ML20235V089

Forwards Responses to NRC 890203 Request for Addl Info Re SSAR for Advanced Bwr.Response Pertains to Chapters 9,11,12 & 13

ML20235V089
Person / Time
Site:	05000605
Issue date:	03/07/1989
From:	Marriott P GENERAL ELECTRIC CO.
To:	Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
15-89, NUDOCS 8903100033
Download: ML20235V089 (92)

Text

{{#Wiki_filter:.

.'7.

M GerreralEkz'ric Company. 175 Curtrw kwue, Sen Jose, CA 9G125 March 7,1989 MFN No. ' 15 Docket No. STN 50-605 Document Control Desk U.S. Nuclear Regulatory Commission L Washington,D.C. 20555 Attention: Charles L. Miller, Director Standardization and Non-Powcr Reactor Project Directorate ~

Subject:

Submittal of Responses to Additional Information as Requested in NRC Letter from Dino C. Scaletti, Dated February 3,1989

Dear Mr. Miller:

Enclosed are thirty four (34) copies of the Responses to Request for Additional Information (RAI) on the Standard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABW.R). Hese rer,ponses principally pertain to Chapters 9,11,12, and 13. Also included are other committed responses to previous RAls. It is intended that GE will amend the SSAR with these responses at the end of April 1989. Sincerely, L P. W. Marriott, Manager Licensing and Consulting Services cc: D. R. Wilkins (GE) F. A. Ross (DOE) J. F. Quirk (GE) D. C. Scaletti (NRC) f 'pd 8903100033 890307 g FDR ADOCK 05000605 PDC y i 1

g I s GE Nuclear Energy bt'l's [f!'.

:.~ o - :,.e:c u :-rs March 7,1989 MFN No.15-89 Docket No. STN50-605 Document Control Desk U.S. Nuclear Regulatory Commission Washington,D.C. 20555 Attention:

Charles L Miller, Director Standardization and Non Power Reactor Project Directorate

Subject:

Submittal of Responses to AdditionalInformation as Requested na NRC Letter from Dino C. Scaletti, Dated February 3,1989

Dear Mr. Miller:

Enclosed are thirty four (34) copies of the Responses to Request for AdditionalInformation (RAI) on the Standard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABWR). These responses principally pertain to Chapters 9,11,12, and 13. Also included are other committed responses to previous RAIs. It is intended that GE will amend the SSAR with these responses at the end of April 1989. Sincerely, P. W. Marriott, Manager Licensing and Consulting Services cc: D. R. Wilkins (GE) F. A. Ross (DOE) J. F. Quirk (GE) D. C. Scaletti (NRC)

MM 23A6100AT Standard Plant arv. s ~ QUESTION 100.1 l In light of the recent interest in ABWR thermal hydraulic stability following the LaSalle instability event, it appears to be highly desirable to ar.sste that in an advanced BWR design the possibility of instability is precluded, both in normal and anticipated abnormal operating conditions; this should be the case without requiring the prompt intervention of the operator. If actions are required, they should be automatic. If operator attention is required, suitable monitoring capability should be readily available. Please discuss the extent to which this is provided for in I the ABWR. This discussion should consider (1) the various potential problem areas which have been identified in i the current BWR stability review (particularly asymmetric oscillations), (2) the relevant stability related i characteristics of the ABWR core such as fuel entrance loss coefficients, void reactivity coefficients, fuel l conductivity, and including extremes of conditions in both the initial core and potential reload cores with different l fuel, (3) accessible stability significant regions of the power-flow map, involving both normal and abnormal events l (including multiple out of service or tripped recirculation pumps), (4) the selected control rod run-in (SCRRI), describing its relevant characteristics including provisions for automatic initiation, speed of operation compared to need for rapid action, boundaries of operation (on power-flow map), flexibility of these boundaries as need for change may arise, (5) the existing relevant instrumentation and possible need for improved or augmented instrumentation such as on line stability measurement or easily available relevant LPRM readings and automaGe l action based on these measurement., (6) the need for frequent mapping of boundaries of operational map regions to be avoided (Chapter 4). RESPONSE 100.1 General ABWR Desien The stability design for ABWR follows the same design and licensing philosophy outlined in Thermal Hydraulic Stability Amendment to GESTAR II", NEDE 24011-P A, Rev. 6, Amendment 8 (Reference 1), and ' Compliance of the GE BWR Fuel Design to Stability Licensing Criteria," NEDE-22277-P-1 (Reference 2). Both of these two references have been approved by the NRC (Reference 3). Specifically, the ABWR design assures the stability performance in the normal operating region (Regions I and IV in Figure 4.41) is more stable than the ] current operating BWRs by incorporating the following design features: 1 (1) Smaller jnlet orifices, which increase the inlet single-phase pressure drop, and, consequently, improve the core and channel stability. (2) Wider control rod pitch, which increases flow area, and, consequently, reduces the void reactivity coefficient and improves both core and channel stability. (3) More steam separators, which reduce the two phase pressure drop, and improve the stability, and (4) Automatic logics which prevent plant operation in the region with the least stability margin. Furthermore, automatic startup logic may be programmed to avoid the region with the least stability margin. (See the startup path in Figure 4.4-1.) Therefore, it is expected that the ABWR operation in the normal operating domain (which is currently defined only below the rated rod line) is always stable. In addition, regional (LPRM) and core wide (APRM) neutron flux time histories are available for display to enabic the reactor operator to detect any neutron flux oscillations. This capability is an addition to current conven-tional design. Therefore, the operator can easily detect any flux oscillations, which are very unlikely in ABWR. Amendment Futur 1 l

. ~ 31 r 23A6100AT gudard Pinnt nev. s Caaeh ABWR Ceahility Prevention Features Operation in the high power / low flow region, which the system is the least damped (Region III of Figure 4.4-1) is precluded by the selected control rod run in (SCRRI) logic. This automatic logic is shown in Figure 20.3-21. The setpoints shown in this figure are analytical limits determined from conservative fuel nuclear characteristics. The nominal setpoints will be determined based on the approved setpoint methodology (e.g., nominal setpoint for the flow setpoint is about 48% of rated). The SCRRI function is automatically initiated to avoid stability concerns when a trip of two or more RIPS occurs. The SCRRI function is bypassed when the power is below a specified setpoint, or when the core flow is above a specified setpoint. The automatic initiation of the signal when a trip of two or more RIPS occur, takes place when core Dow is less than or equal to 36 percent 'AND" when the power level is greater than or equal to 30 percent. This is to assure power level below 80% rod line at natural circulation and to assure flow rate is higher than that of eight RIPS operating with the minimum pump speed. The SCRRI stability control function is accomplished by the automatic initiation of electrical insertion of selected control rods and the total rod worth for the selected control rods to bring the reactor power from the 100% rod line to below 25% power to assure stable operation following a trip of two or more RIPS. An operator has the capability in the control room to make calculations based on previous operations and determine the rod worth necessary and to preselect control rods for the SCRRI function. The preselected control rods for the SCRRI function are stored in memory in the RC&IS circuitry. The initiation of the SCRRI function controls the stability concern region based on the power level and the core flow level. Since no single failure in any trip logic, including RIT logic, can cause more than one reactor internal pump (RIP) to trip, the probability of initiating the SCRRI logic is very low. For further protection, a scram trip based on core flow coastdown rate is implemented. Currently, a setpoint is chosen such that a simultaneously trip of more than 5 RIPS with the initial power higher than about 80% of rated power would initiate an automatic scram trip. Therefore, operation in Region III is highly unlikely. In add: tion, a control rod withdrawal block is also irsplemented as shown in Figure 20.3-21 to prevent the plant operation in Region III during startup. The operator can clear this block only by increasing the core flow. In summary, the ABWR design assures stable operation in the normal operating domain. Automatic logic is also provided to prevent the plant operation in the region with the least stability margins. Therefore, the ABWR design meets the stability design and licensing criteria. l ABWR Features which Preclude imRalle-Tyne Event initiators A question has arisen whether the laSalle type of event could occur in the ABWR design. Specifically, could a single failure in one reference leg, cause the plant to initiate trips or take control action based on false water level indications? The answer is no, this type of event is not possible in the ABWR design because of the two-out-of-four (2/4) trip logic adopted in ABWR for all water level related trips, and the use of triplicated water level signals for use in control algorithms. Anneedment Futur 2 .m

4 I \\ O ~ Mstk mandard Plant arv. s The design of the water level trip logic at LaSalle is such that the asustion of a single pressure switch can < cause the trip to occur. This is complicated by the fact that a common reference leg is used for both of the ATWS redrculation pump trip switches (i.e., the pressure switch that causes pump A to trip uses the same reference leg as the pressure switch that causes pump B to trip). Therefore, the single failure in one reference leg caused both recirculation pumps to trip. The feedwater control system responded to the false levelladu=rian, because it derives its water level signal from a single sensor on the same reference leg also. The ABWR design precludes this type of event by ensuring that all water level related trips and control a:gnals are derived from multiple sensors using different reference legs. The feedwater control system controls level based on a reactor water level signal which is derived from three different transmitters using three different reference legs. The middle value of the three signals is used in the control algorithm. Therefore,if one reference leg fails, causing an erroneous level signal, it will have no effect on the control process, because the controller will be using other signals. Similarly, the recirculation pump tric I sic on low reactor water level is derived from four different level transmitters using four different reference hs. The trip occurs only if two of the four signals indicate low level. Therefore, the failure of one reference leg a noot cause the recirculation pump trip to occur. The type of event which occurred at I aSalle cannot vcu with the ABWR design. SwAfic Resnonse to Question 100.1 (1) The relationship between various stability modes in terms of conventional core and hot channel decay ratios is shown in Figure 20.3-22. Regional oscillations are possible only when both core and hot channel decay ratios are high and close to the limit of LO. For ABWR design, the hot channel decay ratio is reduced by incorporating smallinlet orifices, whose loss coefficient is about double of that for ABWR/5/6s, and wider control rod pitch. The hot channel decay ratio is less than 0.45 even with a very bottom-peaked axial power shape (e.g.,2.0 peaked at node 3) at the intercept of minimum pump speed line and the 102% rod line j (Point 1 of Figure 4.4-1). Therefore, regional oscillations are not very likely to occur in ABWR. (2) The design target for ABWR is to have a calculated core decay ratio less than or equal to 0.8, taking into consideration model uncertainties, future core design and possible operating modes with at least 9 RIPS in operation. This design target is achieved by incorporating smaller inlet orifices, wider control rod pitch and more steam separators. These design changes result in less negative void reactivity coefficients for both initial core and reload cores and consequently enhance both core and channel stability. The calculated core decay ratio for current ABWR core design is less than 0.7 in the equilibrium cycle. The core decay ratio is estimated to be less than 0.8 for future core design. (The procedure used here is similar to that used for LaSalle post event analysis, which calculated core decay ratio of 1.05 at the conditions of oscillations). (3) As discussed above, normal plant operation with all RIPS in operation, or even with one RIP out of service, is stable in the whole operating region. With more than one RIP out of service, the plant operation is automatically limited to be outside the SCRRI region by the SCRRI and rod block logics, and therefore the operation is stable. During transients, the reactor stays in the stable region if a reactor scram is immediately initiated during the event Hence, the only transient which may lead the reactor into the SCRRI region, is a trip of two or more RIPS. In this case, the SCRR1 logic is automatically initiated to prevent the plant from staying in the least stable region. Therefore, stable plant operation is assured during normal and abnormal operations. Amendment Putur 3

e i M 2sA61ooAT Standard PInnt m (4) The selected control rod run-in (SCRRI) is automatically initiated when a trip of two or more RIPS occurs in 5 which the plant enters the least unble region (see the setpoints in P'.gure 20.3 21). The preselected control rods are commanded to be inserted electrically with the normal speed. It takes about 2 minutes to insert these rods to the full-in position if these rods are in the full-out pcsition. However, plant experience shows l that an insertion of a few notches is sufficient to bring the plant to a stable condition when oscillations occur. Therefore, the effectiveness of the SCRR1 initiation will start in a few acconds. De boundaries of operation are changeable. Nonetheless, changes in setpoints are very mdikely since the ABWR stabihty design already takes all operating modes, conditions, and future core design into consideration. (5) As described under " General ABWR Design

• above, a new design is added to provide displays of regional and core-wide neutron flux time histories (i.e., APRM/LPRM histories) to enable the operator to detect any neutron flux oscillations.

(6) %c ABWR stability design takes all operating modes, operating conditions and future core design into consideration as discussed above. Therefore, frequent mapping of boundaries of operational map is not mereacary and is not needed. References for Resnonse 100.1 1. NEDE-24011.P A," General Electric Standard Application for Reactor Fuel" 2. NEDE-22277-P-1," Compliance of the General Electric Boiling Water Reactor Fuel Design to Stability Ucensing Criteria," October,1984 3. Letter, C.O. Thomas (NRC) to H.C. Pfefferlen (GE),' Acceptance for Referencing of Ucensing Topical . Report NEDE-24011, Rev. 6, Amendment 8,* Thermal Hydraulic Stability Amendment to GESTAR 11,* April 24,1985 Arnendment Futur 4 1

i ABM n46xorr j ~ Standard Plant nry n I QUESTION 281.9 l Section 6.4.4.2 (page 6.4-6) discusses personnel respirator use in the event of toxic gas intrusion in the control room. However, the chlorine detection system is not Murd. Also, any control room functions that are automatically triggered by a chlorine detector alarm (closing intake dampers, energizing control room HVAC system recirculation) should be identified. RESPONSE 281.9 The ABWR control room habitability system does not include a chlorine detection system. QUESTION 281.13 Table 11.14 indicates that the N 16 concentration in the steam is four times the normal value when hydrogen water chemistry (HWC) is used HWC tests conducted at BWRs have indicated that N-16 activities have increased in the range of 1.1 to 5 times the N-16 concentrations observed during normal water chemistry operations. What is the basis of the factor of four increase for the ABWR7 Is it based on the model for predicting HWC that was reported in "U.S. Experience with Hydrogen Water Chemistry for Boiling Water Reactors", R.L. Cowan, C.P. Ruiz and J.L. Simpson, April 19887 (11.1) RESPONSE 281.13 A value of 200 uCi/gm was given in Table 11.1-4 for the N-16 concentration (actually the N-16 equivalent of the combined N 16 and C-15 concentrations) in steam which is applicable to operation with hydrogen water chemistry. This was a current best estimate. It was obtained by determining the nominal N-16 source concentration which was necessary to reconcile the observed increase by a factor of 5 in the dose rate at the Dresden 2 main steam line radiation monitor during operation with hydrogen water chemistry and incorporating an additional factor for conservatism. The increase in main steam line monitor dose rate at Dresden 2 was at the upper bound of increases observed at the plants which have operated with hydrogen water chemistry. This value is under continuing review and may be adjusted as more definitive data becomes available. It is anticipated that further refinement of the number would not be expected to result in an increase of more than about 15% in the source term..(11.1) QUESTION 281.14 In Section 11.5.2.1.1, there is no discussion of a dual setpoint for the main steam radiation monitors (MSLRMs) when HWC is used. Below 20% power, the MSLRM setpoint is established to detect high radiation levels in the main steam lines and provide signals for reactor scram and MSIV closure to reduce the release of fission products to the environment in the event of a control rod drap accident. When hydrogen is injected into tue feedwater at power levels above 20%, the MSLRMs may have to be reset due to the increased N-16 activity in the main steam line. (11.5.2) RESPONSE 281.14 The dual setpoint methodology for the main steam line radiation monitor (MSLRM), when utilizing hydrogen water chemistry (HWC),is one of several ways to address the ine ease in N 16. An alternative approach, such as that proposed by Susquehanna Steam Electric Station, and accepted by the NRC (Docket No. 50-387), was to demonstrate that an increase in the trip setpoint would not mask ihe fission product release associated vdth a gross fuel failure. Thus, by raising the trip setpoint a considered amount, the need for a dual setpoint can be avoided. Additionally,it should be noted, that the MSLRM high trip setpoint is not considered in any accident analysis for the ABWR. Amendment Futur 5

r-23A6100AT 1 Standard Plant arv. s QUESTION 410.44 Provide P& ids for the Condensate Storage Faci!: ties and Distribution System (i.e., Makeup Water condensate (MUWC) System). Also, provide a list of tanks (with capacity) and other requirements in the system. (9.2.9) l l RESPONSE 410.44 The MUWC P&ID will be provided by March 31,1989. The only tank in this system is the condensate storage tank which has a capacity of approximately 560,000 gallons. This tank is located outdoors adjacent to the turbine building. The other requirements of this system are provided in Subsection 9.2.9. - QUESTION 410.45 Clarify which portion of the MUWC system is within the ABWR scope. Also, identify the system interfaces which include flow rates, supply pressure and temperature. (9.2.9) RESPONSE 410.45 All of the MUWC system is within the ABWR scope. QUESTION 410.46 Clarify whether the distribution system includes any surge volume and,if so, how much and for suction of which pumps. Also, if applicable, describe how protection against the effects of flooding resulting from possible failure of the surge volume is ensured. Define what 'HPCF pumps" means. (410.46) RESPONSE 410.46 l t The surge volume of the system is within the condensate storage tank (CST). The capacity requirements of { the CST are in Table 9.2-3. Section 3.4 demonstrates that failure of the CST will not lead to unacceptable results. ) 'HPCF pumps" means the high pressure core flooder pumps. l QUESTION 410.47 Describe the design features provided in the system and/or interfacing components to ensure automatic switchover of the suction of the applicable pumps to safety-related water sources,if so required. (9.2.9) RESPONSE 410.47 l Level sensing elements and transmitters are provided for the condensate storage tank (CST). Signals are sent to the HPCF, RCIC and SPCU pumps to provide automatic switchover to the suppression pool when sufficient water is not available in the CST. Amendment Putur 6

ABWR 2mmr ~ Standard Plant REV.B QUESTION 410.48 Discuss conformance of the MUWC systems design with the requirements 10 CFR 50.63,

• Loss of all Alternating Currc.nt Power." Specifically include the system's capacity and capability to ensure core cooling by removing decay heat independent of preferred and onsite emergency ac power in the event of a station blackout for the specified duration, in accordance with guidelines of Regulatory Guide 1.55,* Station Blackout,' Positions C.3.2 through C3.5, es applicable. (9.2.9)

RESPONSE 410AS The condensate storage tank (CST) is designed to provide approximately 150,000 gallons of water for use during station blackout. Other consumers of condent. ate are switched to other water sources so that this volume of water is s.1 ways available during power operation. This volume of water is sufficient for operation of the RCIC system to rernove decay heat during the first eight hours of station blackout. QUESTION 410A9 Discuss compliance of the system with Positions C1 and C2 of Regulatory Guide 1.29. (9.2.9) RESPONSE 410A9 The normal secured source of water for decay heat removal is the suppression pool. The condensate storage tank (CST) is used in preference to the suppression pool because the water quality is normally better. As a result the CST is not required to be Seismic Category 1. QUESTION 410.50 Provide P&lDs for the Demineralized Water Makeup System (i.e., makeup Water System (Purified) (MUWP)). (9.2.10) RESPONSE 410.50 The MUWP P&ID will be provided by March 31,1989. QUESTION 410.51 Clarify which portion of the MUWP is within the ABWR scope. Also, identify the system interfaces which include temperature, chemistry, system capacity (i.e., tank volume) and treatment. (9.2.10) RESPONSE 410.51 See response to Question 410.52. Amendment Futur 7

