ML20067B228

From kanterella
Jump to navigation Jump to search
Forwards Revised Response to Chemical Engineering Branch 820212 Request for Addl Info Re NUREG-0737,Item II.B.3, Post-Accident Sampling Sys. Info Will Be Provided in OL Application,Amend 48
ML20067B228
Person / Time
Site: Seabrook  NextEra Energy icon.png
Issue date: 12/01/1982
From: Devincentis J
PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO.
To: Knighton G
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM SBN-395, NUDOCS 8212060094
Download: ML20067B228 (22)
v • d • e


Text

{{#Wiki_filter:o a smaom sms IPUBLIC SERVICE .;.:.., orme: Companyof New Hampshire 1671 Worcester Road Framinoham, Massachusetts 01701 (617) - 872 - 8100 December 1, 1982 SBN-395 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing

References:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) USNRC Letter, dated February 12, 1982, " Request for Additional Information," F. J. Miraglia to W. C. Tallman (c) PSNH Letter, dated March 12, 1982, " Response to 281 Series RAIs; (Chemical Engineering Branch)," J. DeVincentis to F. J. Miraglia (d) PSNH Letter, dated February 12, 1982, " Implementation of TMI Action Plan Requirements of NUREG-0737," J. DeVincentis to F. J. Miraglia

Subject:

Revised Response to RAI 281.6; Post-Accident Sampling; (Chemical Engineering Branch)

Dear Sir:

We have enclosed a revised response to the subject Request for Additional Information (RAI) which was forwarded in Reference (b). The enclosed response provides detailed information regarding the Post-Accident Sampling System required by NUREG-0737, Item II.B.3. The original response to RAI 281.6 was submitted in Reference (c) and essentially duplicated our position on NUREG-0737, Item II.B.3 which was submitted in Reference (d). The enclosed information will be provided in OL Application Amendment 48. Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY JmA J. DeVincentis Project Manager ALL/fsf cc: Atomic Safety and Licensing Board Service List 8212060094 821201 PDR ADOCK 05000443 A PDR

~ 4 ~~ I. RAI 281.5 (9.3.2. NUREG-0737. II.B.3) -{ Provide information that satisfies the attached proposed license conditions for post-accident sampling.

RESPONSE

TheshieldingandoperationofthereactorcoolantandcontainmentEtmosphere sampling systems has been designed to provide the capability of personnel to promptly obtain (less than 1 hour) a sample under accident conditions without incurring a radiation exposure in excess of the limits delineated for this requirement. A post-accident sampling panel has been designed to NUREG-0737. FSAR Section 9.3.2 is being revised in Amendment 48 to address the concerns of NUREG-0737, Section II.B.3. Procedures to obtain post-accident samples and the radiological and chemical analyses will be developed three months prior to fuel load. l

[ 1. SB 1 & 2 FSAR e 9.3.2 Process Sampling System s-9.3.2.1 Design Bases " The sample system pr 'as representative liquid and "- -m..ples for Nk chemical and radio-chemic' ',boratory anal c' the water chemistry of the reactor coolant system,. rators, secondary steam and water systems and other auxil stems. le 9.3-1 lists the sample sources, analysi ormed, type of samp urpose of the analysis, and the appli - n of the analyses to the contro he plant. Each unit dk# has own sample system, and there are no intercon

[lons.

The seismic and quality group classifications of sample lines and components conform to the classification of the system to which each sampling line and component is connected. Where appropriate, classification to a lower seismic and/or quality group is justified on the basis that adequate isolation valving or flow restriction is provided. Sample lines penetrating the containment are provided with engineered safety features actuation system (ESFAS) isolation valves. Containment isolation and valve descriptions are discussed in Subsection 6.2.4. Heat exchangers, vessels, piping, fitting and valves are designed, procured and installed in accordance with ASME Boiler and Pressure Vessel Code, MObPN Sections Ill, VIII, and ANSI B31.1. Safety class descriptions of tbn various components are indicated on the system P&IDatsga@GT-s@ 9&bj-56M s-for the reactor coolant, steam generator and other auxiliary systems [, j sampling subsystems. The components of the secondary steam and water sampling subsystem are non-nuclear safety class (NNS). Flow in the reactor coolant and steam generator blowdown sample lines is turbulent during purging or sampling, to ensure that any particles remain suspended. The reactor coolant sample lines are provided with a purge path to the chemical and volume control (CVCS) or boron recovery (BRS) systems. Purging of *he lines prior to collecting the sample is required. Gaseous flow from the chemical and volume control tank (CVCT) and pressurizer relief tank (PRT) sample lines are directed through sample vessels and discharged to the equipment vent system. The sampic lines from the residual heat removal (RHR) and demineralized water (DW) systems are directed to the sample sink for " grab" sampics, and are purged by allowing the fluid to drain to the sink prior to taking the sample. The sampling system is designed to direct the reactor coolant sample purge fluids to the chemical and volume control tank or the primary drain tank, in the chemical and volume control tank is not available. Purge flows and sample overflows from the steam generator blowdown and other auxiliary [NSNgT systems sampling subsystems are normally directed to the radioactive liquid waste system via the floor and equipment drains. The steam generator blowdown sampling is continuous, and provides radiation, conductivity and sodium ion monitoring of each blowdown line. See Subsections 9.3.2.5 and 10.4.8 for discussion of steam generator blowdown s, high radiation signals. 9.3-6 ../

