ML20247H394
| ML20247H394 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 03/30/1989 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20247G537 | List:
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| References | |
| NUDOCS 8904040449 | |
| Download: ML20247H394 (28) | |
Text
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i C E S S A R ninnc. m., (Sheet 1 of 1)
( l EFFECTIVE PAGE LISTING CHAPTER 12 Table of Contents j Page Amendment i E ii E 111 l Text PJge Amendment 12.1-1 E 12.1-2 E 12.1-3 E 12.2-1 E )
12.2-2 E 12.2-3 E 12.2-4 E 12.3-1 E 12.3-2 E ,
12.3-3 E l 12.3-4 E 12.3-5 E 12.3-6 E i Tables Amendment 12.2-1 '
12.2-2 12.2-3 12.2-4 12.2-5 12.2-6 E 12.2-7 E 12.2-8 E 12.2-9 E 12.2-10 E 12.2-11 Og40%f[$$0$$kOiPD Amendment E
{ December 30, 1988
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TABLE OF CONTENTS CHAPTER 12 Dection Bubiect Pace No.
12.0 RADIATION PROTECTION 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION 12.1-1 EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS 12.1-1 12.1.2 DESIGN CONSIDERATIONS 12.1-1 12.1.3 OPERATIONAL CONSIDERATIONS 12.1-3 12.2 RADIATION SOURCES 12.2-1 12.2.1 CONTAINED SOURCES 12.2-1 12.2.1.1 Containment 12.2-1 12.2.1.1.1 Reactor Core 12.2-1 12.2.1.1.2 Reactor Coolant System 12.2-1 12.2.1.1.3 Main Steam Supply System 12.2-2 12.2.1.1.4 Spent Fuel Handling and 12.2-2 12.2.1.1.5 Transfer Chemical and Volume 12.2-2 lE Control System (CVCS) 12.2.1.2 Safety Equipment Buildina and 12.2-3 Containment Subschere Structure E 12.2.1.2.1 Shutdown Cooling System 12.2-3 12.2.1.2.2 Component Cooling Water System 12.2-3 12.2.1.3 Fuel Buildina 12.2-4 12.2.1.3.1 Spent Fuel Storage and Transfer 12.2-4 Areas 12.2.1.3.2 Spent Fuel Pool Cooling and 12.2-4 Cleanup System J
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Amendment E i December 30, 1988
CESSAR Enfincamu TABLE OF CONTENTS (Cont'd)
CHAPTER 12 Section Subiect Pace No.
12.2.1.4 Turbine Buildina 12.2-4 12.2.1.5 Auxiliary Buildina 12.2-4 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 FACILITY DESIGN FEATURES 12.3-1 12.3.1.1 Radiation Zone Designation 12.3-1 12.3.1.2 Ecuipment and System Desian 12.3-1 Features for Control of Onsite Exposure O i O
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LIST OF TABLES l
CHAPTER 12 Table Subiect 12.2-1 Maximum Neutron Spectra Outside Reactor Vessel 12.2-2 Maximum Gamma Spectra Outside Reactor Vessel j 12.2-3 Shutdown Gamma Spectra Outside Reactor Vessel 12.2-4 N-16 Activity 12.2-5 Spent Fuel Gamma Source 12.2-6 CVCS Heat Exchanger Soluble Inventories 12.2-7 CVCS Heat Exchanger Crud Plateout Activity, Maximum Values 12.2-8 CVCS Ion Exchanger Inventories O s 12.2-9 CVCS Filter Inventories 12.2-10 CVCS Tank Inventories 12.2-11 Shutdown Cooling System (SCS) Specific Source Strengths l I
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12.0 RADIATION PROTECTION This chapter describes the radiation protection measures incorporated in the station design and in the operating procedures to ensure that internal and external radiation l exposures to station personnel, contractors, and the general population due to station conditions, including anticipated operational occurrences, will be within all applicable limits, and furthermore, will be as low as is reasonably achievable (ALARA). !
12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS The System 80+ Standard Design will meet the intent of Regulatory E ;
Guide 8.8. Detailed policy considerations will be addressed by the site operator.
