ML21294A106
Text
SSES-FSAR QUESTION 021.01 Provide the following additional information for the secondary containment:
(1)Show an appropriate plant elevation and section drawings, those structures and areas that will be maintained at negative pressure following a loss-of-coolant accident and that were considered in the dose calculation model; (2)Provide the Technical Specification limit for leakage which may bypass the Standby Gas Treatment System Filters, (e.g.,
valve leakage and guard pipe leakage); and, (3)Discuss the methods of testing that will be used to verify that the systems provided are capable of reducing to and maintaining a negative pressure of 0.25", e.g., within all secondary containment volumes.
RESPONSE
1)Following a loss-of-coolant-accident, all affected volumes of the secondary containment will be maintained at negative pressure. All these volumes are identified on Figures 6.2-24, 6.2-25, 6.2-26, 6.2-27, 6.2-28, 6.2-29, 6.2-30, 6.2-31, 6.2-32, 6.2-33, 6.2-34, 6.2-35, 6.2-36, 6.2-37, 6.2-38, 6.2-39, 6.2-40, 6.2-41, 6.2-42, and 6.2-43 as ventilation zones I, II and III. Also see Subsection 6.5.3.2 for a discussion of the reactor building recirculation system.
2)See Technical Specifications 3/4.6.5.3 for the limiting conditions for operation and the surveillance requirements for the SGTS. All leakage into the secondary containment is treated by the SGTS. Refer to Subsection 6.2.3.2.3 for a discussion of containment bypass leakage.
3)The Standby Gas Treatment System (See Subsection 6.5.1.1) in conjunction with the reactor building recirculation system (see Subsection 6.5.3.2) and the reactor building isolation system (see Subsection 9.4.2.1.3) is provided to produce and maintain negative pressure within affected volumes of the secondary containment. Actuation and operation of the above systems will be used to verify that the negative pressure is established and maintained.
Each ventilation zone is provided with redundant negative pressure controllers. Low pressure side inputs (low pressure sensing elements) to these controllers are located as follows:
Rev. 51, 02/97 021.1-1
SSES-FSAR Ventilation Zone I - Access are of El.749'-1" (See Figure 6.2-28)
Ventilation Zone II - Access area of El.749'-1" Ventilation Zone III - Refueling Floor, El.818'-1" (See Figures 6.2-30 and 6.2-40).
The quantity of air exhausted from the secondary containment will be such that in each affected ventilation zone the negative pressure will be maintained. The interconnecting ductwork of the recirculation system will equalize the negative pressure throughout each zone.
Rev. 51, 02/97 021.2-1
SSES-FSAR QUESTION 021.02 In accordance with the guidelines of SRP 6.2.3., provide the analysis and input assumptions used to determine the depressurization of the secondary containment for our independent review. This analysis should include a pressure time profile.
RESPONSE
Subsection 6.2.3.2.1 has been revised to include this information.
Rev. 51, 02/97 021.02-1
SSES-FSAR QUESTION 021.03 For bypass leakage of the secondary containment, provide the following information:
(1) The analytical method by which this bypass leakage was calculated.
(2) Where a closed system is relied on as a barrier, discuss how it meets the requirement specified for a closed system in Branch Technical position CSB 6-3.
RESPONSE
FSAR Subsection 6.2.3.2.3 and Table 6.2-15 have been corrected to eliminate the discrepancy mentioned in Question 312.10 and to resolve the above question.
Rev. 51, 02/97 021.03-1
SSES-FSAR QUESTION 021.04 For the post-blowdown depressurization analysis of the drywell, justify the heat transfer and condensation mechanisms assumed in the drywell modeling and the spray effectiveness of the ECCS spillage.
RESPONSE
Subsection 6.2.1.1.3.3.1.5 has been revised to include this information.
Rev. 51, 02/97 021.04-1
SSES-FSAR QUESTION 021.05 Provide a schematic diagram showing the geometric configuration of the downcomer and describe the method used to evaluate the head loss coefficient.
RESPONSE
FSAR Subsection 6.2.1.1.3.2 has been revised to include this information.
Rev. 51, 02/97 021.05-1
SSES-FSAR QUESTION 021.06 Provide a schematic diagram of a vacuum breaker including the size and describe the bases upon which the loss coefficient and vent flow area were determined.
RESPONSE
FSAR Subsection 6.2.1.1.3.2 has been revised to include this information.
Rev. 51, 02/97 021.06-1
SSES-FSAR QUESTION 021.07 Provide the vacuum breaker set point and opening time corresponding to the containment depressurization rate analysis.
RESPONSE
FSAR Subsection 6.2.1.1.3.2 has been revised to include the requested information.
Rev. 51, 02/97 021.07-1
SSES-FSAR QUESTION 021.08 Tables 6.2-9 and 6.2-10 present the short term (50 sec) mass and energy release rate for two postulated breaks. Provide the long-term (106 sec) mass and energy release rate for both recirculation line and main steam line break similar to that information presented in Tables 6.2-9 and 6.2-10.
RESPONSE
Subsection 6.2.1.3.1 has been revised to include this information.
Rev. 51, 02/97 021.08-1
SSES-FSAR TABLE 021.08-1 THIS TABLE WAS DELETED
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SSES-FSAR TABLE 021.08-2 THIS TABLE WAS DELETED
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SSES-FSAR QUESTION 021.09 Provide the minimum spray water temperature used for evaluation of an inadvertent spray actuation for containment depressurization.
RESPONSE
FSAR Subsection 6.2.1.1.4 has been revised to include this information.
Rev. 51, 02/97 021.09-1
SSES-FSAR QUESTION 021.10 With respect to containment steam bypass for small breaks, indicate your compliance with our proposed Branch Technical Position "Steam Bypass for Mark II Containments,"
which is enclosed.
RESPONSE
A comparison of the Susquehanna SES design with your proposed BTP "Steam Bypass for MK II Containments" is presented below. The item numbers correspond with the items in the BTP.
1.a. Bypass Capability, Containment Wetwell Sprays The wetwell spray system electrical instrumentation and controls supplied by GE meet the same ESF standards of quality, redundancy and testability as the RHR system, of which it is a part. The system is manually controlled and actuated.
The consequences of actuation of the wetwell spray on ECCS function are addressed in the response to Question 211.13.
1.b. Transient Bypass Capability Analyses The response to this item is covered by PP&L's response to SER Open Item #25, which was resolved as documented in SSER3, Subsection 6.2.1.7.
2.a. FSAR Subsection 6.2.6.5.1.1 addresses this item.
2.b. FSAR Subsection 6.2.6.5.1.2 addresses this item.
2.c. FSAR Subsection 6.2.6.5.1.2 addresses this item.
2.d. Visual inspection will be required every 18 months.
3.a. The Susquehanna design meets the intent of this item.
See Subsection 6.2.1.1.3.2.
3.b. See Technical Specification 3/4.6.4.
Rev. 51, 02/97 021.10-1
SSES-FSAR With respect to compliance with the proposed Branch Technical Position "Steam Bypass of Mark II Containments," the following Susquehanna SRP position statement is respectfully provided:
Issuance of the Standard Review Plans (SRP) post-dates the Susquehanna construction permit by more than 2 years. Therefore, no attempt was made to design the plant to the requirements of the SRPs. The Susquehanna FSAR was prepared using Revision 2 of Regulatory Guide 1.70 as much as practical for a plant of its vintage, with assurance from NRC management that compliance with this Regulatory Guide assured submittal of all necessary licensing information.
As documented in a letter of August 5, 1977 from G. G. Sherwood to E. G. Case of the NRC, the SRPs constitute a substantial increase in the information required just to describe the degree of compliance of various systems. This increase in turn represents a substantial resource expenditure which is unjustified and which could cause project delays if required of these projects. As stated in the reference letter, General Electric (and PP&L) believes that SRPs should be applied to FSARs only to the extent that they were required in the FSARs.
PP&L and General Electric believe the above position, which is the essence of a directive from Ben C. Rusche, Director of Nuclear Reactor Regulation, to the NRC staff dated January 31, 1977, is the appropriate procedure for review of the Susquehanna FSAR.
Rev. 51, 02/97 021.10-2
SSES-FSAR QUESTION 021.11 The design and proposed operation of the Containment Purge System is not discussed in sufficient detail for our review. Provide the information per our Branch Technical Position CSB 6-4, "Containment Purging During Normal Plant Operations."
RESPONSE
FSAR Subsection 6.2.5.2 has been revised to include this information.
Rev. 51, 02/97 021.11-1
SSES-FSAR QUESTION 021.12 In the unlikely event of a pipe rupture inside a major component subcompartment (e.g.,
the annulus and head region) the initial blowdown transient would lead to nonuniform pressure loadings on both the structure and the enclosed component(s). To assure the integrity of these design features, we require that you modify Appendix 6A and 6B to fully present your compartment, multi-node pressure response analysis, and provide the following information:
(1) Provide and justify the pipe break type, area, and location for each analysis.
Specify whether the pipe break was postulated for the evaluation of the compartment structural design, component supports design, or both.
(2) For each compartment, provide a table of blowdown mass flow rate and energy release rate as a function of time for the break which results in the maximum structural load, and for the break which was used for the component supports evaluation.
(3) Provide a schematic drawing showing the compartment nodalization for the determination of maximum structural loads, and for the component supports evaluation. Provide sufficiently detailed plan and section drawings for several views, including principal dimensions, showing the arrangement of the compartment structure, major components, piping, and other major obstructions and vent areas to permit verification of the subcompartment nodalization and vent locations.
