Coolers 1/2 RR-V75C Manual-LC. 6 RR-V25C Air Disabled-L.C. R-582 Reactor Building RR NO 6 RR-VIIB Air Disabled-L.C. Air Coolers 1/2 RR-V75B Manual-LC. 6 RR-V25B Air Disabled-LC. R-583 Reactor Building BS NO 8 BS-VIB De-energized-L.C. Spray !nlet Line 3 BS-Vl30B Manual-LC. 7.2-41 UPDATE 2 - AUGUST 1997
TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operationsi Size Isolation , Penetration Service System System (inches) Valves Status R-584 Reactor Building RR NO 8 RR-V6D De-energized-L.C. Air Coolers i RR-V28D Manual-LC. 1/2 RR-V95 Manual-L.C. I RR-V98 Manual-L.C. R-585 Reactor Building RR NO 6 RR-VIID Air Disabled-L.C. Air Coolers 1/2 RR-V75D Muual-LC. 6 RR-V25D Air Disabled-LC. R-586 Reactor Building BS NO 8 BS-VIA De-energized-L.C. Spray inlet Line 3 BS-V130A Manual-LC. R-587 Reactor Building RR NO 8 RR-V6E De-energized-LC. Air Coolers 1 RR-V28E Manual-LC. 1/2 RR-V96 Manual-LC. R-588 Reactor Building RR NO 6 RR-VIIE Air Disabled-LC.
% Coolers 1/2 RR-V75E Manual-L_C.
6 RR-V25E Air Disabled-LC. R-589 Decay lleat Dil NO 10 DII-V4 A De-energized-L.C. Coolant Supply R-590 Decay IIcat DI{ NO 10 Dil-V4B Deenergized-LC. Coolant Supply R-591 liigh Pressure MU NO 2 1/2 MU-Vl6C De-energized-LC. Injecti m Line NO 1/2 MU-V366 Manual-LC. R-592 Ifigh Pressure MU NO 2 1/2 MU-V16D De-energized-LC. Injection Line 1/2 MU-V368 Manual-LC. O 7.2-42 UPDAT AUGUST 1997
t TABLE 7.2-2 (Cont'd) [' CONTAINMENT ISOLATION TABLE Line : Operational Size isolation .
- Penetration Service System System (inches) . Valves Status t
t R-593 Sump Pene- Dli YES 18 DH-V6A - Operational tration Sleeve
& Drain Line R-594 Sump Pene- Dil YES 18 DH-V6B Operational tration Sleeve
'& Drain Line l t
R-616 .- Emergency EF NO 6 - EF-V12B De-energized-LC.. i Feedwater to 4 EF-V33B De-energized-LC. OTSG "B" 3/4 EF-V36 Manual-L.C. 6 Blind Flange Installed
- 6 Blind Flange Installed ,
- R-617 Feedwater - FW NO 20 FW-Vl7B De{.K,pzxd-L.C.
to OTSG "B" ' FW-Vl9B De-energized-LC. , 3,4 FW-V68B Manual-L.C. 3/4 FW-V35B Manual-L.C. t 2 GR-V7B Manual-L.C. i 10 Blind Flange Installed 1 t I t I 7.2-43 UPDATE 2- AUGUST 1997 *
-r
TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE Line Operational Size Isolation Penetration Service System System (inches) Valves Status R-618 Feedwater to RV NO 20 RV-Vl7A De-energized-L.C. OTSG "A" 6 FiV-Vl9A De-energized-L.C. 3/4 RV-V68A Manual-LC. 3/4 RV-V35A. Manual-LC. ' 2 GR-V7A Manual-L.C. 10 Blind Flange Installed R-619 Main Steam MS NO 24 MS-V7B De-energized-LC. to Turbine 10 MS-VISB De-energized-LC. 3/4 MS-V224 Manual-LC. I MS-VSIB Mcnual-LC. R-620 Main Steam MS NO 24 MS-V4B De-energized-LC. to Turbine 6 MS-VID Manual-L C. 3/4 MS-V2B Manual-LC. 4 MS-Vi iB De-energized-LC. I MS-V50B Manual-LC. 3/4 MS-V225 Manual-LC. 2 GR-VIB Manual-LC. R-621 Main Steam MS NO 24 MS-V7A De-energized-LC. to Turbine 10 MS-VISA De-enurgized-LC. 1 MS-V51 A Manual-LC. 4 MS-VII A De-energized-LC. 3/4 MS-V227 Manual-LC. G 7.2-44 UPDAT AUGUST 1997
r' TABLE 7.2-2 (Cent'd) - . CONTAINMENT ISOLATION TABLE Line Operational Size Isolation 1 Penetration Service Systems Systens (inciees) Valves Status R-622 Main Steam to MS NO 24 MS-V4A . Dew ~.simi-LC. Turbine 6 MS-VIA Manual-LC. * , 3/4 MS-V2A - Manual-L,C. ; I MS-V50A Manual-L.C. 3/4 MS-V226 Manual-LC. , 2 GR-VIA Manual-LC. R-623- Ewmiss.~y Feedwater EF NO 6 EF-V12A De-energized-LC. to OTSG "A" 4 EF-V33A De-eneq!itzed-LC. 6 Blind Flange 6 Blind Flange i R-626 Spare N/A NO 2 SWS-FV-1 Manual-L.C. ; 12 Flange Installed R-401 . RB Basement RBS 'NO 1/2 RBS-IV-1009. Manual-LC. j Ixvel Indication 1/2 RBS-IV-1011 Manual-L,C. 1/2 RBS-IV-1013 Manual-LC. !- 1/2 RBS-IV-1014 Manual-L.C. > 12 Flange Installed [ i j R-508 Electrical Pene- N/A NO II Flange Installed tration ROSA /CCTV Coax Cables i l i 1 7.2-45 UPDATE 2 - AUGUST 1497
& " w L_ -
A -__ -_ _. _ __.____-L---..._._. .m_;:.._Ami-.xhw- _a_..__
r l TABLE 7.2-2 (Cont'd) CONTAINMENT ISOLATION TABLE l Line Operational Size Isolation - Penetration Service Svstem System finches) Va.ves Status N/A RB Personnel RDA YES 1/2 RBA-V-2 Opcrational (Air Sampling) 1/2 RBA-V-3 Operational l N/A RB PersonnelIlatch N/A YES N/A Personnel Operational l (Airlock #2) Door N/A RB Equipment Ilatch N/A NO 31'6" N/A Bolted Flange N/A RB Equipment flatch N/A YES N/A Personnel Operational (Airlock #1) Door l 1 9 7.2-46 UPDAT AUGUST 1997
- - - - .. .. -_ - . _ ~ . _ . . - - .. ..- ~ . . -. .. -- - . . - . - -
W g 1 TABLE 7.2 3 OPERATIONAL SUMP SYSTEMS FOR PDMS Sumos Associated With SD System SWDD location Turoine Building Sump Turbine Building Control Building Area Sump M-20 Area West ~ Control & Senice Building Smice Building Sump Tendon Access Gallery Tendon Access Gallery Sump Air intake Tunnel Normal Air in .c Tunnel , Sump Sumos Associated With WDL System Smag Location (O j Containment Basement Sump Containment Building i Auxiliary Buildmg Sump Auxiliary Building Auxiliary Building Decay Heat Removal Pump Room Sumps (2) Reactor Building Spray Auxiliary Building Pump Room Sumps (2) Contammated Drain Senice Building Tank Room Sump i I O 7.2-47 UPDATE 2 - AUGUST 1997 l l
.ueJ 4a..g_- *,- s S 4..M---a'-Ms. .Mwa.4.. -- see.,--o-mW us,,,m-m eme.,_% _ ue 4.4L--, 4 .ha.. ms-,ees.-.... m4-e,,+,a am. es m a e m,4 r,,a.wa..4.ee-4.,45&*4-mm4ew.*s--Wh-d* =pe8.-*8*a -
P i Table 7.2 4 DELETED i i t h l l l 4 9 I l l I i i 4 s 4 4 7.2 48 WDATE 2 - AUGUST 1997 t i 1
. - ~ . .
J.'. - . . - b
/ l[/ HEPA FILTER d u
>2 EOUIV ALENT g -
LEAK _ _ l /
/
[4 30 FT
/ <
o V l 1 l
~
ASSUMPTIONS:
- 1. LAMINAR FLOW IN VENT AND LEAK.
- 2. NEGLIGIBLE TURNING, ENTRANCE LOSSES.
- 3. NO LOSSES IN V At VEG.
UPDATE 2 AUGUST 1997 f CONTAINMENT ATMOSPHERIC
\ .