) SB 1 6 2 FSAR The secondary steam and water sampling is, in general, continuous, and N" serves as an aid to preventing corrosion, inhibiting formation of scale and minimizing deposits on heat transfer surfaces and turbine blades in the seconda g system. r SetE: r14701%$ W)-Sas;-5 b ;* Sc. M on F N.3 3 for those samples routed to a central sampling point. The locations 01 the sample points are shown on the appropriate system piping and instrumentation diagrams for the system to be sampled. Sample points for the secondary steam and water sampling subsystem are also shown diagramatically on Figure 9.3-6. 9.3.2.2

System Description

The sample subsystems for the reactor coolant, steam generators and other auxiliary systems provide representative gas and liquid samples for laboratory analysis, in accordance with Regulatory Guide 1.21, positions C.6 and C.7. Typical information obtained includes: reactor coolant boron, sodium ion and halogen concentrations, fission product radioactivity level, hydrogen, oxygen, and fission gas content, corrosion product concentration, and chemical additive concentration. The sampling subsystem for secondary steam and water systems provides representative samples for measuring specific and cation conductivity, concentrations of ; odium ion, dissolved oxygen, silica and hydrazine, and pil. 2 N., a. Subsystem Description The system is divided into four subsystems: reactor coolant campling, steam generator blowdown sampling, auxiliary syste sa, ling, and secondary steam and water sampling. 1. Re,' tor Coolant Sampling Subsystem The reac r coolant is sampled at four loc lons in the ggggg reactor coo ne system. A steam sample s taken from the pressurizer st m space. The remai - g three liquid samples are taken from th ressurizer li d space and reactor coolant Loops 1 and l Each of the four reactor lant sample lines inside containment are provi with utomatic ESFAS isolation valves. The three quid sample ines are joined together in a common head before Icaving ie containment. This common line ir provided with an auto. tic exterior containmen solation valve, as is the mple line used for sam ng the pressurizer steam space.

  • ic sampling line nuections to the reactor coolant loops re sized t

cet the small leak analysis of Subsection 1.' 2. l l Each reactor coolant loop sample line has a delay coil d manual flow valve to limit the flow to less than 373 lb/hr s, f The delay coils are provided in order to permit decay of short-lived radionuclides. The delay coils are designed CohI.. 9.3-7 I

3 SB 1 & 2 FSAR o provide a minimun 45-second delay within containme 45-second delay time allows the short-lived

otopes, Th prima v N-16 (7.4 second half-life) to dec.

cufficiently to minim 1m the hazard to personnel. Del-coils are not required for pressurizer samples, ause the pressurizer is a relatively s "nant volume and e ef fective half-life lnl$dESL9f is great enough to d y the N-1 Additional shielding is pr ed, where necessary, to reduce described in Section 12.3. potential personnel exp re, I Each pressurizer mple line has a illary tube to limit CONT S the flow to lese han 373 lb/hr (0.75 g, sample rate) gg we with all valt > in the line fully open. e capillary tube also _rmits a small flow of 50 lb/hr be purged or intermittently to the volume con 1 tank, constan thus moving non-condensible gases, ample heat exchangers are provided in the pressurizer steam space sample line and the common reactor coolant linoid e,mnlo line to enol the sample to 950F,g riow is controlled in either line by adjusting the pressure reduction or the block valve in the respective line, and is then routed to the sample sink for grab samples. For operator safety, these lines are double-valved at the sink. Over pressurization of these heat exchangers.Is controlled administratively by assuring that either the valve upstream This prevents fluid from being or downstream is left open. isolated in these lines. Both of these boundary valves may be closed if the process line is drained. A ten milliliter sample vessel is provided for obtaining volume control tank and pressurizer relief tank gas samples. The vessel is made of austenitic stainless steel and is equipped with quick-disconnect couplings with poppet-type check valves and integral isolation valves at the sample sink. l l 2. Steam Generator Blowdown Sampling Subsystem flow path for each sample is typical, therefore, only The one path is discussed. The steam generator blowdown (SGBD) isolation valves is sampled downstream of the containment and upstream of the blowdown system pressure-reducing See Subsection 10.4.8 for discussion of blowdown valves. isolation. Each sample heat exchanger pair reduces the sample temperature to 109 F at 373 lb/hr flow rate. The flow is then routed through a flow regulating valve which reduces pressure to 50 m psig. All instrumentation in located downstream of this valve. The radiatinn instrumentation v be bypassed. provides cont inuous monit or ing and cannot 9.3-9 1

f [ SB 1 & 2 FSAR The strumentation is protected against overpressurizatio I ef valve venting to the floor drain on the com-i by a re his ne downstream of the radiation monitors. discharge flash could occur m high pressure in the steam genera downstream block valve. T lowdown tank or closure a d to one of the two e aple sinks sampling lines are ru for grab samples. g ed basins housed The sample sink is actually two e in a hooded enclosure equip sith an ex st fan to provide ein air flow in e hood at all time The sinks are stainless stee Ath raised edge to contain Tashed liquid. They in to the waste disposal system. basin is steam generator blowdown samples and the o r for remaining grab sample lines. A wall is employed 1 I isolate the two basins. g the sink are provided with Sample line discharges at quick-disconnect adaptors such that a sample vessel may be used to collect samples. 3. Auxiliary Systems Sampling Subsystem This subsystem consists of sampling lines which run from auxiliary system to the sample sink or local the plant component sample stations. s, The sampling lines f rom the chemical and volume control removal systems to the sink are provided and residual heat the sink. These lines also have with double valving at bypass connections to the chemical and volume control tank and primary drain tank through the reactor coolant sampling The following auxiliary system sample line discharge. taps are provided at the sample sink: Type Sample System Origin Grab Chemical and Volume Letdown Heat Exchanger Control System Grab Chemical and Volume Cation and Mixed Bed Demineralizer Control System Grab Chemical and Volume Letdown Degasifier Control System Trim Cooler Grab Chemical and Volume Thermal Regeneration Control System Demineralizer Grab Residual Heat Removal hesidual lieat Renoval Heat Exchanger 9.3-9