(4) Provide a tabulation of the nodal net-free volumes and interconnecting flow path areas. For each flow path provide an L/A (ft-1) ratio, where L is the average distance the fluid flows in that flow path and A is the effective cross sectional area. Provide and justify values of vent loss coefficients and/or friction factors used to calculate flow between nodal volumes. When a loss coefficient consists of more than one component, identify each component, its value, and the flow area at which the loss coefficient applies.
(5) Describe the nodalization sensitivity study performed to determine the minimum number of volume nodes required to conservatively predict the maximum pressure load acting on the compartment structure. The nodalization sensitivity study should include consideration of spatial pressure variation; e.g., pressure variation circumferentially, axially, and radially within the compartment. Describe and justify the nodalization sensitivity study performed for the major component supports evaluation, where transient forces and moments acting on the components are of concern.
Rev. 51, 02/97 021.12-1
SSES-FSAR (6) Discuss the manner in which movable obstructions to vent flow (such as insulation, ducting, plugs, and seals) were treated. Provide analytical and experimental justification that vent areas will not be partially or completely plugged by displaced objects. Discuss how insulation for piping and components was considered in determining volumes and vent areas.
(7) Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for a selected number of nodes. Discuss the basis establishing the differential pressure on structures and components.
(8) For the compartment structural design pressure evaluation, provide the peak calculated differential pressure and time of peak pressure for each node. Discuss whether the design differential pressure is uniformly applied to the compartment structure or whether it is spatially varied. If the design differential pressure varies depending on the proximity of the pipe break location, discuss how the vent areas and flow coefficients were determined to assure that regions removed from the break location are conservatively designed.
(9) Provide the peak and transient loading on the major components used to establish the adequacy of the supports design. This should include the load forcing functions [e.g., fx(t), fy(t), fz(t)] and transients moments [e.g., Mx(t), My(t),
Mz(t)] as resolved about a specific, identified coordinate system.
RESPONSE
Appendix 6A has been revised to include this information.
Rev. 51, 02/97 021.12-2
SSES-FSAR QUESTION 021.13 Provide the projected area used to calculate these subcompartment loads and identify the location of the area projections on plan and section drawings in the selected coordinate system. This information should be presented in such a manner that confirmatory evaluations of the loads and moments can be made.
RESPONSE
Appendix 6A has been revised to include this information.
Rev. 51, 02/97 021.13-1
SSES-FSAR QUESTION 021.14 With regard to all Class 1E equipment located inside the containment building such as CRD hydraulic system, reactor vessel supports and all incore instrumentation leads, we require that the environment is maintained within the temperature range for which the equipment is qualified to operate.
Indicate if the Reactor Building Ventilation System (RBVS) is required to assist in the maintaining of an acceptable temperature range. If it is, provide the following information on the RBVS:
(1) Justification for not treating this system as an ESF system.
(2) The results of an analysis that the RBVS will not be a potential source for missiles and meets our pipe whip criteria.
(3) A discussion on the operating procedures to be initiated should the RBVS be unavailable.
(4) The location of all temperature sensors associated with the operation of the RBVS.
(5) The requirements imposed on this system in order to perform all Appendix J testing.
RESPONSE
The environment inside the primary containment is controlled by the Primary Containment Ventilation System which is described in Subsection 9.4.5. The information requested in the subparts of Question 021.14 is contained in the following paragraphs:
Question FSAR Subsection 021.14(1) 9.4.5.1(d) 021.14(2) 9.4.5.2 & 9.4.5.1(d) 021.14(3) 9.4.5.3 021.14(4) 9.4.5.2 & 7.6.1b.1.2.4 021.14(5) 9.4.5.2 Rev. 51, 02/97 021.14-1
SSES-FSAR QUESTION 021.15 We note that there is an approved topical report on your recombiners. Provide the proper references on this subject.
RESPONSE
FSAR Subsection 6.2.5.4 has been revised to include the proper references.
Rev. 51, 02/97 021.15-1
SSES-FSAR QUESTION 021.16 Provide the seismic class and quality group of the hydrogen monitoring system.
RESPONSE
See revised Table 3.2-1.
Rev. 51, 02/97 021.16-1
SSES-FSAR QUESTION 021.17 Section 6.2.5.2, states that all hydrogen mixing will be accomplished by the containment building ventilation system. In order to evaluate this system, we will require detailed layout drawings showing all ventilation ductwork, including intake and exit ports in order to establish circulation paths that will support your mixing assumptions.
RESPONSE
Subsection 9.4.5.2 has been revised to include this information.
Rev. 51, 02/97 021.17-1
SSES-FSAR QUESTION 021.18 Identify (1) the location of the hydrogen sample points in the drywell and suppression chamber and (2) location of CGCS suction and discharge points, with respect to local structures and equipment.
RESPONSE
FSAR Subsection 6.2.5.2 has been revised to include the requested information.
Rev. 51, 02/97 021.18-1
SSES-FSAR QUESTION 021.19 Discuss and schematically show the design provisions that will permit the personnel airlock door seals and the entire air lock to be tested. Discuss the design capability of the door seals to be leak tested at a pressure of Pa; i.e., the calculated peak containment internal pressure. If it will be necessary to exert a force on the doors to prevent them from being unseated during leak testing, describe the provisions for doing this and discuss whether or not the mechanism can be operated from within the air lock.
Also, discuss how the force exerted on the door will be monitored.
RESPONSE
Subsection 6.2.6.2 and Table 6.2-22 have been revised and Figures 6.2-58-1, 6.2-58-2 and 6.2-59 have been added to supply the requested information.
Rev. 51, 02/97 021.19-1
SSES-FSAR QUESTION 021.20 Identify the types of insulation used within the containment (e.g., reflective metal insulation, mass insulation, and encapsulated (sheathed) mass insulation) and discuss the methods of attachment to piping and components. Estimate, for a representative break location, the amount of insulation material that would be removed from the pipes by a LOCA. On the basis of the properties and characteristics of this material determine the locations it would accumulate and in what form. Discuss the potential for loose insulation and other debris to clog drains leading to the sump and the sump screening.
RESPONSE
FSAR Subsection 6.2.2.3 has been revised to include this information.
Rev. 51, 02/97 021.20-1
SSES-FSAR Q,UESTION 021. 21 We are aware_that revision 3 to the DFFR is to be submitted this Summer and that Revision . 2 which is now referenced is out-of-date, *as it does not adequately reflect the status of current pool dynamic loads. Discuss how the DAR will be updated to reflect this status and discuss any other reports you intend to submit to document your plant design.
RESPONSE .:
PP&L is working with the other Mark II owners to deve l op methodologies, analytical programs and test data which wil l provide .. improved definitions of hydrodynamic loads. This effort has resulted in Revision 3 to the DFFR, and is expected to ,result in further revision to that report.
Future revisions to the DFFR are expected to have no effect on the SSES DAR, since plant specifics as well as generic Mark II methodologies applicable to SSES will be incorporated into the DAR. The DAR has been updated to reflect the .current des i gn assessment methodologies used at SSES.
Rev. 51, 02/97 . 0:2L21-l
SSES-FSAR QUESTION 021.22 Based on our review of the information presented in subsection 6.2.1.5 of the FSAR, we find that the discussion of steam bypass from the drywell to the wetwell for a small break is incomplete and does not conform to the enclosed branch technical position (BTP) titled "Steam Bypass for the Mark II Containment." Accordingly, provide the appropriate discussions, justifications, and analyses to demonstrate compliance with the BTP.
RESPONSE
Compliance with this Branch Technical Position was assessed in the response to Question 021.10. Please refer to this response for the information being requested.
Rev. 51, 02/97 021.22-1
SSES-FSAR QUESTION 021.23 The response to Item 021.05 is inadequate. Provide a detailed calculation of the friction loss coefficient for the entire vent system. Discuss whether the results of the 4-T tests have been used to confirm the vent loss coefficient calculated. State the margin applied to the friction loss coefficient to account for any difference between the Susquehanna vent design and that of the 4-T facility.
RESPONSE
The calculation method for the friction loss coefficient is provided in a new FSAR subsection 6.2.1.1.3.2.1. The results of the 4T tests were not used to confirm the vent loss coefficient used for SSES. The vent system used in the 4T test is not prototypical of the SSES design and thus can not be applied directly to analyze the SSES design.
Rev. 51, 02/97 021.23-1
SSES-FSAR QUESTION 021.24 The response to Items 021.6 and 021.7 regarding the vacuum breaker is incomplete with regard to the vacuum breaker set point and opening time and the bases upon which the loss coefficient and vent flow area were determined. Provide the requested information and in addition:
(1) Describe the preoperational and inservice tests that will be performed to verify proper pressure setpoint and opening time; and (2) Provide the sensitivity limits and hysteresis characteristics of the switches.
Provide a discussion and the results of analysis performed to determine the maximum opening between valve disc and seat from when the position indicator system indicates that the valve is closed.
RESPONSE
Subsection 6.2.1.1.3.2 has been revised to address this question.
Rev. 51, 02/97 021.24-1
SSES-FSAR QUESTION 021.25 Section 6.2.1.1.4 of the FSAR states that the containment negative pressure is addressed in the Design Assessment Report. However, the information as provided is insufficient to allow an independent evaluation. Therefore, provide the analysis, including the thermodynamic model assumptions and the parameters used for the drywell cooldown transients that were performed to establish the containment negative pressure.
RESPONSE
This analysis is provided in revised Subsection 6.2.1.1.4.
Rev. 51, 02/97 021.25-1
SSES-FSAR QUESTION 021.26 Discuss in detail the design provisions incorporated for periodic inspection and operability testing of the containment heat removal systems' components such as pumps, valves, duct pressure-relieving devices and spray nozzles.