1 23A6100AT Standard Plant l may a [ QUESTION 410.52 Provide the water quality skaracteristics for the MUWP water (SSAR Section 9.2.10.1, Item 3, refers to section 9.2.8 which in turn refers to section 9.2.16. However, Section 9.2.16 does not give the water quality characteristics). (9.2.10) RESPONSE 410J2 j The response to this question h provided in new Table 9.2-2a. i QUESTION 41033 Discuss compliance of the system with Position C1 (e.g., containment penetration portions) and Position C2 of Regulatory Guide 1.29. (9.2.10) RESPONSE 41033 The MUWP line which enters primary containment has a locked closed manual valve outside of containment and a check valve inside of containment. The containment penetration is Seismic Category I and Quality Group B. (9.2.10) QUESTION 41034 Verify that floodmg analyses have been performed for a failure of the nonseismic Category I demineralized . water makeup system where the piping runs through safety-related structures and tunnels containing safety-related equipment. (9.2.10) RESPONSE 41054 Section 3.4 studies of MUWP piping run failures throvf)h safety-related structures and tunnels containing safety related equipaent have shown that flooding wil! not have an adverse effect on these structur(s and equipment. Amsedment Putur 8 I

i 21A6100AT Standard Plant nev. s QUElrI10N 410J5 With respect to the capability of the Reactor Building Cooling Water System for h:~i. control, and isolat on of system leakage, and radioactive leakage: (9.2.11) i 1. Identify the isolation valves which irmte the non-essentialloads from the essential supply headers and describe their isolation function in the event of a LOCA or in the event of a leak detected in the non-essential system piping.. 2. Identify and describe instrumentation used to detect leakage in the non--atM system piping. 3. Identify the valves which are activated by the surge tank level switch to isolate a leaking system train. t Identify all radiation monitors provided and describe their individual function. Also, ciarify whether the system design includes any radiation m%itor in the pump suction header to detect inleakage from radioactive systems. l RESPONSE 410J5 1. In the event of a IDCA signal, valves F075, F080, F081, F101 and F141 shall close. If a pipe break occurs in the non-essential portion of the piping, valves F072, F074 and F082 shall be closed. These three valves are not directly closed by a LOCA signal. The operator may close these valves if desired. 2. Leakage in the entire system can be detected by level monitors in the surge tank. To determine if the leakage is in the non-essential portion of the system, valve F082 is closed and the flow is monitored using FI642. If flow continues, this indicates that a leak is present. Then, valves F072 and F074 are closed. 3. Valves F072, ft)74 and F082 are closed by surge tank level switches. 4. The radiation monitor is located after the RCW pumps and heat exchangers in each division. If j detectable radioactivity is observed, grab samples will be obtained from sampling lines downstream of each heat exchanger cooling highly radioactive water. QUESTION 410.56 1 Identify the functional performance requirements associated with water hammer and address the design, provisions and procedures provide to meet these requirements. (9.2.11) l RESPONSE 410.56 i One of the functions of the surge tank is to provide adequate pressure for pump suction. Additionally, each isolatable portion of the system is provided with high point vents. The safety-related portions of the sprem are isolated by closing only one valve. Thus, these portions remain pressurized during normal operation. The operational procedures will require that isolated systems be filled and vented before being placed in operation to avoid water hammer. Amendment Putur 9 a

23A6100AT Standard Plant arv. r QUESTION 41037 Identify the system requirements for water makeup, and address the capacity of the surge tanks to accommodate expected leakage from the system or that a seismic source of makeup water can be made available within a time frame consistent with surge tank capacity. (9.2.11) RESPONSE 41037 The surge tanks are sized to operate for thirty days without makeup at allowable leakage rates. During this period, adequate makeup water can be made available. QUESTION 41038 Provide the design characteristics for the system pumps, tanks and heat exchangers. (9.2.11) RESPONSE 41038 The response to this question is provided in revised Subsection 9.2.11.2 and new Table 9.2-4d. QUESTION 41039 Define the terms; FCS, CAMS, LCW, HSCR, HWH hot water heat exchanger, and HCW. (9.2.11) RESPONSE 41039 FCS = Flammability Control System CAMS = Containment Atmospheric Monitoring System IEW = Inv Conductivity Waste HSCR = Heating Steam Condensate Receiver HWH hot water heat exchanger = Hot Water Heating System HCW = High Conduaivity Waste QUESTION 410.60 Discuss how the Reactor Building Cooling Water (RCW) system complies with Position C2 of Regulatory Guide 1.29 with respect to the non safety related portions, and with respect to GDC 2 for safety-related portions (e.g. physicallocation to protect against earthquakes and floods) (9.2.11) RESPONSE 410.60 The safety-related portion of the RCW system is Seismic Category I up to and including the valves which isolate it from the non safety related portion of the RCW system. The RCW system meets the GDC 2 requirements as dise=d in Subsection 3.1.2.1.2.2. l Amendment Futur 10 .A

~ ABM 23461oorr Standard Plant arv. s QUESTION 41041 Clarify whether the flows indicated for the components servi.:ed by the RCW system in SSAR Tables 9.2-4a, 4b and de represent the minimum Dow requirements at the inlet of each component. Also, specify the maximum allowable RCW temperature at the inlet of each component under different operating M*iaa< (9.2.11) RESPONSE 41041 ne flows in Tables 9.2-4a,4b and 4c are design values which include an allowance for instrument error. The design temperatures at the inlet of each component under different operating conditions will be provided in a process Dow diagram to be submitted by March 31,1989. QUESTION 410.62 Clarify whether availability of only gn division of the RCW system is sufficient to provide cooling water to the drywell coolers and the RIP coolers (SSAR Tables 9.2-4a and 4b list only Division A and B servicing above. Further Table 9.2-4b lists only the Drywell B cooler as being serviced by Division B). (9.2.11) RESPONSE 410.62 The availability of only one division of RCW system is sufGcient to provide cooling water to the drywell coolers. Studies have shovm that the increase in drywell temperature will not be sufficient to require shutdown of the plant. Division A of RCW system cools drywell coolers A and C and five of the RIP coolers. Division B of RCW system cools drywell cooler B and the other five RIP coolers. Revised P& ids for the system will be provided by March 31,1989. QUESTION 410.63 Regarding the HVAC Normal Chilled Cooling Water System, provide information on the following: (9.2.12) a) Compliance with GDC 2 for safety-related components (i.e., physical location for complying with the .GDC) 1 I b) Compliance with Position C2 of Regulatory Guide 1.29 for the non-safety related portion. c) Automatic features to provide cooling water to the equipment serviced by the system in the event of its failure on loss of offsite power (specify the system that will provide cooling water in the above situation). d) Description of the turbine building cooling water system which provides condenser cooling (refer to SSAR section 9.2.12.2) if it is within the ABWR scope. Otherwise, identify it as an interface requirement. 11 Amendment Futur

,e MM 23A61ooAT fdandard Plant arv. s RESPONSE 41043 a) N HVAC normal chilled coohng water (HNCW) system meets the GDC 2 requirements as d;u -d in Subsection 3.1.2.1.2.2. b) & HNCW system is non-safety related. Failure of this system to provide cooling water will not adversely affect any safety related equipment. c) During loss of offsite power, cooling water will not be provided to the equipment serviced by the HNCW system. d) N description of the turbine building coolin.g water system will be provided by March 31,1989. - QUESTION 41044 Regarding the HVAC Eraergency Chilled Cooling Water system, provide information on the following. (9.2.13) a) Compliance with GDC 2 for safety related portion (i.e., physicallocation for complying with the GDC). b) Compliance with Position C2 of Regulatory Guide 1.29 for the non-safety related portion, '.' there is any such portion. c) Compliance with GDC 4. d) System active component failure analysis. RESPONSE 41044 m) The HVAC emergency cooling water (HECW) system meets the GDC 2 requirements as discussed in Subsection 3.1.2.1.2.2. b) There are no non-safety related portions of the HECW system. c) N HECW system meets the GDC 4 requirements as d;<ct-d in Subsection 3.1.2.1.4.2. d) See new Table 9.2-10. i Amenoment Futur 12 ) i

M 23A6100AT RianAmed Plant nev. s QUESTION 430.13 Section 6.2.1.1.3 of the SSAR states that the containment functional evaluation is based upon the consideration of neveral postulated accident conditions including small break anM*=*= Provide the assumptions, analysis and results of the small break accidents considered, and demonstrate that the identired (in the SSAR) feedwater line and steam line breaks are the Smiting accident. RESPONSE 430.13 Figures 20.3-11a through 20.311f demonstrate that the maximum drywell pressure occur during feedwater line break accident, and the maximum drywell temperature condition would result from a main steam line break accident. All of the analyses assume the primary system and containment system are initially at the maximum normal operating conditions. QUESTION 430.27 Describe the design features of the suppression pool suction strainers. Specify the mesh size of the scrcens and the maximum particle size that could be drawn into the piping. Of the systems that receive water through the suppression pool suction strainers under postaccident conditions, identify the system component that places the limiting requirements on the maximum size of debris that may be allowed to pass through the strainers and specify the limiting particle size that the component can circulate without impairing system performance. Discuss the potential for the strainers to become clogged with debris. Identify and discuss the kinds of debris that might be developed following a loss-of coolant accident. Discuss the types ofinsulation used in the containment and describe the behavior of the insulation during and after a LOCA. Include in your discussion information regarding compliance with the acceptance criteria associated with USI A-43 as documented in NUREG 0897. (6.2) RESPONSE 430.27 The strainers are sized to satisfy pump NPSH pressure drop requirements while being 50% plugged. The strainers and mesh are designed to prevent passage of particles larger than a certain size, as determined by the pump design requirements, to prevent clogging the HPCF main pump cyclone separators or seal cooling water orifices. The ECCS suction strainers are designed to block passage of debris which could potentially degrade the performance of these systems while considering the potential plugging effects of suppression pool debris, especially under post IDCA conditions. The ECCS suction strainers are designed and located so that, considering the type and amount of pool debris, they will not become so plugged during the post LOCA operating period as to result in degraded performance of these systems. Containment protective coatings are required to withstand radiation, temperature and pressure, and not flake to present problems of plugging ECCS suction strainers in the suppression pool. j l nimary system piping and reactor vessel insulation are designed to minimize adverse effects on containment function following a IDCA. l 13 Amendment Futur 1

y MN ~ zwtooAT hadard Plant nev a For each postulated break location, the type and estent of insulation debris are =#1=med and evaluated for the debris' effect on the potential plugging of the drywell connecting vents and for potential suction strainer plugging in the suppression pool. The mesh size to have a dimension on one side of approximately 1.5 mm. His size is to protect the ECCS pump cyclone separators. Metal pipe insulation is used where 151 (in service inspection) of welds is perfanned. He metal insultion I willsink in the pool. The major pipe insulation material is potentially calcium silicate, which will float for a very short time and then sink in the pool. De vents from the drywell to the wetwell are provided with screens to block pipe insulation or other debris from entering the wetwell. Suppression pool suction strainer compliance with the acceptance criteria associated with USI A-43 is addressed in Appendix 19B. QUESTION 430.30 Provide a tabulation of the design and performance data for the secondary containment structure. Provide the types of information indicated in Table 6-17 of Regulatory Guide 1.70, Revision 3. l RESPONSE 430.30 Response to this question is provided in revised Subsection 6.2.3.2 and new Table 6.2 2d. QUESTION 430.46 i According to SRP 6.2.5 specific acceptance criteria related to the concentration of hydrogen or osygen in the containment atmosphere among others are the following: a) The analysis of hydrogen and oxygen production should be based on the parameters listed in Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design basis for combustible control systems. b) The fission product decay energy used in the calculation of hydrogen and oxygen production from radiolysis should be equal to or more conservative than the decay energy model given in Branch Technical Position ASB9 2 in SRP 9.2.5, Provide justification that the assumption used in the ABWR in establishing the design basis for the combustible gas control systems are conservative with respect to the criteria a. and b. above. (6.2) RESPONSE 430.46 GE has been evaluating the need for additional hydrogen control during finalization of the ABWR design. A recombiner, based on the source terms specified in Table 1 of Regulatory Guide 1.7, will be added to Subsection 6.2.5. Asieedinent Futur 14

MM 2sA6 MAT Dandard PInnt arv. s j QUESTION G.47 Provide an analysis of the production and accumulation of combustible gases within the containment following a postulated loss-of-coolant accident including all applicable information specified in Section 6.2.53 of Regulatory Guide 1.70, Revision 3. l RESPONSE G.47 l l An analysis of the production and accumulation of combustible gases within the primary containment will be included in the recombiner of Subsection 6.2.5 referred to in response to Question 430.46. QUESTION G34 Regarding Control Room Habitability systems. (6.4) QUESTION G34a Provide the minimum positive pressure at which the control building envelope (which includes the mechacical equipment room) will be maintained with respect to the surrounding air spaces when makeup air is suppbed to the envelope at the design basis rate (295 CFM). RESPONSE M J4a As stated in Section 6.4.2.1, the control building will be maintained at a positive pressure of + 0.1 to + 0.5 in. I of water gage pressure at all times. The mechanical equipment room is maintained at +0.0 to +0.5 in. of water l QUESTION G54b Provide the periodicity for verification of control room pressurization with design flow rate of makeup air. RESPONSE 430.54b Subsections 9.4.1 and 73.1.1.8 will be provided by March 31,1989 concerning instrumentation of essential HVAC. QUESTION G34c l Clarify whether all potentist leak paths (to be provided in Section 9.4.1) include dampers or valves upstream of recirculation fans. I RESPONSE 430.54c Figure 9.4-1 (to be provided by March 31,1989), will show that the control room HVAC system provides two motor operated valves between all vent openings and their respective fans. l Amendment Putur 15 l l I

4 23A6100AT Eenndard Plant ms QUESTION G34d Identify the action to be taken when there is no Dow of the equipment room return fan and consequently the equipment room is overpressurized (Table 6.41 contains no information on the above). RESPONSE GJ4d Table 6.41 has been revised accordingly. QUESTION G3de Provide the actual minir.ium distances (lateral and vertical) of the control room ventilation inlets from major potential plant release points that have been used in your control room dose analysis. Also, provide a schematic of the locaten of control room intake vents. RESPONSE 43054e See response to Question 43034g. Actual distances will be provided in the reevaluation of the control room dose analysis scheduled for March 31,1989. QUESTION G34r Provide Figure 6.4 5 (plan view) which you state shows the release points (SGTS vent). RESPONSE 43054f i Reference changed from Figure 6.4-5 to 6.4-1 and Figure 6.4-1 (to be provided by March 31,1989) will be j revised to show reacf or building stack, QUESTION 43034g Section 6.4.2.4 and Figure 6.41 indicate Da'3 ong air inlet for supplying makeup air to the emergency zone. However, Table 6.4 2 and 15.6-8 and Section 15.6.5.5.2 indicate that there are two automatic air inlets for the emergency zone. Correct the above discrepancy as appropriate. Also describe the characteristics of these inlets with respect to their relative locations and automstic selection contr01 features. State how both flow and isolation in each inlet assuming single active component failure will be ensured. RESPONSE 43054g Response to this question provided in revised Subsection 6.4.2.4. QUESTION 43054h Describe the design features for protecting against confined area releases (e.g. multiple barriers, air flow patterns in ventilation zones adjacent to the emergency zone). RESPONSE 43054h Response to this question is provided in revised Subsection 6.4.2.4. Amendment Futur 16

ABWR msmr Standard Plant REV.B \\ QUESTION 430.541 Describe the specific features for protecting the control room operator from airborne radioactivity outside the control room and direct shine from all radiation sources (e.g., shielding thickness for control room structure boundary, two door venibules). i RESPONSE 430.541 The control room structural boundary has been designed to provide the necessary shielding for control room operators from direct shine. The control building is located between the reactor and turbine buildings. The caerior walls and ceiling of the control building are 0.9m (35") thick. In addition, enra shielding is provided by the reactor and turbine building exterior walls. The floor of the computer room is 2.0m (78*) thick for shielding between the control structure and the steam tunnel. The control room is maintained at positive pressure with respect to neighboring zones. In addition all doors are of the double door vestibule type. QUESTION 430.54j Cla#@ what you mean by

• Sustained occupanef (See SSAR Section 6.4.1.1, Item 3) for 12 persons.