BREATHER *'MOST PROBABLE" PATHWAY MODEL
' FIGURE 7.2-1 PAGE 7.2 49
f l . O i I I i 9 . ... -
- a -
;;g/
b
~ay
/c\
t I UPDATE 2 - AUGUST I
, RADIATION SURVEY LOCATION - 347' ELEV.
FIGURE 7.211 PAGE 7.2.50 1 e
l 4 O IWIDE AIg g
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l l I b-INSIDEAIgg t O '
, UPDATE 2- AUGUST 997 LOCATION - 305' ELEV.
FIGURE 7.212 l PAGE 7.2.51
73 REFERENCES Provided below is a list of reference documents that provide further information. Relevant additional information can be found in these documents (e.g.; drawings numbers, procedure numbers, etc). REF DOCUMENT # TITLE 73-1 OPM Section R-4 Unit-2 PDMS Ventilation System 73-2 OPM Section R-3 Unit-2 Sump Pump and Discharge System 73-3 OPM Section R-9 PDMS Cunpressed Air Supply System 73-4 GPM Sectio.3 R-6 PDMS Electrical 73-5 OPM Section M-6 Flood Protection 73-6 OPM Section M-8 Plant Communications System Systems 73-7 OPM Section R-1 PDMS Alarm Monitoring 73-8 OPM Section R-8 Radiation Monitoring in PDMS 73-9 OPM Section R-5 Reactor Building Ventilation / Breather 73-10 OPM Section R-2 Unit-2 Liquid Radwaste Disposal in PDMS 73-11 OPM Section R-10 PDMS Miscellancous 73-12 GPU NucIcar Letter, LL2-81-0191 " Design Pressure for Containment and Future Mechanical And Electrical Penetration Modifications," dated December 04,1981. 73-13 TMI Operating Procedure Number 110S-22, " Response to PDMS Alarms (SAR 7.2.6.11)." 7.3-14 GPU Nuclear SDD T2-680A, "TMI-2 IIcat Sensitive Wire Fire Detection System. i l 7.2 - 52 UPDATE 2- AUGUST 1997 1 1 O O O
i l i J 4 CIIAPTER 8
- ROUTINE AND UNANTICIPATED RELEASES l
l l I E i 4 0
CHAPTER 8 ROUTINE AND UNANTICIPATED RELEASES - h/ TABLE OF CONTENTS SECTION - TITLE PAGE 8.1 GENERAL 8.11 8.1.1 ROUTINE RELEASES 8.1 1 8.1.2 SOURCE TERMS 8.12 8.1.2.1 Airborne Releases 8.1 8.I'2.2
. Liquid Releases 8.1 3 8.1.2.3 Airborne Source Terms During PDMS for 8.1-3 Routine Releases ,
8.1.3 OFF-SITE DOSE ESTIMATES 8.1-5 ! 8.1.3.1 --Routine Releases 8.1-6 8.1.3.2 Unanticipated Releases 8.1-6 8.2 UNANTIClPATED EVENTS ANALYSIS 8.2-1 8.2.1 VACUUM CANISTER FAILURE 8.2-1 8.2.2 ACCIDENTAL SPRAYING OF CONCENTRATED 8.2-3 CONTAMINATION WITH HIGH PRESSURE SPRAY 8.2.3 ACCIDENTAL CUTTING OF CONTAMINATED PIPE 8.2-6 4 8.2,4 ACCIDENTAL BREAK OF CGNTAMINATED PIPING 8.2-8 8.2.5 FIRE INSIDE THE CONTAINMEhT 8.2-8 8.2.6 OPEN PENETRATION - 8.2.10 4
-. i UPDATE 1 -JUNE 1995
l l CHAPTER 8 I TABLE OF CONTENTS (Cont'd) LIST OF TABLES TABLE TITLE ' PAGE 8.1-1
SUMMARY
OF SOURCE TERMS 8.1-9 8.1 2 ESTIMATES OF Cs AND Sr INVENTORY OF 8.1 10 SELECTED REACTOR BUILDING LOCATIONS 8.1 3 Ci FRACTIONS IN RESIDUAL FUEL 8.1 11 8.1 4 MAXIMUM POTENTIAL DOSES ESTIMATED 8.1 12 FROM ROUITNE EFFLUENT RELEASES 8.15 ESTIMATED DOSE COMMITMENTS RESULTING 8.113 l FROM ROUFINE AND UNANTICIPATED EVENTS 4 DURING PDMS
- O a
1 W- { h O e --
1 1 l 1 1
)
f3 d CriAPTER 8 ROlTTINE AND UNANTICIPATED RELEASES 8.1 GENERAL l The primary objective of the Thfl-2 Cleanup Program was the elimination of the radiological hazards to the public resulting from the March 28,1979 accident and nummization of on-site ; worker exposure. The program progressed from the initial efforts to stabilize the plant conditions l through the fmal major cleanup efforts, including the removal and shipment of fuel and decontammation of major portions of the AFHB and the Reactor Building. The cleanup efforts have progressed to the point where the plant is in a stable and benign condition suitable for passive I storage with trinimum active maintenance. The potential for release of significant quantities of radionuclides during PDMS is substantially reduced from that during normal power plant operation, or any of the post accident cleanup phases. His results from the reduced radionuclide inventory (see Tables 5.3 4 and 5.3-5) and the absence ofinherent driving forces for transport processes. The assessment of any radionuclide release during PDMS, thercibre, hinges on the identification of processes or events that could either alter the potential for transport of the remaining radionuclide inventory or proside unanticips.ted transport m-chanisms to the environment. A range of potential unanticipated events has been I postulated to establish the bounding conditions of potential off-site releases. The radiological consequences associated with routine releases as well as the bounding conditions are then es:imated O in subsequent sections. N
" In addition to the evaluation of routine releases, evaluations of the emironmental effects are presented for each unanticipated event which results in off-site radiation exposures in excess of those which result from routine releases.
8.1.1 ROlJTINE RELEASES Atmospheric releases to the emironment during routine PDMS operations will be limited to airbome contamination released as a result of operating the Reactor Building ventilation systems or through the Passive Breather System sia the Auxiliary Building. Ventilation discharges will be through controlled, HEPA filtered, and monitored paths. The source terms for the routine releases are tabulated on Table 8.1-1. Liquid systems, except for systems needed to occasionally process batches of contanunated liquids, were dramed to the extent pmetical and deactivated for PDMS. The major sources of contaminated liquids requiring processing during PDMS are expected to be groundwater inleakage, collected precipitation, and occasional small quantities of fluids used for local decontamination. Rainwater and groundwater inleakage are anticipated. Such inleakage will be collected and analyzed for any contanunation. The canability of processing this liquid will be available to ensure that discharges are well within regulatory requirements. The source terms for routine liquid releases are tabulated on Table 8.1-1. t l
,( 3, 2 8.1-1 UPDATE 2 - AUGUST 1997
l I 3.1.2 SOURCE TERMS The im entory of radionuclides remaining mite during PDMS is greatly reduced from that existing prior to the accident or during any of the phases of the recovery operations. This results prunvily from (1) removal of the fuel, which represents the largest conc, ntration of radionuclides, . (2) processing and shipping radioactive waste, and (3) natural decay. The remaining radioactivity can be charactenzed as residual contammation located primarily in either closed piping systems that were drained, but not aggressively decontanunated, on surface films tightly adherent to equipment or structural wrfaces. An exception is the Reactor Building basement (elevation 282'). He largest source of radioactivity in the Reactor Building basement is the block wall enclosing the stairwell and elevator. Radionuclides (primarily cesium and stromium) have been absorbed into the concrete structure of the blocks during the period when the wall was partially submerged in the highly-contammated water which collected in the Reactor Dailding basemert during and following the accident. Since the radioactive material is embedded in the concrete, it is not readily available as a source for airlo ne release in the near term. However, over longer periods of time, mechanisms related to diffusion and leaching by cyclic changes in moisture content may transport a fraction of the radionuclides in the block wall to the surface where it can become available fer suspension. Even though this fraction is expected to be small, the large inventory of the block wall (i.e., an estimated 19,000 Ci of Cs and 750 Ci of Sr) could make any suspension of radionuclides reaching the surface a significant airborne source term. Other major sources of radioactivity in the Reactor Building, which could make a significant contribution to the airborne source term, include the remaining wall and floor areas that were submerged in the highly-contammated water located in the Reactor Buildmg basement following i the accident, the interior of the D-rings, and the sediment remaining in the Reactor Building basement subsequent to the completion of the sediment removal activities. The current estimates of the inventory of Cs and Sr activity in these areas are listed on Table 8.1-2. Since these potential sources, as well as other less-significant potential sources of airbome radioactivity, have existed for a number of years, their effect on Reactor Building atmospheric particulate concentrations can be deduced from existing measurements. The denvation of a maumum airborne particulate level for use as a source term for routine releases during PDMS is given in Section 8.1.2.3. Another important factor in the consideration of residual contamination is the transuranic content. Although the quantity of fuel remaining after completion of defueling is insufficient to be of concem with respect to criticality, it is necessary to examme the potential contribution it could make to radiological source terms. He relative fractions of the significant transuranic elements remaining in the residual fuel are given in Table 8.1-3. The Ci fractions for the residual fuel were calculated based on the original core inventory corrected for 8 years decay. On the basis of the samples analyzed to date, as well as the analyses of the course of the accident, the transumnic elements can be assumed to be associated with residual fuel. Most of the residual fuel remaming during PDMS will be fixed in the form of fine and granular debris that is inaccessible to defueling, tightly adherent surface deposits not readily removable by available dynamic defuelmg techniques, and resolidified material 8.1-2 UPDATE 2 - AUGUST 1997
(d) that is either tightly adherent to the RV components or inaccessible to d> fueling. Approximately 1.3 kg of residual fuel fmes ternain in the Reactor Building basement sediment that could conceivably become airbome, in addition, there is residual fuel in other areas open to the RB atmosphere that could also become airbome and residual fuel in areas that is essentially contained (e.g., the RCS), covered (e g , the RV), or bagged (i.e., the defueling tools) that was also considered to some degree. As a reference point for the calculations of potential off-site dose consequences,it is conservatively assumed that 2.8 kg of fuel remains as a suspendable airborne source during PDh1S. 8.1,2.1 Airbome Releases The method of suspension and hence the fraction of the inventory contr.buting to the source term is a function of the scenario postulated. For routine releases, the airbome concentratisn was estimated from the most applicable observations of Reactor Building airbome concentrations to date (see Section 8.1.2.3). It was assumed that the Reactor Building airbome concentrations will reach equilibnum between intermittent operations of the Reactor Building purge system. Each
- purge operation is assumed to result in the discharge of the entire contents of the Reactor Building atmosphere. A total of fifty discharges of Reactor D2iilding atmosphere particulate content per year was conservatively assumed fcr the routine release calculations. For the postulated unanticipated events inventory, experimentally determined suspension fractions were used as described in Section 8.2.