RESPONSE
The design provisions incorporated for periodic inspection and operability testing of the pumps and valves in the containment heat removal system are discussed in Subsection 6.2.2.4.
Preoperational testing of the containment spray nozzles is discussed in Section 6.2.2.2.
The spray nozzles will not be tested periodically.
There are no ducts, and hence no duct pressure-relieving devices, in the containment heat removal system.
Rev. 51, 02/97 021.26-1
SSES-FSAR QUESTION 021.27 Provide a detailed analysis of the available net positive suction head for the RHR pumps that are used as part of the containment heat removal system to demonstrate compliance with Regulatory Guide 1.1, "NPSH for Emergency Core Cooling and Containment Heat Removal Systems Pumps." Specify the required NPSH of the pumps.
RESPONSE
The requested information is contained in revised subsection 6.3.2.2.4.1 of the FSAR.
Rev. 51, 02/97 021.27-1
SSES-FSAR QUESTION 021.28 Describe the sizing analysis performed for the RHR suction screens. Provide a drawing that shows the suction screen assembly.
RESPONSE
The requested information is given in revised FSAR Subsection 5.4.7.2.2.
Rev. 51, 02/97 021.28-1
SSES-FSAR QUESTION 021.29 Estimate, for a representative break location, the amount of insulation that would be removed from pipes as a result of a LOCA. On the basis of the properties and characteristics of these materials, determine the locations it would accumulate and in what form and whether or not there is a potential for inhibiting suction flow due to clogging of the strainers.
RESPONSE
For a complete response to this question see the response to Question 021.20.
Rev. 51, 02/97 021.29-1
SSES-FSAR Question Rev. 52 QUESTION 021.30 Provide an analysis of the pressure and temperature response in the secondary containment due to a postulated LOCA in the primary containment. Discuss and justify the assumptions made in the analysis and specify the design leakage rate of the reactor building.
RESPONSE
This question is similar to NRC question 021.02 which requested a post-LOCA pressure profile and design leak rate for the secondary containment. Question 021.02 has been answered in full in FSAR Section 6.2.3.2.1.
In addition, Question 021.30 requests the post LOCA temperature in the secondary-containment. The temperature response is given in Dwgs. C-1815, Sh. 4, C-1815, Sh 5, C-1815, Sh. 6, C-1815, Sh. 7, C-1815, Sh. 8, C-1815, Sh. 9, C-1815, Sh. 10, C-1815, Sh.11 and C-1815, Sh. 12.
FSAR Rev. 58 021.30-1
SSES-FSAR I
Question Rev. 52
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QUESTION 021.31:
Identify all openings provided for gaining access to the secondary containment. and discuss the administrative controls that will be exercised over them. Discuss the instrumentation to be provided to monitor the status of the openings and whether or not the position indicators and alarms will have readout and alarm capability in the main control room.
RESPONSE: \
- 1) Secondary Containment Access Openings:
Door Nos Elev. Col. Coordinates Security Monitored 101 670 U/29 Yes 102 670 U/37.4 Yes 103-0 670 U/20.6 Yes 104-0 670 U/29 Yes 119A 676 P/20.6 Yes 120A 676 P/37.4 Yes 571-0 818 P/32 Yes Roof Hatch@ Elev. 872, coordinates: P/37.4 (Security Monitored)
- 2) Doors #119A, 120A and 571-0 provide access into the secondary containment through the use of card reader/cipher keyboard control. Doors 101, 102, 108-0, 104-0 and the roof hatch (#4001) will not normally be used to gain access into the secondary containment. All transactions will be logged into the Susquehanna Security Computer System. All alarms generated will annunciate at both the Security Control Center (SCC) and Alternate Security Control Center (ASCC). The plant control room will not have a readout or alarm capability. Both the sec and ASCC are, however, manned continuously 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.
Instrumentation to control and monitor the status of secondary containment is described in Chapter 7.0 of the Susquehanna SES Physical Security Plan.
FSAR Rev. 58 021.31-1
SSES-FSAR QUESTION 021.32 The following additional information related to potential bypass leakage paths is needed to provide an adequate response to Item 021.03.
(1) Expand Table 6.2-15 to include any branch lines which penetrate the secondary containment and connect to system lines which penetrate primary containment.
(2) For each line in (a), identify each of the potential leakage barriers; (3) For each air or water seal, perform an analysis that will demonstrate that a suf,ficient inventory of the fluid is available to maintain the seal for 30 days, and describe the testing program and proposed entries for the Technical Specifications that will verify the assumptions used in the analysis. Provide the basis fo~
the valve fluid leakage used in the analysis; (4) For each of these paths where water seals eliminate the potential for bypass leakage, provide a sketch to show the location of the water seal relative to system isolation valves;
( 5) Table 6. 2 -15 does indicate that the combustible gas sampling system is eliminated as bypass leakage paths .
Show how this system meets each of the requirements specified in Branch Technical Position, CSB 6-3, Section 9a-f, for a closed system; (6) Table 6.2-15 does not appear to list all potential bypass leakage paths (e.g., steam to RCIC system) .
Therefore, provide a list of all containment penetrations and seals which do not terminate in the secondary containment and an evaluation of these lines as delineated in the Branch Technical Position, CSB 6-3, "Determination of Bypass Leakage Paths in Dual Containment Plants 11 ;
(7) Table 6. 2-22 indicates that the feedwater lines and purge exhaust are secondary containment bypass leakage paths; Table 6.2-15 indicates the opposite. Discuss the discrepancy.
Rev. 51, 02/97 021.32-1
SSES-FSAR (8) The statement is made in Section 6.2.3.2.3 that no bypass leakage will occur following the design basis LOCA. Table 6.2-15 identifies those lines penetrating the primary containment which do not terminate inside the secondary containment and are considered as potential bypass leakage paths. Explain the inconsistency.
RESPONSE
Each of the eight parts of the question is discussed on the pages indicated below:
(1) Branch lines are discussed in subsection 6 . 2.3.2.3 and listed in Table 6.2-15.
(2) See Table 6.2-15.
(3) Water inventory for water seals used to prevent bypass leakage is discussed in subsection 6.2.3.2.3.1.
(4) The locations of water seals relative to the containment isolation valves are shown on system P&IDs. Appropriate cross-references are provided in Tables 6.2-15 .
(5) The combustible gas sampling system is located entirely within the secondary and primary containments and thus can not lead to bypass leakage.
( 6) Al 1 potential bypass leakage patches are listed in Tables 6.2-15.
(7) See Subsection 6.2.3.2.3.
(8) Subsection 6.2.3.2.3 has been amended.
Rev. 53, 04/99 021.32-2
SSES-FSAR QUESTION 021.33 The penetrations taken from Table 6.2-12 and listed below and the corresponding valve arrangement in Figure 6.2-44 are not consistent. Please clarify these inconsistencies.
The penetrations are: X - 10, 11, 13A, 19, 23, 24, 36, 53, 54, 55, 56, 85 - A & B, 86 -
A & B, and 215.
RESPONSE
Table 6.2-12 and Figure 6.2-44 have been revised to resolve all of the discrepancies noted.
Rev. 51, 02/97 021.33-1
SSES-FSAR QUESTION 021.34 Section 6.2.4 of the FSAR, "Containment Isolation System," should be augmented to provide the justification for any penetration including branch lines which do not conform to the requirements of the General Design Criteria. In addition, provide the containment isolation rationale for your design (e.g., RHR pump suction).
RESPONSE
The "Remarks" column of Table 6.2-12, along with the corresponding notes at the end of the table, discusses non-conformances to the General Design Criteria.
See also the responses to NRC Questions 021.38 and 021.39. The response to Question 021.50 discusses the rationale for non-conformance to the leak rate testing requirements of 10CFR50 with respect to certain penetrations.
Rev. 51, 02/97 021.34-1
SSES-FSAR QUESTION 021.35 Standard Review Plan 6.2.4, "Containment Isolation Systems," states that provisions should be made to allow the operator in the main control room to know when to isolate systems that require remote-manual isolation. Expand Table 6.2-12 to identify for those systems that rely on remote-manual isolation and the leakage detection provisions for these systems to assure that adequate information is available to the operator for identifying the affected line and for isolating it.
RESPONSE
FSAR Subsection 6.2.4.2 has been revised to include this information.
Rev. 51, 02/97 021.35-1
SSES-FSAR QUESTION 021.36 (1) The statement is made in Section 6.2.4.1 of the FSAR that instrumentation lines are designed to the provisions of Regulatory Guide 1.11. Provide the analysis performed which demonstrates that in the event of a rupture of the instrument lines and/or any component in the line outside the primary containment, the integrity and functional performance of secondary containment and its associated filtration systems are maintained.
(2) Revise Table 6.2-12 to include the isolation provisions for instrumentation lines penetrating the primary containment.
RESPONSE
(1) Instrument lines which penetrate reactor containment incorporate design features provided for complying with Regulatory Guide 1.11, as discussed in subsection 6.2.4.3.5.
FSAR Section 15.6.2 analyzes the consequences of an instrument line break outside primary containment. Barrier performance and radiological consequences are discussed in Subsections 15.6.2.4 and 15.6.2.5, respectively.
As mentioned in Subsection 15.6.2.3.1, instrument line breaks are considered to be bounded by the steam line break analyzed in Subsection 15.6-2, the postulated steamline break occurs within secondary containment; consequently, the steam tunnel blowout panels would relieve the pressure to the environs.