RESPONSE 430.54j Food, drink, lavatory facilities are provided for in the emergency zone to provide life support for up to 12 operating personnel during an emergency for a period of 5 days. QUESTION 430.54k Provide justification for not specifying any unfiltered infiltration of contaminated air into the control room in SSAR Table 156-8. RESPONSE 430.54k The control room is maintained at a positive pressure with respect to atmospheric pressure. The only leakage paths are the doors. Because of the positive pressure maintained in the control room allleakage will be outward resulting in no inleakage to the control room. Amendment Futur 17

n

/V' a m. ,(, te MM 2sA61ooAT - Standard Plant

nev. a QUES 110N G.541 I

Provide Subsection 63.1.1.6 which you state (SSAR Section 6.4.6) eaatains a complete description of the required instnunentation for ensuring contral rcom habitability at all times. RESPONSE W.541. Reference changed from Subsection 63.1.1.6 to 73.1.1.8. QUESTION G.54m Give schematics for control room emergency mode of operation during a postulated IDCA (this is required for calculating control room IDCA doses). t RESPONSE G.54m For detailed information on the control room emergency zone HVAC operation see Subsection 9.4.1 and 73.1. QUESTION G.54n The source terms and control room atmospheric dispersion factors (X/Q values) used in the control room dose analysis (see $ lAR Tables 15.6-8 and 15.6-12) to demonstrate ABWR control room compliance with GDC 19 are non-conservative. Therefore, reevaluate control room doses during a postulated LOCA using RG 13 source terms and assumptions and the methodology given in Reference 4 of SSAR Sec.'.on 15.6.7. Include possible dose ] contributions from containment shine, ESF filters and airborne radioactivity outside the control room. Also check 3 and correct as appropriate the recirculation rate in the control room (22.4 m /sec) given in Table 15.6-8. 1 RESPONSE 430.54n The analysis for evaluation of control room doses to show compliance with GDC 19 has been done in a .j conservative manner but not in strict compliance with existing NRC guidelines. However, the methodology j employed utilizes the alternate calculational methodology of Paragraph 8.9 of the ABWR Nada: Review Bases l issued by Tom Marley on August 7,1987. A reanalysis using this alternate methodology will be redonc consistent ) with a redesign of the control room. Containment shine will be included in this reevaluation scheduled for j submittal by March 31,1989. ) QUESTION 430.54o Section 6A.7.1, External Temperature,'provides design maximum external temperature of 100 F and 10 F. How are these values used in the design and assessments related to the ABWR7 What factors, such as insulation, heat generation from control room personnel and equipment and heat losses, are taken into account? Do these j values represent ' instantaneous" values or are they temporal and/or spatial averages? i H RESPONSE 430.540 a Ther: values represent the summer maximum dry bulb air temperature. They are used in sizing the HVAC o essential chilled water system chillers and the control room HVAC system. I l Amendmest Futur 18

') s ABM t zu i aur Standard Plant _,.sg.a QUESTION 430.54p Garify your position on potentia! hazardous or toxic gas sou*ces onsite of an ABWR. If applicable, indicate the special features provided in the ABWR design in this regard, to ensure control room hahinahility. RESPONSE W.54p Response to this question is provided in revised Subsection 6.4.73. QUESTION G.54g Identify all the interface requirements for control room habitability systems (e.g., instrumentation for protection against toxic gases in general and chlorine in particular; potential toxic gas release points in the environs). RESPONSE G.54q 1he ABWR control room habitability system has no interface requirements. QUESTION W.56 Regarding Fission Product Control Systems and Structures. (6.53) QUES 110N G.56a Provide the drawdown time for achieving a negative pressure of 0.25 inch water gauge for the secondary - , containment with respect to the environs during SGTS operation. Clarify whether the unfikered release of redbedi dy to the environs during this time for a postulated LOCA has been considered in the LOCA dose analysis. (Note that the unfiltered release need not be considered provided the required negative pressure differentials achieved within 60 seconds from the time of the accident.) RESPONSE,G.56a No release of unfiltered fission products other than those released via the MSIVs is considered since based upon the conditions outlined in the response to Question 470.4 (Part 1) no release of fission products is expected from the core for one hour following accident initiation. QUESTION W.56b Provide justification (see SRP Section 6.53,114) for the decontaraination factor assumed in SSAR Tables 6.5 2 and 15.6 8 for iodme in the suppression pool, correct the elemental, particulate and organic iodine fractions given in the tables to be consistent with RG 13, and incorporate the correction in the LOCA analysis tables. - Ahernatively, taking no credit for iodine retention in the suppression pool, revise the IDCA analysis tables. Note l that the revision of the I.OCA analysis tables (this also includes the control room doses) mentioned above is strictly i in relation to the iodine retention factor in the suppression pool (also, there may be need for revision of other parameter (s) given in the tables and these will be identified under the relevant SRP Sections questions). Aswadesat Putur 19 i (

i l 6 ~ 21A6100AT IEf asadard Plant nev. s RESPONSE m.56b The LOCA analysis found in Section 15.6 is performed in accordance with paragraph 8.9 of the IJcensing Review Bases document. An evaluation of suppression pool scrubbing using the MAAP3B code for LOCA conditions shows a scrubbing factor of 6001000 (Subsection 19E.2.1). Therefore the use of a scrubbag factor of 100 is sufficiently conservative. The variance between the current calculations and the prior evaluation methodologies in found in Table 20.31. QUES 110N m.56c Identify the applicable interface requirements. RESPONSE W.56c %ere are no interface requirements. QUESTION &J8 The accident analyzed under this section considers only the airborne radioactivity that may be released due to potential failure of a concentrated waste tank in the radweste enclosure. The SRP acceptance criteria, however, requires demonstration that the liquid radwaste concentration at the nearest potable water supply in an unrestricted area resulting from transport of the liquid radwaste to the unrestricted area does not exceed the radionuclides concentration limits specified in 10 CFR Part 20, Appendix B, Table II, Column >2. Such a demonstration will require information on possible dilution and/or decay during transit which, in turn, will depend upon site specific data such as surface and ground water hydrology and the parameters governing liquid waste movement through the soil. Additionally, special design features (e.g., steelliners or walls in the radwaste enclosure) may be provided as - part of the liquid radwaste treatment systems at certain sites. The staff will, therefore, review the site specific characteristics mentioned above individually for each plant referencing the ABWR and confine its review of ABWR, only to the choice of the liquid radwaste tank. Therefore, provide information on the following: (15.73) a) Basis for determining the concentrated waste tank as the worst tank (this may very well be the case, but in.the absence of information on the capacities of major tanks, particularly the waste holdup tanks, it is hard to conclude that the above tank both in terms of radionuclides concentrations and inventories will turn out to be the worst tank). b) Radionuclides source terms, particularly for the long-lived radionuclides such as Cs-137, and Sr 90 (these may be the critical isotopes for sites that can claim only decay credit during transit) in the major liquid radwaste tanks. RESPONSE W.58 The scope of the ABWR SSAR has recently been extended to include the Radwaste Facility. A new Subsection 15.73 analysis will be submitted along with the completely revised Chapter 11 submittal scheduled for March 31,1989. It is anticipated that only airborne releases will be considered since the radwaste tanks will be located in a Seismic Category I steel lined Radwaste Building substructure which will prohibit any liquid release. Asundment Futur 20 ~

vs MM 2sA61ooAT Standard Plant ann QUESTION 440J0 Steam isolation valves P063 and P064 are to be opened it sequence to reduce water hammer and for slow warm up of the piping. P064 and P076 are opened first. The valves logic should prevent the operator from opening the valves out of sequence. Confirm that the valves controllogic includes an interlock.. RESPONSE 440J0 GE does not consider that an interlock is necessary since opening of these valves is governed by strict administrative and procedural control. The ABWR a lesign specification requires these valves to be opened sequentially as stated below: ' 'In order l', prevent damage from water hammer, neither steam innIntian valve is opened automatically by an initiatics signal. Should either or both of these valves be closed, they must be r@ by first closing both valves completely. With both valves closed, the outboard isolation valve P064 can be rW to allow any moisture in the line to drain. Then, moisture ahead of the inboard isolation valve P063 is drained slowly as line pressure across inboard isolation valve is equalized and the downstream line is warmed by slowly opening the inboard isolation valve bypass valve P076. Pically, the inboard isolation valve P0f3 may be reopened.' QUESTION 440J1 Describe how the system design reduces water hammer. Confirm that a condensing sparger will be provided at the turbine exhaust to reduce water hammer. Add a necessary note in the P&ID to indicate that the steam supply and exhaust lines are to be sloped to reduce water hammer. RESPONSE 440J1 The RCIC steam supply piping is sloped downwards with drain pots and steam traps upstream and downstream of the turbine. During normal reactor operation steam condensate are continuously drained and the steam traps automatically isolated when RCIC is initiated. Also,in the turbine exhaust line a condensing sparger is provided to reduce water hammer, minimize line vibrations and reduce noise levels. In addition, the turbine exhaust line is installed above the maximum water level of the suppression pool. It is also provided with vacuum breakers, to relieve the exhaust line from pressure instability that can cause water hammer. Addition of ac4es in the P&ID is not required since these requirements are part of the ABWR RCIC design specification as given below: " System piping shall be arranged to provide a continuous dov nward slope as follcws: Steam supply line from the main steam line to the drain pot, just ahead of the turbine. RCIC steam a. inboard isolation valves " Bypass, Drain and Warm-up line" shall be sloped parallel with section of line and valve it is bypassing to prevent a dead leg from occurring b. Turbine exhaust line - from the upstream side of the check valve to the turbine exhaust drain pot and downstream of the check valve to the werwell. i Amendment Futur 21

M\\ 23A6100AT EfenAmrd Plant arv n

e.. Vacuum pump discharge line - from the upstream side of the check valve to the vacuum pump and

] &mastream of the check valve to the suppression pool.

d.. The gland seal equipment shall be located such that there are no water pockets between this equipment and the turbine. Paping between the turbme and the gland seal equipment shall be kept to a minimum."

QUESTION 44040 In SSAR Table 13 2, it is stated that the RHR heat exchanger duty for suppression pool coohng is based on assuming they are placed in operation 20 hrs after reactor shutdown. This statement is not consistent with the normal assumption that suppression pool coohng is stated within ten minutes after a LOCA. What is the basis for sizing the RHR Hx? In SSAR Chapter 5.4.73.2, it is stated that ATWS was considered for RHR beat enchanger sizmg. But a Feedwater line Leak (PWLB) is the most limiting event. Describe in detail why FWlE is the knuting event and not ATWS. RESPONSE 44040 The note at the bottom of page 13-13 of Table 13-2 is incorrect. It has been corrected to read: *... Heat =A-

er duty at 20 hours following reactor shutdown."

%e basis for sizing the RHR heat =Amaper I. to limit the long term pool temperature to a maximum of 207 F for the most limiting events. Analym show that the hedwater line break (FWLB) is the most limiting event with the assumption of a 10 minute operator action time to start pool cooling. The ATWS ewnt is less limit *mg than the FWLB because the ATWS high-pressure signal causes the electricalinsertion of the fine-motion control rod drives which leads to a hot shutdown condition within less than 3 minutes. W;th 10 minute operator action time to initiate pool cooling, the maximum pool temperature is much less than 200 F. QUESTION 44041 In SSAR Chapter 5.4.7.3.2, Section 2, it is stated "because it takes 4 to 6 hrs to reach the peak pool temperature, shutdown coohng will be initiated before peak pool temperature. The energy release from the reactor will be controlled by the shutdown cooling system, and there is no need to release the reactor energy to the pool." Which scenarios are postulated for the assumption stated above? For most scenarios, suppression pool coohngis started within a short time. Shutdown cooling is started at a much later stage. Describe in detail the assumptions made for sizing the RHR heat exchangers. RESPONSE 44041 The assumption stated in subsection 5.4.73.2(2) is not used in any of the analyses for determining the required heat ~4 ager capacity. It is a statement which points out one of the conservatism of the analyses by not accounting for this effect. As stated in the response to Question 440.60, the analyses use the assumption that suppression pool cooling is initiated 10 minutes after start of the event. Other assumptions utilized in the FWLB analyses (which is the limiting event for the heat evaanger sizing) are: 1-22 Amendment Futur

J 1 MM 2sA6tooAT Standard Plant mzvy (1) Reactor power is 102% of rated, (2) One RHR heat exchanger failure, (3) ANS 5+20/10 decay heat, (4) 164% rated feedwater flow for 120 seconds, (5) MSIC closure 3.5 seconds following scram, (6) Reactor water level at normal operating level, (7) Suppression pool water level at minimum level, (8) Initial suppression pool temperature at the maximum operating value of 950F, (9) Maximum service water temperature at 84.20F, (10) Maximum pool temperature limit of 2070F. QUESTION 440.65 In Section 5.4.7.2.3.1(3) it is stated that " redundant interlocks prennt opening the shutdown connections to and from the vessel whenever the pressure is above the shutdown range? RSB 5-1 requires that the suction and discharge valves interfacing with the RCS shall have independent - drverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR design pressure. Confirm that the high/ low pressure interface with RCS satisfies the requirements of RSB 5-1. RESPONSE 440.65 The valves that interface with the RCS are the shutdown cooling valves on the suction side and the injection valves on the discharge side. RSB 5-1 has different requirements for the suction and discharge conditions. Diverse interlocks are mentioned for the suction side valves. The ABWR RHR design does not explicitly meet this requirement for diversity; it does meet the intent of the requirement to provide high reliability against inadvertent opening of the valves. The pressure signal that provides the interlock function is supplied from 2-out-of-4 logic, which has four independent pressure sensor and transmitter inputs. The independence is provided by each being in a separate instrument division. The 2-out-of-4 logic provides equal or better reliability than the existing licensed BWR plants. Furthermore, the inboard and outboard valves of a common shutdown cooling suction line are operated by different electrical divisions. The discharge side requirements of RSB 5-1 allow a choice between four different configurations. ABWR utilizes the second selection, which is one or more check valves in series with a normally closed power-operated valve. l Amendment Futur 23

MM 23A6100AT Flandard Pinnt arv. s QUESTION 440.72 l NRC Bulletin 88-04 dated Iday 5,1988, *=- the potential safety related pump loss. The first concern involves the potential for the dead-heading of one or more pumps in safety related systems that have a minin w line o common to two or more pumps or other configurations that do not preclude pump-to-punp interaction during miniflaw operation. A second concern is whether or not the installed miniflow capacity is adequate for even a single pump in operation. In the ABWR design, HPCS pump miniflow lines and test return lines to the suppression pool are routed through the RHR 'c' loop test and minimum flow lines. How does the ABWR design satisfy the concerns given in NRC Bulletin No. 88-047 RESPONSE 440.72 The ABWR RHR and HPCF systems use a common minimum flow return line to the suppression pool, which is one of the topics addressed by NRC Bulletin No.b88-04 The ABWR design has a very low pressure drop associated with the pipe down stream of the junction where the RHR and HPCF begin to share the common pipe. Because of this, a pressure cannot be created in the down stream shared line that could result in either one of the pumps discharging into a pressure that could dead head a pump. When both RHR and HPCF pumps are running simultaneously at minimum flow the Dow losses are approximately 1 psi and the static head for the discharge pipe's submergence in the suppression pool is approximately 7 psi. Even for the case (which is unplanned) where both pumps are simultaneously operating at rated flow, the total pressure drop down stream of the unction is approximately 25 psi. The RHR and HPCF approximate pressure heads at rated flow are 178 psi and 270 psi respectively, which are significantly above the common down stream pressure that could potentially dead head a pump. The other concern related to NRCB 88-04 is the minimum flow percentage of rated flow. The ABWR values are 15.5% and 10% for RHR and HPCF respectively. BWR/6 plants have specified ECCS pumps at 10% minimum flow and operational problems have not been encountered. QUESTION 440.73 In RHR process diagram 5.4-11b, RHR heat exchanger heat removal capacity for different modes is not given. Revise the process diagram to include the heat removal capacity. RESPONSE 440.73 Figures 5.4-11b and 63 3b will be revised to show the RHR heat exchanger removal capacities for the different modes at their next revision as indicated in Figures 203-19a and 203-19b, respectively. QUESTION 440.77 Confirm that the HPCF system meets the guidelines of Regulatory Guide 1.1 regarding pump Net Positive Suction Head (NPSH)(63). RESPONSE 440.77 Response to this question is provided in revised Subsection 6.2.23.1 and new Table 6.2-2c. Amendment Putur 24 i

MN 22A61ooAT Sisadard. Plant Prv n QUESTION 440.95 List the capacity and setting of all relief valves to be grovided for the ECCS to satisfy system overpressure. RESPONSE 440.95 The ECCS relief valve capacities and settings are provided below: A. RHR-Valves F031A, B, C - Discharge Line Capacity: linch pipe size Setting: 500psig Valves F008A, B, C - Near Heat Exchanger Capacity: 1 inch pipe size Setting 500psig Valves F068A, B, C - Suction to Reactor Capacity: linch pipe size Setting: 200 psig Valves F015A, B, C - Suction to Suppression Pool Capacity: 1 inch pipe size Setting: 200 psig Valves F035A, B, C - Heat Exchanger Secondary Inlet Capacity: 1 inch pipe size Setting: 200 psig B. HPCF Valves F018B, C Discharge Line Capacity: 1 inch pipe size Setting: 1565 psig Valves F014B, C Suction to Suppression Pool Capacity: 1 inch pipe size Setting 200psig l i I Amendment Futur 25

+ l. MM 2M6100AT l Standard Plant arv. s l l-C. RCIC 1 Valve P017. Pump Suction Capacity: ' inch pipe size Setting: 200 psig Valves P018 - Cooling Water Capacity: 2 inch pipe size Setting: 75 psig Valves P033-Vacuum Tank Capacity: 1 inch pipe size Setting: 5 to 7 psig QUESTION 440.97 SSAR Tabic 5.4-2 gives the design parameters for RCIC system components. Provide similar information for RHR and HPCF systems. (63) RESPONSE 440.97 Response to this question is provided in revised Subsection 63.2.2 and new Tables 63-8 and 63-9. Anwedment Futur 26 I J

ABM 2mmr Standard Plant nrv. n QUESTION 460.1 With respect to radioactive source terms and the calculations of subsequent release to the environment, discuss your position in terms of the regulatory guidance provide in NUREG 0800, SRP 11.1, such as NUREG-0016,

• Calculation of Releases of Radioactive Materials in Gaseous and Uquid Effluents from Boiling Water Reactors," Revision 1 and Regulatory Guide 1.112, ' Calculation of Release of Radioactive Materials in Gaseous and Uquid Effluent from Light Water-Cooled Power Reactors." (11.1)

RESPONSE 460.1 NUREG-0800, Section 11.1 and Regulatory Guide 1.112 identify the BWR-GALE computer code as an acceptable method of calculating releases of radio-etive materials for demonstration of compliance with applicable regulatory requirements. GE does not use the BWR GALE computer code per se; however, the methods and assumptions embodied in the code are used to evaluate expected releases in gaseous and liquid effluents except in those specific instances where alternate parameters or modelt are regarded as more appropriate. QUESTION 460.2 Clanfy whether the radioactive source terms given in ABWR SSAR Tables 11.11 through 11.15 have been adjusted to the maximum core thermal power of the ABWR evaluated for safety consideration in the SSAR. (11.1) RESPONSE 460.2 The data contained in Tables 11.1-1 through 11.15 are design basis source terms for use in plant design. As such they are not predicted source terms at a specified power level. These are intended to be conservative source terms chosen to provide substantial margin relative to expected average source terms for long term opert, tion at rated power. These tables have been revised in their entirety. The revised tal;les contain radionuclides mixes derived from the American National Standard Radioactive Source Term for Normal Operation of Light Water Rea: tors (ANSI /ANS-18.1 1984). The bases for the new tables include normalization of the magnitpdes of the noble gas release rates to a total of 100,000 uCi/see as evaluated at 30 minutes decay and normalization of the magnitudes of reactor water radioiodine concentrations based on an 1-131 release rate from the fuel of 700 uCi/sec. i QUESTION 460.3 Check and correct as appropriate the following: 1. Caption for Column 3 of SSAR Table 11.1 1, 2. Kr-87 value given in Column 4 of SSAR Table 11.1-1. 3. N-16 steam and reactor water concentrations given in SSAR Table 11.1-4. (11.1) RESPONSE 460.3 The two items noted in Table 11.1 1 have been corrected. As indicated in response to Question 460.2, Table 11.1 1 has been revised in its entirety. Table 11.1-4 has similarly been revised in its entirely, and now contains a revised concentration of N-16 in l redor water. I l Amendment Putur 27 l

M 23A6100AT Standard Plant arv. s QUESTION 460.4 The staff requires the values of some parameters for performing an independent evaluation of the ABWR Reactor Coolant System (RCS) radioactive source terms. These are used in conjunction with radwaste management systems applicable for specific plants referencing the ABWR to determine the adequacy of the specific radwaste management systems (see NUREG-0016, Rev.1, Chapter 4). Therefore, provide information on the following parameters or provisions: (11.1) 1. Thermal Power (Mwt) 2. Total steam flow rate (Ib/ht) 3. Mass of water in the RCS (lbs) 4. Steam / water concentration ratio,i.e., reactor vessel carry over factor for halogens and particulate 5. Main condenser tubing material of construction (stainless steel or copper) 6. Powdex or deep bed condensate treatment 7. Air ejectro offgas holdup time (hr) 8. Charcoal delay system for treating offgases: Operating and dew point temperatures of the delay system a. b. Mass of charcoal (Ibs) 3 c. Dynamic adsorption coefficients (cm /g) for Kr, Xe and Ar 9. Clean or radioactive steam for gland seal

10. Mechanical vacuum pump iodine release fraction if within the ABWR scope

11. Provisions incorporated to reduce radioactivity releases through the ventilation or exhaust systems that come within the ABWR scope (e.g., HEPA filters, charcoal adsorbers and their thickness)

12. Release points characteristics (see NUREG 0016, rev.1, Chapter 4, section 4.7, Item 4). Include description of the main stack.