8.1.2.2 Liquid Releases O Inicakage of groundwater and precipitation is anticipated to be the major scurce ofliquids during ("/ PDhtS. Such inleakage, which has occurred in the past, is kept under control by periodic maintenance. Based on the experience to date, an annual inleakage of 5000 gallons is conservatively estimated. To the extent *. hat such inleakage becomes contaminated by any residual contamination on floors and sumps, it will be processed before discharge. Experience to date has shown that twical release concentrations of Cs 137 and St 90 are 4E-6 Ci/ml and 1E-5 pCi/ml, respectively. The liquid effluent source term, therefore, is based on the assumption of a discharge of 5000 gallons annually at the above mentioned radionuclide concentrations. The source terms resulting from these considerations, as well as several of the controlling parameters, are summarized in Table 8.1-1. 8.1.2.3 Airbome Source Terms During PDhtS for Routine Releases The largest sources of radioactive material which could cause significant off-site releases were removed prior to entering PDhtS. He remaining radioactivity can be characterized as residual contamination which is either deeply t.mbedded in solid materials (e g., activation products in the reactor vessel and s metural materials), or distributed in thin films adherent to surfaces which have been flushed, but not aggressively decontaminated (e g., by chemical solutions or mechanical surface removal). His remaini .g inventory of radionuclides, therefore, is mostly fired and does not represent a potential airborne source tenn in the near future. Over the longer term, howcwr, a small fraction of this inventory may become available for suspmsion as a result of the aging of surface conditions (e.g., rustmg of steel surfaces, c' d iag of
;"/ 3.1-3 UPDATE 2 - AUGUST 1997 l
I
paints, or flaking of concrete surfaces). The mechanisms of suspension of surface contammation resulting from these aging processes are diffusion, air motion and evaporation. In addition to these sources of radioactive material, the sediment remaining in the Reactor Building basement following cleanup activities and the residual fuel in other areas open to the RB atmosphere prodde a source ofloose activity available for suspension. While it is not practical to attempt quantification of each of the processes described above, the upper bound of their combined effect can be determined empirically from the observation of airborne contammation levels in the Reactor Building. He Reactor Buildmg atmosphere has experienced conditions spanning a broad range of the variables affecting the formation of airborne contanunation (e g., temperature, humidity and concentration gradients and changes in air Row) which readily encompasses the conditions emisioned for PDMS, ne most important factors affectiag airborne contammation levels today are hu, nan activities associated with decontanunation, plant modification, and defuelmg activities (e s., foot traffic, sibrations from mrchinery, cutting, grirvling and weldmg) Since such actisities will not routinely occur during PDMS, the Reactor Duilding air samples from the period prior to routine Reactor Buildmg entries are likely to be more representative of PDMS conditions thar. tbr current Reactor Building atmospheric concentrathm Since Reactor Building entries commenced in late July of 1980, a review of Reactor Buildmg air sample results from June and early July, 1980 was performed. 71.e meumam airbene concentrations of Cs 137 and Sr-90 were 3.8E 10 and 1.3E-10 uCi/cc, respectively. He rate of build-up of actisity following purging can be estimated by the T order rate equation: dC/dt = (ISN)-Iy,C, - [(Q/V)C.) Where IS, is the net sum of all sources (pCi/ day) of radionuclide i, Iyi is the sum of all concentration-dependent removal rate constants and Q is the net exhaust flow of the ventilation (purge) system from the Reactor Buildmg free volume (Y). For the initial conditions ofinterest (i.e., negligibly small concentrations et the beginning and no purge flow during this period), this equation has the simple solution: C,(t) = (Is/Iyi)(1-e*i') Where s, = S/v The source and depletion parameters in this equation which approv.imate the observed behavior following the krypton purge are: Is = 5.0 E 11 ( Ci/cc/ cay) for Cs-137 Is==.005 Iy 1.5(day' E-11 for (p)Ci/cc/ both isotopesday) for St-90, and if the Reactor Building atmosphere was allowed to reach equilibrium (i.e., previously prevented by personnel entries, associated " mini-purges," and other activities) the equilibrium concentration would be: 8.1-4 UPDATE 2 - AUGUST 1997
() C4 = (Es/Iy) ne atx>ve source and sink estimates would predict equilibria at 1.0E4 and 3.0E-9 pCi/ce, respectively, for_Cs 137 and Sr 90. he values are considered conservative as they exceed the actually observed levels at any time after irutial accident condstions. He removal of many of the potential sources of Cs-137 and St 90 in the Reactor Building reculting from the decontammation oflarge surfaces areas, removal of the basement water and sedunent, and flushing and sealing of contaminated systenu prior to PDMS, has resulted in further ruiuctions in airborne concentrations during PDMS. I 8.1.3 OFF-SITE DOSE ESTIhMTES l i Emironmental doses from postulated releases were quantified with the SEEDS code. This is the model used to estimate routine doses for annual reports to the NRC. The accuracy of the SEEDS code was documented in Reference 8.1 1. This model calculates the dispersion of radionuchdes in three ways: normal airbome dispersion (X/Q), depleted dispersion (i.e., which accounts for various removal processes like settling out), and deposition (D/Q). The SEEDS calculations use the " delta T", or the difference in temperature between sensors at 33 feet and 150 feet, of the on site meteorological tower to determine the atmospheric (Pasquill) [ stability class. He model then uses the stability class with other meteorological parameters st,eb as wind speed and direction along with plant parameters including stack height, stack diameter, and (' stack flow rate to determine the atmospheric dispersion. The meteorological data used by SEEDS are automatically collected from the on-site tower and stored electronically for lutare use. Each meteorological tower sensor (there are about 20)is polled by a computer every 10 seconds. The ten-second tesults are averaged into a 15 minute r.verage. The fifteen minute averages centered on the hour are used as hourly values. The hourly values are used for routine dose calculations. ! To caiculate the doss to the public, SEEDS employs numerous data files which describe the area around TMI in terms of population distribution and foodstuffs production. The area around TMI is subdivided into rixteen equal meteorological sectors (N, NNE, NE, ENE, E, ESE, etc.). Each sector is then represented in each set of data files and dispersion is calculated for each sector separately. Data files include such information as the distance from the station vent to the site boundary in each sector, the population groupires, and the location of milk cows, milk goats, gardens of more than 500 square feet and meat ammals. SEEDS also contains dose comersion factors for 75 radionuclides for each of four age groups (aduits, teenagers, children and infants), seven pathways (inhalation, ground deposition, plume direct dose, and ingenion of cow milk, goat milk, vegetables, and meats), and eight organs (total body, thyroid, liver, skin, kidney lung, bone, and GI). The pathways, organs and age groups are j those specified in Regulatory Guide 1.109. i l S.1-5 UPDATE 2 - AUGUST 1997 l. l l l 1