12.1.2 DESIGN CONSIDERATIONS A Experience from past designs and operating reactors has been
( employed in the establishment of radiation protection design guidelines. A program of data acquisition and retrieval has been employed to establish an equipment and system design bases. The engineering effort is directed toward characterizing the mechanisms of radiation level buildup, evaluating the performance of plant systems in mitigating radioactive buildup, and establishing the role of operating procedures in reducing radiation level buildup. Data and experience gained from this effort is directly applied to all disciplines in the design and I development of equipment employed in the System 80+ design.
Systems and equipment employed in the System 80+ design have been designed with the objective of reducing the need for maintenance within radiation areas. Whenever possible components requiring frequent maintenance are separated for location in low radiation zones or are flanged to facilitate ease of removal to a low radiation zone. Whenever possible materials are selected to )
withstand a 60-year service life thus minimizing the need for l E l replacement and reducing maintenance frequencies. Controls are remotely mounted for location in a low radiation zone. Equipment ;
such as heat exchangers and valves are designed for ease of access during maintenance. Equipment is environmentally qualified to meet their performance requirements under the l environmental and operating conditions in which they will be j
[ required to function. The overall objective in equipment design l
is to insure the occupational radiation exposures are ALARA by ensuring that operators are required to spend a minimum amount of time in a radiation environment. ,
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Experience fror. past designs and inservice inspection programs have resulted in design features being incorporated into the System 80+ design that reduce occupational radiation exposure. ;
The most significant improvement for performing inservice j inspection is the reduction of linear feet of weld in the major components.
I The reduction in weld footage has been accomplished by component )
redesign, use of forged sections versus forged-welded plate !
sections, and increasing the size of certain sections. l Systems and equipment employed in the System 80+ NSSS have been 1' designed with the objective of ensuring that occupational exposure due to decommissioning procedures will be ALARA.
Decommissioning can be facilitated in the design stage through features which will minimize the buildup of in-plant radiation and contamination. It is anticipated that decommissioning will be accomplished through the application of one of several available alternative methods, e.g., mothballing, entombment, ;
immediate or delayed dismantling. The experience gained in the 1 continued application of these methods, and any developing variations, will further minimize occupational radiation exposures.
The System 80+ design incorporates many of the design features recommended in Regulatory Guide 8.8 inadditiontootherspecificlE designs and established guidelines to keep in-plant exposures ALARA. The following design featuren are specifically effective in reducing in-plant exposures during decommissioning. l A. Components containing radioactive material, such as primary j coolant, resin and concentrates are provided with j connections for flushing with water or decontamination !
chemicals.
B. Equipment is designed to minimize crud buildup and I facilitate decontamination.
C. Spaces are provided where appropriate to place shielding for the purpose of reducing neutron activation.
D. Activated corrosion product buildup over the life of the design is minimized in the design stage through appropriate selection of corrosion resistant materials, specification of an appropriate chemistry control program and limitation applied to the cobalt content of materials exposed to the primary coolant.
Established radiation protection guidelines are employed to meet the intent of Regulatory Guide 8.8. These guidelines are Amendment E 12.1-2 December 30, 1988 w___--______ _ _ _ _ _ _ _ _ -
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provided to design engineers in each discipline to ensure occupational radiation exposures are maintained ALARA and are consistent with radiation protection guidance provided in Regulatory Guide 8.8. Radiation protection design reviews are performed, as necessary, based upon these guidelines and the guidance provided in Regulatory Guide 8.8. The general design objective for systems and equipment is to reduce exposure to operating personnel as low as reasonably achievable to meet the intent of Regulatory Guide 8.8 and to operate within the limits of radiation protection in restricted areas given in 10 CFR 20 and 10 CFR 50.
Radiation protection reviews are conducted as necessary throughout the design. Guidelines are provided to engineers working in other disciplines involved in the design. Since all E l disciplines involved in the design are covered by the guidelines, every discipline is involved in the ALARA reviewing. Engineers responsible for radiation control work directly with engineers and designers in other df"qiplines to ensure that all radiation r
protection consideration are taken into account. These engineers advise on the most desirable design option for radiation protection when alternate designs are possible in satisfying the process ' requirements. Such recommendations ir.volve a study of the exposures which are likely to be derived from alternate designs and the recommendation of' the option resulting in the lowest exposure.
E Radiation protection design reviews take place, as necessary, prior to the release of design drawings, system design requirements or component design requirements. Comments from the engineers performing the review ce ransmitted by appropriate documentation to the applicable demie2 engineers for resolution.