l SB 1 & 2 FSAR L to minimize interference of ammonia and hydrazine (added to the secondary system for corrosion control) with the sampling process. Deviations of measured quantities from specified values are alarmed at a local panel in the turbine building. In the event of leakage of reactor coolant into the secondary system, radioactivity may be present in the SSW samples. A high radiation alarm from the steam generator blowdown NgY sampling subsystem radiation monitors is available at the i local panel to alert the operator to manually stop all. sample streams. sr-The required analysis and frequencies are given in Chapter 16. b. Equipment Location and Description The system equipment is situated primarily at three locations: The sample coolers, sample sink and flow control valve a erature and pressure instruments f NYIO oc-reactnr coolant, aenerator blowdown ther auxiliary systems sampling a te.c located on the j grade level of the. primary au 'ng. d' l v 2. The reactor oop and pressurizer sample he ca ubes and delay coils are located in the react ontainment annulus area. 3. The secondary steam a,nd water sampling subsystem equipment and components are located in the turbine building. O The equipment design parameters for the reactor coolant, steam generator blowdown and other auxiliary systems sampling subsystems are summarized in Table 9.3-2. 9.3.2.3 Safety Evaluation Tne sample system has no emergency or safety function, nor is its performance required to prevent an emergency condition. No components are shared between units. Isolation of those samples originating within the containment is accomplished by: a. Manual valves near the sample point. ElastricaHy operarted. solenoid b. ' calves which automatically close on a containment g_ isolation signal, or can be closed by remote manua l swi t ches on the main control board. c. Manual valves at the sample sink. 9.1-11

SB 1 & 2 DEL ETE "a from hi b t. c ..s u oy a s_ j delay coil to allow si r' ,nd supplemental shiciding ..m '~ where rm 9.3.2.4 Tests and Inspections Prior to initial criticality, the system will be operationally tested and samples drawn including appropriate purging from each sample point. 9.3.2.5 Instrumentation and Control Local instrumentation for monitoring pressures, temperatures, and flows are provided in the sample sink area and at the sample panel in the turbine building to provide for safe manual operation and to verify sample flows.

  • h>W

,EsWY stean generator blowdown sample lines are continuously monitored for high radiation. If a high radioactive level is detected, alarms are triggereo on the local panel, the main panel _(turbine building) and in the control toom. In addition, the blowdown tank discharge line is automatically isolated. Administrative overrides allow blowdown flow to continue after isolation of the system, for evaporation processing and/or sampling on an individual line basis. See Subsection 10.4.8 for additional information on blowdown system operation. The blowdown samples are physically separate from the other samples at the sampic sink to prevent any radioactive ,, cross-contamination. The steam generator blowdown portion of the system also contains inline specific conductivity and sodium ion elements to monitor for condenser lea ka ge. Each steam generator blowdown line is monitored separately. A high conductivity sample is alarmed at the sample room control panel and at the main control board. Sample system lines penetrating the containment have appropriate containment isolation valves which automatically close on a "T" (Phase A containment isolation) signal and also fail closed. These valves, being safety-related, are also controlled from the main control board. See Subsections 6.2.4 and 7.3 f or additional information on containment isolation. Globe type valves are used for interior containment isolation. The interior and exterior isolation valves are equipped with operators for automatic or remote operation. The valves are actuated by a containment isolation signal or manually from the control room. See Section 6.2.4 for the types of operators used and discussion of containment isolation l signal. h isaies%, locatd in W sangle rum oF%. yrimary auxdan p digital p i e.,s ure _ is n. sed to Mg g accurately measure reactor coolant system h pressure for control J ca1ibration. This instrument is inoperable during containment isolation. .Wasured quant i t ies from the secondary steam and water sampling subsystem are indicated and/or recorded at local panels in the turbine building. 9.3-12

SB 1 & 2 FSAR TABLE 9.3-2 N-(Sheet 2 of 3) 680 Desigr. Temperature, F 2485 Design Pressure, psig Pressure Drop Pressurizer (1 of 2), psi 13.1 Pressurizer (2 of 2), psi 15.3 13.5 Steam Generator, psi 2. Capillary Tubes Pressurizer Liquid Sample Line 150 Tube Length, f t. O.25 Tube 0.D., in. 0.065 Tube Wall Thickness, in Pressurizer Steam Space Sample Line 100 Tube Length, ft. 0.25 Tube I.D., in. Tube Wall Thickness, in. 0.065 Austenitic Stainless Steel Material 2 Safety Class I Seismic Category ( f3. Delay Col l S ar tube length Len gth inside containment is 170-200 feet. 0.375 0.D., in 0.065 Wall Thickness, in. Austenitic Stainless Steel I Material " III Design Code 2 Safet-os I Se smic Catego ry Valves, Piping and Tubing l Reactor Coolant Sample Lines 2485 Des ign Pressure, psig Design Temperature, F 680 0.375 0.D., in. Wall Thickness, in. 0.065 Steam Generator Blowdown Sample Lines 1285 Des ign Pressure, psig 600 Design Temperature, OF 0.375 l 0.D., in. g,, 0.065 Wall Thickness, in.