Therefore, it can be seen that secondary containment is not required for pipe breaks outside containment. Note also that the pressure-temperature analyses in Appendix 3.6A which verify that, with blowout panels, structural integrity is maintained for high energy pipe breaks outside primary containment.
(2) See new Table 6.2-12 (a).
Rev. 51, 02/97 021.36-1
SSES-FSAR QUESTION 021.37 Revise Table 6.2-12 to include the isolation provisions for the following penetrations X-35B, 36, 37A, 38A, 41, 42, 88 and 93.
RESPONSE
Table 6.2-12 has been updated to include the requested information.
Penetration X-35B is a capped spare penetration.
The CRD hydraulic system return line has been deleted. Therefore, no containment isolation provisions are required for penetration X-36.
Rev. 51, 02/97 021.37-1
SSES-FSAR QUESTION 021.38 Table 6.2-12 indicates that the isolation provisions for the containment spray system (X-39A, 205A), the floor drain (X-72A), the equipment drain (X72B), the RHR pump suctions (X-203A,C), the RHR pump test line and containment cooling (X-204A), the core spray pump suction (X-206A), core spray pump test and flush (X-207A), core spray min. recirculation (X-208A), HPCI pump suction (X-209) RCIC pump suction (X-214),
RHR min. recirc. (X-226A), suppression pool clean up and drain (X-243) and RHR relief valve discharge (X-246A) conform to the requirement of General Design Criteria 54. It is our position that the isolation provision to these lines should meet the requirement of GDC 56. However, a single isolation valve outside containment is acceptable as discussed in Standard Review Plan 6.2.4, II.3.e. Revise Table 6.2-12 to reflect our position and indicate if the other acceptable alternative for meeting the requirement of the GDC as specified in the SRP could be applied to any of these lines.
RESPONSE
The isolation provisions for the lines listed in Question 021.38 meet the requirements of GDC56 as modified by Standard Review Plan 6.2.4. Table 6.2-12 has been revised to reflect this position. Subsections 6.2.4.3.3.7, 6.2.4.3.3.8, and 6.2.4.3.6 have been added to discuss the alternative isolation provisions pursuant to SRP 6.2.4, paragraphs II.3.d and II.3.e.
Rev. 51, 02/97 021.38-1
SSES-FSAR QUESTION 021.39 Table 6.2-12 indicates that a check valve outside the containment is considered as a containment isolation valve for the standby liquid control (X-42), the HPCI pump minimum flow recirculation (X-211), the HPCI turbine exhaust (X-210), RCIC pump recirculation (X-217), the RCIC vacuum pump discharge (X-245). Provide justification for this approach.
RESPONSE
For the standby liquid control system, the simple check valve is inside, vice outside, containment. See Figure 6.2-44, detail K.
The penetration numbers for the RCIC vacuum breaker (X-245), RCIC pump recirc (X-216), and the RCIC vacuum pump discharge (X-217) were previously given incorrectly in the table but are now correct.
The justification for the approach taken for the RCIC penetrations is given in Subsection 6.2.4.3.3.2. The justification for the approach taken for the HPCI penetrations is given in Subsection 6.2.4.3.3.3.
Rev. 51, 02/97 021.39-1
SSES-FSAR QUESTION 021.40 The statement is made in Section 6.2.5.2 that nitrogen gas will be used for primary containment atmosphere control. Discuss the reasons which necessitate inerting the primary containment, since the hydrogen concentration does not exceed 3.5 volume percent.
RESPONSE
FSAR Subsection 6.2.5.2 has been revised to include this information.
Rev. 51, 02/97 021.40-1
Question Rev. 52 SSES-FSAR QUESTION 021.41 The statement is made in Section 6.2.5.3 of the FSAR that the recombiners and purge systems are activated when the hydrogen concentration reaches 3.5 volume percent. It is our position that the combustible gases resulting from a postulated loss of coolant accident should be controlled without release of radioactive materials to the environment. Therefore, the hydrogen purge system should only be used if, as a result of a post-LOCA event, both recombiner systems fail. Revise your state to indicate conformance with this position.
RESPONSE
Subsection 6.2.5.2 of the FSAR has been revised to indicate conformance to the above NRC position. As a result of NRC SER (NRC 2006-0026) which removed the hydrogen recombiners from the Technical Specifications and eliminated the requirements for the hydrogen recombiners, LDCNs 5295 and 5296 changed the design basis hydrogen generation to reflect SSES commitment to Regulatory Guide 1.7 Revision 3. The Unit 1 and 2 hydrogen recombiners are abandoned in place.
FSAR Rev. 70 021.41-1
SSES-FSAR QUESTION 021.42 Section 6.5.3.1 of the FSAR states that the containment purge system is manually operated from the control room at the discretion of the operator. It is our position that the purge system design should satisfy Branch Technical Position, CSB 6-4, "Containment Purging During Normal Plant Operation," if it is used during the reactor operational modes of power operation, startup, hot standby and hot shutdown, or the purge line isolation valve should be locked closed. Therefore, propose a purge system design that complies with the design provisions of the BTP. Also, provide the analysis identified in the BTP if the purge system is used during the operational modes specified above.
RESPONSE
The information requested in this question was provided in response to NRC question 021.11 in Rev. 1 (8/78) to the FSAR.
Rev. 51, 02/97 021.42-1
SSES-FSAR QUESTION 021.43 The response to Question 021.14 is incomplete with regard to the requirement imposed on the Reactor Building Ventilation System in order to perform all Appendix J testing.
Provide this information.
RESPONSE
In the response to NRC question 21.14 in Rev. 1 (8/78), it was pointed out that primary containment ventilation is provided by the Primary Containment Ventilation System and not the Reactor Building Ventilation System. Part (5) of the response to 21.14 referenced the FSAR section containing the Appendix J provisions of the Primary Containment Ventilation System. The Reactor Building Ventilation System provides ventilation for the secondary containment which is not subject to Appendix J requirements.
Rev. 51, 02/97 021.43-1
SSES-FSAR QUESTION 021.44 Table 6.2-22 identifies certain valves for which test pressure is not applied in the same direction as the pressure existing when the valve is required to perform its safety function, as required by Appendix J to 10 CFR 50. Demonstrate that the valve leakage rate is equivalent to or conservative with respect to that which would occur if the test pressure was applied in the direction when the valve is required to perform its safety function.
RESPONSE
Because of generic BWR valve arrangements, certain isolation valves are tested in the reverse direction as indicated on Table 6.2-22. Other BWR's with similar arrangements have tested their valves in a similar manner. Table 6.2-22 has been amended to provide justification for reverse flow testing.
Rev. 51, 02/97 021.44-1
SSES-FSAR QUESTION 021.45 Identify those fluid lines penetrating the containment which will be vented and drained to ensure exposure of the system containment isolation valves to the containment atmosphere and the full differential pressure during the containment integrated leakage rate (Type A) test. Discuss the design provisions that will permit this to be done. Those systems that will remain fluid filled for Type A test should be identified and justification provided.
RESPONSE
Venting and draining of those fluid lines penetrating the containment is done in accordance with the Leak Rate Test Program described in Section 6.2.6 Rev. 53, 04/99 021.45-1
SSES-FSAR QUESTION 021.46 Provide plan and elevation drawings of the air locks, and identify all mechanical and electrical penetrations. Discuss and schematically show the design provisions that will permit airlock door seals and the entire airlock to be tested.
RESPONSE
Subsection 6.2.6.2 has been revised to include this information.
Rev. 51, 02/97 021.46-1
SSES-FSAR QUESTION 021.47 Discuss the design capability of the door seals to be leak tested at a pressure of Pa; i.e., the peak calculated containment internal pressure. If it will be necessary to exert a force on the doors to prevent them from being unseated during leak testing, describe the provisions for doing this and discuss whether or not the mechanism can be operated from within the air lock. Also discuss how the force exerted on the door will be monitored.
RESPONSE
The information requested by this question was provided in response to NRC question 021.19 in Rev. 1 (8/78) of the FSAR.
Rev. 51, 02/97 021.47-1
SSES-FSAR QUESTION 021.48 Discuss your plans including the reactor building pressure sensing lines, that will become extensions of the containment boundary following a LOCA, in Type A test.
RESPONSE
The reactor building pressure sensing lines penetrate the secondary containment boundary only. They are 1/2 inch in diameter and are used to sense the outdoor atmosphere static pressure to provide high pressure side input to differential pressure transmitters. These transmitters control the reactor building ventilation system or the SGTS, in order to maintain the secondary containment at a negative pressure. The reactor building provides the secondary means of containment after the primary containment. Only primary containment (specified by Appendix J) is subject to the requirements of Type A testing.
The subject lines would, however, be tested as part of the secondary containment negative pressure tests.
Rev. 51, 02/97 021.48-1
SSES-FSAR QUESTION 021.49 Closed systems outside containment having a post accident function, become extensions of the containment boundary following a LOCA. Certain of these systems may also be identified as one of the redundant containment isolation barriers. Since these systems may circulate contaminated water or the containment atmosphere, system components which may leak are relied on to provide containment integrity.
Therefore, discuss your plans for specifying a leakage limit for each system that becomes an extension of the containment boundary following a LOCA, and leak testing the systems either hydrostatically or pneumatically. Also discuss how the leakage will be included in the radiological assessment of the site.
RESPONSE
Closed systems outside containment which become extension of the RCPB post-LOCA are listed in Table 6.2-21. Containment leak rate testing of these systems is discussed in Subsection 6.2.6.1 and Table 6.2-22. Leakage limits are established in the Leak Rate Test Program to ensure that leakage is maintained within that assumed in the DBA LOCA Dose Analysis.