Followup reviews are conducted to ensure resolution within the established radiation protection guidelines.
12.1.3 OPERATIONAL CONSIDERATIONS This Section to be provided by site operator. E
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U 1 12.2 RADIATION BOURCES This section discusses and identifies the sources of radiation that form the basis for shield design calculations for the design of personnel protective measures and for dose assessment.
12.2.1 CONTAINED SOURCES The shielding design source terms are based on full operation with 0.25% fuel cladding defects. Sources inpower the lE j primary coolant include fission products released from fuel clad J defects, activation and corrosion products. The sources in_the reactor coolant are discussed in Section 11.1. Throughout most i of the reactor coolant system, activation products, principally nitrogen-16 (N-16), are the primary radiation sources for shielding design during power operation. j The design sources are presented in this section by system. E 12.2.1.1 Containment 12.2.1.1.1 Reactor Core The primary radiation emanating from the reactor core during d normal operation are neutrons and gamma rays. Tables 12.2-1 and 12.2-2 list neutron and gamma multigroup fluxes in the reactor cavity at the side of the reactor vessel; these tables are based i on nuclear parameters discussed in Chapter 4. Table 12.2-3 lists core gamma sources after shutdown for shielding requirements during shutdown and inservice inspection.
12.3.1.1.2 Reactor Coolant System Sources of radiation in the reactor coolant system are fission products reimased from fuel and activation and corrosion products I that are circulated in the reactor coolant. These sources are listed in Section 11.1 and Table 12.2-4 and their bases are discussed in Section 11.1.
Tables 11.1.2-7 and 11.1.2-9 list maximum expected activities due I to crud deposits oa steam generator tubing and primary system l piping.
The activation product nitrogen-16 is predominant activity in the reactor coolant pumps, steam generators, and reactor coolant piping. The N-16 activity in each of the components depends on (
the total transit time to the component. The derivation of N-16 activity is shown in Section 11.1.3.
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12.2.1.1.3 Main Steam Supply System See Section 11.1.8.
12.2.1.1.4 Spent Fuel Handling and Transfer The spent fuel assemblies are the predominant long term source of radiation in the containment after plant shutdown for refueling.
A reactor operating time necessary to establish near-equilibrium fission product buildup for the reactor at rated power is used in j determining the source strengtl.. The initial fuel composition that produced the maximum decay source is used. The spent fuel decay gamma source is given in Table 12.2-5. lg 12.2.1.1.5 Chemical and Volume Control System (CVCS)
The shielding design is based on the maximum expected activity in each component. These sources are listed in Tables 12.2-6 through 12.2-10.
lE A. Heat Exchangers (Table 12.2-6 and 12.2-7)
Activities are provided on a normalized volumetric basis. E B. Ion Exchangers (Table 12.2-8)
- 1. Purification Ion Exchanger Total curie inventory is based on a resin buildup of 1.2 effective years. This ion exchanger is used for lithium removal and normal purification of RCS letdown.
When it is used for lithium removal it is on line an average of 58 days prior to placing it in service as a purification ion exchanger for 292 days.
All nuclides except Xe, Kr, Rb and Cs have a decontamination factor (DF) of 10 and efficiency of 90%, Xe and Cs have a DF of 1.0 and efficiency of 0%,
Rb and Cs have a DF of 2.0, and efficiency of 50%.
- 2. Preholdup Ion Exchanger l i
Total curie inventory is based on resin buildup of 1.0 j effective year (292 days). All nuclides except Xe, Kr, Rb, Cs, have a decontamination factor (DF) of 10 and an officiency of 90%, Rb and Cs have a DF of 100 and efficiency of 99%, Xe and Kr have a decontamination factor of 1 and an efficiency of 0%.
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Sources processed Igy the prehold-up Ion-exchanger include: 1.1 x 10 gallons of letdown previously processed through the purification Ion exchanger and purification filter, 200 gpd from the Reactor Drain Tank (RDT) and 50 gpd from the Equipment Drain Tank (EDT).