The atmospheric dispersion is combined with the dose conversion factors and applied to each organ, age group, and pathway to estimate the dose to an individual by integrating by sector, distance, and time (meteorology changes) to determine the dose, distance and direction to the maximally-exposed individual. 8.1.3.1 Routine Releases ne off-site doses resulting from rot. tine releases of gaseous fission products and transuranics were estimated by combining the annual release source term discussed in Section 8.1.2.3 and summartzed in Table 8.1 1, averaged over the year, with the hourly 1985 meteorological data from the Thil tower. He calculation, therefore, is based on about 8600 different sets of meteorological conditions. Liquid pathway doses were evaluated using htIDAS (a precursor to SEEDS), based on the 94 year average Susquehanna River flow. MIDAS liquid dose calculations consider three pathways: fish ingestion, water ingestion, and shoreline exposure The eight organs and four age groups of Regulatory Guide 1.109 also were used. For compariron, the larEest potential doses which could be postulated on the basis of the actual effluent measurements for Thil Units 1 and 2 are shown in Table 8.1-4 for several recent years. It should be noted that the calculated values on this table are hypothetical doses which could be accrued to a maximally-exposed individual on the basis of the measured effluents from TM1-2. The comparison of these calculated doses with the estimated PDMS doses on Table 8.1-4 shows that actual airbome releases from TMl-2 during recent cleanup activities were generally smaller than the projections made for PDMS. It is concluded, therefore, that the estimated values of Table 8.1-5 are conservative estimates of the likely impact of PDMS on the emironment. Nevertheless, the estimated doses for PDMS are such small fractions of the normal background doses that they can be considered insignificant. 8.1.3.2 Unanticipated Releases ne off-site doses from unanticipated events have been calculated using SEEDS and other Regulatory Guide 1.109 based calculation methods (e g., MIDAS). Off-site doses from releases which are described as instantaneous puffs in Section 8.2 were calculated for acute inhalation exposure only. The calculation is based on the actual expected station ventilation flowTate and assumes a very stable G class stability to maximize the off-site concentration estimates. The actual calculated X/Q for these conditions of 7.67E-4 sec/ cubic meter was used. Due to the short duration, the other pathways included in the Regulatory Guide 1.109 Appendix ! calculation are not considered for instantaneous releases. He off site doses from other than puff releases (i.e., 80 minute or longer unanticipated airbome releases) have been evaluated using SEEDS. A 14-hour period of very stable conditions was selected at random from the 1985 TMIN3 meteorological data. The selection of this data maximized off-site concentration estimates and allows the inclusion of the ingestion pathways into the total dose estunate. He release source terms given in Section 8.2 were averaged over the 14-hour period. Finally, the long-term (i.e., 3 month) release for the open penetration event was also evaluated 8.1-6 UPDATE 2- AUGUST 1997
. _ . _ _ _ . _ . . _ . _ . . . _ . . . _ _ _ _ _ _ _ - . . . _ _ . . . . . . . _ . _ _ _ . . . . _ _ _ _ . . _ . _ _ _ . _ _ . _ _ _ . _ . ~ . _ _ _ . _ _ .
n e using SEEDS. In this case,1936 meteorological data from the TMI tower (July I to September.
- 30) was used. The selection of mis data provides a best estimate of the expected dose from a -
calendar quaner while maximizing some effects such as animal grazmg periods.- The release
. source terms given in Section 8.2 were averaged over the 3 month period.
i 'l 5 i i-o i l e 4 4 i 1 1 J h i 1-4 i 8.1-7 UPDATE 2- AUGUST 1997 4 s n ,e-- - ,. va,
----__-__.--------a,w. +, mc n +w- -, .-w ,>,e., -- ,a,- w-- -g,- er ,
FEFERENCES 8.1-1 GPU Nuclear memorandum 6650-90-143, from D. W. Ballengee to T. D. Murphy,
" Safety /Emironmental Determination and Review of SEEDS," dated August 10,1990.
t i
\
l l 4 4
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8.1-8 UPDATE 2 - AUGUST 1997
+
-i t
( ,
. TABLE 8,1 1
SUMMARY
OF SOURCE TERMS
-l RottrrNE RELEASES:
4
- Airborne Concentration Cs 137 = 1 x 10 pCi/ml Sr-90 = 3 x 10* pCUml >
Reactor Buildmg Atmosphere 50 Containment volumes
- Particulate Content per year p
[ Fiher efficiency -99 % Particulate soura term Cs-137 = 2.8 x 10" CUyear Sr-90 = 8.5 x 10 Ci/ year 4
- Transuranic source term 3.2 x 10 Ci/ year ,
. (See Table 8.1 3 for isotopic composition) i L10UID RELEASES: 4 Quantity 5000 gaUyear r ! Cs scurce term 7.6 x 10'8 Ci/ycar i i Sr source term 1.9 x 10" Ci/ year i 7
\
8.1-9 UPDATE 2 - AUGUST 1997 i
- 1. -
- w e m,,.,--e, .--w-_v__y y ,,.,_,.iweg., , , ,vp,,,,yq.;p,,,, ,, ,%, ,,, , _ , _ 9 g,,4,, .,, g..m _,.,_ _,,Y,, m_,, ,
TABLE 8.12 ESTIMATES OF Cs AND Sr INVENTORY OF SELECTED REACTOR BUILDING LOCATIONS i Cs 137 ACTMTY St 90 ACTIVITY LOCATION _(CURIEji)_ (CURIES)
"B" D-ring 15,000 750 "A" D-ring 1,660 80 Previously Submerged 7,000 300 Floors and Walls Sediment 460 450 Openitions Deck 29,000 8,200 (305' and 347')
NOTE: These numbers are pre-decontanunation, are considered consenutive, and have error bands of up to 50% 8,1-10 UPDATE 2 - AUGUST 1997
TABLE 8.13 Ci FRACTIONS IN RESIDUAL FUEL ISOTOPE Ci FRACTION Pu-238 1.12 E 2 Pu-239 6.81 E-2 Pu-240 3.32 E-2 Pu 241 7.73 E 1 Am-241 1.15 E-1 TOTAL 1.00 8.1-11 UPDATE 2 - AUGUST 1997
TABLE 8.1-4 MAXIMUM POTENTIAL D7SES ESTIMATED FROM ROMTINE EFFLUENT RELEASES (mrem / year) PDMS PDMS Administrative 1983 1984 1985 1986 (ESTIMA1ED) Limit
- Liquid Releases Critical Organ Dose 2E-3 1.4 E-2 4.lE-3 3.2E-3 5.0E-3 10 Critical Organ Liver Liver Bone Bone Bone Total Body Dose IE-3 6.8 E-3 1.4 E-3 1 3 E-3 2.0E-3 3 Gaseous Releases Skin Dose 9E-3 1. l E-2 13E-5 5.0E-5 15 Total Body Dose 8.5 E-2 9.3 E-5 4.5 E-6 4.2 E-7 5 Airborne Particulate / lodine Critical Organ Dose I . l E-2 1.8 E-3 22E-3 1.7E-2 2.2E-2 15 Critical Organ Liver Total Body Total Body Bore Bone
- NOTE: The PDMS dose limits were derived from applicable dose limits established in 10 CFR 50 Appendix 1.
8.1-12 UPDATE 2- AUGUST 1997 O O O
TABLE 8.15 ESTIMATED DOSE COMMITMENTS RESULTING FROM ROUTINE AND UNANTICIPATED EVENTS DURING PDMS I. POPULATION DOSES Person-rem /vear Population doses from routine releases sia , airbome pathway:. Bone Dose 0.12 Total Body Dose 0.03 Population doses from routine releases sia liquid pathway: Bone Dose 0.67 Total Body Dose 0.24 Total population dose: Bone Dose 0.21 Total Body Dose 0.05 ( II. DOSES TO MAXIMAI.LLEXPOSED II DIVIDUAL mrem /vear Dose from routine releases via airbome pathway: Bone Dose 0.02 Total Body Dose 0.01 Dose from routine releases via liquid pathway. Bone Dose 1.18
- Total Body Dose 0.78 III. MAXIMUM UNANTICIPATED RELEASE Dose to maximally exposed indisidual from unanticipated event release 13.5 mrem (bone dose limiting - inhalation only) i l
Y 8.1-13 UPDATE 2 - AUGUST 1997
( 8.2 UNANTICIPATE0 EVENTS ANALYSIS Unant h.ipated radiological releases could occur from unanticipated occunences arising from the conditions %- or activities postulated during PDh15. Since there are no major actisities planned for PDhtS, an accidental event invohing a major fraction of the remaining inventory of radionuclides is not likely. However, a number of unanticipated events have been postulated, based on the types of actisities considered within the scope of PDhtS. ncse events were reviewed to id ntify the bounding event. hiinor radiological events were postulated, including leaks during vacuum operations, rnishaps dunng local decontamination operations, accidental cutting of contaminated piping and a fire in a enntaminated area.