SB 1 6 2 FSAR I TABLE 9.3-2 s,, (Sheet 3 of 3) Chemical and Volume Control Demineralizers Sample Lines Design Pressure, psig 200 Design Temperature, OF 300 0.D., in. 0.375 Wall Thickness, in. 0.065 Volume Control Tank Gas Space Design Pressure, psig 75 Design Temperature, oF 250 0.D., in. 0.375 Wall Thickness, in. 0.065 Residual Heat Re' oval Sample Line Design Pressure, psig 600 Design Temperature, OF 400 0.D., in. 0.375 Wall Thickness, in. 0.065 Material Austenitic Stainless Steel Design Codes ANSI B31.1.0, except inside containnent and containment isolation which are designed to ASME III s_, Safety Class 2 and NNS 4 Seismic Category Non-Seismic except inside containment and containment isolation which i are seismic Category I h Sample Vessel Volune, ml 10 Design Pressure, psig 200 Design Tempe rature, OF 250 Mate rial Austenitic Stainless Steel Design Code ASME VIII Div. I j Safety Class NNS Seismic Category Non-Seismic t i l ss l I L

INSERT _l_ @ g 3_ (,) The sample system provides representative liquid and gas samples for chemical and radio-chemical laboratory analysis of the water chemistry of the reactor coolant system, steam generator blowdown, secondary steam and water systems and other auxiliary systems under normal operating conditions. The sample system also provides the capability to obtain gas samples of the containment atmosphere and liquid samples from the reactor pressure vessel, containment recirculation sump, pressurizer relief tank and ECCS pump room sumps under post-accident operation. - Table 9.3-1 lists the sample sources, analyses performed, type of sample, purpose of the analyses, and the application of the analyses to the control of the plant. Each unit has its own sample system and there are no interconnections, i INSERT 2 @.9.~5 The post-accident sample system is designed such that the flow through the sample lines is turbulent in order to reduce plate out. In addition, these lines can be flushed w1th demineralized water after a sample is taken. The post-accident sample system also allows for collection of an adequate volume of fluid which re-sults from purging the sample lines in order'to obtain representative samples. Provisions exist which enable samples to be returned to the containment, even if pressurized. (j? 9,3_7).d l INSERT 3 l a. Subsystem Description i l The system is divided into five subsystems: reactor coolant sampling, l steam generator blowdown sampling; auxiliary system sampling, secondary steam and water sampling and post accident sampling. 1. Reactor Coolant Sampling Subsystem i Reactor coolant is sampled at four locations in the reactor coolant i system. Liquid samples are taken from the pressurizer liquid space and reactor coolant loops 1 and 3. The remaining sample is a steam sample and is taken from the pressurizer steam space. Provisions exist to enable sampling of reactor coolant loops 1 and 3 under post-i accident conditions. 1 l Each of the four reactor coolant system sample lines inside contain-ment are equipped with automatic ESFAS isolation valves. The pressur-izer steam and liquid sample lines are joined together in a common header before leaving the containment. This common line is provided with an automatic exterior containment isolation valve. The sample lines from reactor coolant loops 1 and 3 are also provided with auto-i matic exterior containment isolation valves. The sample line connec-tions to the reactor coolant loops are sized to meet the cmall leak j analysis of Subsection 15.6.2. =---*e-yryr-=--s + - --wg-+- m+74-my eae+e.s-+a-.-ga.+-=-- - - - w +w-----a eg g, y,y.-,n,-m,.igwwwww re-,--r,+ -r.-%#----w-

e-+--av.--,yw-ee-

b Each reactor coolant loop sample line has a manual flow valve to limit the flow to less than 373 lb/hr. The length of each reactor coolant loop sample line inside containment is sufficient to permit decay of short-lived radionuclides. The length of these lines is adequate to provide a minimum 45-second delay within containment. This 45-second delay time allows the short-lived isotopes, primarily N-16 (7.4 second half-life) to decay sufficiently to minimize the hazard to personnel. The pressurizer is a relatively stagnant volume aad the effective half-life is great enough to decay the N-16. Additional shielding is provided, where necessary, to reduce potential personnel exposure, as described in Section 12.3 Each pressurizer sample line has a capillary tube to limit the flow to less than 373 lb/hr (0.75 gpm sample rate) with all valves in the line fully open. The capillary tube also permits a small flow of 50 lb/hr from the pressurizer steam space to be purged constantly or inter-mittently to the volume control tank, thus removing non-condensible gases. Sample heat exchangers are provided in the common pressurizer steam and liquid space sample line and in the common line from reactor cool-ant loops 1 and 3. These sample heat exchangers are sized to cool the sample to 950F. INSERT 4_ g,93 The instrumentation is protected against overpressurization by a relief valve venting to a floor drain in the sample room of the primary auxiliary building. Venting could occur from high pressure in the steam generator flash tank or closure of a downstream block valve. The blowdown sampling lines are routed to one of the two sample sinks for grab samples. l The sample sink is actually two segregated basins housed in a hooded enclosure equipped with an exhaust fan to provide venting from the hood at all times. The sinks are stainless steel with a raised edge to contain splashed liquid. They drain via a floor drain to the waste disposal system. One basin is for l steam generator blowdown samples and the other for the remaining samples. Demineralized water is supplied to both basins. Both basins are equipped with vacuum gauges to assure correct operation of the exhaust fan. A stainless steel barrier is employed to isolate the two basins. l l l l