Rev. 53, 04/99 021.49-1
SSES-FSAR QUESTION 021.50 Table 6.2-22 of the FSAR indicates that exemptions to 10 CFR 50 are required for certain lines. However, the nature and the rationale for the exemption are not given.
Provide this information for the following penetrations: X-10, 21, 23, 24, 85, 86A&B, 87, 93 and 218.
RESPONSE
Table 6.2-22 has been updated to show that no exemptions are required for fluid lines at the following penetrations: X-21, 23, 24, 85 A & B, 86A&B, 87, 93 and 218. See Table 6.2-22 for nature and justification of exemption requested for penetration X-10.
Rev. 51, 02/97 021.50-1
SSES-FSAR QUESTION 021.51 Based on our review of the information presented in Subsection 6.2.1.1.5 of the FSAR and the responses to Questions 021.10 and 021.22, we find that your discussion of steam bypass from the drywell to the wetwell for a steam line break to be unacceptable.
In your response you indicated that the requested information represents a substantial resource expenditure, which is unjustified because the information is required only to describe the degree of compliance of various systems. We find this not to be the case.
The staff's position that was attached to Item 021.11 is intended for implementation on all Mark II containments because of its safety significance.
In addition, you stated in your response to 021.10 that Ben C. Rusche's directive to the NRC staff dated January 31, 1977 is the appropriate procedure for review of the Susquehanna FSAR. It should be noted that the referenced letter concerns documentation of departures from the Standard Review Plan. Question 021.10 was forwarded to you specifying our position that the Susquehanna containment should be designed to have a steam bypass capability as characterized in Appendix I to Standard Review Plan 6.2.1.1.c. (It should be noted that Appendix I has been previously forwarded with Question 021.22 as Branch Technical Position CSB 6-X). Accordingly, provide the appropriate discussions, justifications and analyses to demonstrate compliance with the Appendix I to SRP 6.2.1.1.c.
RESPONSE
See revised response to Question 21.10.
Rev. 51, 02/97 021.51-1
SSES-FSAR QUESTION 021.52 Section 6.2.1.1.3.2 of the FSAR indicates that the loss coefficient of the vacuum breaker was calculated based on actual flow measurements conducted in the manufacturer's shop. Discuss the applicability of the test performed (e.g., flow regime) considering the conditions that are expected in the containment when the vacuum breaker is required to operate.
Provide a diagram showing the locations of the vacuum breakers relative to the downcomer and the floor slab. Discuss your plan to comply with the requirement (Item 3.b) of Appendix I to SRP 6.2.1.1.C.
RESPONSE
A. The applicability of the vacuum breaker loss coefficient test is discussed in Subsection 6.2.1.1.3.2.2.
B. Figure 6.2-56 shows the vacuum breaker flange connection to the downcomer.
The location of the vacuum breaker relative to the diaphragm slab is also shown in this figure.
C. The degree of compliance with Item 3.b of Appendix I to SRP 6.2.1.1.c was discussed in response to Question 021.10.
Rev. 51, 02/97 021.52-1
SSES-FSAR QUESTION 021.53 The response to Item 021.32(4) references system P&ID's which do not show the location of the water-seal relative to the system isolation valves; provide a sketch to show these elevations for each path where water seals eliminate the potential for bypass leakage.
RESPONSE
Figures 6.2-66B, 6.2-66C, 6.2-66D, 6.2-66E, 6.2-66F have been provided to show the location of water-seals relative to the system isolation valves. The elevations, pipe quality group and the seismic classification have been provided in the sketches.
Rev. 51, 02/97 021.53-1
SSES-FSAR QUESTION 021.54 The statement is made in Subsection 6.2.3.2.3 of the FSAR that closed systems are not relied upon as barriers to eliminate bypass leakage. It was further stated that isolation valves inside or outside the primary containment are considered to limit but not to eliminate bypass leakage. It appears that some of the lines listed in Tables 6.2-15a (e.g., RBCCW) have been eliminated as potential bypass leakage paths because of either or both of the above mentioned statements. Provide clarification for this apparent discrepancy. In addition, provide the quality group and seismic qualification of the closed systems that are relied upon to eliminate bypass leakage.
RESPONSE
See Subsection 6.2.3.2.3 Rev. 53, 04/99 021.54-1
SSES-FSAR QUESTION 021.55 Item 2 on page 6.2-36a of the FSAR provides justification for elimination of the RWCU line (DBA-101) as a bypass leakage path. Similarly, provide the justification for line EBC-104.
RESPONSE
See Subsection 6.2.3.2.3 and Table 6.2-15.
Rev. 53, 04/99 021.55-1
SSES-FSAR QUESTION 021.56 Note 8 in Table 6.2-12 of the FSAR indicates that the valve isolates two piping penetrations. Provide a sketch to show the typical arrangement and discuss how such an arrangement meets the General Design Criteria.
RESPONSE
Details have been added to Figure 6.2-44 to show the arrangement of the valves, associated with the penetrations to which Note #8 applies.
The General Design Criteria do not forbid the use of a common isolation valve; the Criterion met by each of the subject penetrations is listed in Table 6.2-12.
Rev. 51, 02/97 021.56-1
SSES-FSAR QUESTION 021.57 Provide justification for using a check valve outside the containment as a containment isolation valve for the following penetrations; X-210, 211, 214, 215, 216 and 217.
RESPONSE
As stated in the response to NRC Question 021.39, the justification for using check valves outside containment for the identified penetrations is given in Subsections 6.2.4.3.3.2 and 6.2.4.3.3.3.
Note that the outside containment isolation valve for penetration X-214 is a gate valve, not a check valve.
Rev. 51, 02/97 021.57-1
SSES-FSAR QUESTION 021.58 Item II.6 of Standard Review Plan 6.2.4, "Containment Isolation Systems," requires diversity in parameters sensed for initiation of containment isolation. Provide justification for not having diversity in the parameters sensed to initiate isolation of the following lines; X-121, 35-B, 208A, 211, 215, 216, 217, 226A and 246A.
RESPONSE
The requested information has been provided in the appropriate subparagraphs of FSAR Subsection 6.2.4.
Rev. 51, 02/97 021.58-1
SSES-FSAR Question Rev. 53 QUESTION 021.59 With regard to the control rod drive system provide the following information:
(1) The piping integrity test to detect any leakage from the hydraulic control unit; (2) The type and number of valves and method of actuation on the charging water, drive water and cooling water; (3) The type of indicators available to the operator to indicate any leakage; (4) Whether the CRD system is vented during the performance of the type A test; and (5) The proposed Technical Specification limit on leakage through the hydraulic control unit.
RESPONSE
Response to (1)
Prior to shipment, all Hydraulic Control Units (HCU) are hydrostatically tested in accordance with the applicable code(s) of construction (Reference Table 3.2-1).
Appropriate portions of the CRD system are ASME Code Class 2, and piping integrity is demonstrated in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.
See the Susquehanna SES Technical Specifications.
Response to (2)
The number and types of valves on the charging water, cooling water and drive water headers are shown on Dwgs. M-146, Sh. 1 and M-147, Sh. 1.
Response to (3)
FSAR Subsection 4.6.1.1.2.4.2 has been revised to include this information.
Response to (4)
This information is provided in FSAR Table 6.2-22 Note 20.
FSAR Rev. 59 021.59-1
SSES-FSAR Question Rev. 53 Response to (5)
There are no Technical Specification limits necessary for the possible leakage from the HCU.
The design of the CRD system, of which the HCU are a part, is such that a loss of fluid will not prevent or inhibit the execution of the system's safety function (i.e., scram) as long as the scram accumulator operability is under Tech. Spec. control. The operational status of the accumulator is provided by the instrumentation at each hydraulic control unit which monitors the accumulator pressure and the total amount of water which may have collected on the gas side of the accumulator due to internal leakage.
FSAR Rev. 59 021.59-2
SSES-FSAR QUESTION 021.60 The statement is made in Subsection 6.2.4.3.2.1 of the FSAR that the feedwater valve is remote manually closed from the control room upon operator determination that continued makeup from the feedwater system is unavailable or unnecessary. We find this approach acceptable, however, discuss the information that will be available to the operator to alert him of the need to isolate the feedwater, the time when this information would become available, and the time it would take the operator to complete this action.
RESPONSE
For response, refer to Subsection 6.2.4.3.2.1.
Rev. 51, 02/97 021.60-1
SSES-FSAR QUESTION 021.61 Question 021.42 requests certain information regarding the containment purge system addressed in Section 6.5.3.1 of the FSAR. Your response to that question and to Question 021.11 is related to the containment hydrogen purge system. Provide the information requested in 021.42.
RESPONSE
Subsection 6.5.3.1 has been revised to provide this information. Refer to this Subsection for response to this question.
Rev. 51, 02/97 021.61-1
SSES-FSAR QUESTION 021.62 The response to Question 021.44 does not provide enough justification for the testing of certain containment isolation valves in the reverse direction. Therefore provide the following information:
(1) The method by which these penetrations are to be tested and how the leakage will be assigned to that penetration (i.e., if test pressure is between the valves and the total leakage is assigned to that penetration);
(2) Justification that the isolation valves have similar leakage characteristics in both the forward and reverse direction for those penetrations discussed in item a above; and (3) Justification that these testing methods will yield results at least equivalent to the case when the valve is tested in the forward direction for any other valves that will be tested in the reverse direction.