Note that item Nos. 5 and 6 are required to determine the carry over factor for radiohalogens from reactor water to steam. Also, note that when a summary of building ventilation system and mechanical vacuum pump releases are provided in December 1988 (see SSAR Section 12.33), additional information on the releases may be requested. RESPONSE W.4 Two new tables have been a3ded to revised Section 11.1 to better describe the basis for the liquid and gaseous source term. The remaining items requested pertain to Sections 11.2 and 113 which will be provided by March 31,1989. I i Amendment Putur 28

23raoorr ABM . REV B Standard Plant QUESTION 471.1 Section 12.1.1.2 of the submittal states that operational policies are out of the Nuclear Island scepe. Following Section 12.1.13, the report states that

• Compliance of the Nuclear Island design with Title 10 of the Code of Federal Regulations Part 20 (10 CFR 20),is ensured by the compliance of the design and operation of the facility within the guidelines of Regulatory guides (R.G.) 8.8,8.10, and 1.8" Further, Section 12.1.13.2 states that R.G. 8.10 is out of the Nuclear Island scope and section 12.1.133 states that R.G.1.8 is out of the Nuclear Island scope. The applicant should clanfy these statements by describing to what degree the guidance in R.G. 8.8,8.10, and 1.8 incorporated into the ABWR design.

RESPONSE 471.1 Section 12.1 has been revised. Regulatory Guide 8.10 deals solely with operational aspects and personnel and is not related to the design activity. Therefore since the designer has no authority over the training and operation of the faciFty, regulatory Guide 8.10 is out of scope for purposes of this submittal. IJkewise Regulatory Guide 1.8 conceres the qualification and training of operating personnel and is beyond the authority of the designer and therefore out of scope for this submittal. The intsal of Regulatory Guide 8.8 has been met, but those areas concerning actual operational management, training, and procedures have not been discussed as out of scope. QUESTION 471.2 Section 12.1.2.2.l indicates that in lieu of specific instructions, design engineers were instructed to incorporate the applicable design criteria in R.G. 8.8. What mechanism in the design process ensured that d applicable criteria were considered in the individual design. RESPONSE 471.2 To insure that the criteria of Regulatory Guide 8.8 were followed during the design of the ABWR, a continuing review and analysis of occupational exposure was conducted throughout all design phases of the ABWR. These reviews consisted of a team of engineers from GE and its technical associates who during each phase established projected maintenance times and exposures based upon experience with similar tasks and equipment from BWR's thfoughout the world. From these reviews recommendations were made for changes in location, shielding, equipment, and automation to reduce overall occupational exposure. The word d has been removed from Subsection 12.1.13.1 since it implies that the design activity has authority over administrative procedures which is clearly in the domain of operation policier. QUESTION 4713 Section 12.1.2.2.2, paragraph (4) states *Past experience has been factored into current designs. The steam relief valves have been redesimed as a result of inservice testing. Access for inservice inspection has been changed." state in what respect the access has been changed, and what is the impact of this change on occupational radiation exposure (ORE). 29 Amendment Futur - ' ~ - - - - -

i l ABWR ~ m ur i Standard Plant l ann l l RESPONSE 4713 The SRVs employed are direct action SRVs as opposed to pilot valve operated SRVs. The SRVs have been placed circumferential around the pressure vessel in an access hallway with dedicated hoist for removal. The overall reduction in occupational exposure on SRV maintenance (due to all factors including lower radiation due to recirculation removal and pipe placement) is estimated at slightly less than a factor of four, i l QUESTION 471.4 Section 12.1.2.23, paragraph (4), last sentence is not completed. It states: 'These systems are designed to limit the radioactive'. Complete the sentence. RESPONSE 471.4 The paragraph has been completed. QUESTION 471.5 Section 12.1.2.23 should address the reduction of personnel exposure due to the elimination of external primary coolant recirculation loops in ABWR design. RESPONSE 471.5 Response to this question is provided in new Subsection 12.1.2.23(5). 1 QUESTION 471.6 Section 12.1.23.2, paragraph (5) refers to

• packaged units". State whether or not this includes skid mounted components; if not, define what is meant by " packaged units".

RESPONSE 471.6 Packaged units refer to pieces of equipment or components which can be removed as a single unit from areas of higher radiation to maintenance areaa for reconditioning. A good example would be the reactor internal pumps which can be removed for maintenance. Such components need not be skid mounted. QUESTION 471.7 Section 12.1.23.2, paragraph (8) refers to providing means for decontamination of service areas; clarify the statement by providing examples of means of decontamination and of service areas referred to. RESPONSE 471.7 Specific areas of the reactor building are allocated for equipment maintenance. Those areas are provided with radwaste drains and sumps so that they may be decontaminated after use. Such an area would be the control rod drive maintenance area. Amendment Futur 30 ~

MM 2sA6100AT Standard Plant ma QUESTION 471J Table 12.2-3 part A, shows gamma ray sources in the core during operation (MeV/see W) for various energy bound (McV). Provide the basis for these data. RESPONSE 4712 The gamma spectral data of GESSAR II (Chapter 12) was utilized. The ABWR core is rated at. approximately the same power density as GESSAR II and therefore would exhibit very nearly the same core gamma spectra. A separate two dimensional transport calculation for ABWR is be'ag made and these tables will be revised accordingly in a future amendment. QUESTION 471.9 Table 11.14, Coolant Activation Products in Reactor Water and Steam, indicates that values in steam for N-13, N 16 and N-17 should be multiplied by a factor of four wher, hydrogen water chemistry is used. Explain why the concentrations of these isotopes remain unaffected in reactor water when hydrogen water chemistry is used. State whether the effect of hydrogen water chemistry on the nitrogen activation product concentration (increase by a factor of four) was incorporated into the plant shielding design. RESPONSE 471.9 The design basis reactor water concentration is based on normal water chemistry. The concentration is somewhat reduced'during operation with hydrogen water chemistry due to the increased steam carryover; consequently, use of the concentration for normal water chemistry is conservative for this condition. Plant chielding is being designed to accommodate the increased coolant activation product source.. concentration in steam associated with the use of hydrogen water chemistry QUESTION 471.10 Address'in Section 12.1.2.2.3 the selection of materials and instrut entation with respect to radiation exposure damage, frequency of maintenance, and AIARA personnel radiation exposure (i.e., reactor coolant pump component and material sele: tion). RESPONSE 471.10 A radiation environment envelope is included as part of the purchase specification for the pump. This specification details the environment which the pump must survive along with a frequency of maintenance schedule. Evaluation of radiation exposure for this type of recirculation pump was based primarily on European experience as well as maintenance practice. QUESTION 471.11 Provide drawings of cross sections R4 (0 180 ) and RD (90 270 ) of the ABWR reactor buNing for better orientation of radiation zones. l Amendenent Futur 31

MM 2sA61ooAT. Es*=dard Plad - nev. a e RESPONSE 471.11 These drawings will be provided by March 31,1989. j QUESTION 471.12 ' In Table 11.1 1, Noble Radiogas Source Terms (Steam), (Page 11.19) revise the description of first column: Source Term (t = 30 min) to Source Term (t = 0 min). RESPONSE 471.12 The indicated column description was erroneous. As noted in the response to Questions 460.2 and 460.3, Table 11.1 1 has been revised in its entirety. QUESTION 471.13 - In accordance with section 12.2.2 of R.G.1.70, provide a descriptica of radioactive sources in the spent fuel. pool. This description should _ include the expected radioactive concentrations in the spent fuel pool water, as well as contained sources within the pool. RESPONSE 471.13 l Subsection 12.2.2 will be updated to incorporate revised Section 11.1 and the requested data by March 31, 1989. QUESTION 471.14 In accordance with R.G.1.70, Section 12.2.2, Airborne Radioactive Material Sources, provide average expected annual airborne concentrations, at normal operating and anticipated operational occurrences, in various areas of the plant nosmally occupied by operating personnel. RESPONSE'd1.14 See response to Question 471.13. QUESTION 471J5 Provide the expected N 16 source strength increase in steam leaving ABWR pressure vessel due to the eliminatica of the external reactor coolant loops in comparison to BWR plant with emernal coolant loops of the j same power level. Provide the melbod of calculation. RESPONSE 471.15 The N 16 source concentration in steam leaving the ABWR pressure vessel is not expected to increase relatne to existag BWRs as a direct consequence of eliminating the external recirculation loops. This is consistent with a model in which the volatile porten of the N-16 produced in a single pass through the core is stripped out and carried from the vessel with the steam. The source term provided in the American National Standard Radioactive Source Term for Normal Operation of Light Water Reactors (ANSI /ANS-18.1 1984) is taken to be applicable. Amendment Putur 32 l

.e4 - 4 = ABWR msmr Standard Plant man QUEST 70N 471.16 Address TMI issues in accordance with NUREG-0737 as it relates to ABWR design hrelan 12. RESPONSE 471.16 - .TMI action plan items are addressed in Appendix 1A. This includes the f:41owing pertaining to Chapter 12: H.B.2, D.F.1(1) and III.DJ3(3). QUESTION 471.17 Provide the missisg Table 12.2-1, A, B, C, D; Table 12.2-3, C; and Table 12.2-4, B; Tables 12.2-5 through tables 12.2-21; Section 1233; Section 123.4; Table 123-3; Figures 12.3-8 through 123-23; and Section 12.4. RESPONSE 471.17 'Ihis information will be provided by March 31,1989. QUESTION 471.18 Describe the radiologicalimpact of each of the advanced design features of the ABWR design. Show how .AIARA considerations were engineered into these features by describing the source term, reliability, maintenance and surveillance associated with these components. Features discussed should include, but not be limited to, the i internal reactor circulating pumps, the control rod drive mechanisms, hydrogen water chemistry and reactor vessel l bottom head design. I RESPONSE 471.18 The ABWR employs a number of advanced features designed to improve plant operation and reduce operator exposure. The discussion below highlights those features which reduce exposure. (1) MELY maintenance is expected to be reduced by use of MSIV overhauling device, use of main steamline plugs, automatic MSIV lapping system, elimination of the external recirculation pip *mg, and l RHR return through the feedwater system. Use of automated systems will result in overall reduction in maintenance times of 40%. Removal of recirculation lines will reduce this contribution to zero while it is estimated that the average field will be 4 mrem /hr for this activity. (2) SEY maintenance will primarily be affected by better accc:.s to SRVs and a reduction in the radiation field by elimination of the recirculation piping. Improved access has not been accounted for in this estimate with the radiation field estimated at 5.5 mrem /hr. (3) CEI2 maintenance will be significantly reduced with an assumed maintenance schedule of 2 drive and spool pieces overhauled per year and 20 drive motors inspected per year. A semi automated handling machine will be used with the maintenance crew reduced from five to four. A effective (averaged over all conditions) dose rate of 17 mrem /hr is estin;ated. (4) RIP /RHR / Heat exchanners assume 3 pumps and 1 heat exchanger will be serviced per year. Estimates are based upon European experience assuming an effective dose rate of 40 mrem /hr. Amendment Putur 33

1 1 23A610MT Standard Plant arv. s RESPONSE 471.18 (continued) (5) la-vice Inspection is reduced by removal of recirculation line, elimination of 2 = ante inspections per year, elimination of shield penetrations and plug removal associated with lines, improved automation equipment, and reduced vessel welds. -(6) General Drvwell work is reduced by removal of recirculation line, reduced

• specion of snubbers, and m

associated components. (7).Ysag] access and assembly will be reduced by the use of stud tensioner over % volts. Estimated dose rate is 3 mrem /hr. (8) Refuehng will be reduced by use of an automated refueling platform where no personnel are located on the platform itr, elf. Overall reduction is estimated at 20%. (9) Fuel Sionine is delete based upon improved fuels. (10) CRD rebuildine is reduced based upon rebuilding 2 drives per year. (11) RWCU assumes 2 pump inspections per year with canned motors based upon use of an improved motor design. The table below delineates projected occupational exposure for a nominal operating year (does not attempt to evaluate off nominal occurrences). Exposure &ca 11cm (manrem /vri Drywell MSIV 21.8 SRV 1.4 CRD 5.2 LPRM/TIP 6.7 ISI 72 RIP /RHRx 7.6 Misc 4.5 Instruments 3.1 Other 23 Anwedment Putur 34 1 l 1 _--_____J

ABM av.6toorr Standard Plant REV. B RESPONSE 471.18 (continued) Exposure MC2 lism (manem /vr) R/B Vessel 2.4 Refueling 5.1 RPVinspection 0.8 Fuelinspection 03 Fuelsipping 0.0 CRD rebuild 03 RWCU 1.0 RHR 5.6 Other 13.8 T/B Turbine overhau! 3 Valves / pumps 5 Condensate 3 Other 12.9 .RW/B Radwaste 20 Work at Power 33 Total 177 QUESTION 471.23 The last paragraph in Sadon 123.13 states that all arets with radiation levels greater than 100 mrem /hr will be locked. This implies that the ABWR design will not incorporate Standard Technical Specification 6.12 which allows areas to remain unlocked up to 1000 mrem /hr. If this is not the case the radiation zone maps should be revised to identify zones with radiation levels greater than 1000 mrem /hr. RESPONSE 471.23 'Ibe cutoff for ABWR is 100 mrem /hr. Amendment Futur 35

l i ABM 23462o041 Standard Plant arv. s QUESTION 471.24 Figures 12.3-3 and 12.3-4 show two stairwells on the cast side of valve rooms B and C where the access to and from the stairs is in a low radiation zone (_ Smrem/hr), but the stairwell itself is in a high radiation area (/100 mrem /hr). Justify why additional shielding in the stairwell to prevent an unwarranted high radiation traffic area is not reasonably achievable. RESPONSE 471.24 1 I Access in the stairwell to the valve rooms is via a labyrinth passageway. Revised radiation zone drawings (to be provided by March 31,1989) show this passageway more clearly. QUESTION 471.25 ~ Figure 12.3-2 also shows an arrangement where one would have to traverse a high radiation area to get to a l low radiation area. The insert diagram in Figure 12.3-2 shows that access to the TIP drive room (_5 mrem /hr) is through a room on the south side which is a high radiation area. Justify the planned access to the TIP drive room. Also provide the zoning layout for the entire 1500 mm level. I RESPONSE 471.25 i The figure is incorrect. It will be revised and submitted by March 31,1989. QUESTION 471.26 The first paragraph at the top of page 12.3-13 states that in the event of a complete TIP retraction, egress from the TIP room is possible with less than 100 mR radiation exposure. What features has the ABWR design incorporated to ensure that exposures received from the recovery from this event are ALARA7 RESPONSE 471.26 The giaragraph will be revised to move clearly indicate egress and the associated drawing revised by March 31,1989 indicating access. QUESTION 471.27 Page 12.3-13 of the submittal has a statement that a concrete CRD storage vault, used for storing CRD parts and assen, bled units,is provided in the CRD maintenance room. This design feature is not indicated in Figure 12.3-2. Provide a figure depicting this design feature. Discuss the anticipated source term within this vault and associated shielding requirements. RESPONSE 471.27 The CRD storage tanks will be shown on the revised Figure 12.3-2 to be provided by March 31,1989. 1 J Amendment Putur 36 I i 1

ABM 23AM00AT Standard Plant arv n QUESTION 471.29 Table 12.3-1 lists five computer shielding codes used in ABWR design. The last entry in the table states 1 ' Additional Codes to be added by Applicant.' Identify these codes and give a full description of their application to the ABWR design, or clarify the use of the term ' Applicant". RESPONSE 471.29 Applicant refers to the A/E responsible for designing the balance of plant. In regards the ABWR submittal, I this involves codes and procedures used primarily in the turbine building which will be submitted when available. l QUESTION 47131 Sections 12.2.2.1 and 123.2.2.2 state that the ABWR shielding design is based on a fission product release rate of 50,000 mci /see of noble gas after a 30 minute decay time. The standard assumption (See Standard Review Plan p.12.2-4) is 100,000 mci /sec. Justify the use of this much lower source term. RESPONSE 47131 The noble gas release reference of 0.05 Curie per second was based upon improved fuel performance for BWRs employing only 8X8 improved fuel and is in closer agreement with NUREG-0016. However, to provide a more conservative basis for shielding, the reference has been revised to the older 0.1 Curie per second noted in SRP 12.2-4. This has been changed throughout the SSAR. QUESTION 47133 The acceptance criteria for radiation streaming through reactor shield wall penetrations (on page 123-12) is unclear. Describe the radiation streaming through reactor shield wall penetrations during refueling operations for all feasible fuel configurations. RESPONSE 47,133 Subsection 123.23 has been revised to clarify the shielding requirements and will be provided by March 31, 1989. QUESTION 47134 L Section 123.5 of the submittalidentifies areas requiring access to mitigate the consequences of an accident. Indicate whether this is a complete list of the vital areas (as described in item 11 B.2 of NUREG-0737) of the facility. If not, identify the vital areas of the facility; and if so, justify why the post accident sampling station (PASS) and the counting rooms are not considered vital. RESPONSE 47134 This list is valid for the reactor building only. Since the scope of the SSAR has been expanded, the list has been revised and will be provided by March 31,1989. Amendment Futur 37

MM 2sAsuxwr ma=dard Plant arv. s QUESTION 47136 Between pages 12.3-6 and 12312, it appears that the designators RWSC, RWCU and CUW are all being used interchangeably to refer to the Reactor Water Clean Up System. Verify which is correct and delete or define the other acronyms used. RESPONSE 47136 These pages will be revised accordingly and provided by March 31,1989. QUESTION 47137 The fourth paragraph at page 123-9 indicates that drains from the SGTS filter housing will be piped directly to a floor drain sump. Industry experience has shown that these housing drains can provide bypass pathways around fiker/ absorber beds. Provide a description of the ABWR SGTS and Control Room filter housing drains showing how fiker bypass is prevented. RESPONSE 47137 The fourth paragraph on page 123-9 requires clarification,'A connection to tbc filter train is provided so that a water source may be correded in the event the charcoal catches fire. However, the design basis for the cause of the fire is not failure to remove decay heat, but rather accidental ignition dering maintenance or other periods not coincident with a LOCA. Decay heat (from iodine) is removed by forced convection from the redundant process fans located upstream of the filter train. These process fans are powered from separate divisions of diesels." As such there is no open line piped to a flow drain sump, as stated in Subsection 123.1, from the filter train. There are loop seals located in the drains from the moisture separators in the Dryer Units and also in the line at the base of the stack. The downstream loop rend is provided with redundant levelinstrumentation to provide assurance that a net flow out of the secondary containment is maintained. QUESTION'47'1.39 Provide layout drawings of the control room showing radiation zones during normal operation, anticipated operational occurrences, and design basis accidents. Shield wall thickness, calculational parameters (and assumptions), and the models used to determine compliance with GDC 19 should be indicated. RESPONSE 47139 These drawings will be provided by March 31,1989. QUESTION 471.41 Figure 1234 shows several small A zones ( 0.6 mrem /br) completely surrounded by higher level zones (C rones). What are the purposes of these areas and justify why continuous A zones cannot be provided. RESPONSE 471.41 Revised drawings will be provided by March 31,1989. Amendment Futur 38