- 3. Boric Acid Condensate Ion Exchanger Total curie inventory is based on resin buildup of 1.0 effective year (292 days). Anion decontamination factors of 10, and efficiency of 90% were used. All other ions have a decontamination factor of,1 and ag efficiency of 0%. Total liquid processed is 1.83 x 10 gallons. ,
C. Filters (Table 12.2-9) l Total curie inventories on all CVCS filters are based on I crud buildup of 292 days. All CVCS filters remove crud with a decontamination factor of 10 and an efficiency of 90%.
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q D. Tanks (Table 12.2-10)
U Activities are provided on a normalized volumetric basis.
12.2.1.2 Bafety Eculptrent Buildinc and Containment subschere Structure
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12.2.1.2.1 Shutdown Cooling System The pumps, heat exchangers, and associated piping of the shutdown cooling system (SCS) are potential carriers of radioactive materials. For plant shutdown, the SDCS pumps and heat exchanger i sources of radioactivity result from the radioactive isotopes '
carried in the reactor coolant, discussed in Section 12.2.1.1.2, after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown and dilution.
Table 12.2-11 Provides a listing of the maximum specific source strengths (MeV/gm-sec) in the SCS.
12.2.1.2.2 Component Cooling Water System (LATER) E
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12.2.1.3 Euel Buildinq l 12.2.1.3.1 Spent Fuel Storage and Transfer Areas l
The predominant radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. Spent Fuel assembly sources are discussed in Section 12.2.1.1.4. The spent fuel decay gamma source to be used in shielding design is given in Table 12.2-5. 1 12.2.1.3.2 Spent Fuel Pool Cooling and Cleanup System 1
Activity levels in the Spent Fuel Pool Cooling and Cleanup !
Systems are determined based on the activities present in the Spent Fuel Pool. Spent Fuel Pool Activities are discussed in Section 11.1.7.
12.2.1.4 Turbine Buildina (LATER) 12.2.1.5 Auxiliary Buildinc (LATER) l l
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rr TABLE 12.2-1 MAXIMUM NEUTRON SPECTRA OUTSIDE REACTCR VESSEL
- Average NeutronSpecgra Neutron Enercy (Mev) (neutrons /cm -s) 13.60 5.90 x 10+6 11.10 1.86 x 10+
9.10 3.79 x 10+
7.27 6.87 x 10+
5.66 1.08 x 10+8 4.51 9.00 x 10+
3.53 1.58 x 10+8 2.73 2.01 x 10+
2.40 6.69 x 10+
2.09 3.86 x 10+
1.47 1.48 x 10+
~1 8.30 x 10 5.20 x 10+9 3.30 x 10 1.47 x 10+ 0 5.70 x 10~ 1.05 x 10+10 1.96 x 10~ 3.24 x 10+
~4 3.42 x 10 2.96 x 10+
-5 6.50 x 10 2.01 x 10+
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1.98 x 10 1.27 x 10+'
-6 6.90 x 10 1.44 10+'
-6 2.09 x 10 1.09 x 10+
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7.60 x 10 9.67 x 10+8
-8 2.50 x 10 (thermal) 6.22 x 10+
(a) At core midplane, one half foot from vessel surface
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O TABLE 12.2-2 I
MAXIMUM GAMMA SPECTRA OUTSIDE REACTOR VESSEL *I Average Gamma Spgctra Gamma Enerav (Mov) Gamma /cm -s) 9.00 2.17 x 10+8 7 25 1.27 x 10+9 5.75 8.42 x 10+
4.50 7.21 x 10+
3.50 9.90 x 10+8 2.75 6.60 x 10+
2.25 1.11 x 10+
- 1. 8 :' 8e19 x 10+8
\j 1.50 8.43 x 10+8 1.16 1.07 x 10+
0.90 8.06 x 10+8 0.70 1.03 x 10+9 i 0.50 2.46 x 10+9 0.35 1.56 x 10+9 0.25 2.60 x 10+
0.15 4.14 x 10+
0.075 1.05 x 10+9 0.025 4.99 x 10+
(a) At core midplane, one half foot from vessel surface.
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U i TABLE 12.2-3 SHUTDOWN GAMMA SPECTRA OUTSIDE REACTOR VESSEL ("'
l Average Decay Gag a MaterialActjvation i Gamma Enerav (Mev) (Gamma /cm -s) (Gamma /cm -s) 2.75 1.18 x 10+4 l 2.25 3.72 x 10+4 1.83 8.24 x 10+4 ]
1.50 1.37 x 10+ q 1.16 1.92 x 10+5 7.39 x 10+
0.90 1.48 x 10+ 8.50 x 10+4 0.70 1.84 x 10+ 1.00 x 10+5 0.50 2.44 x 10+ 1.48 x 10+5 0.35 2.08 x 10+ 1.29 x 10+5 .