\
Events of these types were postulated and analyzed in a Beneric study of a PWR decommissioning following an accident (NUREOiCR 2601). His generic study was used as a basis for the initial selection of evems to be postulated during PDhiS. %cse events were reviewed and events were added, modified, or deleted as appropriate based on applicability to Th142 during PDhtSt 8.2.1 YACUUhi CANISTER FAILURE It is anticipated that floor areas may be contaminated during PDhtS by personnel tracking contamination , from one area to another or through the settling of airbome contarnination on the horizontal floor surfaces. It may be necessary during PDhtS to centrol the spread of contamination in the various plant areas. One metbl of controlling this cot 2 .ination is by tl.c use of a llEPA vacuum to remove and trap the contamination ir a vacuurr. q lt can be postulated that the vacuum canister could fail due to a defcet in the hardware or an operator error. in either case, an event could occur involving a canister loadsd with a substantial amount of contamination ( which could be expelled into the local area. He event could occur with the purge system operating, the building at negative pressure and isolated, or with the building under pasuve ventilation. He vacuum canister has a volume of about 21E+5 cm' of contamination with :he following isotopic distnbution. IS.0 TOPE GRIES Pu 238 1.1E-3 Pu 239 1.3E 2 Pu 240 3.4E 3 Pu 241 1.4E 1 Am 241 2.IE 2 Sr 90 3.8E+0 Cs 137 6.0E l i Upon failure of the vacuum canister, a portion of the contamination becomes airborne. The airbome release fraction is assumed to be IE 2. This is a higher release fraction than nonnally postulated due to the potentially high elevation release and the slightly pressurized expulsion of the contamination. Using the above scise term, the following three cares have been analyzed.
'In response to an NRC question generated during the review of the PDhtS SAR an evaluation of the powntial impact of radiolytic generation of nitric acid and hydrogen in the Reactor Vessel was performed, p The evaluation concluded the impact was negligible.
\'
8.2 -l UPDATE 2 - AUGUST 1997
-_.--..,...__m .----_.__ ,,,__,__,,,,,__,,_,_,_,
Case 1 Reactor Building Purce system operating For this case, it is assumed that the Reactor Building Purge is operating at the nominal fluw rate and all effluents will be filtered through the Reactor Buildmg Purge llEPA filters with a 99% filter efficiency. It is postulated that the vacuum canister fails and the total source term release fraction is exhausted into the Reactor Buildmg Purge, filtered througl. the llEPA filters, and exhausted to the emironment. That portion of the source term which is exhausted to the emironment is treated as a puff release. This results in a release to the emironment of:
]SOTOPE CURIES Pu 238 1.1E 7 Pu 239 1.3 E-6 Pu 240 3.4E 7 Pu 241 1.4E 5 Am 241 2.1E-6 Sr90 3.8E-4 Cs 137 6.6E 5 This postulated reluse results in a calculated bone dose to the maximally exposed indnidual of 6.9 mrem.
This dose includes the inhalation patimay only. Cue 2 - Repetor Buildinn at negathr Ditmtts For this case, it is assumed that the Reactor Building is isolated and is being held at a slightly negative pressure. As in Case 1, it is assumed that the vacuum bag ruptures and the contamination contamed in the vacuum canister i.ohen expelled to the Reactor Building atmosphere. A portion of the contarnmation is suspended and diffused in the Reactor Building atmosphere. It is then assumed that the Reactor Buildmg Purge is actuated and the source term release fraction is exhausted over a period of time through the Reactor Buildtng Purge IIEPA filters assuming a 99% filter efliciency. It is assumed that no settimg of the suspended contamination occurs and the total release fraction of the vacuum camster source term is exhausted through the Reactor Building Purge dunng the first Reactor Buildmg air change subsequent to the actuation of the purge system. In addition to the contamination released from the vacuum canister, the source tenn attributable to one normal Reactor Bailding volume must be included in the efiluent release calculations. This results in a release slightly higher than Case 1 over an 80 minute tune period Tais results in a release to the emironment of: 8.2 - 2 UPDATE 2 - AUGUST 1997
i C'T (U/ ISOTOPE Pu 238 CURIES 1.lE 7 Pu 239 1.3E-6 , Pu 240 3.4E 7 I Pu 241 1.4E 5 Am-241 2.1E-6 Sr 90 3.8E 4 Cs 137 6.6E 5 "This postulated release results in a calculated dose to the maximally-exposed indhidual of 1.1 mrem. This dose includes all applicable Appendix I pathways. Case 3 Reactor Buildine under passive ventilation Although it is not anticipated that personnel will be performing vacuuming operations in the Reactor Buildmg with the building under passive ventilation, this case is considered in the event the situation should occur. For this case, the assumptions for the source term from the failure of the vacuum canister are the same as for Cases 1 and 2. This case is very similar to Case 2, in that there will be no specific transport mechanism to exhaust the source term from the containment. Ilowever, it can be assumed that subsequent to the failure of the vacuum canister, the containment is isolated and the Reactor building Purge is activated. This then would be the same as Case 2, assuming the purge is actuated within a reasonable time after the failure of the p) ( V vacuum canister. However, the longer the delay after vacuum canister tilure before actuation of the Reactor Building Purge, the less the quantity of contamination available for off-site release. The clapsed time would allow a portion of the source term to settle and redeposit on the horizontal surfaces. Thus, the longer the time the greater the amount of source term redeposited and the less the amount of source term available for transport to the emitonment. Therefore, Case 3 results in a maximum release no greater than that calculated for Case 2. Since the manmum release for Case 3 can be no greater than that calculated for Case 2, no further analysis of Case 3 is requirrd. 8.2.2 ACCIDENTAL SPRAYING OF CONCENTRATED CONTAMINATION WITillilGli PRESSURE SPRAY lt is not anticipated that there will be major decontamination efforts during PDMS; however, it may be necassary to undertake some decontamination activities, This could arise from the necessity to undertake activities in areas that are contammated to levels greater than desired for the planned actisities. High , pressure spray and low pressure flush are two primary methods of surface decontammation. With either high pressure spray or low pressure flush it is possible to direct the spray fluid to an area with a level of contamination much higher than anticipated. In either case, some contamination may become airborne and be released to the emironmat, llowever, due to the higher pressures and velocities associated with the high pressure spray, the amount of
.O j \
v 8.2 - 3 UPDATE 2 - AUGUST 1997 l
contaminated rnatenal which may become airbome is greater with the high pressure spray than with the low pressure flush Therefore, the high pressure spray case is analyzed. It is assumed that a 25 gpm spray at 10,000 psi is directed at a representative area of casement contammation for a penod of 10 rninutes.
- lids results in the following irutial airborne source term.
]SQlOfE CURIES Pu 238 1.0E 5 Pu 239 2.6E-4 Pu 240 6 9E 5 Pu 241 9.2E 3 Am 241 1.1 E-4 Sr 90 1.5E+0 Cs 137 9.6E 1 The amount of contamination that remains airborne for this event is a function of the initial airbome fraction and plate out, it is assurned that a 9036 platcout will occur from an initial airbome fraction.