RESPONSE
See Subsection 6.2.6.3 Rev. 53, 04/99 021.62-1
SSES-FSAR QUESTION 021.63 Provide the rationale for not including leakage from valves identified in Table 6.2-22, with notes 14 and/or 26 in the 0.60 La total Type B and C tests.
RESPONSE
Refer to Subsection 6.2.6.3 for response.
Rev. 51, 02/97 021.63-1
SSES-FSAR Question Rev. 52 QUESTION 021.64 Discuss the method by which water seals will be maintained for 30 days following LOCA.
Specify the quality group and the seismic qualification of all components that are relied upon to perform this function.
RESPONSE
The method by which water seals will be maintained for 30 days post-LOCA is discussed in Subsection 6.2.3.2.3.1. The quality group and seismic category of piping relied upon to provide a water-seal is shown on the water-seal sketches (Figure 6.2-66 B through F), and the associated P&ID's. Refer to Dwg. M-100, Sh. 2 for a key to pipe line numbers.
FSAR Rev. 58 021.64-1
SSES-FSAR QUESTION 021.65 The statement is made in Subsection 6.2.6.3 of the FSAR that a factor will be applied to contaminated liquid to determine the airborne fraction that will be added to Type B and C test totals. Provide the methods by which this factor is determined.
RESPONSE
The liquid leakage from tests of lines designed to remain filled with liquid following a LOCA will be reported as total liquid leakage. This is consistent with 10 CFR 50, Appendix J (IIIC.3) and the proposed Revision 4 of ANS N274, "Containment Systems Leakage Testing Requirements." A factor to determine the fraction of the leakage which becomes airborne cannot be determined with a large degree of accuracy due to the uncertainty of the leakage location, the water temperatures, the varying activity levels in the leaking water and the dependence on the time after the accident when the leak occurs. The dose effects of liquid leakage from lines designed to be filled with liquid for the duration of the LOCA are bounded by the analysis for ECCS system leakage given in Subsection 15.6.5.
Subsection 6.2.6.3 has been revised to delete the reference to the factor for determining the airborne fraction.
Rev. 51, 02/97 021.65-1
SSES-FSAR QUESTION 021.66 The statement is made in Subsection 6.2.6.5.1.2 of the FSAR that the low pressure test to determine drywell to suppression chamber atmosphere bypass area will be conducted at each integrated leakage rate test interval. This approach is unacceptable.
Our position is stated in Appendix I to SRP 6.2.1.1.C. Revise the FSAR to indicate compliance with our position.
RESPONSE
See revised Section 6.2.6.5.1.2.
Rev. 53, 04/99 021.66-1
Question Rev. 70 SSES-FSAR QUESTION 021.67 With regard to the analysis of hydrogen production and accumulation within the containment following a postulated loss-of-coolant accident:
(1) Provide the corrosion rates for the zinc base paint and galvanized steel in this environment. In so doing provide a copy of references 6.2-7 and 6.2-8 for our review and discuss the applicability of these referenced data considering the environmental conditions that are expected following a LOCA.
(2) The staff is currently undertaking additional effort toward better defining the various sources of hydrogen, including zinc-rich paints and organic materials.
The attached figure depicts the hydrogen generation rates as a function of temperature that the staff currently uses for confirmatory analysis. Provide a sensitivity study based on this figure that shows that the hydrogen concentration inside the containment will not exceed the acceptance criterion of 4 volume percent. In so doing provide the time the hydrogen recombiner should be turned on and the time needed to heat up the recombiner.
RESPONSE
Subsection 6.2.5, "Combustible Gas Control in Containment" has been revised to provide the information required by this question. The zinc-based paint and galvanized steel corrosion rates have been provided in Table 6.2-13. The references requested by NRC have been replaced with references to non-proprietary data; thus, they have not been supplied. As a result of NRC SER (NRC 2006-0026) which removed the hydrogen recombiners from the Technical Specifications and eliminated the requirements for the hydrogen recombiners, LDCNs 5295 and 5296 changed the design basis hydrogen generation to reflect SSES commitment to Regulatory Guide 1.7 Revision 3. The Unit 1 and 2 recombiners are abandoned in place.
FSAR Rev. 70 021.67-1
SSES-FSAR QUESTION 021.68 With regard to the secondary containment's functional capability:
(1) Discuss if there is any connection between the Unit 1 and Unit 2 secondary containments. If there is a door that separates the two secondary containments, discuss if the SGTS is capable of maintaining a 1/4" water gauge negative pressure in the affected unit's secondary containment assuming the door was open at the time of LOCA.
(2) Discuss the design provision incorporated to prevent such doors from being inadvertently opened.
(3) Discuss the test that will be performed to verify the inleakage assumption and the drawdown time for reestablishing -0.25 inches of water gauge following LOCA.
RESPONSE
(1),(2) There is one door on elev. 779'-1" between Unit 1 and Unit 2. Since this is in Ventilation Zone III which is common to both units, it has no effect on secondary containment operation.
(3) The test procedure outlined in Subsection 4.6.5.1 of the Technical Specification will be used to verify the inleakage assumption and the drawdown time for reestablishing -0.25 inches of water gauge following a LOCA. The difference between the simulated and actual LOCA is the absence of heat transferred from the drywell head. This heat is insignificant when compared with that generated by operating equipment and lights and can be discounted. (See FSAR Subsection 6.2.3.2.1 for the analysis of the post-LOCA pressure transient in the secondary containment.)
Rev. 51, 02/97 021.68-1
SSES-FSAR QUESTION 021.69 The LOCA and SRV related pool dynamic loads that are currently acceptable to use are discussed in NUREG-0487. Table IV-1 of NUREG-0487 summarizes these Mark II pool dynamic loads. By letter, dated February 2, 1979, you indicated on Table IV-1 the LOCA related dynamic loads acceptable to the staff that will be adopted for SSES.
Revise the DAR to incorporate this information and provide the same information for the SRV related pool dynamic loads. For both the SRV and LOCA loads indicate the alternative criteria that will be used for each item for which an exemption is proposed, and provide references that discuss these alternative criteria.
RESPONSE
The information requested by this question was provided to the NRC by letter dated November 2, 1979, identified as PLA-417, from Mr. N.W. Curtis to Mr. Olan D. Parr.
Rev. 51, 02/97 021.69-1
SSES-FSAR QUESTION 021.70 Subsection 4.2.1.1 of the DAR state that the drywell pressure transient used for the pool swell portion of LOCA is based on the methodology described in NEDO-21061.
Subsection III.B.3.a.6 of NUREG-0487 requires that a comparison similar to those presented in reference1 be made if the model used is different from the model described in NEDM-10320. We require the model prior to completion of review of the pool swell calculations.
RESPONSE
There is no methodology for calculating drywell pressures in the DFFR (NEDO-21061) and the DAR does not say that the pressure transient was based on such methodology.
The Mass and Energy Release Report results were used for calculating the drywell pressure transient using NEDM 10320 methodology.
1 Reference (1) Letter Response to NRC Request for Additional Information (Round 3 Questions), to J. F. Stolz (NRC-DPM) from L.J. Sobon (GE) dated June 30, 1978.
Rev. 51, 02/97 021.70-1
SSES-FSAR QUESTION 021.71 Subsection 4.2.2.2 of the DAR states that the chugging loads on submerged structures and imparted on the downcomers will be evaluated later. Provide the present status of these evaluations and the schedule for your submission of the completed evaluation.
RESPONSE
The calculation of submerged structure loads due to chugging use the improved chugging load methodology developed under March II Owners Group Task A16. The appropriate design sources are used with the Green's function solution for the SSES annular containment to provide the pressure distribution in the suppression pool. The pressure around a structure is integrated to determine the net pressure load on the structure. A description of this methodology and verification is included in the DAR.
The chugging source used are developed from the pressure time histories provided by KWU for the design assessment (see SSES DAR Section 9.5.3).
The downcomer has been assessed for the chugging loads and the results is incorporated into the DAR. The other submerged structures have also been evaluated.
Rev. 51, 02/97 021.71-1
SSES-FSAR QUESTION 021-. 72 Statements are made in Subsections 4.2.3.2 and 4.2 . 3.3 of the DAR that plant unique data of the Susquehanna SES intermediate bre*ak accident ( IBA) and small break accident (SBA) are estimated from curves for a typical Mark .II containment.
Discuss the applicability of the-se analyses (e.g., power level, initial. conditions, downcomer configuration, etc.) to Susquehanna SE.S.
RESPONSE: -
The Susquehanna SES FSAR Subsections 6.2.1.1.3.3.4 and 6.2.1.1-3.3.5 give consideration to . the SBA and IBA conditions and FSAR figures 6.2-14 and 6.2-15 give the IBA unique curves for
- Susquehanna SES based on the initial conditions as given in FSAR Tables 6.2~1 thru b.2-4. In comparing the IBA Susquehanna SES unique FSAR figures with t_he Susquehanna SES DAR figures 4-50, 4-51 and 4-52, which are .. generic DFFR, SBA and IBA, it can be seen that basically the generated curves provide similar design basis mass energy data._ (See figures 021. 72-1 and 021.72-2) Thus we expect the Susquehanna SES.SBA conditions to be similar also to the DFFR {or DAR) figures. The parameters for the DFFR 'and Susquehanna SES are provided in DFFR {NEDO 21061) Table 4-1 arid Susquehanna SES FSAR Tables 6. 2-1 and
- 6. 2-4 respectively. The differences in parameters are as follows:
QEEE ~
Drywell Free Air Volume ft 3 2.029 X 10 5 2.396 X 10 5 Wetwell Free ..Air Volume ft"3 1.455 X 10 5 1. 4859 X 10 5 Wetwell-Water Volume ft 3 1.2 X 10 5 1. 224 X 10 5 Drywell Initial Pressure psig.-- 0.75 1.5 Wetwell Initial Pressure psig. 0.75 1.5 Vent Submergence ft. 11 11. 5 Number of vents 108 87 I Rev. 5:L, 02/97 021. 72 COMPARISON OF SSES & DFFR IBA CONDITIONS 50.---------------------------------------,
40 1 ssES
"*~
a.