~ 23A6100AT Standard Plant arv, n QUESTION 910.7 The ABWR IJcensing Review Bases document states in its section 7.1 that the importance of such potential contributors to severe accident risk as sabotage should be carefully analyzed and considered in the design of new plants. To permit our review of this analysis and considerations, please provide a Aar===iaa of the insider and outsider sabotage actions that would be necessary to cause significant core damage or Part 100 release levels. This discussion should include identification of the ABWR design features that decrease relian:e on physical security programs for sabotage protection. (13.6.1) RESPONSE 910.7 Appendix 19C,' Design Considerations Reducing Sabotage Risk

• will be submitted with the balance of Chapter 19. Thir appendix includes identification of the ABWR design features that decrease reliance on physical accurity programs for sabotage protection. It will consider both insider and outsider sabotage actions.

i

1. .g e l

l i I l l l l l Amendment Futur 39

MM 21A6100AC W Standard Plant mEv c TABLE 13-2 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTIC (Continued) His Plant GESSAR NMP2 Gened Gulf ABWR BWR/6 BWR/S BWR/6 Eralgm/comnonent 278472 &M-JR 351:26d 351:M0 Low Pressure Coolant Inhelon* Number ofloops 3 3 3 3 Number of pumps 3 3 3 3 Flow rate 4200 at 7100 at 7450 at 7450 at (gpm/ pump) 40 psid 20 psid 26 psid 20 psid Auxillary Systems Residual Heat Removal System (Subsection 5.4.7) Reactor shutdown coolina mode Number ofloops 3 2 2 2 Number of pumps" 3 2 2 2 Flow rate 4200 7100 7450 7450 (gpm/ pump) i Duty (MBTU/hr*" 29.0 46.9 41.6 50.0 beat exchanger) l Number of heat 3 2 2 2 exchangers Primary contain. 4200 7100 7450 7450 ment cooling mode Flow rate (gpm) l ABWR design referred to as Low Pressure Flooder i The design of the pumps is in part based on the required capacity during the Utactorflooding mode. Heat exchanger duty at 20 hours after reactor shutdown. i Anandsent. Peter 1.313 l ._________-________________D

ABM siasiona. Sta=dard Plant any. c Subsection 5.4.7.4 for further discussion of initial suppression pool temperature and the RHR preoperational testing.) acrvice water temperature are at their maximum values. This assumption maximizes the heat sink 6.2.2.3 Design Evaluation of the Containment temperature to which the contalement heat is Coeling System rejected and thus maximizes the containment temperature. In addition, the RHR heat 6.2.2J.1 System Operation and Sequence of exchanger is assumed to be is a fully fouled Events condition at the time the accident occurs. This conservatively minimizes the heat exchanger heat In the event of the postulated LOCA, the removal capacity. Even with the degraded abort. term energy release from the reactor pri-conditions outlined above, the maximum mary system will be dumped to the suppression temperature is maintained below the design limit pool. Subsequent to the accident, fission pro-specified in Subsection 6.2.2.1. duct decay beat will result in a continuing en-ergy input to the pool. The RHR LPFL mode and It should be noted that, when evaluating this suppression pool cooling mode will remove this long term suppression pool transient, all heat energy which is released into the primary contain-sources in the containment are considered with i I ment system, thus resulting in acceptable sup-no credit taken for any heat losses other than pression pool temperatures and containment through the RHR heat exchanger. These heat pressures. sources are discussed to Subsection 6.2.1.3. In order to evaluate the adequacy of the RHR It can be concluded that the conservative system, the following is assumed: evaluation procedure described above clearly demonstrates that the"RHR system in the (1) With the reactor initially operating at 102% suppression pool cooling mode limits the of rated power, a LOCA occurs. post LOCA containment tesaperature transient. (2) A single failure of a RHR heat exchanger is 6.2.2.4 Test and Inspections the most limiting single failure. The containment cooling system is required (3) The ECCS flows assumed available are 2 HPCF, to have scheduled maintenance. The system 1 RCIC, and 2 LPFL (RHR). testing and inspection will be performed periodically during the plant normal operation R (4) Containment cooling is initiated after 10 and after each plant shutdown. Functional minutes. (See Response to Question 430.26) testing will be performed on all active components and centrols. The system reference Analysis of the net positive suction head characteristics will be established during (NPSH) available to the RHR and HPCF pumps in preoperational testing to be used as base points g,i accordance with the recommendations of Regulatory for checking measurements obtained from the gg Guide 1.1 is provided in Table 6.2 26. system tests during the plant operatio:2. General compliance for Regulatory Guide 1.26 The preoperational test program of the may be found in Subsection 3.2.2. containment cooling system is described in Subsection 14.2.12. The following functional Failure modes and effects analyses for the RHR tests will be performed. The RHR pump will be and RCWS are provided in Appendix 15B. tested through the suppression' pool cooling loop operation by measuring flow and pressure. Each 6.2.23.2 Summary of Containment Coollag pump will be tested individually.,y Analysis Containment spray sparsers will be tested When calculating the long-term, post LOCA pool during reactor shutdown by air, and by visual temperature transient, it is assumed that the inspection to verify that all the nozzles are Amanament Pwear 6.2 17

ABWR nom. Standard Plant nry e 6.2 28 Containment Boundaries in the Reactor fuel storage pools, RWCO, FP/SPCU and other Building-Plan Section A A (0-1800) potentially radioactive sources in the secondary containment. The NVAC exhaust systems and SGTS 6.2 29 Containment Boundaries in the Reactor ere also Ioeated withim the secondary Building Plan Section B B (900-2700) eontainaeat to assure eolleetion_ of any leakage. The RHR and HPCF pump seals and valve 6.2-30 Containment Boundaries in the Reactor packings and RCIC system components are a Building Plan at El(-) 13200 mm potential source of radioactive release and are located within the secondary containment. 6.2 31 Containment Boundaries in the Reactor Building Plan at El(-)6700 mm During refueling operations, the drywell head is removed and the secondary containment becomes 6.2-32 Containment Boundaries in the Reactor the containment envelope. Therefore, entry into Building Plan at El(-) 200 mm the secondary containment is provided via double door vestibules, or, in the case of the main 6.2-33 Containment Boundaries in the Reactor equipment hatch, a double door entry. This Building Plan at E17300 mm assures the integrity of the secondary containment envelope with effluent monitoring 6.2 34 Containment Boundaries in the Reactor and treatment of airborne radioactive material l Building-Plan at E113100 mm resulting from normal plant or refueling operations or from abnormal events such as a - 6.2 35 Containment Boundaries in the Reactor fuel drop accident. Building Plan at El18500 mm The airborne fission product is contained by 6.2 36 Containment Boundaries in the Reactor maintaining all portions of the secondary Building-Plan at E126700 mm containment at a negative 1/4 in. of water gage relative to the lowest pressure boundary "outside g Secondary containment design and performance the secondary containment. This negative a data is provided in Table 6.2 2c. pressure is achieved following an accident by the SGTS. During normal operation, the secondary containmer>t system is operated at a slightly The airborne fission product leakage from the negative pressure relative to the atmosphere. primary containment is processed by the SGTS. This assures that any leakage from the primary The SGTS achieves a 99.99% removal of halogen containment will be collected and can be treated (stable and radioactive) and a 99.9% of airborne before release if its radioactivity level is particulate prior to discharge to the above prescribed limits. The secondary contain-environment. This removal efficiency will be ment HVAC system operates on a feed and bleed periodically tested in accordance with principle with internal recirculation. Air flow regulatory requirements. The dose limit is from clean to potentially contaminated areas. evaluation takes credit for 99% airborne halogen and particulate for this type of leakage. A The building effluents are monitored for ra-99% removal credit is allowed even though the bioactivity by the area radiation monitoring sys. design will achieve 99.99% removal capability. tem. If the radioactive level rises above set levels, the secondary containment discharge can The SGTS will maintain the secondary contain-be routed through the SGTS for treatment before ment air flow pattern from potentially low to release. The operation of the secondary contain-high level contaminated areas. The potentially ment SGTS and HVAC are discussed in more detail high level contaminated areas are the following: in Section 6.5.1 and Chapter 9, respectively. (1) RWCU System Rooms During normal operat!on, the secondary containment is the envelope that forces (2) RCIC System Room collection of airborne radioactive material from Assanamnet Putur 6.2 20

ABM 2m62aars Standard Plant nev. c 1 TABLE 6.2-2c NET POSITIVE SUCTION HEAD (NPSH) AVAILABLE TO HPCF PUMPS A. Suppression poolis at its minimum depth, El.-3740mm ( 12.27 Ft). B. Centerline of pump suction

• is at El. -7200mm (23.62 Ft).

I C. Suppression pool water is at its maximum temperatue for the given operating mode, 97*C (207'F). D. Fressure is atmospheric above the suppression pool. E. Maximum suction strainer losses are 2.0 psi. NPSH = HATM + HA -HVAP Hp where: HATM atmospheric head = HS static head = HVAP vapor pressure head = Frictional head including strainer Hp = R Minimum Exoected NPSH 3 ,. HPCF Pump Runout is 890 m /h (3918 gpm). Maximum suppression pool temperature is 97'C (207'F) HATM 10.73m (35.20 Ft) Hs 3.46m (11.35 Ft) = HVAp 9.74m (31.95 Ft) = Hp 1.82m (5.97 Ft) = Strainer head loss = 2.0 psi = 1.46m = 4.80 Ft NPSH avallable = 10.73 + 3.46 - 9.74 1.82 = 2.63m (8.63 Ft) NPSH required = 2.2m (7.22 Ft)

• NPSHreferencepoint i

Amendment Putur 6.244

i ABM 2mm. Standard Planf nev s Table 6.2-2d SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA DESCRFTION UNIT VALUE A. W Containment Desip j 3 6 1. Free Vohane ft 3.0 x 10 2. Pressure, inches of water, gauge inch H O (-)0.25 2 3. leak Rate at Pa.s.eeldent Pressure %/ day

• 50 (e r c.,n.d ry t'n e.i.wat Free a

Volume) 4. Exhaust Fans 2 Number Centrifugal Type 5. Filters (a) Basic specification g 1 2 Number of filter train Dust Type (b) Componcat specification (1) PreSker Number of set 1 Dry Type - (2) HEPA filters 2 Number of set Type (Material) Glass fiber (3) Charcoal absorber 1 Number of set Deep bed Type 6.2 46e

1. h =1 Putur

J ABM zwicais. Etandard Pfarrt arv. h Tabic 6.2 2d ] SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA ']' (Continued) DESCRIPTION UNTT _ VALUE B. Transient Analysis 1. Initialconditions l (a) Primary C='a:=ent 1 (1) Presture psia 15.45 (2) Temperature HF 135.0 (3) Outside air temperature ) Summer operation: MF 115.0 Winter operation: MF (-)40.0 (b) &Me y Containment E (1) Pressure inch H O (-)0.25 g-2 (2) Temperature Max value in summer HF 104.0 Min valuein winter %F 50.0 2. Thickness of Secondary emedament Wall. Wall thickness range from inch (11.81 - 59.06) 3. Thickness of Primary Containment Wall (a) Concrete Wall inch 79.0 (b) Uner Plate inch 0.25 p m.paa L1x1

ABM 234.too4. Standard Plant arv s Table 6.2-2d SECONDARY CONTAINMENT DESIGN AND PERFOPJ4ANCE DATA (Continued) DESCRIPTION UNTT VALUE C. *nermal Characteristics L Primary Containment Wall (a) Coefficient of Linear Espansion Concrete Wall in/in-%F 0.55x10'3 Liner Plate in/in KF 0.73x10'# (b) Modulus of Elasticity Concrete Wall psi 3.41x1f IJoer Plate psi 27.2x10 (c) DermalConductivity ConcreteWall Bru/hr ft %F 0.941 5 IJaer Plate Bru/br-ft KF 30.91 (d) Dermal Capacitance Concrete Wall Btu /ft -%F 28.84 IlnerPlate Bru/ft %F $3.56 &cW 7 ontainmentWall C 2. (a) DermalConductivity btu /hr-ft -HF 0.941 3 (b) Dermal Capacitance Bru/ft %F 28.84 Amendment Futur

j M MM mim Standard Plant arv. s Table 6.2 2d SECONDARY CONTAINMENT DESIGN AND PERPORMANG DATA (Continued) DFMRWTION UNTT VAME 3. Heat Transfer Coefficients (a) Primary r=#ai==ent l l Annosphere to Primary 2 cmeniament Wall bru/hr-ft -%F 0.246 (b) Primary raaral==ent Wall to Secondary emtmin=ent 2 Annosphere Stu/hr-ft -HF 0414 (c) heaadary emrainment Wall I, 8 j to teraadary Containment f Abnosphere Bru/hr-ft -%F 0414 (d) Primary r=*=ia-cat 0.95 Emissivity 0.95 (c) hA-y Containment Emissivity 1 1 a-. re r aus

MM - n(- anAsicoAs Standard Plant nry e The following discussion provides details of the 63.2.2.1 High Pressor: Cere Spray (HPCS) combined systems; in particular, those design fea. System tures and characteristics which are common to all systems. The HPCS system is composed of two HPCS loops (B and C) spraying water above the core. Each f 63.2.1 Schessatic Piping and Instrumentation of the two loops belongs to a separate division; Diagrams electrical and mechanical separation between the two divisions is complete. Physical separation The P& ids for the ECCS are identified in Sub-is also assured by locating each division in a. section 6.3.2.2. The process diagrams which iden-different area of the reactor building. The two tify the various operating modes of each system loops are both high pressure pumping systems are identified in Subsection 6.3.2.2. (i.e., they are capable of injecting water into the reactor vessel over the entire operating 6.3.2.2 Equipament and Component Description pressure range). Rated flow at both high and low pressure is the same for each loop. The The starting signal for the ECCS comes from piping and instrumentation diagram and process four independent and redundant sensors of drywell diagram are given in Figures 6.3 7 and 6.3-1 re-pressure and low reactor water level. The ECCS spectively. is actuated automatically and requires no op-erator action during the first 30 min following The reference pressure for the operating per-the accident. A time sequence for starting of formance of the system at high pressure is the the systems is provided in Table 6.3 2. lowest spring (safety) setpoint of the SRVs. Elcetric power for operation of the ECCS is Both HPCS divisions take primary suction from from regular AC power sources. Upon loss of the the CST and secondary suction from the suppres-regular power, operation is from onsite emergency sion pool. In the event CST water level falls standby AC power sources. Emergency sources have below a predetermined setpoint or supp'ression sufficient capscity so that all ECCS requirements pool water level rises above a predetermined are satisfied. Each of the three ECCS functional setpoint, the pump suction will automatica:ly groups identified in Subsection 6.3.1.1.3(1) has transfer from the CST to the suppression pool. its own diesel generator emergency power source. Both HPCS system loops have suction lines that Section 8.3 contains a more detailed description are separate from RHR loops. of the power supplies for the ECCS. The HPCS pumps are located at an elevation Regulatory Guide 1.1 prohibits design reliance which is below the water level in the suppres-on pressure and/or temperature transients ex-sion pool. This assures a flooded pump pected during a LOCA for assuring adequate NPSH. suction. The motor operated valve in the The requirements of this regulatory guide are ap-suction line from the suppression pool on each plicable to the HPCS, RCIC and RHR pumps. division is normally closed since primary suction is taken from the CST. This valve auto-The BWR design conservatively assumes O psig matically opens on receipt of either of the containment pressure and maximum expected tem-suction transfer signals noted above. The sup-peratures of the pumped fluids. Thus, no reli-pression pool suction valves on each loop are ance is placed on pressure and/or temperature capable of being closed from the control room if transients to assure adequate NPSH. a leak develops in the system piping downstream of the isolation valves. Overpressure protec-Requirements for NPSH are given in Figures tion of the pump suction line is provided by a l 6.3-1 (HPCS),6.3 2 (RCIC) and 6.3 3 (RHR). Pump relief valve to the pump minimum flow line. characteristics curves are given in Figures 6.3 4 (HPCS),6.3-5 (RCIC) and 6.3 6 (RHR). Each of the two high pressere flooder spray loops discharges water into the core via a The design parameters for the HPCF and RHR separate overhead spray sparger. Internal g system components are provided in Tables 6.3-8 and 6.3 9, respectively. Amendment Purur 6M

ABWR

suum.

Standard Plant mm,. s Table 6.34 DESIGN PARAMETERS FOR HPCF SYSTEM COMPONENTS (1) Main Pumps (C001) Number of Pumps 2 Pump % Centrifugal Drive Unit Motor Flow Rate 800 gpm @ 1192 psia reactor pressure 3200 gpm @ 115 psia reactor pressure Developed Head 2926 ft. @ 1192 psia reactor pressure 725 ft. @ 115 psia reactor pressure Maximum Runout Flow 3800 gpm @ 247 psia reactor pressure Minimuns Bypass Flow 400 gpm S: Water Temperature Range 50% to 212KF NPSH Required 8 ft. (2) Strainer (D001) locatio'n Suppression Pool Size 50% plugged shall meet pump NPSH requirements (3) Restrictig Orifice (D002) location Pump discharge line { Size limit pump flow to values specified i i (4) Condensate Storage Tank 150,000 gal reserve storage for HPCF and RCIC systems combined l I i l Amendment Pwur 63-24a l l I

~ ,.c ABM ziassocia Standard Plant arv. s thbie 6.34 DESIGN PARAMETERS POR HPCF SYSTEM COMPONENIS ((Minued) (5) Nw Elements (FE001) 1 mention Pump discharge downstream of minimum flow bypass line Head Loss 20 ft. maximum @ 3200 gpm Accuracy d.25% combined element, transmitter and indicator at maximum rated flow (6) Core Flooder Sparger Flow Rate 3200 gpm minimum @ 115 psia reactor pressure Pressure Drop 260 ft. maximum @ 3200 gpm g (7) Piping and vaha Design Pressures 70 psig - suction and discharge connected to suppression pool 200psig pumpsuction 1575 psig pump discharge Design Tersperatures 100%F condensate tank suction 212%F pump suction and discharge 575%F-discharge to vessel a==d--t Putur 6.124b

MN n,=.ang pin., 2sA61oors m, Table 634 - DESIGN PARAMETERS POR HPCF SYSTEM COMPONDrI5 (Continued) (s) valve operation Pump Sosion Valve, Normally closed, opens on low water Suppresalan Pool (P001) levelin condensate tank Pump Semion Check Valve, Prevents backDowinto suppression pool Suppression Pool (P002) Pump Section Valve, Normally open, closes on low water raade===se Tank (P009) levelin condensate tank Pump Sedion Check Valve, Prevents backflowinto condensate tank t'aade===se Tank (P010) g g Pump Discharge Valve, Normally closed, opens within [ ReadorInjection Valve (P004) 36 seconds afterinitiation signal Testable Cbcck Valve, Prevent loss of coolant outside drywell Reactor 9,*taa line (P005) for line break r Maintenance Valve, t Normally open, used to isolate system ReactorInjectsonIJne (P006) from reactor for maintenance purposes i PumpTest IJne Valves (P015, Normally closed, throttle valves used P016) to test system flow at rated and runout conditions Pump Maimum Flowline Normally closed, opens on signal when Valves (R)12, P013) pump is running and no flow through flow meter. Used to protect pump from overheating. Amendment Futur 63-24c