di) h, g,~3 -Il INSERT 5 5. Post-Accident Sampling Subsystem The post-accident sampling subsystem provides the capability to obtain liquid samples from reactor coolant loops 1 and 3, ECCS pump room sumps, the pressurizer relief tank, the containment recirculation sumps and gas samples of the containment atmosphere under post-accident conditions. The reactor coolant sample line used during post-accident operation branches off the common line from loops 1 :nd 3 inside the primary auxiliary building. This line bypasses the sample heat exchangers and runs directly to the post-accident sample panel. The configuration of the cont.ainment isolation valves on the sample lines from reactor coolant loops 1 and 3 and the power supply arrange-ment to these valves ensures that a reactor coolant sample can be ob-tained in the event of a power train failure. The post-accident sample panel is located in the sample room on the grade level of the primary auxiliary building. The valving on the panel is operable through a shield wall behind which the panel is mounted. A water bath on the panel adequately cools the sample being taken. The panel provides the capability to extract a gaseous or d/Med/,$4//samph gM e shielded syringe for laboratory analysis. After a sample has oeen removed for analysis, the sample panel can be flushed with demineralized water and retained in a flush tank before being returned to the containment. This return line is provided with automatic ESEAS isolation valves. Reactor coolant is also sampled from the liquid space of the pressurizer relief tank (PRT). The sample is pumped from the PRT, located in the containment, to the post-accident sample panel. The discharge line from this sample pump is equipped with a relief valve which vents back to the pressurizer relief tank to protect it from overpressurization. The sample line from the PRT penetrates the containment and is provided with auto-matic ESFAS isolation valves. The analyses performed on reactor coolant system samples include: gross activity, boron content, chloride content, dissolved hydrogen or total gas and gamma spectrum. Samples from the containment recirculation sumps are taken from the recir-culation lines downstream of encapsulated valves CBS-V8 and CBS-V14. In order to sample either one of the two sumps, each sample line is provided with a remotely operated solenoid valve before joining together in a common header. From this point, the sample ".s pumped to the post-accident sample panel. The ECCS pump room sumps which are sampled during post-accident operation are the primary auxiliary building sump "A" and the two sumps in RHR/CBS equipment vaults 1 and 2. These sumps are sampled to detect any radio-active releases which would result from equipment leakage. Samples from these three sumps are pumped to the post-accident sample panel. The discharge lines from these sumps are all provided with relief valves which vent back to their respective sumps and provide protection from overpressur-ization.

g y//a/cd62=/1from the containment recirculation sumps and the ECCS pump 4 room sumps are analyzedfor gross activity, boron content, chloride content, andJ' ammi sfwchwa. Cas samples of the containment atmosphere are obtained by bypassing the flow to the hydrogen analyzers through sample vessels w/i/c4 are equipped with quick-disconnect couplings ith poppet-type check valves and integral isolation w valves. Once a sample is taken, the sample vessel is removed and its contents are analyzed for hydrogen content, oxygen content and a gamma spectrum. All electrically powered equipment (i.e., solenoid valves and sample pumps) whose operation is required to perform post-accident sampling is powered from an emergency backup power source. INSFST 6 (. h,3 "f 1. The sample heat exchangers, sample sink, sample panel, post-accident sample panel, flow control valves, reach rod operated valves and local flow, temperature and pressure indicators for the reactor coolant, steam generator blowdown, post-accident and other auxiliary systems sampling subsystems are located on the grade level of the primary auxiliary building. 2. The capillary tubes on the pressurizer steam and liquid space sample lines are located inside the missile barrier in containment.

SB 1&2 FSAR ATTACHMENT 281.6-1 NUREG-0737, II.B.3 - POST ACCIDENT SAMPLING CAPABILITY REQUIREMENT a Provide a capability to obtain and quantitatively analyze reactor coolant and containment atmosphere samples, without radiation exposure to any indi-vidual exceeding 5 rem to the whole body or 75 rem to the extremities (CDC-19) during and following an accident in which there is core degradation. Materials to be analyzed and quantified include certain radionuclides that are indicators of severity of core damage (e.g., noble gases, iodines, cesiums and non volatile isotopes), hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron and chloride in reactor coolant samples in accordance with the requirements of NUREG-0737. To satisfy requirements, the applicant should (1) review and modify his sampling, chemical analysis and radionuclide determination capabilities as necessary to comply with NUREG-0737, II.B.3, (2) provide the staff with infor-mation pertaining to system design, analytical capabilities and procedures in sufficient detail to demonstrate that the requirements have been met. EVALUATION AND FINDINGS ~ The applicant has not provided the technical information required by NUREG-0737, Item II.B.3 for our evaluation. Implementation of the require-is not necessary prior to low power operation because only small quan-ment tities of radionuclide inventory will exist in the reactor coolant system and therfore will not affect the health and safety of the public. Prior to exceeding 5% power operation the applicant must demonstrate the capability to promptly obtain reactor coolant samples in the event of an accident in which there is core damage consistent with the conditions stated below. 1. Demonstrate compliance with all requirements of NUREG-Ol37, II.B.3, for sampling, chemical and radionuclide analysis capability, under accident conditions. 2. Provide sufficient shielding to meet the requirements of GDC-19, assuming Reg. Guide 1.4 source terms. 3. Commit to meet the sampling and analysis requirements of Reg. Guide 1.97, Rev. 2, 4. Verify that all electrically powered components associated with post accident sampling are capable of being supplied with power and operated, within thirty minutes of an accident in which there is core degradation, assuming loss of off site power. _ 1-