Using the above source term, it is possible to examine three different cases These are that the postulated event occurs: (1) with the Reactor Buildmg Purge operattng; (2) with the Reactor Building isolated and at negative pressure with the Reactor Buildmg Purge $ccured, and (3) with the Reactor Building under passive , ventilation. Doe.1 Rsactor Buildine Purge sv11cm onerating For this case, it is assumed that the Reactor Buildmg Purge is operatmg at the nominal flow rate and all effluents will be filtered through the Reactor Building Purge HEPA filters with a 99?6 filter efliciency. It is postulated that the high pressure spray redistributes the contamination and a portion is released to the Reactor Building atmosphere, it is a:sumed that this source term release fraction is then exhausted into the Reactor Building Purge, filtered through the HEPA filters, and exhausted to the emironment. That portion of the source term which is exhausted to the emironment is treated as puff release. This results in a release to the emironment of. ISOTOPE CURIES Pu 238 1.0E-8 Pu 239 2.6E 7 Pu 240 6.9E 8 Pu 241 9.2E-6 Am 241 1.1E 7 Sr 90 1.5E 3 Cs 137 9.6E-4 This postulated release results in a calculated bone dose to the maximally-egosed individual of 3.5 miem. This dose includes the inhalation pathway ordy. S.2 - 4 UPDATE 2 - AUGUST 1997
\ Can.2._]ha@tiluildine at necative pressms for this case, it is assumed that the Reactor Building is isolated at a slightly negative pressure. As in Case 1, it is assumed that the high pressure spray redistributes a portion of the contamination to the Reactor Buildmg atmosphere. Ilowever, in this case, the Reactor Building is isolated and the source term release fraction is contained within the Reactor Building and diffused within the Resetor Building atmosphere. It is then assumed that the Reactor Buildmg Purge is actuated and the source tenn release fraction is exhausted over a period of time through the Reactor Building Purge IIEPA filters assuming a 99% filter efliciency. It is also assumed that no settling of the redistributed contamination occurs and the tetal source tenn release fraction is exhausted to the emironment in the first Reactor Building air change subsequent to the actuation of the Reactor Buildmg Purge. In addition to the source term from the high pressure spray event, the source term attributable to one nonnal Reactor Building volume must also be included in the total sclease. This results in a release of the emitonment of: ISOTOPE CL%IES Pu 238 1.0E 8 Pu 239 2.6E 7 Pu 240 6.9E 8 Pu 241 9.2E-6 Am 241 1.1E 7 St 90 1.5E-3 Cs 137 9.6E 4 This postulated release results in a calculated dose to the maximally exposed individual of 4 0 mrem. His y dose includes all applicable Appendix 1 pathways. CAg.),,,,Jkq@Ll}gildtne under tiassive ventilatioB Although it is not anticipated that personnel will be performing high pressurt, spray decontammation operations in the Reactor Building with the buildmg under passive ventilation, this case is considered in the event the situation should occur. For this case, the assumptions for the source tenn resulting from the high pressure spray event are the same as for Cases 1 and 2. Case 3 is very similar to Case 2. He Reactor Building is under passive ventilation and there will be no specific transport mechanism to exhaust the source term from the Contamment. Ilowever, it can be assumed that subsequent to the distribution of contamination by the high pressure spray, the Containment is isolated and the Reactor Building Purge is actuated This would be the same as Case 2 assuming the purge is actuated within a reasonable time after the hp , icssure spray event. However, the longer the delay until actuation of the Reactor Building Purge, the less the severity of the event. The clapsed time would allow a portion of the contarmnation to settle and redeposit on the horizontal surfaces. Rus, the longer the time, the greater the amount of contamination redeposited and the less the amount of source term available for transport to the emironment. O
\ /
V 8.2 - 5 UPDATE 2 - AUGUST 1997 1
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. Gody, t ..se 3 results in a maxirnum release no greater than that calculated for Cas: 2. Since the
. simum release for Case 3 is no greater than that calculated for Case 2, no further analysis of Case 3 is +gnired 8.2.3 ACCIDENTAL CITITING OF CONTAMINATED PIPE Ahhough there are no such specific actisities planned during PDMS, it may be necessary to cut into systern piping, ne piping systems are contammated to various degrees, with many systems contaimng no contamination at all. Although each activity which involves the cutting of piping will be planned and controlled by procedures, it is possible that the degree of contammation in the planned cutting area may be underestimated or the indmdual performing the operation simply cut the wrong pipe. In either case, it is possible to release an unplanned source tenn to the Containment.
It can be postulated that an activity is being conducted which involves the cutting of a contaminated pipe. This cutting may be done by mechanical means, such as metal sawing, or by nonmechanical means, such as a gas cutting torti Regardless of the means, it can be postulated that the cutting is done in a area of higher than anticipated contamination or in an area with significant contamination where none was expected. In either scenario, it is assumed that a source term will be released to the containment. The quantities and distribution of the pnneipal isotopes are given below. ISDIDff! CLURiflS St 90 1.3E 3 Cs 137 4.2E-4 Pu 238 1.5E 8 Pu 239 1.6E.7 Pu 240 4.4E 8 Pu 241 2.0E 6 Am-241 3.2E 8 For this case, the source term is not pressunzed and is not a soose particulate medmm Rather, it is a film associated with the surface of the pipe. Therefore, the release fraction for this event is assumed to be 1.0E-4. Using the above source tenn release fraction, it is possible to examine three different cases. The postulated event occurs: (1) with the Reactor Building Purge operating; (2) with the Reactor Building isolated at a negative pressure and the Reactor Building Purge secured; and (3) with the Reactor Building under passive ventilation. Dse 1 - Reactor lhtildine Purce system operatine For this case, it is assumed that the Reactor Building Purge is operating at the nominal flow rate and all efIluents will be filtered through the Reactor Building Purge liEPA filters with a 99% filter efficiency. It is postulated that the process of cutting the pipe releases the contamination and it is assumed that the source term release fraction is distributed into the Reactor Buildtng atmosphere and exhausted into the Reactor Building Purge, filtered through the IIEPA filters, and released to the emironment. That portion of the source term which is exhausted to the emironment is treated as puff release. This results in a release to the emironment of: 8.2-6 UPDATE 2 - AUGUST 1997
hQ ISOTOPE CURIES Sr90 1.3E 9 Cs 137 4.2E-10 Pu 238 1.5E-14 Pu 239 1.6E 13 Pu-240 4.4E 14 Pu 241 2.0E 12 Am 241 3.2E 14 This postulated release results in a calculated bone dose to the rnaxirnally exposed indisidual of 2.6E 6 mrem. This dose includes the inhalation pathway only. Case 2 - RcE10LiluildintatneratirrEnnitt For this case, it is assumed that the Reactor Building is isolated at a slightly negative pressure As in Case 1, it is assumed that the process of cutting the pipe releases the contamination and a portion of the contamination is then released to the Reactor Building atmosphere.11owever, in this case, the Reactor Building is isolated and the source term release fraction is contained within the Reactor Building and diffused within the Reactor Building atmosphere. It is then assumed that the Reactor Building Purge is actuated and the source term release fraction is exhausted over a period of time through the Reactor Building Purge l{ EPA filters, assuming a 99% filter efficiency. It is assumed that no settling of the redistributed contammation occurs and the source term release fraction is edausted through the purge system to the emironment in the first Reactor Duilding air change subsequent to the actuation of the Reactor Building f] Purge, in addnion to the source term resulting frorn the high pressure spray event, the source term Q attnbutable to one normal Reactor Building volume must also be included in the total release. This results in a re! case to the emironment of. ISOTOPE CURIES Cs-137 5.7E-6 Sr 90 1.7E-6 Pu 238 3.2E 12 Pu-239 3.6E-11 Pu 240 9.6E-12 Pu-241 4.4E 10 Am 241 7.0E-12 This postulated release results in a calculated dose to the maximally exposed individual of 4.8E 3 mrem. This dose includes all applicable Appendix 1 pathways. I V 8.2-7 UPDATE 2 - AUGUST 1997