0.25 3.81 x 10+5 2.63 x 10+5 0.15 5.86 x 10+ 4.06 x 10+5 0.075 1.38 x 10+ 9.80 x 10+4 1
0.025 6.89 x 10+ 5.00 x 10+2 e
NOTES: a. At core midplane, one half-foot from vessel i surface. }
- b. At 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown. .
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TABLE 12.2-4
-l N-16 ACTIVITY Activity Location (disintegrations /cm -s)
Vessel Outlet Nozzle 5.76 x 10+6 Vessel Outlet Line (midpoint) 5.69 x 10+6 Steam Generator (midpoint) 4.61 x 10+0 Pump (midpoint) 3.71 x 10+
Vessel Inlet Line (midpoint) 3.49 x 10+6 4
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,V TABLE 12.2-5 !
SPENT FUEL GAMMA SOURCE l Gamma Source (Mov/ watt-s)
Time After Shutdown l Mean Energy (Mev) 50 hr 200 hr 500 hr 1000 hr -
0.30 1.3 x 10+ 6.5 *' 'a+8 3.1 x 10+8 1.7 :t 10+8 0.63 8.6 x 10+ 5.8 x 10+ 4.4 x 10+9 3.4 x 10+9 1.10 1.2 x 10+ 5.9 x 10+8 3.0 x 10+8 1.6 x 10+8 .
j 1.55 2.9 x 10+ 2.1 x 10+ 1.0 x 10+ 3.6 x 10+8 1.99 2.6 x 10+ 1.7 x 10+ 9.7 x 10+ 5.0 x 10+ ;
2.38 1.4 x 10+8 1.0 x 10+ 5.3 x 10+ 1.9 x 10+
2.75 2.7 x 10+5 2.6 x 10+5 2.6 x 10+5 2.6 x 10+5 ;
3.25 1.0 x 10+4 9.8 x 10+ 9.5 x 10+ 9.2 x 10*3 i
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TABLE 12.2-6 CVCS HEAT EXCHANGER SOLUBLE INVENTORIES Maximum Values (lci/cc)
l Amendment E December 30, 1988
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, Energy Group Normalized Crud Normalized Crud Normalized Crud l MeV Plateout Activity Plateout Activity Plateout Activity 0.25 4.2(+6)* 5.3(+7) 5.1(+6)
'I 0.50 3.9(+7) 4.8(+8) 4.9(+7) l 0.75 1.4(+8) 1.7(+9) 1.6(+8) 1.00 3.7(+5) 4.6(+6) 4.3(+5) E 1.38 7.3(+6) 8.7(+7) 8.3(+6) 2.00 8.2(+5) 1.0(+7) 1.0(+6)
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4.00 - - -
6.00 - - -
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- Number in parentheses denotes powers of ten. j l
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TABLE 12.2-8 CVCS ION EXCHANGER INVENTORIES Maximum values (curies)
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(LATER)
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i TABLE 12.2-10 CVCS TANK INVENTOFJIES Maximum Values (lci/cc)
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O TABLE 12.2-11 SHUT 00WN COOLING SYSTEM (SCS) SPECIFIC SOURCE STRENGTHS Maximum Values (MeV/ gram-sec)
Decay Energy (Mev)
Time (hr) 92 0.63 1 10_
1.55 1.99 2.38 2.75 3.25 3.70 1 3.3(+4)* 2.4(+5) 6.7(+4) 1.9(+4) 4.7(+3) 3.4(+2) 1.6(+2) 9.9(+1) 1.2(+2) 10 2.5(+4) 1.2(+5) 2.9(+4) 7.5(+3) 2.2(+3) 2.9(+1) 6.7(-1) 6.2(-1) 8.9(-3) 100 1.8(+4) 4.4(+4) 6.3(+3) 2.4(+3) 3.5(+2) 2.2(+1) 2.7(-2) 8.7(-3) -
- Number in parentheses denotes powers of ten.