SB 1 & 2 FSAR 5. Verify that valves which are not accessible for repair after an accident are environmentally qualified for the conditions in which they must operate. 6. Provide a procedure for relating radionuclide gaseous and ionic species to estimated core damage. 7. State the design or operational provisions to prevent high pressure carrier gas from entering the reactor coolant system from on line gas analysis equipment, if it is used. 8. Provide a method for verifying that reactor coolant dissolved oxygen is at < 0.1 ppm if reactor coolant chlorides are determined to be > 0.15 ppm. 9. Provide information on (a) testing frequency and type of testing to ensure long term operability of the post-accident sampling system and (b) operator training requirements for post-accident sampling. In addition to the above licensing conditions the staff is conducting a generic review of accuracy and sensitivity for analytical procedures and on-line instrumentation to be used for post-accident analysis. We will require that the applicant submit data supporting the applicability of each selected analytical chemistry procedure or on-line instrument along with documentation demonstrating compliance with the licensing conditions four months prior to exceeding 5% power operation, but review and approval of these procedures will not be a condition for full power operation. In the event our generic review determines a specific procedure is acceptable, we will require the applicant to make modifications as determined by generic review. l l l [ i I l l l t l l 1 I

I \\ I p P P P~ [toest 0 6. ] [ t 0C... / e/ ell e ee69-e- <am Ctlaa.esstpoq g F-604 *C0 I E8'*% g (n*(4 # 55-eas S% s,7.5v,<. O,e s.v,,, J = m.p.330 Men f*0es toutco= 1 I""'" '/ ~1*' - :^ ~ " ' ' ~ ' ' ' ' ' en ss vaO2 4/(-S LDt %- 6,93 Ar-ee64 - e-g y.s 3.ar-F AT (m as esti \\\\ g g,9 / 1 ss w ye'-u'-vao* W f5 h e' 4 ces v512 ' Des v3M' E8a*6'O= d6 d' cn. Ss -e=4% nn sn V mawt57 j i/(es-vaos/- Qe/* ss-vros s];in *s-vro ()vi.4a ];idm* e 6-]' s/s* u vros 1 1_ gp ere'sTvaar e/(W P

  • 1* ELL LSW"-

75 r3 2 viu-vut '[W W F '.A ] g a viss,t,s, C n edW # gg.,,, n-we %) T .--ee 6/4-11 3. o sarm warra ona.= idmaa* id-Evair ~ K) va*ss-vus sis-varo )H came o sA**tt tota samt T o m poons Osam s/*isS *26 t/4Mv2 0 y4

a..a.

two .x Ee. T u tw % - eas - %O g I/4SS = Midss vus e/. .as. ]H c=.e...o= cowestro. edss-van ]H Op.s4m-vur T tocas savett h*C noon ceanu 9 POST ACCIDENT SAMPLE PANEL 55-C P-419 V t/e's01 l e 9 e t ) I )

b see 5

  • lad Cata=ser, west y
  • SS v267 h - g s,e g

/ toc

    • 62 -

feitel 6.sg esitti ,- v2% \\ [ %' 2-Ap s [ g~ - /.. [ N 'g" 3", ,\\, a6* t-l\\' usee-n,,,,,o %.uv263 ) SS-Ta-etr ,,g,,, l / toc -H-v264 -u - v263 .na c.. Simut tuSL

  1. StautTWE H

N

    • %9-+-

p64 L%er te trg.g 4229 . M *264 /

  1. f-9..so. Moo - 3df

/DIPv 4AOPS to 3 v230 - #-'dS% : \\f-toS025 l' ,,w. g. s g = FC FC -u v2%4 '**2-ww...

  • M v2%

x J_ v231 ~2_. g,, j . E-v2M -Stv260 M l(**%8yI [ Td [2u % es** 4 ii two l 4> 'a*+ I __o.,3..,.

      • p' 3 0 m,sa C,

a., SisluC It#1 4 ' N UnuCTutt ue - on -v3F.

  • v. -

....,s, l' s g 64 % 4 2 -

  • M-vDO
  1. 232

'^%El 80,8 e -se%-3 ~cu-693 Ar. M%-t-gg,,, p f.f..T. N.(SAC. ( SWPS W ---v. 2 ,._.s,. - DM-v3FF -Dee-v375 g II b I" . g.,p g f. e444 - 4 j8as.. r - en2 / aege. ea ge.g.

  • SS v246 LOC v233

' ' M1 '8D'E ' + -0.e-693-S- aF. esS4 t

  • M SS te43-e-
  1. s'S, Y l

.x \\F - *o * *'2 1 u eles E -cu-v3Fs -c ' vy. ss e 12 .4 si... p nicas rou.e noti suup r -904994 eegy. p. =M v242 &DC L,3. ' m x ur m. ast-s

  • SD NG 30

_n...,3....,. '*S' " ' " " TH SYSTEM PAErW FOR THf$ n.s sw w a* DIAGR AM G '55 LM.ESS OTHERWISE C vut.n. -so

  • -t/

2 ya n a enTED. ysge_ - 0W-v3F3 - Des-v 3F2 fx w'-3 3 e,

3....oson

.,,, \\ f ', s4%n-4 [si* sin ut twr 'k' dd,,Y / int. = = + . wen. = ri.. c o o. . unc io W- ,NE tu0TE 6 .X.,g3, ef t stSpitfstt upgratas vntog p DZ35 g (. .on.u.v239,743,2*F,2%s,8 2%4) 49 .u. --,,,3...... nnen vui r .-$fr =237 ' M*W LO(

0. v pe

- 0.s - v p0 Jammw_ n-as... s. ~-n2-., n.....-.,- f,,,,,. a c,,;. 3., \\m. 3 l PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE SAMPLE SYSTEM NUCLEAR-POST ACCIDENT SEABROOK STATION - UNITS 1 & 2 F'&l DIAGRAM i FINAL SAFETY ANALYSIS REPORT l FIGURE 9.3-Sa