MM 23A6100AB - Standard Plant ary. n Table 6.3-9 DESIGN PARAMETERS POR RHR SYSTEM COMPONFRIS (1) Main Pumps (0001) Number of Pumps 3 PumpType Centrifugal Drive Unit Motor Flow Rate 4200 gpm ~ Developed Head 400 ft @ 115 psia reactor pressure Maximum Rumout Flow 5000 gpm Maximum Bypass Flow 650 gpm Minimum Shutoff Head 640 ft. Maximum Pump Brake Horsepower 550 kw p I Water Temperature Rang: 40% to 358HF NPSH Required 8 ft. (2) Heat F=ek=a-ers (B001) [ Numbi.r of units 3 Seismic CategoryI design and analysis Types of e=rhangers Horizontal U Tube /Shell Maximum primary side pressure 500 psig Design Point Function Cooling Post LOCA Containment l Aw~at Putur 6.3 24d

g ABM zwman Standard Plant arv. s l Toble 6.3-9 ' DESIGN PARAMETERS POR RHR SYSTEM COMPONEN13 (Continued) Prunary aide (tube side) performance data (1) Flow ' 4200 gpm (2) Inlet temperature 358%F maximum (3) ABowabic pressure drop 10 psi (Max) (4) Type water Suppression Poolor Reador Water (5) Fouhng factor 0.0005 wary side (shell side) performance data (1) Flow 5280 gpm b, (2) Inlet temperature 105KF maximum N (3) ABowable pressure drop 10 psi (Max) (4) Type water Reactor Building Cooling Water (5) Fouhag factor 0.0005 (3) Strainer (D008) Location Suppression Pool Size Meet pump NPSH requirements when 50% plugged l (4) Restricting Orifices l location (D003) Vesselreturn line Si:z limit flow to vessel to 4200 gpm Location (D002) Suppression pool return line l l l A=-=d-at Pmur 63 24c l

( ~ I MM asAsm As Eenndard Plant nev. m Table 63 9 l i DESIGN PARAMETERS FOR RHR SYSTEM COMPONE!frS (Continued) l Size IJmit flow during suppression pool coohng to 4200 spm I=tian (D004) Fuel poolreturnline Size IJmit flow during fuel pool coohng to 4200 gpm I meia= (D001) Pump minimum flowline Size limit pump Dow through the bypass line to 650 gpm i X4 cation (D005) Dischargeline to wetwellspray size IJmit werwell spray sparger flow to 500 gpm Location (D006) Dischargeline to drywellsparger Size Limit drywell spray sparger Dow to 3"/00 gpm E (5) FlowElements (FE009) location Pump discharge line, downstream of heat exchanger bypass return Rated Flow 4200 gpm Head Loss 20 ft maximum @ 4200 gpm Accuracy 12.5% combined element, transmitter and indicator at rated flow (6) Vessel Flooder Sparger Flow Rate 4200 gpm Pressure Drop i Asseadment Putur 4.3-24f ..... ~

ABM =>^62aarn Standard Plant arv. m TnWe 6.M DFJIGN PARAMETERS POR RHR SYSTEM COMPONE!G3 (ransinued) (7) WetweB Spray $parger Pkm Rate - 500 gpm Pressure Drop M ft. @ 500 gym (8) DryweH Spray Sparger Flow Rate 3700 spm Pressure Drop 94 ft. @ 3700 gpm (9) Paping and Yahes Design Pressures "10 psig discharge piping conneaed to suppression pool 100 psig - suction piping connected to suppression pool 125 psig - wetwell and drywell sparger piping 200 psig - pump suction piping 25iP 8 - Pump hge pip *mg 1250 psig - w.ssel suction and return piping Design Temperatures 212%F suppression pool piping and wetwcU and drywell sparger piping 358%F-pump section and discharge piping 575%F-vessel suction and return piping (10) Vahe Operation See Table 5.4-3, RHR PUMP / VALVE LOGIC Ameha Pwur gyg

+ p~ :. : t ABWR msi, Standard Plant muy e system is designed to provide and maintain an environment with controlled temperature and humidity to ensure both Of these spaces, all but the mechanical comfort and safety of the operators and equipment room are maintained at a positive the integrity of the control room pressure of +0.1 to +0.5 la. of water gage ] components.: pressure at all times. The mechanical I equipment room is maintained at 0.0 to +0.5 in. (2) Provisions for periodic inspection, of water gage. Pressure control dampers at the - testing and maintenance of the principal inht of the ventilation system maintain these components shall be a part of the design pressures. These spaces constitute the requirements. operation, living and environmental control areas and can be isolated for an extended 6.4.2 Systesi Design period if such is required by the existence of: l a LOCA or high radiation condition. Figure 9.4-1 provides the flow diagrams describing the control building HVAC system. Frequent access to spaces numbered (1), (2), Heating, cooling and pressurizing the control (3), (5), (6), (7), (8), (9), (10), and (11) building, and filtering the air therein, is fully during both normal and emergency conditions are described in Section 9.4.1, wherein function is safe for extended human occupancy. discussed and equipment is listed. 6.4.2.2 Ventilation System Design 6.4.2.1 Centrol Building Envelope The design, construction and operation of The control building spaces within the the control building HVAC system are described envelope supplied by the HVAC habitability in detail in Subsection 9.4.1. Figure 9.41 is systems includes: a diagram of the control building HVAC system, showing major components, seismic (1) control room proper including the classifications and instrumentation, critical document file; Description of the charcoal filters is given (2) ensputer room. in Subsertion 9.4.1. (3) contrdi equipment room; Description of control room instrumentation for monitoring of radioactivity is given in (4) apper and lower corridors; Subsections 11.5.2 and 12.3.4. (5) elevator shaft and stair wells; A description of the smoke detectors is in Subsection 9.5.1. (6) office and chart room; 6.4.2.2.1 Control Room Drawings (7) kitchen and lunch rooms; Layout drawings of the control room and the (8) instrument repair room; remainder of the control building are given in Section 1.2. (9) aleeping area 6.4.2.2.2 Release Points (10) men'slavatory; Release points (SGTS vent) are shown in (11) wonnen's lavatory and lounge; Figure 6.41 (plan view). The air intakes are E well above grade. Elevation of other : $ (12) HVAC mechanical equipment rooms. structures is seen in Figures 1.2 9 and 1.210.

• • • Fwur 6.4-3

ABM i 23461oarn 4 Standard Plant Rev. c 6.4.23 14ak11ghtness Leakage through the various paths is The control building boundary walls are negligible except for that through the doors. designed with low leakage construction. All boundary penetrations are sealed. The access 6.4.2.4 1steraction Witb Other Zones and doors are designed with self closing devices Pressure-Containing Equipment which close and latch the doors automatically following the passage of personnel The control building is heated, cooled, ventilated and pressurized by a recirculating All potential leak paths in and from the air system using filtered outdoor Jr for control building boundary are tabulated and shown ventilation and pressurization purposes, on Figures 9.4-1. Recirculated air and outdoor air are mixed and drawn through filters, a cooling coil and zone The control room in leakage analysis was electric reheating coils. performed using the methods and assumptions given in NAA SR 10100 (Conventional Building For There are two intakes on the top floor side Reactor Containment), Atomics Internadonal, and walls of the control building, one on each Regulatory Guide 1.78. The leakage rate is end. Radiation monitoring sensors located in calculated using the following: each duct warn the operating personnel (by means of readouts and alarms in the : ain (1) a 1/2 in. WG differential across control room) of the presence of airborne surfaces and components exposed to or contamination. Also, the signal automatically protected from effe:ts of winds; closes down, the contaminated air intake valves (2) maximum design differential for dosed and normal vent dampers, opens the emergency A dampers on the suction side of the vent dampers, and turns on the primary supply fans; and emergency filter unit fans on reduced flow. If both air intakes are contaminated the control (3) Equation: room operator can manually override the system to open eitner air intake to draw makeuup air q = AP + BPg (6.41) when necessary. This makeup air is routed through HEPA and charcoal filtering system for whe re, cleanup before being used for pressurization. q = leakage rate per unit leak ...The control room is rnaintained at positive path (cfm); pressure with respect to atmosphere. In an eraergency the pressure differential will A = empirical constant (cfm per eliminate infiltration of airborne g* unit leak path per inch of contamination. The doors are of the double R water pressure); vestibule type to increase pressure differential between rooms; thereby eliminating B = empirical constant (cfm per infiltration when the doors are opened. unit leak path per 1/2 inch of water pressure); and The control room must remain habitable dur-ing emergency conditions. To make this possi-P = differential pressure (in. ble, potential sources of danger such as steam w.g.). lines, pressure vessels, CO fire fighting 2 containers, etc. are located outside of the The leak paths considered were concrete walls control room and the compartments containing and slabs, wall and slab joints, door frames, control building life support systems. doors, electric cable penetrations, duct penetrations and pipe penetrations. The A tabulation of moving components in the empirical constants A and B for each leak path control building HVAC system, along with the are taken from NAA-SR-10100. respective failure mode and effects, is shown Amendment Putur 6M

MM 2sA61oors Etandard Plant arv. c \\ in Table 6.41. All dampers except the mixing dampers in the air conditioning units are of the two position (open or closed) type. 5A.2J Shielding Design 6A.23.1 Design Basis The control building shielding design is i l l I. l 6.4-da Anneedment Putur

s ' l4 MM 21A6100AB Eeandard Plant arv c Food storage space is provided as a part of operation. All equipment is designed to the kitchen lunchroom adjacent to the control facilitate the above discussed test and equipment room. Water and food storage edequate in:pection functions. for 12 people for 5 days is stored in this area. Tije storage cabinets have a net volume of 28. Failure of any system or component to it useable for food storage. In addition, thg properly perform its assigned function during refrigerator has a met volume of 10 ft any test or inspection is grounds for repair or available. Potable water is stored in scaled replacement. sanitary containers in the kitchen lunchroom. 6.4.6 Instrumentation Requirements All foodstuffs and water intended for emergency use must be so labeled and not be used A complete description of the required for normal conditions, thus ensuring and adequate instrumentation is given in Subsection h supply at all times for emergency use. 63.1.1.6. The sanitary facilities are located across the 6.4.7 Interfaces ball from the control room. The control room habitability system design 6.4.5 Testing and Inspection was based on the following environmental conditions. The system is designed to permit periodic inspection of important components (e.g., fans, 6.4.7.1 External Temperature motors, belts, coils, filters, ductwork, piping, dampers, control instrumentation and valves), to Tge maximum external air temperature is assure the integrity and. efficiency of the 100 F and the minimum external air system. Local display and indicating devices are temperature is -10 F. provided for periodic inspection of vital parameters such as air temperature upstream and 6.4.7.2 Meterology(x/O's) downstream of the heating and cooling coils, cooling water inlet temperatures, filter pressure The X/O's used for evaluation of the drop, duct static pressures, and water pressures control room operator dose to meet General at the inlet and outlet of coils. Design Criterion 19 were derived from Regulatory Guide 1.3 for ground level release. Test connections are provided in the duct work Specific values and assumptions are presented and piping for periodic checking of air and water in Subsection 15.6.5. I flows for conformance to design requirements. All features are periodically tested by 6.4.7.3 Toxic Gases initiating all dampers during normal operation. The operating system is proven operable by its No hazardous or toxic gas sources external performance during normal plant operations. The to the control room were used in the design of g HEPA filters are periodically tested with DOP the control room habitability system. v smoke per ANSI N101.1. The charcoal filters re to be periodically tested with a freon gas for adsorption efficiency. Inspection and sampling connections are provided for on site filter testing. Filter pressure drop is to be routinely monitored and a high differential alarm alerts the operator to switch over to standby system. The systems are to be tested periodically by initiating the changeover sequence during normal Amendment Putur 6.4-7 { \\

ABWR

mum, Standard Plant arv. e Table 6.4-1 IDENTIFICATION OF FAILURE /EFFECT IN THE CONTROL ROOM HVAC SYSTEM Component Nomenclature Ms.dr Failure ElTect kil.nn Main Outside Air Open Contaminated Air Penetration RE Automatically S; arts Emergency Unit EmergencyOutside Air Closed Loss of Control Room Pressurization Flow Switch Starts Redundant Damper Unit Emergency Outside Air No Flow Loss of Control Room Pressurization Flow Switch Starts Redundant Supply Fan Unit Control Room Supply Fan No Flow Loss of Control Room Cooling Flow Switch Starts Standby Unit Control Room Return Fan No Rota-Control Room Overpressurization Flow Switch Starts Redundant Lion Unit Control Room Return Air Closed Partial Loss of Control Room Cooling Flow Switch Starts Redundant Damper Unit Equipment Room Supply Closed Loss of Equipment Room Cooling Flow Switch Starts Redundant Air Damper Unit Equipment Room Return No Flow Equipment Room Overpressurization Flow Switch Starts Redundant 3

Fan Unit g Equipment Room Return Closed Partial Loss of Equipment Room Cooling Flow Switch Starts Redundant Damper Unit Note: Failure mode and effect is indicatedfor each individual component in the system during an emergency operation. 7hepostulation of more than a singlefailure in the system is not considered. Amendment Putur M-8

2sA6100AH ~, Etandmed Plant any. m ' the RCW system shall be designed to Seismic requirements, including a LOCA or a loss of ~ Category I and the ASME Code, Section III, offsite power, or both. Each RCW division is Class 3 Quality Assurance B, Quality Group supplied electrical power.from a different C,IEEE-279 and IEEE 308 requirements. division of the ESP power system. (4) The RCW system shall be designed to limit During normal operation, RCW cooling water leakage to the environment of radioactive flows through all the equipment shown in Tabic contamination that may enter the RCW from 9.2-da, b, and c. - the RHR System. l During all plant operating modes, an RCW (5) Safety-related portions of the RCW system water pump and beat exchanger are normally shall be protected from flooding, spraying, operating in each division. Therefore, if a steam impingement, pipe whip, jet forces, LOCA occurs, the RCW systems required to shut missiles, fire, and the effect of failure of down the plant safely are already in operation. any non Seismic Category I equipment, as The second pump and heat exchanger in each

required, division are put in service if a LOCA occurs.

1 (6) The safety-related portion of the RCW System The monsafety related parts of the RCW system shall be designed to meet the foregoing de-are not required for safe shutdown and, hence, sign bases during a loss of preferred power are not safety systems.. Isolation valves sepa-(LOPP). rate the essential subsystems from the nonsafe. 9.2.11.1.2 Power Generation Design Bases ty related subsystems during a 14CA, in order to assure the integrity and safety functions of the The RCW system shall be designed to cool safety related parts of the system. Some non. various plant auxiliaries as required during: safety-related parts of the system are operated (e) normal operation; (b) emergency shutdown; during all other modes, including the emergency (C) nomal shutdown; and (d) testing. shutdown following an 14PP, or 14CA as shown in Table 9.2 4a, b, and c. 9.2.11.2 System Destdption Surge tank water level is monitored. A level The RCW System distributes cooling water dur-switch detects leakage and isolates the non es- ) ing various operating modes, during shutdown, and sential subsystem, thus assuring continued oper-during post-I'OCA operation. The system removes ability of the safety related services. Instru-heat from plant auxiliaries and transfers it to ments, controls, and isolation valves are locat-the ultimate heat sink. Figures 9.2 la through ed in the safety-related part of the RSW system 9.211 show the piping and instrumentation and designed to safety grade requirements as diagram. Design characteristics for RCW system stated in design base (3) of Subsection 9.2.11. components are given in Table 9.2-4d. 1.1. i The RCW system serves the auxiliary equipment 9.2.11.3 Safety Evaluation listed in Table 9.2-la, b, and c. I 9.2.11.3.1 Fallure Analysis The RCW system is designed to perform its required cooling function following a postulated A system failure analysis of passive and LOCA, assuming a single active failure in any active components of the RCW system is presented J mechanical or electrical system. In order to in Table 9.2 5. Any of the assumed failures of I meet this requirement, the RCW system provides the RCW system are detected in the control room three complete trains, which are mechanically and by variations of and/or alarms from the various electrically separated. In case of a failure system instruments and also from the leak detec-which disables any of the three divisions, the tion system sensing leakage in the ECCS pump and other two division meet plant safe shutdown heat exchanger areas. 1 9.24 Amendment Putur

ABM 2 m io w i Standard Plant Rev. n TABLE 9.2-2a WATER QUALITY CHARACTERISTICS FOR THE MAKEUP WATER SYSTEM (PURIFIED) WATER QUALITY Operating System Maximum PARAMETER Tarnet DC&ign Value Chloride ppb) 10.0 20.0 100.0 Sulfate (ppb) 10.0 20.0 100.0 Conductivity

• at 25 C (uS/cm) 0.2 20.0 100.0 Silica (ppb as SiO )

10.0 20.0 100.0 g 2 pH at 25 C min 6.4 6.2 5.6 max 7.8 8.0 8.6 I Corrosion Product Metals (ppb) { Fe insoluble soluble Cu total 10.0 20.0 100.0 y allother metals sum 10.0 20.0 100.0 Organic Impurities" Equivalent oK(u3/cm) 0.2 0.4 2.0 Does not include an incremental conductivity value of 0.8 uS/cm at 25 C due to carbon dioxide from air in water stored in tanks open to the atmosphere.

• • Organic impurity values apply to fresh makeup water stored in the Demineralized Water Storage Tank.