30 F~--
SSES w oFFB,.-
a:::
(/)
(/)
w a:::
a.. 20 10 1 DRYWELL PRESSURE RESPONSE 2 WETWELL PRESSURE RESPONSE 0
o~~~-~~~~~-~~-~~~~~-~~-~~~~~--~-~~~~~
10 102 10 3 TIME (sec)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PRESSURE RESPONSE FOLLOWING AN IMMEDIATE BREAK ACCIDENT (IBA)
FIGURE 021.72-1, Rev 47 AutoCAD: Figure Fsar_021_72_1.dwg
COMPARISON OF SSES & DFFR IBA CONDITIONS 400---------------------------------------
1 DRYWELL TEMPERATURE 2 SUPPRESSION POOL TEMPERATURE 300 LL w
0::
~
ffi 200 _,,,.A D..
- i1:
w _.,,,,. .,,,,. .,,,,.
I-oFFR \BA _ _ _ _
-~ES UNIQUE \BA 100 SSES UNIQUE IBA 0.1 10 TIME (sec)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT TEMPERATURE RESPONSE FOLLOWING AN IMMEDIATE BREAK ACCIDENT (IBA)
FIGURE 021.72-2, Rev 49 AutoCAD: Figure Fsar_021_72_2.dwg
SSES-FSAR QUESTION 021.73 Provide the information previously requested in 020.44 regarding loads resulting from pool swell waves following the pool swell process or seismic slosh. Discuss the analytical model and assumptions used to perform these analyses.
RESPONSE
The analytical method of calculating the loads resulting from seismic slosh and the assumption used are described in a writeup included in the DAR.
Rev. 51, 02/97 021.73-1
SSES-FSAR Question Rev. 52 QUESTION 021.74 Provide a list and drawing to identify all piping, equipment instrumentation and structures in containment that may be subjected to pool dynamic loads. In addition, provide drawings to show the location of access galleys in the wetwell, the vent vacuum breaker configuration, wetwell grating, vent bracing configuration, vent configuration in the pedestal region of wetwell and large horizontal structures in the pool swell zone.
RESPONSE
The drawings requested in 021.74 are listed below. Symbols utilized on these drawings are identified on Dwgs. M-100, Sh. 1, M-100, Sh. 2 and M-100, Sh. 3.
LARGE PIPING Figure Title 021.74 - 1 Reactor Building, Unit 1, Primary Containment, M.S.R.V. Discharge in Suppression Pool.
- 2 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 704'-0".
- 3 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 719'-1".
- 4 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 738'-11 1/2".
- 5 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 752'-2 1/2".
- 6 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 779'-1
- 7 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Misc. Sections.
- 8 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Section C-C.
- 9 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El 761'-1".
- 10 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Section L-L.
- 11 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 806'-0".
- 12 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Section M-M.
- 13 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Misc. Details.
FSAR Rev. 58 021.74-1
SSES-FSAR Question Rev. 52 SMALL PIPING Figure Title 021.74 - 14 M.S.R.V. Discharge in Suppression Pool, Reactor Building, Unit 1, Primary Containment.
- 15 M.S.R.V. Discharge in Suppression Pool.
- 16 Plant Design Drawing, Reactor Building, Unit 1 Area 26, Plan of El. 704'-0".
- 17 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 719'-0".
- 18 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El.
738'-11 1/2" Sht. 1.
- 19 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El.
738'-11 1/2", Sht. 2.
- 20 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El.
752'-2 1/2".
- 21 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 779'-1".
- 22 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Misc. Plan of El. 719'-1".
- 23 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Misc. Sections
& Details.
- 24 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 761'-1".
- 25 Plant Design Drawing, Reactor Building, Unit 1, Area 26, Plan of El. 806'-0".
FSAR Rev. 58 021.74-2
SSES-FSAR Question Rev. 52 EQUIPMENT LOCATION Figure Title 1.2 - 17 Equipment Location, Reactor Building, Plan of Basement, El 645'-0".
- 18 Equipment Location, Reactor Building, Unit 1, Plan of El. 670'-0".
- 19 Equipment Location, Reactor Building, Unit 1, Plan of El. 683'-0".
- 20 Equipment Location, Reactor Building, Unit 1, Plan of El. 719'-1".
- 21 Equipment Location, Reactor Building, Unit 1, Plan of El. 749'-1".
- 22 Equipment Location, Reactor Building, Unit 1, Plan of El. 779'-1".
021.74 - 26 Instrument Location Drawing, Reactor Building, Unit 1, Area 26, Plan Below El. 704'-0".
Figure 021.74-27 (Reactor Building 1&2 Primary Containment-Suppression Chamber Platform) shows wetwell grating. Vent bracing configuration is presented in the DAR, Ammend. #1, Figures 1-3 and 4-53. Vent configuration in the pedestal region is not applicable, since Susquehanna SES does not have vents in this region. Also, Susquehanna SES does not have access galleys. There are no large horizontal structures in the pool swell zone.
FSAR Rev. 58 021.74-3
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 1 PRIMARY CONTAINMENT M.S.R.V. DISCHARGE IN SUPPRESSION POOL FIGURE 021.741
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 704'0" FIGURE 021.742
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 719'1" FIGURE 021.743
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 738'11 1/2" FIGURE 021.744
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 752'2 1/2" FIGURE 021.745
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 779'1" FIGURE 021.746
Security-Related Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, MISC. SECTIONS FIGURE 021.74-7
Security-Related Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, SECTION C-C FIGURE 021.74-8
Security-Related Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 761'-1" FIGURE 021.74-9
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, SECTION LL FIGURE 021.7410
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 806'0" FIGURE 021.7411
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, SECTION MM FIGURE 021.7412
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, MISCELLANEOUS DETAILS FIGURE 021.7413
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 1 PRIMARY CONTAINMENT M.S.R.V.
DISCHARGE IN SUPPRESSION POOL FIGURE 021.7414
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 1 PRIMARY CONTAINMENT M.S.R.V. DISCHARGE IN SUPPRESSION POOL FIGURE 021.7415
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF E. 704'0" FIGURE 021.7416
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 719'1" FIGURE 021.7417
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 738'11 12" SH. 1 FIGURE 021.7418
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 738'11 1/2" FIGURE 021.7419
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 752'2 1/2" FIGURE 021.7420
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 779'1" FIGURE 021.7421
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 719'1" FIGURE 021.7422
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, MISCELLANEOUS SECTIONS & DETAILS FIGURE 021.7423
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 761'1" FIGURE 021.7424
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT PLANT DESIGN DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN OF EL. 806'0" FIGURE 021.7425
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT INSTRUMENT LOCATION DRAWING REACTOR BUILDING UNIT 1 AREA 26, PLAN BELOW 704'0" FIGURE 021.7426
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REACTOR BUILDING UNIT 1 & 2 PRIMARY CONTAINMENT SUPPRESSION POOL CHAMBER PLATFORM FIGURE 021.7427
SSES-FSAR QUESTION 021.75 Discuss the applicability of the generic supporting programs, tests and analyses to SSES design (i.e., FSI concerns, downcomer stiffeners, downcomer diameter, etc.)
RESPONSE
A complete description of the GKM-IIM test program, test results and evaluation of the test data is provided in Chapter 9.0 of the Susquehanna SES DAR. The GKM-IIM tests were structured to be as prototypical of the Susquehanna SES plant configurations as was practical. As such, concerns related to FSI, downcomers stiffness, downcomer diameter, etc., are fully addressed.
Rev. 51, 02/97 021.75-1
SSES-FSAR QUESTION 021.76 Provide the time history of plant specific loads and assessment of responses of plant structures, piping, equipment and components to pool dynamic loads. Identify any significant plant modifications resulting from pool dynamic loads considerations.
RESPONSE
Time history information for LOCA loads can be found in SSES DAR, Section 4.2.
Similar information due to SRV actuation can be found in SSES DAR, Section 4.1. In addition, the plant specific LOCA and chugging load definition developed from the GKM II-M test program can be found in Subsection 9.5.3. This load definition will be used to evaluate the conservatism of the DFFR LOCA load definition developed from the GKM II-M test program can be found in Subsection 9.5.3. This load definition was used to evaluate the conservatism of the DFFR LOCA load definition.
Assessment of the piping to pool dynamic loads is completed. PP&L interprets this question as requiring:
a) Response of piping in the wetwell to pool dynamic time history loads.
b) Response of piping in the drywell, wetwell and reactor building to response spectra due to SRV and LOCA loads.
Summary of the results of piping analysis has been provided in the DAR upon completion of piping analysis in May of 1981.
Modification of plant design to date a) Addition of quenchers b) Design changes in platform, vacuum breakers, and recombiner Support beams by raising them out of the pool swell zone c) Redesign of downcomer bracing system d) Added 60 reinforcing bars in each suppression chamber e) Added embedments and anchor bolts in suppression chamber walls and diaphragm slab.
f) Diaphragm slab reinforcements changed from 45q to 90q to increase uplift loadings acceptance g) Significant number of pipe supports added or modified.