) } .nsoE Costunutni__ S._ ouTSOE CONTAsNMENT, S _p 4 g STRUCTURE STRUCIURE 4) see eenngaa.S ess AL.z2m CWi'Q [ lll t..tibf .u., ( .m m _,, et-seek 9054// () r,'n,'," "$-E e c.. Na, I L...- a== s, v., ac.. w. I g = = < - b,, =+ u.:

  • = -
}

u u-u.a ,na - .:e on 4c..m.,soe-en.,e () .s.am n-g l' ,a e .A, ---n.m 4> -.S, *.., 4 o m

      • ' 1 09.'. t!!

o ac-we,N g39. =-SP,t at!.EMll [, .,7,*,E_ - Wy, <rac.,e 44*v,'#" m n si p NAs ,,,,,gf g ,, - l ly V8 88-' 60 " 8 we-ac-. t : ;4ar. lc ,._,3 geg..gg,,g. _.r s,.,w ece-a-aw 4w l r,,J4 =-*-.,=<u k ace rep.g aa W-- a ,. _ m. ~+e h vm. rn. in:t t ,,! ngtT

  • T<J

]

  • (
  • *s b,.,.g u T

2=* * *'+vt " %, e V_ >W l< Voow.. ta.c, s ,,.,,.,-,,o,e J .3,, b ,1o. w,.,,,,,,,;;

  • (-**'*"'i g(*'j,*.*,< 'A.g#fs!,,-

's ,u -.. ~ s &#6 cc k 9 (< TUNNEL AREA - (( k L b LOC (t( m>e o su.m v; Y tut.,'t df,,. .,: h 1 b.,., -. L,,, ,e - g y.. t,,:I t"-:* *>*' *e o a. n ':r,<,, :',,,. I.f. ....,t....,.: .. ~., 3 u 5 M )

1 1 m .u.,.w v. u..,., u

u. o,,

,w.. -l l t-u._! ,s.7,Trm. ,p;;,, rn i w m i m <= =~= wy J'..,.., m, ! _1(..,,,,,<, wu.4 w n... m u-... l % =+.' '= ' I l I q:.,;; g.;;i.) I'**" u --** n.u,.4 u....,<

c. u... <

g

?,'y

?,;n.. ; 3;.~ ; .,;gy . 3, rm,.,

m.. 1

.x.,..

  • .*.n s

.~ .c w ~, > .x.. y.~ ~ k.an. e gig..... m u..,,g g M,e.,, Q,,..,,-a ~,t,,,,,,. .<*, nu. w l 0g -[9' r,y y;-~ ~, G,F," } .. =u u-u,} '

  • u.. ",

I 'o" I . < n..,n) p 9 %u.,3

  1. , 3 u a+',.

a mt'**" t,a,x::~. W,, n 5,..u. .m n ,YTN u(,,-..e[ f'g"l.$..a,".'.,.yg.',,".,',i [ b.b.a.. $}'"* ]l,4 u 4, '), m i j ' _'C ,-a. m u. -ti

  • l'"

S wa w ...=.,.r y. > a e*= ..-...,y, a-t, ,a n...... w u4. h.~.n, y u.

.s L

wi= ~,*

r. u....e u......,..,

a' e .u. . J"F,* 'Mtu r m., g p-i 7.d q 6.c.. [,- U. (P-.( F A MP"' / m tu... i i, >u.~;p,, ,,rc,4.c%.,......, p-- w a-" ~< - :(, ,,,,u,,,.. I' y mu, g'l, '."5.,W

  • M,., g

,7 ,,g a u4, m u..,u a u,. m u.... THE SYSTEM PRErix FOR THIS DIAGRAW mqur- ..,g g.. ( .If,7 {.*rd.*'[h l M U.'d,U e". '" " " ') '"' -- IS 'SS' UNLESS OTHEPWISE NOTED. ce M' ,,,,U,,. w.. k " w,1 +,. 1 n,,,,,, FOR FLOW SHEET REFERENCE DRAWINGS. ~ ,$E$U. 7,M..., SYMBOLS.& AE6REviATIONS SEE DR AWING . u.- b.",,*. t.wp mu g g c,76 3 F-80* OOI. I ~ h, -a,. u. SD NO. 30 _ L..,., <..*g ki,,.v.,' m'a. g,3e, py, b# ,/.'., Il l ~"r e' _i 1 ' tyl,8,** " @ 9763 F 804918 g I'.' WORK THIS DRAWING wlTH 9763-F 804977, .4*,',- p - x wr u-a, W,. * /. t u. u.,.... M" 4 "f6, j

w. u....sq m.

m.... h,{,,0.a ca 4 m d.... - ut' m a,.J4,,.. i. t a* n n,,y N,i. u w,vi,h.i 4 u.u ,N NOTES: 1*.L.),5,, 'M ,,.* u m,, u -a, <.u.,u..,a.......,.,,...,....o,..uu i i m.< u. u.... u.u..u u, a...u M P,w *' 7 '+," w ':.e ll m,,. 3 4. u in Y t-T s u.uea 'l ra.',un'- ,.i m u4,m. : t . wa.."(a.'*u'==.. m u... m u... wn u mun ~ =- 4..,s...- w-u..,. f i a,., a, '3 i

  • 8'58 u-d'w'.