The portion of this system which is within the reactor building is within the ABWR scope. The portions involving preparation of the demineralized water, its storage and transport to the reactor building are not within the ABWR scope. Demineralized water l will be provided at a maximum flow rate of approximately 600 gpm at a temperature l between 50 to 100 F. 1 Amendment Futur 9.215a l l

1 l' l 23A61 M - arva mandard Plant ) 1 TABLE 9.2-4d DESIGN CHARACTERISTICS FOR REACIVR BUILDING COOLING WATER SYSTEM COMPONENTS RCW Pumps RCW (A)/(B) RCW (C) Discharge Flow Rate 5,720 gpm/ pump 4,480 gpm/ pump Pump Total Head 82 psig 75 psig Design Pressure 200 psig 200 pig Design Temperature 158 F 158 F RCW Heat Exchangers RCW (A)/(E) RCW (C) 6 6 Capacity 60x10 BTU /h 56x10 BTU /h RCW Surge Tanks Capacity Equalto 30 days of normalleakage Design Pressure Static Head , DesignTemperature 158 F RCW Chemical Addition Tanks l Design Pressure 200 psig Design Temperature 158 F t l RCW Piping i Design Pressure 200 psig Design Temperature 158 F l l l c ) Amendment Futur 9119a

ABM 2 4suo ui Standard Plant arv. s '1 Table 9.2-10 HVAC Emergency CoohngWater System i Actrve FaDure Analysis Fauure of dieselgenerator to start Loss of one refrigerator and j or failure of au power to a single pump in a division would not 4 Class 1E power system bus permit sending chilled water to the main control room. The other HECW division would send l chilled water to the main control room which would maintain adequate cooling. l Failure'of auto pump or Same analysis as above i refrigerator signal - Failure of a single HECW Same analysis as above refrigerator Fauure of a single HECW pump Same analysis as above Failure of HECW pump and Same analysis as above refrigerator room cooling l f l l l l Amendment Putur 9.2 25a

O ABWR nwxa Standard Plant nev s 11.1 SOURCE TERMS negligible because all of the gases released to the coolant are assumed to be rapidly transported out of The information provided in this section defines the vessel with the steam and removed from the the radioactive source terms in the reactor water and system with the other non-condensables in the main steam which serve as design bases for the gaseous, condenser. As a consequence of the immediate re-liquid and solid radioactive waste management sys-moval of all the gases, the expected relative mix of i tems. gases does not depend on the reactor design Radioactive source term data for boiling water The design basis noble gas source term for reactors has been incorporated in American Na!!onal ABWR is selected such that the mix is that of Refer-Standard ANSI /ANS-18.1(Reference 1). The Stan-ence I and the total of the release rates of the 13 dard provides bases for estimating typical concentra-noble gases from the vessel is 100,000 pCi/sec as tions of the prindpal radionuclides which may be an-evaluated at 30 minutes decay. The noble radiogas ticipated over the lifetime of a BWR plant. The source term rate after 30 minute decay has been source term data is based on the cumulative industry used as a conventional measure of the fuelleakage experience at operating BWR plants including mea-rate since it is conveniently measurable and was con-surements at severrd stations throu;;h 1981. It there-sistent with the nominal 30 minute offgas holdup fore reflects the influer ce of a number of observa-system used on a number of early plants. A design tions made during the transition period from opera-basis noble gas release rate of 100,000 pCi/sec at 30 tion with fuel of older designs to operation with fuel minutes decay has historically been used for the of current improved designs. The source terres spec-design of the gaseous waste treatment systems in ified in this section were obtained by applying the BWR plants (Reference 2) with satisfactory results. procedures of Reference 1 for estimation of typical It was selected on the basis of operating experience source terms and adjusting the resuhs upward as ap-with consideration given to several judgemental fac-propriate to assure conservative bases for design, tors including the implications to environmental re-leases, system contamination, and building air con-The various radionuclides included in the design tamination. The design basis value is considered to basis term have been categorized as fission products represent a long term average value. Operation at or activation products and tabulated in the subsec. higher release rates can be tolerated for reasonable tions which follow. The lists do not necessarily in-periods of time. Normal operational noble gas re-clude all radionuclides which may be detectable or lease rates for ABWR are expected to be approxi-theoretically prs;dicated to be present. Those which mately 15,000 pCi/sec as evaluated at 30 minutes have been included are considered to be potentially decay. This may be compared with normal release significant with respect to one or more of the follow-rates on the order of 50,000 yCi/sec based on fuel ing criteria: experience through the mid 1970's (Reference 3). Consequently, continued application of the same (1) plant equipment design, design basis of 100,000 pCi/sec provides increased (2) shielding design, margin relative to expected release rates when oper-(3) understanding system operation and perfor-sting with fuel of modern design. The design basis

mance, noble radiogas source terms are presented in Table (4) measurement practicability, ar.d 11.1 1.

(5) evaluation of radioactivity in effluents to the envi-ronment. 11.1.1.2 Radiolodine Fission Products 11.1.1 Fission Products For many years, desigt basis radiciodine source l ) terms for BWRs have been specified to be consistent l 11.1.1.1 Noble Radiogas Fission Products with an 11311eak rate of 700 pCi/sec from the fuel (Reference 2). Experience indicated that I 131 leak. l Typical concentrations of the 13 principal noble age rates this high would be approached only during gas fission products as observed in steam flowing operation with substantial fuel cladding defects. It i from the reactor vessel are provided in the Source would not be anticipated that full power operation l Term Standard ANSI /ANS-18.1 (Reference 1). would continue for any significant period of time Concentrations in the reactor water are considered with fuel cladding defects as severe as might be, i l Amessmeat Purur 1111 1 1 l l L

I l i 1 \\ l ABM Standard Plant uwoou artil indicated by I 131 leakage in excess of 700 pCi/sec. is expected to be less than 0.001. The design basis concentrations in steam are obtained by multiplying The design basis reactor water radioiodine con-the values in Table 11.13 by 0.001. centrations for ABWR have been based on the rela-tive mix of radioiodines in reactor water predicted by 11.1.2 Activation Products the data of Reference 1 with magnitudes increased such that the 1131 concentration is consistent with a 11.1.2.1 Coolant Activation Products release rate of 700 Ci/sec from the fuel. This pro-vides a substantial margin relative to the expected The coolant activation product of primaryimpor-1131 release rate of approximately 100 pCi/sec. tance in BWRs is N 16. ANSI 18.1 (Reference 1) Reference 1 specifies expected concentrations of the speci5es a concentration of 50 pCi/gm in steam leav-5 principal radioiodines in reactor water for a refer-ing the reactor vessel. This is treated as essentially cnce BWR design and provides bases for adjusting independent of reactor design because both the pro-the concentrations for plants with relevant plant pa-duction rate of N-16 and the steam flow rate from rameters which do not match those of the reference the vessel are assumed to vary in direct proportion to I plant. The concentration adjustment factors were reactor thermal power. The design basis N 16 con. calculated as described in Subsection 11.1.3 using the centration in steam for ABWR is designated to be 50 plant parameters in Table 11.16 and removal pa-pCi/gm. This value has in fact been used as the I t rameters from Table 11.1-7. The scale factor re-design basis concentration for GE BWRs since the quired to increase the 1131 concentration from that early 1970's and operating experience indicates that calculated using Reference 1 to the design basis it is adequately conservative. It should be noted that i value was approximately 6.7. The design basis con-r portion of the source term traditionally identified I centrations are presented in Table 11.1-2. as *N 16' actually represents C 15 which is present to the extent of no more than about 15 pCi/gm. His-The ratio of concentration in reactor steam to torically, gross gamma dose rate measurements concentration in reactor water (carryover ratio) is made to confirm the magnitude of the N 16 concen-taken to be 0.015 for radiolodines (Reference 1). tration have included responses to gamma-r' ys from a Consequently, the design basis concentrations of C 15. Use of the combined 'N 16' source t erm in radiciodines in steam are defined by multiplying the shielding design introduces additional conservatism values of Table 11.12 by the factor 0.015. because the C 15 compcaent has a 2.45 second half life and therefore decays more rapidly with 11.1.1J Otherfission Products transpon time through the system than N 16 which has a 7.1 second half-life. This categoryincludes all fission products other than noble gases and iodines and also includes The design basis N 16 concentrations in steam transuranic nuclides. Some of the fission products and reactor water are shown in Table 11.14. Refer-are noble gas daughter products which are produced ence 1 gives the reactor water concentration at the in the steam and condensate system. The only recirculation system nor.zle as 60 pCi/gm. Since transuranic which is detectable in significant ABWR does not have an external recirculation loop, concentrations is Np-239. Concentrations of those the teactor water concentration has been radionuclides which are typically observable in the decay corrected to the reactor core exit to obtain an coolant are provided in Reference 1 for a Reference estimated value of190 pCi/gm. BWR plant. The Reference Plant concentrations were adjusted to obtain estimates for the ABWR It has been observed that during operation with plant by using the procedure described in Subsection intentionalintroduction of hydrogen to the feedwater 11.13 and appropriate data from Tables 11.1-6 and for the purpose of controlling feedwater oxygen con-11.17. In order to assure conservative design basis centrations (i.e.,'with hydrogen water chemistry), the concentrations for the ABWR the results were N 16 concentration in the steam is significantly cle-increased by the same factor used to obtain design vated. Under these circumstances, conditions for basis radioiodine concentrations (6.7). The design production of volatile nitrogen chemical species are basis reactor water concentrations are presented in more favorable so that a greater portion of the N 16 Table 11.13. The ratio of concentration in steam to produced is narried with the steam. The C 15 concentration in water (carryover) for these nuclides concentration remains approximately the same. For Amsedsment Putur 11.1 2 I

ABWR swoou Standard Plant

a. m operation with hydrogen water chemistry, the Reference 3 specifies an Argon-41 release rate from recommended design basis N 16 concentration in the vessel of 40 pCi/sec for a 3400Mw Reference steam is 4 times the value for natural water BWR. This value bounded the available experimen-chemistry or 200 pCi/gm.

tal data base. Based on adjusting to the ABWR thermal power, a design basis Argon-41 release rate 11.1.2.2 Nomeoelaat Activation Products of 46 pCi/sec is =pW for the ABWR. Radionuclides are produced in the coolant by 11.1.3 Radionuclides Concentration neutron activation of circulating impurities and by Adjustment corrosion ofirradiated system materials. Typical re-actor water concentrations for the principal activa. In order to determine the estimated concentra-tion products are contained in Reference 1. The tions of radionuclides in the groups classified as values of Reference I were adjusted to ABWR con-iodines, other non volatile fission products, and ditions by using the procedure described in subsec. non coolant activation products using the tion 11.1.3 and appropriate data from Tables 11.16 ANSI /ANS-18.1 Source Term Standard (Reference and 11.17. These results were arbitrarily increased

1) it is necessary to apply appropriate adjustment by the same factor used for the design basis factors to the Reference Plant concentrations pro-radioiodine concentrations (6.7) to obtain the con-vided in the Standard.

l servative design basis reactor water concentrations shown in Table 11.15. The steam carryover ratio for Equilibrium concentrations in Reactor water are these isotopes is estimated to be less than 0.001. A assumed to satisfy the relationship: factor of 0.001 is applied to the Table 11.1-5 values to s obtain the design basis concentrations in steam. C= (11.1 1) M(A + R) i 11.1.1.3 Tritium where: Tritium is produced by activation of naturally oc-C radionuclides concentration = curring deuterium in the primary coolant arad, to a s radionuclides input rate to coolant lesser extent, as a fission product in the fuel (Refer. M = reactor water mass ence 2). The tritium is primarily present as tritiated A = radionuclides decay constant oxide (HTO). Since tritium has a long half-life (12 R = sum of removal rates of the radionuclides years) and will not be removed by demineralizers or from the system. other process in the system, the concentration will be controlled by the rate ofloss of water from the Consequently,if the radionuclides input rate is system by evaporation or leakage. All plant process taken to depend primarily on the reactor thermal water and steam will have a common tritium concen. power, the adjustment factors to be applied to the tration. The concentration reached will depend on Reference Plant reactor water concentrations are the actual water loss rate; however, References 1 and given by-3 both specify a typical concentration of 0.01 pCi/gm which is stated in Reference 3 to be based on BWR P. M,- ( A + R,) experience adjusted to account for liquid recycle. Adjustment Factor (11.1-2) = This value is taken to be applicable for ABWR. P,. M-(A + R) 11.1.2.4 Argon-41 where the subscript *r* refers to the Reference Plant, P is the reactor thermal power and M, A, and R are Argon-41 is produced in the reactor coolant as a as defined above. The removal rate from the system consequence of neutron activation of naturally occur. is the sum of the removal rates due to the reactor ring Argon 40 in air which is entrained in the water cleanup system and the condensate feedwater. The Argon-41 gas is carried out of the demineralized andis given by: vessel with the steam and stripped from the system with the non condensables in the main condenser. FE + F. A*B E Observed Argon-41 levels are highly variable due to R= (11.1 3) the variability in air in-leakage rates into the system. M d-^- Punst 11.13

L i, l ABWR msmx Standard Plant' am n l 1 where: water and may plate-out on metal surfaces, concrete, cleanup system flow rate and paint. Radioiodines are found in ventilation air F = E' fraction of radionuclides removed in as methyl iodide and as inorganic iodine (particulate, l = cleanup demineralized elemental, and hypoiodous acid forms). steam flow rate F = A' ratio of radionuclides concentration in As a consequence of normal steam and water = steam to concentration in water leakage in to the drywell, equilibrium drywell con-(carryover ratio) eentrations will exist during normal operation. fraction of radionuclides in steam Purging of this activity from the drywell to the i B = which is circulated through the con-environment will occur via the Standby Gas densate demineralized Treatment System and will eake minor fraction of radionuclides removed in contributions to total plant releases. i E = 8 condensate demineralized. Airborne release data from BWR building ventila-i The Reference Plant and ABWR plant parame-tion systems and the main condenser mechanical ters are sbown in Table 11.16 and the vacuum pump have been compiled and evaluated in nuclide dependent removal rate parameters used for Reference 4, which contains data obtained by utility ABWR are shown in Tabie 11.1-7. The personnel and from specialin plant studies of oper-nuclide dependent parameters are the same as those ating BWR plants by independent organizations and used for the Reference Plant except for the fraction the General Electric Company. Releases due to circulated abrongh the condensate demineralized. process leakage are reflected in the airborne release The Reference Plant data is given for a plant without estimates discussed in Section 11.3. 1 pumped forward heater drains so that the fraction of l condensate treated by the demineralized is 1.0. In 11.1.6 References ABWR, which has pumped forward drains, the radi-onuclides are assumed to preferentially go with the

1. American National Standard Radioactive Term pumped forward flow (Reference 3). The effective for Normal Operation of Light Water Reactors, treatment fractions are.18 for iodines and.01 for ANSI /ANS 18.1.

other fission products and non coolant activation products (Reference 3).

2. Skarpelos, J.M. and R.S. Gilbert, Technical Deri-vation of BWR 1971 Design Basis Radioactive M-

l 11.1.4 Fuel Fission Production Inventory seria/ Source Terms, March 1975 (NEDO 10871). I Fuel fission product inventory information is used

3. Calculation of Releases of Radioactive Materials in establishing fission product source terms for acci-in Gaseous and Liquid Effluents from Boiling dent analysis and is discussed in Chapter 15.

Water Reactors, NUREG 0016, Revision 1, Janu-11.1.5 Process I2akage Sources

4. Martero, T.R., Airbome Releases From BWRsfor Process leakage results in potential release of Environmentallmpact Evaluations, March 1976 noble gases and other volatile fission products via (NEDO 21159).

ventilation systems. Liquid from process leaks is col-iccted and routed to the liquid-solid radwaste system. With the effective process offgas treatment systems now in use, the ventilation releases are relatively sig-mirinne contributions to total plant releases. Leakage of fluids from the process system results in the release of radionuclides into plant buildings. In general, the noble radiogases will remain airborne and will be released to the atmosphere with little delay via the building ventilation exhaust ducts. Other radionuclides will partition between air and A-.-a---e pgsen 11.1-4

Standard Plant M]M ua Table 11.11 I NOBLE RADIOGAS SOURCE TERMS IN STEAM Som m Tenn Decay Constant t = 30 ada holang ilhour) fuC1/sec) ) K-83m 3.73 E 1 2.9 E3 L-85m 1.55 E 1 5.5 E3 L-85 7.37 E-6 2.4 El L-87 5.47 E 1 1.5 E4 Kr-88 2.48 E-1 L7 E4 L-89 132 El 1.7 E2 Xe-131m 2.41 E 3 2.0 E1 Xc 133m 130 E 2 2.9 E2 Xe 133 5.46 E 3 8.4 E3 Xe-135m 2.72 E0 6.8 E3 { Xc135 7.56 E-2 2.2 E4 Xe 137 1.08 El 7.0 E2 Xe138 2.93 E0 2.1 E4 TOTAL 1.0 ES l Amendment Putur 11.1 5

'ABM imimx Standard Plant wa Table 11.12 IODINE RADIOISOTOPES IN REACTOR WATER Decay Constant Concentration Isot0De f / hour) fgCg*/,gl I I 131 3.59 E-3 14 E 2 1 132 3.03 E 1 1.4 E 1 1133 333 E 2 1.1 E 1 I134 7.91 E 1 -2.4 E 1 1 135 1.05 E 1 1.5 E 1 I l l l l ^^ Pwur i 11.1 4

ABWR uui. Standard Plant ua Table 11,13 NONVOLATILE FISSION PRODUCTS IN REACTOR WATER Decay Constant Concentration holong ( / hour) fuci/mi Rb-89 2.74 E0 2.1 E-2 Sr-89 5.55 E-4 33 E-4 Sr-90/Y-90 2.81 E-6 23 E-5 Sr-91 731 E 2 1.4 E-2 Sr-92' 2.56 E 1 ~.7 E-2 3 Y-91 4.93 E 4 13 E-4 Y-92 1.96 E 1 2.2 E-2 Y 93 6.80 E 2 1.4 E-2. Zr 95/Nb-95 4.41 E-4 2.6 E-5 Mo-99/Tc-99m 1.05 E-2 6.6 E-3 Ru 103/kh 103m 7.29 E-4 6.6 E-5 Ru 106/Rh-106 7.83 E-5 9.9 E-6 Te:129m 8.65 E 4 13 E-4 ) Te 131m 231 E-2 33 E-4 Te 132 8.89 E-3 33 E 5 I Cs 134 3.84 E 5 8.9 E-5 Cs-136 2.22 E-3 6.0 E-5 Cs-137 2.63 E-6 2.4 E-4 Cs 138 1.29 EO 4.1 E-2 Ba-140/La-140 2.26 E-3 13 E-3 Ansedment Putur 11.17 ('

1 L ABWR ~ Standard Plant

• 17 1

Table 11.13 - 1 i NONVOLATILE FISSION PRODUCTS IN REACTOR WATER (Continued) i i { Comessernsea jaalang Half-Life IgCd/g) ) l l Cc-141 8.88 E-4 ' 9.9 E-5 l Cc 144/Pr 144 1.02 E-4 9.9 E-6 Np 239 1.24 E-2 2.7 E 2 ' i NOTE: Nuclides shown as pairs are assumed to be in secular equilibrium. The parent decay constant and concentration are shown. Table 11.1-4 COOLANT ACTIVATION PRODUCTS IN REACTOR WATER AND STEAM Steam Reactor Water Concentration Concentration IEfdl.El IECdlld N-16 7.13 sec 5.0 E+ 1* 1.9 E + 2 1 1

• Use 200 pCi/gnfor operation with hydrogen water chemistry.

i j l Amenement Putur 11.14

1 ABWR maux Standard Plant wa q Table 11.15 ) NONCOOLANT ACTIVATION PRODUCTS IN REACTOR WATER Decay Constant Ceeesstration lantear i/ hour) laCi/.8) l Na-24 4.63 E 2 3.4 E-2 P 32 2.02 E 3 6.6 E-4 Cr 51 1.04 E 3 2.0 E-2 Mn 54 9.53 E-5 23 E-4 Mn 56 2.69 E 1 1A E 1 Co-58 4.05 E 4 6.6 E-4 Co-60 1.50 E-5 13 E-3 Fe-55 3.04 E 5 33 E-3 Fe 59 633 E-4 9.9 E-5 Ni-63 7.90 E 7 33 E 6 Cu-64 5.42 E-2 1.0 E-1 Zn-65 1.18 E-4 6.6 E-4 Ag-110m 1.16 E-4 33 E-6 W 187 2.90 E 2 1.0 E 3 l.- I l

• *
• Putur 11.1 9

,a r'! gy. Standard Plant ua Table 11.1-6 PIANT PARAMETERS FOR SOURCE TERM ADJUSTMENT Reference Parameter BanL. ABB1 Thermal Power. Mw 3400 3926 Reactor Water Mass,Ib 3.8 E+5 6.74 E +5 Cleanup System Mow Rate,Ib/hr 13 E+5 334 E+5 Steam Mow Rate 1.5 E + 7 1.68 E+ 7 Ratio of Condensate Demineralized How Rate to Steam Mow Rate 1.0