Ds) e Regtor Building under nassive ventilatim Although it is not anticipated that personnel will be performing pipe cutting operations with tne Reactor Building under passive ventilation, this case is considered in the event the situation should occur. For this case, the assumpti a for the source term resulting from the process of cutting the pipe are the same as for Cases I and 2. Case 3 is very similar to Case 2. Since the Reactor Building is under passive ventilation, there will be no specific transport mechanism to exhaust the source term frorn the containment. Ilowever, it can be assumed that subsequent to the redistribution of contamination by the cutting operation, the contamment is isolated and the Reactor buildmg Purge is activated. His would be the same as Case 2 assuming the purge is actuated within a reasonable time after the pipe cutting event. Ilowever, the longer the delay until actuation of the Reactor Building Purge, the less the severity of the event. The clapsed time allows a portion of the source tenn to settle and deposit on horizontal surfaces. He longer the time, the greater the amount of source term deposited and the less the amount of source term available for transport to the emironment. Herefore, Case 3 results in a maximum release no greater than that calculated for Case 2. Smce the maxirmun release for Case 3 is no greater than that calculated for Case 2, no further analysis of Case 3 is required. 8.2.4 ACCIDENTAL BREAK OF CONTAh11NATED PIPINO lt may be necessary dunng PDhtS to separate a piping system to remove or mstall a component, inspect a component, or to decontaminate internal portions of a pipmg system or component. It can be postulated that the piping system may be broken at a point that contains a higher level of contamination than expected or is contaminated in an area expected to be not contaminated. In either case, an unanticipated source term may be released to the containment. This sequence of events is very similar to those postulated by the accidental cutting of a contaminated pipe and it can be assumed that the source term will be the same. The consequence of a break in a contanunated pipe will be less than that for the accidental cutting of a contaminated pipe because the drising mechanism for the source term (i c., the velocity of the gas from a gas cutting torch) will not be present. Herefore, it can be assumed that the off-site consequences of accidental breaking of a contammated pipe will be no greater than that for the accidental cutting of a contaminated pipe and no further analysis is required. 8.2.5 FIRE INSIDE TiiE CONTAINhiENT Re risk of a fire in the Containment will be greatly reduced dunng PDh15. Pnor to entering PDh1S, all unnecessary and removable combustibles will be removed from the Containment and ignition sources will be reduced. For example, all unnecessary electric power sources will be deactivated at locations outside Containment. In addition, personnel population of the Containment will be at a nummum dunng PDhlS. Although the risk of a fire in the Containment during PDh1S is very small, there is the possibility that some personnel error or electncal short could lead to a limited fire. Dse 1 - Reactor Buildine Purce Ontalms For Case 1, it is assumed that the fire occurs while the Reactor Building Purge is y~iating. Therefore, the Reactor Building Purge is assumed to continue to operate at the nommal flowrate throvgh the entire duration 8.2 8 UPDATE 2 - AUGUST 1997
o of the fue. All fire zones in the Containment were analyzed with respect to resultant off site dose. The analysis of a fire on the operations deck (El. 30$' and 347) produced the highest entical organ dose to the maximally exposed individual, i.e., a 13.5 mrem bone dose. It was assumed that the operations deck contaned 29,000 Ci of Cs 137 and 8200 Ci of Sr 90 prinutily on the stored RV head assembly. One hundred percent (100%) of the contamination was assumed to be loose, surface actisity available to become airborne in a fire. An airbome suspension factor of IE 3 was used for both the contanunation and fuel. Any platcout of the airbome source term was conservatively ignored. A 99% efficiency is assuraed for the llEPA filters in the Reactor Building Purge. Bds results in a release to the emironment of. 11QLQEE JURIES St 90 9.1E-4 Cs 137 3.0E 3 Pu 238 3.5E 7 Pu 239 2.2E 6 Pu 240 1. l E-6 Pu 241 2.5E 5 Am 241 3.6E-6 This postulated release results in a calculated bone dose to the maxirnally-exposed individual of 13.5 mrem (Reference 8.2 1). Bds dose includes the inhalation pathway only. Case 2 - Reactor Building at nerative pressure Q For this case, it is assumed that the Reactor Building is isolated and is bemg held at a slightly negative pressure. It is assumed that the postulated fire distnbutes a portion of the contamination to the Reactor Building atmosphere. However, in this case, the Reactor Buildmg is isolated and the source term release fraction is contained within the Reactor Building and diffused within the Reactor Building atmosphere, it is then assumed that the Reactor Building Purge is actuated and the source term release fraction is exhausted over a period of time through the Reactor Building Purge HEPA filters assuming a 99% illter efficiency. It is also assumed that no settimg of the distributed contammation occurs and the total source term release fraction is exhausted to the emironment in the first Reactor Building air change subsequent to the actuation of the Reactor Building Purge. In addition to the source term from the postulated fire, the source term attributable to one normal Reactor Building volume must also be included in the total release. All fire zones in the Containment were analyzed with respect to resultant off site dose. For this case, the analysis of a postulated fire on the operations deck produced the highest dose to the maximally exposed individual. He postulated release to the emironment is as follows: 8.2-9 UPDATE 2 - AUGUST 1997
ISOTOPE CIRUES Sr 90 9.1 E-4 Cs-137 3.0E-3 Pu 238 3.5E.7 Pu 2h 2.2E-6 Pu-240 1. l E-6 Pu 241 2.5E 3 Am 241 3.6E-6 nis postulated release results in a calculated bone dose to the madnally-exposed indnidual of 3 0 miem (Reference 8.2 2). His dose includes all applicable Appendix ! pathuys. fase 3 - Reactor Buildjantr.idrippssive ventihti2D For this case, it is assumed that the postulated fire occurs during a tune when the Reactor Buildmg is under passive ventilation, he postulated fire could either be small or large in the event of a small fire, it is assumed that there is a slight Reactor Building pressurization (i.e., less than that necessary to initiate automatic closure of the Breather isolation va!ve) and some airborne contammation is expelled through the filtered breather, With the RB under passive ventilation, the most significant motive force available to expel the airborne contarnination would be the pressure differential created by the fire itself, flowever, the quantity of expelled contanunation would be less than that expelled in Case I through the RB Purge llEPA filters because of the size and number of the ventilation pathways. In addition, the Breather IIEPA filter is assurned to be the same efficiency as the RB Purge IIEPA filters, i e.,99% Therefore, the off site dose consequences for the Case 3 srnall fire are bounded by the off-site dose consequences of Case 1. In the event of a significant fire, the Breather isolation valve would close and effectively seal off the Reactor Buildmg. Although the RB is not airtight, the amount of unfiltered leakage, as shown in Section 7.2.1.2.3, would be much less than 1/100 of the amount released through the 99*4 efficient IIEPA filter in the above small fire analysis. Herefore, less :ontamination would be expe!!cd during the time when the isolation valve is shut than when it is open. He Breather isolation valve could subsequently be reopened when RB pressure reached % psi Re remainder of the event scenario for this large fire case would then follow the above small fire analysis. Herefore, the off site dose consequences for the Case 3 large fire are also bounded by the off. site dose consequences of Case 1. 8.2.6 OPEN PENETRATION lt can be postulated that during PDMS surveillance or maintenance activities, a penetration isolation mechanism is inadvertently removed or left open. This would result in a potential unfiltered pathway to the emironment for the Containment atmosphere. It also can be postulated that the mechanical failure of a piping penetration or failure of the IIEPA fiher in the passive breather system could result in an unfiltered pathway. The result of any one of the above postulated events can be characterized as an tmfiltered path to the emironment. Therefore, instead of analyzmg each of the events separately, a representative event is analyzed. 8.2 - 10 UPDATE 2 - AUGUST 1997
For the purposes of this analysis, the following conditions are assumed: 1
- 1. He contairunent is under passive ventilation l 1
- 2. Here is a 6 inch diameter tmfiltered pathway to the emironment
- 3. He unfiltered pathway is open for one quarter of a year (i c., one surveillance inten al)
- 4. 2.5 Reactor Building air changes with the emironment
$. 100% of the total release is through the unfiltered pathway
- 6. "Nonnal" Reactor Building source term His results in a release to the emironment of:
ISOTOPE CURIES St 90 4.2E-4 Cs 137 1.4E 3 Pu 238 1,8E 9 Pu 239 1.1 E-8 Pu 240 5.3E 9 Pu 241 1.2E 7 Am 241 1.8E 8 This postulated release results in a calculated bone dose to the inaximally exposed indhidual of 4.6E 1 mrem (Reference 8.2 3). This dose Meludes all applicable Appendix ! pathways. (G c 8.2 - 11 UPDATE 2 - AUGUST 1997
EffERENCES 8.21 GPU Nuclear memorandum 6615 92 0160 from S. Acker to E. Schrull, " Dose Calculation Results per memo C312 921045, PDMS SAR Rev.16," dated October 27, 1992 8.2 2 GPU Nuclear memorandum 6615 92-0162, from S. Acker to E. Schrull, " Additional Dose Calculations per memo C312 92 1045, PDMS SAR Rev.16," dated October 30, 1992 8.23 GPU Nuclear memorandum 6510-93-0077, from S. Acker to E. Schrull, " Dose Calculation Results per memo C312 93 1019, PDMS SAR Rev.12," dated May 21, 1993. 8.2-4 GPUN memorandum 4240 92-093 from R. Freeman to E. Schrull, "Radiolytic ilydrogen & Nitric Acid Generation in RV," dated June 29,1992. 8.25 GPU Nuclear calculation 4240-3220 91029, " Volume of Water Remaining in RV After Fmal Pumpdown," dated December,1991. 8.2 6 GPU Nuclear calculation,4440-9390-90002, " Condensation h, RV during PDMS," dated Febmary,1990. O e.2 12 - A m m ecSr1,,, 9
1 4 1 l 1 l i j i t 4 l i i l 1 ) CIIAPTER 10 4 i ADMINISTRATIVE FUNCTIONS i I 1 I l
CHAPTER 10 ADMINISTRATIVE FUNCTIONS (V TABLE OF CONTENTS SECTION TITLE PAGE
10.0 INTRODUCTION
10.0 1 10.1 QUALIT'l ASSURANCE PLAN 10.1 1 10.2 SECURITY PLAN 10.2 1 10.3 EMERGENCY PLAN 1031 t d 10.4 RADIATION PROTECTION PLAN 10.4 1 1 ~ 10.5 ORGANIZATION 10.5 1 l 10.5.1 CORPORATE ORGANIZATION 10.5 1
'10.5.1.1 President . GPU Nuclear 10.5 1 10.5.1.2 TMI Division 10.5 1 10.5.1.3 Engineering Division 10.5 1 4-g 10.5.1.4 Nuclear Safety and Technical Senices Disision 10.5 1 10.5.1.5 Financial and Planning Senices Disision 10.5 1 i 10.5.1.6 Human and Administrative Services Division 10.5-2 10.5.1.7 Other Functions 10.5 2 10.5.2 ONSITE ORGANIZATION 10.5 2 10.5.2.1 PDMS Manager 10.5 2 O
i - UPDATE 2 - AUGUST 1997
_ _ ___ _ _ _ __- __ . __= ___ _ - - . . - - - - _ _ - _ . . . _ - _ - 2 i i I ;
. i i CHAPTER 10 1
i i TABLE OF CONTENTS (Cont'd) i ! LIST OF FIGUkES i ! FIGURE I N.E I.LT.LE f.MiE i 10.5 1 GPU NUCLEAR CORPORATION ORGANIZATION PLAN 10.5 3 l 4 l 10.5 2. ORGANIZATION PLAN .TMI 2 10.5 4 4 i i. 1 . I j i I t i< l i i i l [- !