"d%*j F M Og m.u..u l 3 we, m u.. m u..o ; u.., . O n...e .t....H,e.. e,.e. i, s,e i u, e. _w_,.,. u.,,,.....a..... i ,..,.,..n.,.o,.u. .m...a.. f ,,....,, s.n.. u, wu,o ..u,,.,..,,.....,,,, .e.,,. mu..,,, "*} [' .u.. ,.D b.., .e.e.t.,.e s,.,e,.... t. w I ..,w.,, .n.= w.. . <,.ru m. ~ T.-. ne.,<,..O II I wa= a,...a n. o.n. e.u,a ~ ~.a*.** w*a.uta, " a a.'.- . - on. . u..m u. u 1.,,. k. ,s, r.c..,,,,,,.

....a,,,

o~. ..u.u._......... .....u<.... .,.,,,u,. '( *s.. / . oru. . cs.. v aw y.w w-,,~,n =s...,,, .~m. um.~, a ... u..u.. u -. .m...........u.........-..u.....~. ,,,,.....i,,,.,, ; m m, .u... n., =a...., um,w": n :*.,,. _.._ w e ,..;,3,...... p"D"m, 7 I.

  • u~'.

./. e -='m e g.g, [q,;im.,,w, e J n i,e, ne, 3.'. g._ w, g. g ....m g _, e.g.,. -.....,.;... 7,, e w n.......,, u,. y.g, m u.n.{ y{ '*"***'S'* t.t, uma volta s. w u..e L.<..,, ,_m.. q),; SAMPLE SYSTEM NUCLEAR P&l DIAGRAM PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE SEABROOK STATION - UNITS 1 & 2 SHEET 1 FINAL SAFETY ANALYSIS REPORT l l FIGURE 9.3 5b

p p 8 f,Wlo \\ e%0li / (NTi g \\ a..... ~ >WS / . (4 g fd D. m e auPLI Sr5TEu\\ $.4 / ,t e 5f5TEMA yren / LM S4 auptf Ef5T(s\\ SJ#A /

  • * ^ ~

m-sses -ssa. s 4ses s aptg s SY4SWF Euw'es. u '[ .... w. w '" "....'4*!#6 = = 2

  • A3.2ME naa 2 n11'.

L 4 9 -p p p-p u ' no selD es O( m # e, ATE \\ Deg e4-a?-3/e* g gg serf-E P90A-5/44 a WP lMG 3'l f k."...... .......,4 ,=r' a ,," *.7 ,.a..,-,w <.1,oa x

  • ='I" N

.S$4440-5 a r-f,w+se e a \\ De.4-4 49-1/4* g g g / .. tow 30 R Au %u 4

p..., s a

neo 1 Ss 3M' 'e's 2 fve va = e4aF-8 Os** WhW b 'l 'l NI.mi. LM 9 4 mm u...,.,v.- e n 9 e-mePgt S v$TE W (h 7.b"_ ' ^ u.- I -s a.s.t... 1 - S gt.t.0 Sg2, St5 tE 8 a m*L E LM 8 j<= e4 } a .,,,. x ng;s,,g.h {^w*o*j',,l',l'8" LR -. w>s <a.4 y Mw / e l ta a-.s g 1 \\ i

I f-? i,e'ss.wss ry..,, 2 -. f.1/[ x e i'- os o... ) = c ste stvSa V Q ses'Ss yst '" i. M'a yg[D, ..i,,-2 50 3/s',- 3 gr 1, o.s.,ea L XILW h9T' '3 s s s s E.:.!. :. As. t. a c_ e. a A .m. 7 ,IsififI.I a a a a a y,, ,r h una 1.M J .\\ b es. 8,','.M es e-,,3.< t - -e me.,.. fgo,.g.,. I

x..,> m.

mc ...,.ou s, toc. c. ua.n u.a mo,a a ,sw \\ N,-:.as,s.cu.,, m Ress.,-,* ve f le,.,,.,y. > x ew.. I ' 'e SS-C,, w cs-s.e-e-aso..s/u v toc a-a

1n0'11;

..s, = ..... s,,,,.,::., s.,,, 'Ni,r x,. wm I 4 ,5-s-2S06-s/S"-T tX c-ua x u l es -,,me,r a,.,3 x es..,....,- eg xf.eg,3 cmI _ __ g,e

m..

toc. can D us no. a q ,_ es.. 3,v. Q es c..,v. 3,, es..,.,...- fi s/u:cs.v.se xfg op... j co 1 a.m. na.x v W s >-t@a-

o...w..,....,, f f.x ##?J I

. o. - nv.,., 8 13 11 g ~ toc. c-s L..S,so W ... o u r - .h3.. s,. ss, is v,h g it w FOA GENERAL NOTES SEE DRAMNG e 3 7 9763-F-805025. a T s t 3 1 S 3 0 d im: !a l. I. Ia-I,! I* I- +seio vur I-I.2 s.w!;\\ '- a 6 s o 2 a a e E 1 a. e. t i. ) ls V. V. V. V. V. V, X,, te / v: v,: ": v.: In I: 1: Ig n I.: H-p I.: s s-D Ta s a da 4 / &a la 4a la la "a. "a la /) ...a... i.s.s... t....a. w... hss p 4.or... rv .t va'ss vsa gk w.. e i er.- e T w. Fla*_ .m o.M,3...o* 56 , a s,s,,,. e s-ex.4 <x .I i T r i SAMPLE SYSTEM NUCLEAR P&l DIAGRAb PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE SEABROOK STATION - UNITS 1 & 2 SHEET 2 FINAL SAFETY ANALYSIS REPORT l FIGURE 9.3-Sc I}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20067B228&oldid=4430995"