Rev. 51, 02/97 021.76-1
SSES-FSAR QUESTION 021.77 Provide figures showing reactor pressure, quencher mass flux and suppression pool temperature versus time for the following events:
(1) A stuck-open SRV during power operation assuming reactor scram at 10 minutes after pool temperature reaches 110qF and all RHR systems operable; (2) Same as event (1) above except that only one RHR train available; (3) A stuck-open SRV during hot standby condition assuming 120qF pool temperature initially and only one RHR train available; (4) The Automatic Depressurization System (ADS) activated following a small line break assuming an initial pool temperature of 120qF and only one RHR train available; and (5) The primary system is isolated and depressurizing at a rate of 100qF per hour with an initial pool temperature of 120qF and only one RHR train available.
Provide parameters such as service water temperature, RHR heat exchanger capability, and initial pool mass for the analysis.
RESPONSE
The Susquehanna unique SRV mass and energy release analysis is presented in Appendix I of the DAR.
Rev. 51, 02/97 021.77-1
SSES-FSAR QUESTION 021. 78.
With regard to the pool temperature limit, provide the following additional information:
(1) Definition of the 11 local" and 11 bulk 11 pool temperature and their application to the actual containment and to the scaled test facilities, if anyi and (2) The . data base that support any assumed difference between the local and the bulk temperatures.
RESPONSE
(1) The terms
- 11 local 11 and 11 bulk 11 temperature are used as defined in Subsection III.C.1.a of NUREG 0487, "Mark II Containmenf Lead Plant Program Load Evaluation and Acceptance Criteria, 11 United States Nuclear Regulatory Commission, October, 1978.
Because of .the design features of quenchers and their orientation in the suppression pool (as discussed in the SSES DAR, Subsection 8.5.5), the differences between 11 local" and "bulk" pool temperatures are expected to small. . Therefore, the difference should not exceed the value which was previously derived for ramshead discharge devices in Mark I plants (10°F). It is intended to verify the numbers using data from in-plant tests which are presently under preparation for LaSalle and Zimmer.
Rev. 51, 02/97 021.78-1
SSES-FSAR QUESTION 0*21. 73, For the suppreasion pool temperature monitoring system, provide the following additional information:
(1) Type, number __ and location of temperature instrumentation that will be installed in the pool; and (2) Discussion and justification of the sampling or averaging technique that will be applied to arrive at a definitive pool temperature.
RESPONSE: _
( 1} Please refer to revised Section 7.6.lb.1.2.
Susquehanna SES has* completed evaluation of the suppression_ pool . monitoring criteria as d~fined in NUREG-0487 and has developed a basic system as follows:
0 Number and Location *gr Temperature Instruments i remote temperature detectors {see Figure 021.74-32 in 2O each auppression*pool
-16 remote temperature detectors located just below the min. water level and arranged to provide 2 each on 8 locations around the pool.
-4 remote temperature __ detectors (see Figure 021.74-32-TE's 15769, 15761, 15756, 15751) distribut.e d around the pool at 11 Q 11 centerline location 0 Type: Class lE Instrument - Divisionalized with one f rem each location . in each di vision, except for 4 remote .. temperature detectors at the. 11 Q" centerline.
All sensors will be redundant, Seismic category I and supplied from onsite emergency power; The technique*issued to arrive
- at an average, or bulk, pool temperature is congervative due to the placement of the 16 pool temperature ____detectors. These 16 detectors are evenly distributed near the pool surface, where the hottest water will rise.
Rev. 51, 02/97
SSES-FSAR QUESTION 021. 8.0 The statement . is made in response to Question 021. 51 that operator action (containment spray, ADS) will limit suppression chamber*pressure to 45 psig.
The staff position s *e t forth in Appendix I to Standard Review Plant (SRP} 6.2.1.1.c requires automatic actuation of the spray system. However, if it is demonstrated by analyses that there is sufficient .time (minimum of 30 minutes) between the time the . _
operator becomes aware of the leakage path and the time containment design pressure is reached, manual operator action could be acc-epted as an arternati ve to automated spray actuation. * .To complete our review of .the Susquehanna Steam Electric ~station {SSES), please provide the following information:
- a) Graphically show the containment pressure following small steam break assuming a bypass leakage path with a A/K = ~OS ft 1 .a nd containment spray actuated 30 minutes from the time the suppr,e ssiort. chamber pressure reaches the 30 psig setpoint. The analysis should be based on the assumptions set forth in Appendix I
- to SRP 6.2.1.1.C.
b) Specify the operator action that will be taken and the time to complete the action, i.e., containment spray or ADS and discuss the consequences of each action. Modify subsection 6.2.1.1.5.2
- of the FSAR to address the specific action to be taken.
c) If the analysis stated in item a, above, shows that containment design pressure will be . reached, discuss your plan for including automated actuation of the spray system.
RESPONSE*:
a) The Susquehanna SES construction perm.it was issued in 1973. An automatically actuated spray system was not then. nor is it now in the plant design. The Standard Review Plans are not a design or construction basis for Susquehanna SES.
Question 021. 80 presumes that no effective operator action is performed for thirty (30) minutes afte r becoming aware of a leakage path. This is not the basis for Susquehanna SES design nor the supporting analysis.
b) This information is provided in FSAR Subsection 6.2.1.1.5.2.
c) There *are no plans for including automated actuation of the spray system. The Susquehanna SES configuration is proved *:reliable and effective in licensed, operating plants.
\ Rev . 51, O2 / 9 7 021.80-1
SSES-FSAR QUESTION 021.81 Note-6 in Table 6.2~15 indicates that excess flow check valves outside containment eliminates the bypass leakage path.
Discuss the advisability of using excess flow check valves to perform satisfactorily during the entire course of transient.
RESPONSE
See Subsection 6.2.4.3.2.2 for*** a discussion of the design of these__ lines.
Rev. 51, 02/97
SSES-FSAR QUESTION 021.82 The response to question 021. 64 _is incomplete; discuss whether or not jockey pumps are used to maintain water seal for 30 days. Discuss if the system is single failure (active) proof.
RESPONSE:_.
This information has been provided in FSAR Subsection 6.2.3.2.3.1.
I Re.v. 51, 02/97 021. 82-1
SSES-FSAR QUESTION 021; 83 The response to. Question 021. 67 is incomplete; provide the analysis requested in Item 2 of that question.
RESPONSE:* . .
Susquehanna will use an inerted primary containment during power operation. At the maximum oxygen concentration of 4%,
any hydrogen concentration forms a non-explosive mixture. Hence the hydrogen generation from
- z,inc is non-consequential.
Rev. 51, 02/97 021.83-1
SSES-FSAR QUESTION 021.84 Table 6. 2-12, a list of the containment isolation valves, references Figure 6.2-44 for valve configuration . Some of the valve arrangements on Figure 6.2-44 do not match those presented in Table 6.2-12; e.g., penetrations 9A, 244, 245, and 246A do not match the referenced arrangement.
In addition, Subsection 6.2.4.3.3.5 does not agree with the arrangement of penetrations 23 and 24. Review Table 6. 2-12 fo::::
completeness and correctness.
RESPONSE
The valve arrangements of Figure 6.2 - 44 have been revised to match those in Table 6.2-12 for penetrations 9A,B, 244, 245, and 246 A, B.
Subsection 6.2.4.3.3.5 and the valve arrangements of penetrations 23 and 24 have been revised to agree with Table 6.2-12.
Rev. 53, 04/99 021.84-1
SSES-FSAR QUESTION 021.. as .
Provide justification for having a check valve outside the containment as containment isolation for the seismic pump seal water supply lines.
RESPONSE
See Subsection 6.2.4.3.2.2.*
Rev. 51, 02/97 021.85-1
SSES-FSAR Question Rev. 52 QUESTION 021.86 Arrangement e in Figure 6.2-44 shows that the 2-inch bypass line relies on the 24-inch purge valve Hne to perform its intended function. We will require both the 2-inch and the 24-inch valves to meet the requirements set forth in CSBBTP 6-4 or provide another valve arrangement
, that satisfies GDC 56.
It should be noted that while in modes 1 through 4, purging operatrons are permitted up to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year provided that requirements set forth in BTP 6-4 are met.
RESPONSE
Dwgs. M-157, Sh. 1, M-157, Sh. 2, and M-157, Sh. 3 for the piping diagram, Subsection 6.2.4.2 for a description of the containment isolation system, and Subsections 6.5.3.1 and revised 6.2.5.2 for a description of containment purging.
During power operation, the purge valves will be controlled as specified in the Technical Specrfications.
FSAR Rev. 58 021.86-1
SSES-FSAR Question Rev. 54 QUESTION 021.87 With respect to the leakage test program .
a) It is our position that feedwater isolation valves should be Type C tested utilizing air.
b) RHR shutdown supply and return should be Type C tested utilizing air.
c) ECCS injection valves should be tested utilizing air.
d) All containment isolation valves should be Type C tested . Hydrostatic testing is acceptable if it can be demonstrated that the water inventory is sufficient to maintarn a water seal for at least 30 days following a LOCA.
e) CRD insert and withdraw line should be vented during Type A test.
RESPONSE
a) Refer to revised Subsection 6.2.3.2.3.
b) See Table 6.2-22.
c) See Table 6.2-22.
d} Refer to Table 6.2-22 for a description of the testing of containment isolation valves .
e) Refer to Table 6.2-22. Note 20.
FSAR Rev. 59 021.87-1
SSES-FSAR QUESTION 021.88 Provide the projected areas used in_ the calculation of forces on the RPV and supports.
RESPONSE
This information is contained in the . revised Section 6A, specifically the new Table 6A-7.
I Rev. 51, 0 2 / 9 7 021.88-1