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C E S S A R Minne. m u 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES 12.3.1.1 Radiation Zone Designation (LATER) lE 12.3.1.2 Eauipment and System Desian Features for Control of Onsite Exposure Following are some of the specific design features which are used to assure that occupational radiation exposure due to operations and maintenance of the System 80+ design will be ALARA.
Domineralizers are addressed under the heading of ion exchangers.
E A. Pumps
- 1. Most pumps and associated piping are flanged to 3
facilitate case of removal to a low radiation area for maintenance or repair. Pump internals can be removed to a low radiation area for maintenance.
- 2. All pump casings are provided with drain connections to facilitate decontamination.
B. Ion Exchangers
- 1. Ion exchangers are designed for complete drainage.
- 2. Spent resin removal is designed to be done remotely by hydraulically flushing the resin from the vessel to the Solid Waste Management System.
- 3. The fresh resin inlet is designed to e'xtend into a low radiation area above the shielded compartment housing the ion exchanger.
- 4. Ion exchangers are designed with a minimum of crevices in order not to accumulate radioactive crud.
C. Liquid Filters
- 1. Filter housings are provided with vent connections and designed for complete drainage.
- 2. Filter housings are designed with a minimum of crevices
-s in order not to accumulate radioactive crud.
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- 3. Filter housings and cartridges are designed to permit remote removal of the filter elements.
D. Tanks
- 1. Tanks are designed to be isolated for maintenance and provisions will be made for complete drainage.
- 2. Tanks are provided with at least one of the following means of cleaning the tank internals for decontamination purposes:
- a. Ample space is provided to facilitate cleaning from the tank manway.
- b. Internal spray nozzles are provided on potentially highly contaminated tanks for internal decontamination.
- c. The ability to back flush or drain inlet screens hydraulically will be provided (on tanks or vessels with these screens) to facilitate decontamination.
- 3. AlltanksexcepttheIRWSTareventedtoeitherthegaslE collection header or the gas surge header which will facilitate removal of potentially radioactive gases during maintenance.
- 4. Non-pressurized tanks are provided with overflows, routed to a floor drain or other suitable collection point to avoid radioactive fluids spilling to the floor or ground.
- 5. Tanks are designed with a minimum of crevices in order not to accumulate radioactive crud.
E. Package Units ,
- 1. Each package unit is skid mounted with all motors and pumps located on the periphery of the skid for free access and for quick removal to a low radiation area for maintenance or repair.
- 2. Space is provided on the skid for placement of portable shielding.
- 3. All package components are provided with provisions for flushing, drawing and chemical cleaning.
Amendment E 12.3-2 December 30, 1988
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t j 4. Heat exchangers are readily accessible for maintenance.
- 5. Controls are remotely mounted and the package will be able to be remotely monitored. As many control elements as possible are mounted remotely from the components.
- 6. Components are designed with a minimum of crevices in order not to accumulate radioactive crud.
- 7. Radioactive gas is collected and sent to the Gaseous Waste Management System.
F. Valves
- 1. Radiation resistant seals, gaskets and elastomers are employed when practical to extend the design life and reduce maintenance requirements.
- 2. Power operated valves in the primary system are provided with double packing, a lantern gland and stem leakoffs to collect leakage and to direct radioactive fluid away from access areas. All valve packing glands (o
k_)
- have provisions to adjust packing compression to reduce leakage.
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- 3. Valves are designed so that they may be repacked without removing the yoke or topworks. j
- 4. Remotely operated valves are utilized where practical and necessary. l
- 5. Valve wetted parts are made of austenitic stainless j steel or other corrosion resistant material.
- 6. Low leakage valves with backseats are employed wherever possible. Packless diaphragm valves are employed in highly contaminated systems.
G. Piping Design recommendations and information fcr keeping in-plant personnel exposures ALARA are provided. The following information and recommendations are provided.
- 1. Interface criteria and radiation source terms are provided to insure that field-run piping carrying radioactive material is either run in shielded pipe (n')
Q ,1 chases or within shielded cubicles.
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- 2. Whenever possible, pipe runs should be sloped to O'i prevent accumulation and to assist in the removal of l
radioactive corrosion deposits.
- 3. The number of elbows, tees, deadlegs, etc., should be minimized to reduce corrosion deposits. Where elbows are required, they should be a large radius type (or minimization of deposits.