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1 CilAPTEll10 ADMINISTRATIVE FUNCTIONS 3
10.0 INTRODUCTION
i ! "Ihe prirnary administrative functions necessary for the management of TMI 2 duriag i PDMS are referencal in this chapter. i I i
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1 i i s 6 10.0 - 1 UPDATE 2 - AUGUST 1997 a -
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10.1 QUALITY ASSURANCE PLAN Upon completing the TMI 2 Cleanup Program, the plant is in a safe, stable condition that can be maintained efficiently and poses no risk to the health and safety of the public. Until GPU Nuclear determines a final disposition for TMI 2, it is intended that the plant remain in this condition. This phase of;:: ant life has been termd Post Defueling Monitored Storage. During the PDMS period, GPU Nuclear is licensed under 10 CFR $0 to "pessess but not to operate" the TMI 2 facility. Since the plant is in a non operating and defu led status, there are no structures, systems, or components that perform a safety function. Therefore, the gaality assurance requirements of 10 CFR 50 Appendix B do not specifically apply. Ilowever, this Plan has been developed to provide TMI 2 with a limited scope PDMS QA Program based t'pon the guidance of Appendix H requirements. The specific requirements for QA durmg PDMS are documented in the "TMI 2 PDMS Quahty Assurance Plan." ; i 1 (~ 10.1 - 1 UPDATE 2 - AUGUST 1997
10.2 SECURITY PLAN 1he Code of Federal degulations 10 CFR $0 and 10 CFR 73 define the security requirernents for nuclear power plants. Due to the defueled and non operatmg condition of TMI 2 during PDMS, the Secunty requirernents applicable to the facility are less than those that are applicable to an operating nuclear power plant. TMI 2 complies with all appbcable secunty requirements. The specific security provisions for TMI 2 are in the "TMI Modified Amended Physical Security Plan." O
\ \j 10.2 - 1 UPDATE 2 - AUGUST 1997
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'd 10.3 EhiERGENCY PLAN 10 CFR $0 47 establishes requirements for the content and enteria for acceptance of emergency plans. Emergency planning requirements are based on the assumption of the rotential necessity to notify the public of the Lstence of, or potential for significant off.
site releases.10 CFR $0 Appendix E recognizes that emergency plarming needs are different for facihties that present less risk to the public. Due to the non operating and defueled status of Thil 2 during PDhtS, there ;s no potential for any significant off site radioactive releases and, due to the cunence of TMI l on the same site, emergency planmng requirements for the site are dominated by Thil l. Herefore, the limited emergency planning necessary to acconunodate the existence of TMI 2 on the same site as Th11 1 has been incorporatw! in the integrated corporate emergency plan. Emergency planning implementing requirements are given in the "GPUNC Corporate Emergency Plan
- and have been reviewed and approved by the NRC.
%e emergency plan for the Thil site incorporates all of the essential emergency planning requirements established by 10 CFR $0 Appendix E and other regulatory guidance Since '
there are no events associated with TMI 2 which could result in a telease approaching d.e levels estabhshed in the Protection Action Guide, the site emergency action levels are based on potential events which could occur at TMI 1. The site emergency fxilities, such as the Emcrgency Control Center, the Technical Support Center, and the Operations f,upport Center are located in or in convenient proximity to TMI l. All site perscnnel are trained and diilled to respond to any declared site emergency event. O
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10.3 - 1 UPDATE 2 - AUGUST 1997
I 10 4 RADIATION PROTECTION PLAN GPU Nuclear maintains a Radiation Protection Plan which rnects or exceeds standards for i protection against exposures to radiation and radioacthiry arising out of GPU Nuclear owtiership ad operation of nuclear facilities. The unplementation of the Radiation , Protection Plan at TMI 2 ensures that the facility will be managed and maintained during PDMS in a manner which immrmr.cs risks to employees, contractors, sisitors, and the public frorn exposure to radiation and radioactivity at the facility. The implementation of the plan also ensules a radiologically safe working emironment for employees and visitors at TMI.2. The GPU Nuclear Corporate Policy and Procedure Manual contains the GPU Nuclear Corporation Radiation Protection Plan. v IOt
\ 10.4 - 1 UPDATE 2 - AUGUST 1997
1 10.5 ORGANIZATION l 10.5.1 CORPORATE ORGANIZATION 4 he corporate organizational elements responsible for the PDMS phase of TMI 2 are i ' I shown on Figure 10.51. He specific responsibilities are discussed below. Additionally, the PDMS Technical Specifications prescribe specific requirements for staff I ', qualifications, training, and the review and audit of TMI 2 activities. , 1011.1 President - GPU Nuclear The President tiPU Nuclear has the metall responsibibty for the mangement of TMI 2 during PDMS. His responsibility is administered through the management staff, including:
' Director, TMl Division
' Director, Engineering Disision
- ' Director, Nuclear Safety and Technical Services Division
'Direc'or, Financial and Planning Senices Disision
' Director,iluman and Administrative Services Division 1011.2 TMI Division b The TMI Disision has responsibility for operating and maintaining TMI l and V maintaining TMI 2 in the PDMS condition. The Director, TMI Disision is charged with assuring consistent implementation of policies and procedures at TMI 2.
10.5.1.3 Engineering Division De Engineering division prcnides a centralized technical capability to suppon GPU Nuclear facihties and, when requested, will prmide such support to maintain the PDMS condition he disision provides the general mechanical, cisil, elect-ical and instrumentation, engineering mechanics, and chemistry / materials disciphnes. Funher, the division also provides technical support in the areas of nuclear fuel management, computer applications, human engineering, risk analyses, emironmental qualification, and 3 plant analysis, 10.5.1.4 Nuclear Safety and Technical Services Division
%e Nuclear Safety and Technical Servir.es Division prosides several corporate wide functions, including Licensing and Regulatory Affairs. The Nuclear Safety and Technical Services Division also provides the Radiological ard Emironmental Controls, Emergency Preparedness, Occupational Safety, and Training and Education function in support but independent of the GPU Nuclear Operational Disisions 10.5 - 1 UPDATE 2 - AUGUST 1997
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10.5.1.5 Financial and Planning Senices Disision O The Financial and Planning Senices Division provides corporate-wide support in the areas of hinterials hianagement, Espense Analysis, Budget and Rate Case support and Long Range and Strategic Pir.nning. 10.5.1.6 Iluman and Administrative Senices Disision The iluman and Administrative Senices Disision provides corporate wide support in the areas ofIluman Resources including Labor Relations, Information Resource hianagement, Nuclear Security and the GPU Nuclear hiedical Program. 10.5.1.7 Other Functions in addition to the dnisions listed above, the following five smaller independent functions trport to the Office of the President: Nuclear Safety Assessment - Quality Assurance, Audits, and Ombudsman, including the General Office Resiew Board (GORD)
- Communications Corporate Counsel and Secretary including Security senices Continuous improvement System & Industry Senices 10.5.2 ONSITE ORGANIZATION ne onr tee organizational elements responsible for the PDhtS phase of Thil 2 are shown on Figure 10.5 2.
10.5.2. . PDhtS hianager Ac PDhtS hianager has the first level management responsibility for maintaining the Th11-2 PDh!S condition. He PDhtS hianager is directly responsible for the operations and maintenance activities associated with the Thil 2 PDhtS. Reporting to the PDhtS hianager are the subordinate supenisors responsible for execution of the above-listed functions. 10.5 - 2 UPDATE 2 - AUGUST 1997 j
O O O GPU NUCLEAR CORP. i 8;[,'F53J"so;;5a?i";i ORGANIZATION PLAN I I
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