• See Table 11.1-7.

i l l 1 ( ) l h Petur 11.1 10

o ABWR 234uooux Standard Plant no a Table 11.17 REMOVAL PARAMETTIS FOR SOURCE TERM ADJU.7IMENT Other Radionuclides

• Parameter lodines (Excent Rh. Cs)

M Fraction removed by cleanup system 0.90 0.90 0.50 Fraction removed by condensate demineralizers 0.90 0.90 C.30 Ratio of concentrations in steam and reactor water 0.015 0.001 0.001 Fraction of radionuclides in steam treated by condensate demineralized 0.18 0.01 0.01 1 Including all non-coolant activation products and all non-volatile fission products except Rb and Cs Amendeneet Putur 11111

d. -l ABWR ~ m.1 i Standard Plant u. Figurn 11.1 1 through 11.13 have been deleated 4 ^ ^ 'Putur 11.1 14 thru 11.146

MM 2nA6100A1. Simaderd Plant nev. n 12.1 ENSURING THAT OCCUPATIONAL 12.1.1 3.1 Compliance with Regulatory Guide RADIATION EXPOSURES ARE AS 8.8 IhW AS REASONABLY ACHIEVABLE (AIARA) The design of the ABWR Standard Plant fully meets the intent of Regulatory Guide 8.8, and 12.1.1 Policy Considerations reflects the commitment of General Electric. Examples of compliance with Section C.2 of N Administrative programs and procedures, in Regulatory Guide 8.8 are delineated in 6 conjunction with facility design, ensure that the Subsection 123.1. Design features of the ABWR occupational radiation exposure to personnel will allow easy compliance with the recommendations l be kept as low as r asonably achievable (ALARA). of Subsections C.2 and C.4 of the guide. For instance, provisions are made in systems such as 12.1.1.1 Design and Construction Policies the reactor water cleanup system (RWCS) to allow flushing of the piping in shielded cubicles The AIARA philosophy was applied during the before entry, and to use remote reach rods. initial design of the plant and implemented via Portable breathing air is utilized in those internal design reviews. The design was reviewed areas where past experience indicates airborne in detail for ALARA considerations and was re-radioactivity has been a problem. Design viewed, updated and modified as necessary during provisions allow for remote operation of fuel the design phase as experience was gained from handling and radwaste cask filling. operating plants. Engineers reviewed the plant design and integrated the layout, shielding, ven-12.1.1 3.2 Compliance with Regulatory Guide tilation and monitoring instrument designs with 8.10 traffic control, security, access control and health physics aspects to ensure that the overall Out of ABWR Standard Plant Scope. See design is conducive to maintaining exposures Subsection 12.1.4.1 for interface requirement. 5 ALARA. 12.1.133 Compliance with Regulatory Guide All pipe routing containing radioactive fluids 1.8 were reviewed as part of the engineering design effort. This ensured that lines expected to Out of ABWR Standard Plant Scope. See g costsin significant radiation sources are Subsection 12.1.4.2 for interface requirement. = adequately shielded and properly routed to minimize exposure to personnel. 12.1.2 Design Considerations Operating plant results were continuously in. This subsection discusses the methods and tegrated during the design phase of the ABWR features by which the policy considerations of Standard Plant. Subsection 12.1.1 are applied. Provisions and designs for maintaining personnel exposures 12.1.1.2 Operation Policies ALARA are presented in detailin Subsections 123.1 and 123.2. Out of ABWR Standard Plant Scope 12.1.2.1 General Design Consideration for 12.1.13 Compliance with 10CFR20 and Regulatory ALARA Exposures Guldes 8.8,8.10 and 1.8 General Design considerations and method Compliance of the ABWR design with Title 10 of employed to maintain implant radiation exposures the Code of Federal Regulations, Part 20 ALARA, consistent with the recommendations of (10CFR20),is ensured by the compliance of the Regulatory Guide 8.8, have two objectives: design and operation of the facility within the guidelines of Regulatory Guides 8.8,8.10, and (1) minimizing the necessity for and amount of 1.8. personnel time spent in radiation areas, and A - a h - a' Putur 12.1 1 l

e e l 23A6100At. Ihandard Plant nev.a >(2) minimizing radiation levels in routinely 12.1.2.2.1 General Design Criteria occupied plant areas in the vicinity of plant equipment expected to require No specific instructions have been given to personnel attention. component designers and engineers regarding A1 ARA design as provided by specific Acceptance Both equipment and facility designs are Criterion 11.2 of SRP Section 12.1. However, k considered in maintaining exposures ALARA during the engineering design procedures require that plant operations. Events considered include the component design engineer consider the normal operation maintenance and repairs, applicable Regulatory Guides as a part of the refueling operations and fuel storage, inservice design criteria. This includes Regulatory Guide inspection and calibrations, radioactive waste 8.8. In this way, the radiation problems of a handling and disposal, etc. component or system are considered. A summary survey of the components designs was made to The features of the plant design which ensure determine the factors considered. The following that the plant can be operated and maintained paragraphs cite some examples of design with ALARA exposures will also serve to assist in considerations nade to implement ALARA, achieving ALARA exposures during the i decommissioning process. Examples of features 12.1.2.2.2 Equipment Design Considerations to which will assist in maintaining low occupational Unit Time Spent in Radiat!on Areas exposures during decommissioning include the j following: (1) Equipment is designed to bs operated and have its instrumentation and controls in i (1) Provisions for draining, flushing, and accessible areas both during normal and decontaminating equipment and piping. abnormal operating conditions. Equipment i such as the RWCS and the fuel pool cleanup (2) Design of equipment to n;inimize the buildup system (FPCS) are remotely operated, in-of radioactive material and to facilitate cluding the backwashing and precost opera-flushing of crud traps. tions. (3) Shielding which provides protection during (2) Equipment is designed to facilitate maintenance or repairs and during maintenance. Equipment such as the RHR decomatissioning operations. heat exchanger is designed with an excess of tubes in order to permit plugging of (4) Provision of means and adequate space for some tubes. The heat exchanger has drains utilization of movable shielding. to allow draining of the shell side

• water.

Some of the valves have stem packing of the (5) Separation of more highly radioactiu cartridge type that can be easily equipment from less radioactive equipment replaced. Refueling tools are designed for and provision of separate shielded drainage and with smooth surfaces in order compartments for adjacent items of to reduce contamination. Vessel and piping radioactive equipment. insulation is of an easily removable type. (6) Provision of access hatches for installation (3) The material selected for use in the system or removal or plant components. have been chosen to fulfill the environ-mental requirements. Valves, for example, p) Provision of design features such as the use grafoil stem packing to reduce leakage RWCS and the condensate demineralized to and maintenance. minimize crud buildup. (4) Past experience has been factored into 12.1.2.2 Eqnipment Design Considerations for current designs. The steam relief valves ALARA Esposures have been redesigned as a result of Ameedment Putur 12.1-2

MM 2sA61ooA1, meandard Plant an a have been redesigned as a result of (1) locating equipment, instruments, and inservice testing. sampling stations, which require routine maintenance, calibration, operation, or 12.1.2.2 3 Equipment Design Considerations to inspection, for case of access and minimum IJanit Component Radiation Icels required occupancy time in radiation areas; (1) Equipment and piping were designed to reduce (2) laying out plant areas to allow remote or the accumulation of radioactive materials in mechanical operation, service, monitoring, the equipment. The piping, where possible, or inspection of highly radioactive was constructed of seamless pipe as a means equipment; and to reduce radiation accumulation on the seam. The filter demineralizers in the RWCS (3) providing, where practicable, for and FPCS are backwashed and flushed prior to transportation of equipment or components maintenance. requiring service to a lower radiation sten. (2) Equipment designs include provisions for limiting leaks or controlling the fluid that 12.1.2.3.2 Minimizing Radiation Imels in does leak. This includes piping the Plant Access Aress and Vicinity of Equipment released fluid to the sumps and the use of drip pans with drains piped to the floor Facility general design considerations drains. directed toward minimizing radiation levels in plant accesa areas and in the vicinity of (3) The materials selected for use in the equipment requiring personnel attention include primary coolant system consist mainly of the following: austenitic stainless steel, carbon steel and low alloy steel components. (1) separating radiation sources and occupied areas where practicable (e.g., pipes or (4) The system design includes a RWCS and a ducts containing potentially high condensate demineralized system on the radioactive fluids not passing through reactor feedwater. These systems are occupied areas); g designed to limit the radioactive isotopes in the. coolant. (2) providing adequate shielding between radiation sources and access and service (5) External recirculation pumps and areas; recirculation piping were replaced by internally mounted recirculation pumps. (3) locating equipment, instruments, and n 5 Such pumps can be removed easily as an sampling sites in the lowest practicable integral or package unit for maintenance radiation zone; outside the lower drywell radiation zone. (4) providing central control panels to permit 12.1.2.3 Facility tayout General Design remote operation of all essential Considerations for Maintaining Radiation instrumentation and controls from the Exposures AIARA lowest radiation zone practicable; 12.1.2.3.1 Minimizing PersonnelTime Spent in (5) where practicable for package units, Radiation Artes separating highly radioactive equipment from less radioactive equipment, Facility general design considerations to instruments, and controls; minimize the amount of personnel time spent in radiation areas include the following: Am*=d==*nt Putur 12.1-3

t 21A6100AL ElemmAmed Plant mn (6) providing means and adequate space for utilizing moveable shielding for sources within the service area when required; (7) providag means to control contamination and to facilitate decontamination of potentially j contaminated areas where practicable; (8) providing means for decontamination of service areas; ) A W Putur 1113a

o j ABM 2m6 aasr Standard Plant arv. s 4 t, 4 40 S /s, y 35 b $30 Q $25 ( 20 l 15 0 1 2 3 4 5 TIME (SEC) 2 2 Break Area 1: DBA; 2: 0.5ft ; 3: 0.lft ; 4: 0.01ft Figure 20.311a DRYWELL PRESSURE VERSUS '!1ME (Response to Question 430.13) 20S72i Amendment Putur

e-i ABM 2sA6 MAT Standard Plant REV.B 400 4 / f M m Es 300 I O 54 O w EQ j /1 p 200 P 1

• C M

t.c __ S L i M 4 g4 100 H 0- - 0 6 10 15 TIME (SEC) 2 2 Main Steam line Break Area 1: DBA; 2: 0.5ft ; 3: 0.1ft ; 4: 0.01ft I Figure 203-11b DRYWELLTEMPERATUREVERSUS11ME (Response to Question 430.13) Amendment Putur i _ ____________-______a

o r y i '4 MN zwtooAT Standard Plant arv. s 400 4 m Ex< gnn p v i 0WO w W 200 b 2 m A. \\ .X M w 100 Ea 0 'O 1 2 3 4 5 TIME (SEC) Main Steam Line Break Area 1: DBA; 2: 0.5ft ; 3: 0.1ft ; 4: 0.01ft l Figure 20311c DRYWEIL TEMPERATURE VERSUS TIME j (Response to Question 430.13) I Amendsneat Futur 20.3-72k f

g, s b ABWR 2WimAT 1 ' fdandard Plant REV. B j l 2f30 --4 -2 5 J' L9 y210 w h1. 4 st; $160 [ xs 110 0 1 2 3 4 5 TIME (SEC) 2 Break Area 1: DBA; 2: C.Sft ; 3: 0.1ft ; 4: 0.01ft F ure20.311d DRYWEliTEMPERATUREVERSUSTIME 4 (Response to Questloo 430.13) DD Amendment Putur

MM 2sA61oorT Rimadard Plant nev. s l es 23 Ad v2 b 21

7 W

Q:; D $19 w M CG 'f/ 17 s IS 0 1 2 3 4 5 l TIME (SEC) Break Area 1: DBA; 2: 0.5ft ; 3: 0.lft ; 4: 0.01ft Figure 203-11e WE'!WEli PRESSURE VERSUS TIME (Response to Question 430.13) I A==d~nt Putur 20.3 72m I i

r o l

3 23A6100AT Standard Plant 170 4

150 [ w 00 130 D: Ea3 %X EaJ 110 90 0 1 2 3 4 5 TIME (SEC) Break Area 1: DBA; 2: 0.5ft ; 3: 0.lft ; 4: 0.01ft Figure 203-11f WET %T1L TEMPERATURE VERSUS TIME (Response to Question 430.13) Amcodment Futur 20.3-72n i

9 O' = 4 nb y e 9 a i l 8 I i l 1 3 X e 3 = i I 1 1 1 1 1 1 1 1 1 1 jl,i..p..i i sls< ll l i

t g,, j i

1. -.i 1, 1,., i,,.i X iI o n i i,

11.. i. ip 'lli x

j!g i,ii., m.s 1 i -"- I i 8lII' t ~: j gsg4 1 i,l.l.yi,{in ,a i ..jg I i., 1 il, ,,is. i g lf f. p T i i

l.gll il.l:1 f

u i i Mi 4w 3 I il,.Jl, i ....Tr a y i

u..r r

x = 1--f 2 1 11!!ll l v l. i .. m. i g,, j ' .! i...:::..!1-i

.$x i

. v . lf j '- l l ijt{!l/ X f t l u u s -. x-s o, Ii i l i l i i i -{11Ji U - X ' i_ a .. g .iTT il .l. g x ,1 T!! - - X li X i i h:

_Xi i

i i t h. n x 1 L i n.X p -g g i . i i . 1 X iTT--!! i i . x t n.n x l --TX

2 g

,i 'I X l l i - X X.

X X

i i i l i = hi R i i g R v i ni . x 4 i i n i r X 77 i

~

SI' Lif

i:g i n

i l X i i j x / . x X ii:Ri

iTp-'

X ~ 7 i x i t l ":g

l

i-I X

~i T; R

gi 9 :":g i 8 i

~

X 1 x,. ..._X. ,X x p. 1 .._g,. o . x !,* i. . X N, a :~-X~ l:- ~ X .i X X - n~j-.._ gg, l

- 7 i

---gi x gi i n a x x - i l X X e l;-" b :- ~X 4 X X I X. i l ! j [ K,: u g

• , j x

~iy x 1 .x l

"Jg y

j i i s n x j x ig l x . j x p :"ig

' ~i j

i X !' Y l :. h j:-..8; l$ :..>,i ((M-l!

-li.i X

i )l. l g:-. 8 ; l l t, s i 1 i 3 x; 1 :. . x

. 3 is i:

i i i i s l8 1 x x I I i .gl p x-l1l. x ,t. 1 n. l x g, ,j i i g J; - i i

jj g

pg h, IL __ i i x Jl. i t jj x J; . i i j i i

j

. i. ,.i t i i i,; a

t..

i-t i.

l ai a g. l i i l l o i liln1J! lij n ili l!oi ., fi ili i f fi l iJ n ifi i io. 4 m o i l i;l n i !i i l n i fili iluihi t so n no, ~ oJ n i i n i i s ii r. I EP:5 35 l l l l l

i s 0 lg( gj I ilI. hl I i a l l l i l i l l i g n_ g i i lii ..i c5 =- o x

.__ 3I g.

j lijig i.,i n i X I

7pi r ail

,[ g% I i u.. 1 ll 3 k, lllli i l ya li.,3,lii u i l X _ j e& l, fe d5 2221.l.l..l l T ii. f i.l.l R i s. f i a sil g gll,ll g = y ,j""TJ g, g

iip X

__ %.,.3 c X ..n:1 [ ! l $ [1 j 7

I ! 3*

I l l p. .h_. i T.' -1 1 l'> hl) . X [ i ! X J 1 iiT i s I i l i s i li i l ' ' l I l X , i t .'i ' = . i . X s i..l.

II

.i .e s -.]I li f . i j i h k f [ j l i - X [ g_R { l !iT i i Xi

Og

3 1 X

TT--h l i i - X i X x

j j si X

i - X X - X i . X s i iiTh ETX $A __2 R i i l g . i . X: X i .a X !iTf!

~ ~gf j i}}.

l i - X ! 8 i i - X It - Xj t i - X1 X. 1 - x X ,/ X 7 9

i gh i

. x ; e i mr mg- ~g irg-

i.y x

.g71 i X - X

~ X v

x l i r . X ! 8 2 :-~~

g. 's!-~~7:1 :

i. i "ib-$ i ~ 2~ W

7X '! :~-R.l:.7X 0

x i

~-

x en g E!!- - X S i X. R 4f ..g._.. R_

• _-.._-g e.

i._-.._ X ~[...._- ~ A i i X N.i l e n = a l A I i - k l

j, i

7 3, fr\\ g m x. j X I *; ;~T j

~iRi

~i; I l l A

X j T X i l' ~g r I :~ ~ :ila l l l :--}7 X X -V .i i I :~ ~g 1 j i l 6

- -g;l iE-l e

~g . e j ;:~ f l j j . i il lj i r ii.i lp-j x j x i is x l; ...g gel x l - i l i.

j jl x

1

i j

j ll. ru_ gll - i i X Jj irr-i l X 1 - i t ii X li i i i i 'lI l i , l lo i .. ; ih i l li ; llud..i ;!a i fi ihuh ; i >l

2 j.

i. aj [. i i i 3 i n i o fi i i p i n j , o i o i ili a r it ili i t,iti s i l i i i ii T ' ihi t i l i i r i i i . i i i I sr s a.

'r-y i e. 21A6100AT mandard PInnt mn ~ l

REGIONAL, APRMs CORE WIDE TIME HISTORY DISPLAY CORE PLOW SENSOR POWER

~ k 25% 'I a CONTROL RODS SLOCK ASD TRIP SENSOR FLOW 1 PUMP s 36 % ' - 1 PUMP I POWER { m 30% -j 1 PUMP O OR MORE V TRIP LOGIC SELECTED 1 PUMP / CONTROL ROD RUN.IN {d 1 PUMP IPUMP MANUAL 1? UMP 1 PUMP 10 PUMPS esOTES:

1. POWER a 30%: TO ASSURE POWER LEVEL SELOW SD% ROD LINE AT NATURAL CIRCULMION

3. PLOW s 30%: TO ASSURE PLOW RATE IS HIGNER THAN THAT OF EIGHT IWPs OPERATIONS WITH THE MINIMUM PUMP SPEED Figure 20.3-21 STABIL11Y CONTRO1E AND PROTECTION LOGIC (Response to Question 100.1) 203-X Amenda. cat Patur

.e

• 1.

o- ) 1 ABM 2s462004r 9mndard Plant Rev. e 0.9 - ) l CORI MDI I G(00NAL 0.s - Osewn0Ns ose m w0Ns 0.7 - 0.5 - a 0.5 - { 0.4 - p STABLE 0.3 - 0.2 - MANNEL ) NfDRODYNAWO.I 0.1 - 0$CILLAfl0N 0.0 e e a n 0.0 0.2 9.4 0.6 0.8 1.0 CMANNEL OCCAY RATIO 0 LOW CHANNEL / LOW CORE DR STABLE 0 LOW CHANNEL /HIGH CORE DR CORE WIDE O HIGH CHANNEL / LOW CORE DR CHANNEL 0 HIGH CHANNEL /HIGH CORE DR REGIONAL Figure 20.3 22 RELATIONSHIP BETWEEN MODES (Response to Question 100.1) Amendment Futur ?0.3-X}}