H. Heat Exchangers
- 1. Heat exchangers are designed to accommodate the requirements of inservice inspection and for ease of access during maintenance to reduce the time opera'cors are required to spend in a radiation environment.
- 2. Materials are selected to minimize the need for replacement and to reduce maintenance frequencies corrosion resistant materials are employed.
I. Material Selection Material is selected as described below to reduce exposures ,
by reducing maintenance frequencies and by providing less circulating crud as a source of exposure where maintenance l will be necessary.
- 1. Materials of construction for components containing radioactive materials will be selected with consideration of potential release of activated corrosion products from these materials.
- 2. Radiation exposure levels were considered selecting materials for 60-year service.
whenlE
- 3. Material selection was made with consideration given to other fluid conditions which could lead to premature material failure.
- 4. Other materials considerations are discussed in Section !
5.2.3.
J. Reactor Vessel Head Vent A vent nozzle and line is provided on the reactor vessel head. Utilization of this design feature will allow a reduction of exposure during the head removal process by minimizing the gases discharged directly to the containment atmosphere while the head is being removed.
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K. Safety Depressurization System (SDS) E The 3DS discharge is sparged into the IRWST where potential radioactive discharges are captured for subsequent processing as necessary.
L. Reactor Coolant System Leakage Control Exposures from airborne radionuclides to personnel entering the containment will be minimized by controlling the amount of reactor coolant leakage to the containment atmosphere.
Examples of such controlled leakage are listed below:
- 1. Primary pressurf zer safety and safety depressurization E
system valve leakage is directed to the IRWST.
- 2. Valves larger than 2" in diameter are provided with a double-packed stem with an intermediate lantern ring with a leak-off connection to the. Reactor Drain Tank.
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- 3. Instrumentation is provided to detect abnormal reactor- ,
coolant pump seal leakage. The reactor coolant pumps i are equipped with two stages of seals plus a vapor or backup seal as described in Section 5.5. The vapor or backup . seal will prevent leakage to the containment atmosphere and allow sufficient pressure to be maintained to direct the controlled seal leakage to the Volume Control and Reactor Drain Tanks. The vapor seal is designed to withstand full Reactor Coolant System pressure in the event of failure of any or'all of the two primary seals.
l M. Refueling Equipment
- 1. All spent fuel transfer and storage operations are designed to be conducted underwater to insure adequate shielding and to limit the maximum continuous radiation levels in working areas.
- 2. The equipment is designed to prevent the fuel from being lifted above the minimum safe water depth, thereby limiting personnel exposures and avoiding fuel damage.
- 3. The equipment ' design limits the possibility of inadvertent fuel drops which could cause fuel damage and personnel exposures.
- 4. The refueling equipment design will facilitate the transfer of new and spent fuel at the same time to i l
l Amendment E 12.3-5 December 30, 1988
i CESSAR Ennricari:n O
reduce overall fuel handling time; and, therefore, personnel exposures during refueling.
- 5. Underwater cameras are used to facilitate safe handling and visual control, thus minimizing errors and potential exposures.
- 6. Portable hydraulic cutters are provided to cut expended Control Element Assemblies and in-core instrumentation i leads. The cutters allow underwater handling of these l items.
- 7. Equipment is provided to allow for the underwater i determination of leaking fuel elements. j N. In-service Inspection Equipment Inspection of the reactor coolant pressure boundary can be done with remote equipment to keep personnel exposures to a minimum. A detailed discussion of the In-service Inspection l Program is provided in the site-specific SAR. E O. Remote Instrumentation All systems containing radioactive fluids are designed to be controlled remotely to the maximum extent practical. This ?
will allow personnel radiation exposures from the normal operation of these systems to be minimized.
P. In-service Inspection of Reactor Vessel Nozzle Welds The design of welds joining the reactor vessel nozzle to reactor coolant pipe permits in-service inspection to be accomplished from the I.D. of the reactor vessel. Automated equipment normally used for reactor vessel pressure boundary inspections can be utilized in this area.
In the event that in-service inspection of this area is performed from the outside, insulation for the reactor vessel and reactor coolant piping utilizes removable sections for access. These removable sections are lightweight and are held in place mainly by quick actuation type buckle fasteners. After the necessary panels are removed, remote equipment can be utilized to perform the l required inspections.
O Amendment E 12.3-6 December 30, 1988